ML21130A066
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USAR APPENDIX C STRUCTURAL LOADING CRITERIA PAGE 1.0 SCOPE C-1-1 2.0 CONCRETE AND STEEL STRUCTURES C-2-1 2.1 Description C-2-1 2.2 Loads C-2-1 2.3 Load Combinations and Allowable Limits C-2-2 2.4 Method of Analysis C-2-5 2.5 Implementation of Criteria C-2-6 2.5.1 Reactor Building Foundation C-2-6 2.5.2 Reactor Building Floor System C-2-6 2.5.3 Reactor Building Concrete Walls C-2-6 2.5.4 Reactor Building Steel Superstructure C-2-10 2.5.5 Drywell Biological Shield Wall C-2-13 2.5.6 Reactor Pedestal C-2-15 2.5.7 Primary Containment C-2-18 2.5.7.1 Mark I Containment Program C-2-24 2.5.7.1.1 Program History C-2-24 2.5.7.1.2 Short-Term Program C-2-25 2.5.7.1.3 Long-Term Program C-2-25 2.5.7.1.4 Loads C-2-26 2.5.7.1.5 Load Combination & Allowable Limits C-2-26 2.5.7.1.6 Method of Analysis C-2-27 2.5.7.1.7 Modifications Completed C-2-27
- 2. 5. 7. 2 Pipe Whip and Jet Force Analysis C-2-29 2.5.7.2.1 General C-2-29 2.5.7.2.2 Jet Force Analysis C-2-29 2.5.7.2.3 Pipe Whip Analysis C-2-30 2.5.7.2.4 Pipe Whip Energy Absorbing Material C-2-33 2.5.8 Main Steam Tunnel C-2-34 2.5.9 Sacrificial Shield Wall C-2-35 2.5.10 Concrete Attachments C-2-36
- 2. 5 .11 Control Building C-2-42 2.5.12 Elevated Release Point C-2-42 3.0 COMPONENTS C-3-1 3.1 Intent and Scope C-3-1 3.1.1 Components Designed By Rational Stress Analysis C-3-1 3.1.2 Components Designed Primarily By Empirical Methods C-3-1 3.2 Loading Conditions and Allowable Limits C-3-1 3.2.1 Loading Conditions C-3-1 3.2.1.1 Normal Conditions C-3-2 3.2.1.2 Upset Conditions C-3-2 3.2.1.3 Emergency Conditions C-3-2 3.2.1.4 Faulted Conditions C-3-2 3.2.2 Allowable Limits C-3-2 3.3 Method of Analysis and Implementation of Criteria C-3-74 3.3.1 Reactor Pressure Vessel C-3-74 3.3.1.1 Reactor Pressure Vessel Fatigue Analysis C-3-74 3.3.1.2 Reactor Pressure Vessel Seismic Analysis C-3-74 c-1-1 01/02/15
USAR APPENDIX C STRUCTURAL LOADING CRITERIA (CONT'D)
PAGE 3.3.1.3 Reactor Pressure Vessel Support Structure C-3-74 3.3.1.4 Miscellaneous Associated Components and Supports C-3-74 3.3.2 Reactor Pressure Vessel Internals C-3-75 3.3.2.1 Internals Deformation Analysis C-3-75 3.3.2.2 Internals Fatigue Analysis C-3-75 3.3.2.3 Internals Seismic Analysis C-3-76 3.3.3 Piping C-3-77 3.3.3.1 Piping Flexibility Analysis C-3-77 3.3.3.2 Piping Seismic Analysis C-3-77 3.3.3.3 Pipe Rupture Loading C-3-78 3.3.3.4 Drywell Penetration Limit Stops C-3-78 3.3.3.5 Other Piping Analyses C-3-79 3.3.3.5.1 Main Steam Piping C-3-79 3.3.3.5.2 SRV Discharge and Torus Attached Piping C-3-80 3.3.3.5.3 Replacement Piping C-3-82 3.3.3.5.4 Intake Structure Piping C-3-82 3.3.4 Equipment C-3-82
4.0 REFERENCES
FOR APPENDIX C C-4-1 c-1-2 02/18/05
USAR LIST OF FIGURES (At End of Appendix C)
Figure No. Title C-2-1 Reactor Bldg. Dynamic Analysis Model C-2-2 Control Bldg. Dynamic Model C-2-12 Reactor Bldg. Total Acceleration c-2-1 03/08/01
USAR LIST OF TABLES Table No. Title Page C-2-1 Reactor Building Foundation C-2-7 C-2-2 Reactor Building Floor System C-2-8 C-2-3 Reactor Building Concrete Walls C-2-9 C-2-4 Reactor Building Structural Steel Columns C-2-11 C-2-5 Reactor Building Steel Roof Truss-Frame Above El. 117' C-2-12 C-2-6 Drywell Shielding Concrete C-2-14 C-2-7 Reactor Concrete Pedestal C-2-16 C-2-8 Drywell Membrane Stresses C-2-19 C-2-9 Jet Impingement Force Stresses C-2-20 C-2-10 Drywell Stabilizer Shear Lugs C-2-22 C-2-11 Summary of Mark I Containment and Piping Modifications C-2-28 C-2-12 Original Tabulation of Critical Pipe Stresses Due to C-2-32 Dead Weight Plus Pressure, Plus Operating Basis Earthquake, Plus Thermal Loads C-2-13 Control Building Foundation C-2-39 C-2-14 Control Building Floor System C-2-40 C-2-15 Control Building Concrete Walls C-2-41 C-2-16 Elevated Release Point Foundations and Tower C-2-43 C-3-1 Loading Condition Probabilities C-3-4 C-3-2 Deformation Limit C-3-5 C-3-3 Primary Stress Limit C-3-6 C-3-4 Buckling Stability Limit C-3-8 C-3-5 Fatigue Limit C-3-9 C-3-6 Minimum Safety Factors C-3-10 C-3-7 Loading Criteria C-3-11 Reactor Vessel Internals and Associated Equipment C-3-11 Main Steam Piping C-3-24 Recirculation Loop Piping C-3-26 Class IN/IS - Core Spray Discharge Piping CS-ID C-3-27 Class IN/IS - Clean-Up Recirculation Pump Suction C-3-28 Piping CU-IS Class IN/IS - Main Steam Piping to HPCI Turbine and C-3-29 Residual Heat Exchangers lA and lB Class IN/IS - Reactor Feed Piping C-3-31 Class IN/IS - Residual Heat Removal Pump Suction C-3-33 Piping Class IN/IS - Residual Heat Removal Pump Discharge C-3-35 Piping c-3-1 01/02/15
USAR LIST OF TABLES Table No. Title Page Recirculation Pumps C-3-37 RHR Pump C-3-41 Core Spray Pump C-3-43 HPCI Pump C-3-45 RCIC Pump C-3-47 Standby Liquid Control Pump C-3-49 HPCI Turbine C-3-51 RCIC Turbine C-3-55 Main Steam Isolation Valves C-3-59 Main Steam Safety Valves C-3-62 Main Steam Relief Valves C-3-67 Recirculation Valves C-3-70 Standby Liquid Control Tank C-3-73 c-3-2 03/08/01
USAR APPENDIX C - STRUCTURAL LOADING CRITERIA 1.0 SCOPE This appendix provides additional information pertinent to the preceding sections concerning the structural loading criteria applied to Class I structures and components. Station structures and components are classified according to service and location. Class I Seismic structures and components are listed in USAR Section XII-2.1.2.
The loads, loading combinations and allowable limits described in this appendix apply only to Class I structures and components as defined in USAR Section XII-2.1.1. The criteria in this appendix are intended to supplement applicable industry design codes where necessary. Compliance with these criteria is intended to provide design safety margins which are appropriate to extremely reliable structural components when account is taken of rare event potentialities associated with a Safe Shutdown Earthquake or postulated loss-of-coolant accident or a combination thereof.
Class I components may not always be designed to satisfy the criteria using analytical techniques; alternately, the design of some components may be based upon test results, empirical evidence, or experience, including the use of earthquake experience and test data according to the SQUG Generic Implementation Procedure (GIP). This method was used in the resolution of USI A-4 6 and may be used for the design and modification of Class I equipment within the scope of GIP-3. GIP-3 is further described in this Appendix.
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USAR 2.0 CONCRETE AND STEEL STRUCTURES 2.1 Description The Class I concrete and steel structures are designed considering three interrelated primary functions for the design loading combinations described in Subsection C-2. 3. The first consideration is to provide structural strength equal to or greater than that required to sustain the combination of design loads and provide protection to other vital Class I structures and components. The second consideration is to maintain structural deformations within limits such that Class I components will not experience loss of function. The third consideration is to preclude excessive leakage by preventing excessive deformation and cracking, when containment integrity is required.
In general, the load combinations considered and their allowable limits are formed on a "quasi-probabilistic" basis. This means that the higher the probability that a given set of conditions could occur, the lower the allowable limits. This also forms the basis for neglecting some combinations of loads. In general, only one highly improbable event is considered in any load combination, since the probability of two unrelated, highly improbable events occurring simultaneously is vanishingly small.
2.2 Loads The loads considered in the design of Class I concrete and steel structures include the following:
D Dead load of the structure and related equipment plus any other permanent loads contributing stresses, such as soil or hydrostatic loads; live loads expected to be present when the station is operating; and the loads due to thermal expansion under normal operating conditions. This load takes into account any deviations from normal operating conditions which are reasonably expected to occur during the design lifetime of the station.
R Loads resulting from jet forces and pressure and temperature transients associated with rupture of a single pipe within the primary containment.
R' Loads due to a High Energy Line Break (HELB) outside containment. This includes the effects of pressure and temperature transients, pipe whip impact forces, and reactions from pipe anchors. [25 l E Loads due to the Operating Basis Earthquake (OBE)
(0. l0g horizontal ground acceleration; 2/3[ll of horizontal ground acceleration applied simultaneously for vertical seismic acceleration) These loads are also known as Maximum Probable Design Earthquake loads.
E' Loads due to the Safe Shutdown Earthquake (SSE)
(0.20g horizontal ground acceleration; 2/3[ll of horizontal ground acceleration simultaneously applied C-2-1 03/08/01
USAR for vertical seismic acceleration) . These loads are also known as Hypothetical Maximum Possible Design Earthquake loads.
Flood Loads due to flooding the Drywell up to 23 feet above the Drywell flange level.
w Design wind loading conditions. Refer to USAR Section XII-2.3.3.1.
T Loads due to the effects of a tornado. Refer to USAR Section XII-2.3.3.2.
Primary Containment System components associated with the Mark I program are evaluated for the loads as described in the CNS Plant Unique Analysis Report and in accordance with NUREG 0661.
2.3 Load Combinations and Allowable Limits The loads combinations and allowable limits considered in the design of Class I concrete and steel structures include the following (also see USAR Table XII-2-1).
Load Combination Limits D+E Stresses remain within normal code allowable stresses (AISC 6th Edition for structural steel, ACI 318-63 for reinforced concrete, ASME B&PV Code Section III
[Class BJ, S67 for Primary Containment). The customary increase in design stress for earthquake loadings is not permitted.
D+W Maximum allowable stresses may be increased one-third above normal code allowable stresses.
D+R+E Stresses remain within normal code allowable stresses.
The customary increase in design stress for earthquake loadings is not permitted. In the case of jet impingement loading on Primary Containment, where it is backed up by concrete, it may be assumed that local yielding may take place but it shall be established that a rupture will not occur. For jet impingement loading on Primary Containment (including containment penetration assemblies) where Primary Containment is not backed up by concrete, the primary stresses must not exceed 90% of the yield strength of the material at 300°F.
D+E+Flood Local membrane stresses in Primary Containment may exceed the yield point, but with a margin against rupture.
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USAR Load Combination Limits D+T Maximum allowable stresses are as follows:
Steel - AISC 6 th Edition allowable yield stresses with a maximum of 90% of the allowable yield stress.
Concrete - ACI 318-63 "Working Stress Design" method with allowable stresses in concrete of O. 85 f' c, and allowable stresses in reinforcing steel of 0.9 fy.
Alternatively, when the "Ultimate Strength Design" method is used a load factor of 1.0 is applied to this load combination with appropriate reduction factors as described in ACI 318-63.
D+R+E' Maximum allowable stresses are as follows:
Steel - AISC 6 th Edition allowable yield stresses with a maximum of 90% of the allowable yield stress. In the case of jet impingement loading on the primary containment, where it is backed up by concrete, it may be assumed that local yielding may take place but it shall be established that a rupture will not occur. For jet impingement loading on the primary containment (including containment penetration assemblies) where the primary containment is not backed up by concrete, the primary stresses must not exceed 90% of the yield strength of the material at 300°F.
Concrete - ACI 318-63 "Working Stress Design" method with allowable stresses in concrete of O. 8 5 f' c, and allowable stresses in reinforcing steel of 0.9 fy.
Alternatively, when the "Ultimate Strength Design" method is used a load factor of 1.0 is applied to this load combination with appropriate reduction factors as described in ACI 318-63.
D+R'+E' The allowable stresses are generally limited to 90% of their ultimate values, i.e., flexural stresses in concrete are limited to 0.85 f'c, and stresses in reinforcing steel are limited to 0.90 fy (note in certain instances, stresses in reinforcing steel up to approximately 1.0 fy were considered acceptable for this extreme load combination) . [251 Primary Containment System components associated with the Mark I program were evaluated for the load combinations and allowable limits as described in the CNS Plant Unique Analysis Report, and in accordance with NUREG 0661. Allowable stresses for these components are based on the requirements of the ASME B&PV Code,Section III, S77.
Governing Codes This appendix identifies the governing codes and supplementary requirements for the design and installation of structures. Repairs, C-2-3 03/19/18
USAR replacements or modifications of structures may be performed to these requirements or to the requirements of later editions of the construction codes provided the safety design bases described in the USAR are maintained.
The following codes are the codes of record for Class I structures:
For structural steel, the governing code is the American Institute of Steel Construction Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, adopted April 17, 1963 (referred to herein as AISC 6 th Edition).
For concrete structures, the governing code is the American Concrete Institute Standard ACI 318-63, Building Code Requirements for Reinforced Concrete, as amended March 6, 1963 (referred to herein as ACI 318-63).
For the original design for Primary Containment, the governing code is the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Vessels, Subsection B ( for Class B Vessels), 1965 Edition using addenda up to and including the Summer 1967 Addenda (referred to herein as ASME B&PV Code,Section III [Class B], S67).
For the reevaluation of the portions of the Primary Containment System associated with the Mark I program, the governing code is the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Division 1, 1977 Edition using addenda up to and including Summer 1977 Addenda (referred to herein as ASME B&PV Code,Section III, S77).
Allowable Stresses The allowable stresses for Class I structures are generalized in USAR Table XII-2-1. The definition of f'c is the specified compressive strength of concrete in pounds per square inch at a specified age (normally at 28 days) . At CNS the predominant concrete mix is specified to have a minimum 28 day strength of 3,000 psi using Type II cement. £4 l} The actual strengths used are listed in the following paragraphs.
The concrete mix specified for the following structural parts of buildings has a minimum 28 day strength of 4,000 psi: Reactor Building foundation mat and substructure, Control Building foundation and substructure, Radwaste Building foundation mat and substructure, substructure of the Intake Structure.
The concrete mix specified for the following structural parts of buildings has a minimum 28 day strength of 3,000 psi: Reactor Building superstructure, Reactor Pedestal, Control Building superstructure, Radwaste Building superstructure, superstructure of the Intake Structure, all parts of the Diesel Generator Building and Control Corridor.
The concrete mix specified for the Containment Pedestal and leveling slabs of all buildings have a minimum 28 day strength of 2,000 psi.
Allowables for load combinations involving E', T, R' and certain combinations involving R are based on a criterion for "no loss of function" such that safe shutdown can be achieved. The criterion for "no loss of function" for structures is that stresses will remain in the elastic range. E45 l Concrete compressive stresses due to bending are limited to O. 85f'c, reinforcing steel stresses are limited to 90% of yield, and tensile and bending stresses in structural steel members are limited to 90% of their respective yield stresses. l 61 Structural deformations are not the controlling criteria for safe shutdown structures (see USAR Section XII-2 .1 for a list of safe shutdown structures) . Concrete crack control is maintained by limiting reinforcing steel stresses to those given in Table XII-2-1.
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USAR The book "Design of Concrete Structures," [2 J shows stress-strain curves for a member under axial compression. Figure 2. 2 of this reference shows that the concrete stress-strain curve for fast rate of loading (typical for earthquake or tornado) does not deviate appreciably from a straight line below a stress level of 0.85f c (2,550 psi).
1 In addition, concrete gains additional strength with age.
Figure 22 from the "Concrete Manual" [3 l shows the effect of age on compression strength. A mix having a minimum 28 day compressive test strength ( f' c) of 3,000 psi will approximate minimum test strengths of 3,700 psi at 90 days, 3,900 psi at six months, 4,100 psi at one year, and 4,200 psi at two years.
At the time of fuel loading of the reactor the concrete had attained a minimum age of at least nine months, and at full operation more than 14 months. Thus, the allowable design stress of 0.85f'c (2,550 psi) is in actuality only approximately 64% to 67% of the ultimate concrete strength.
Based on the above reasons, it is concluded that stress distribution is linear, and the use of O. 8 5f' c for the allowable concrete stress for the working stress design method is a valid approach.[ 4 l All structures, except the Drywell Biological Shield Wall, are designed considering uni-axial stress conditions. As such, the allowable stresses for shear, as a measure of diagonal tension, for bond and anchorage are as given in the ACI 318-63 Code. [SJ 2.4 Method of Analysis A dynamic seismic analysis was performed for Class I structures shown in Burns and Roe Drawings 2050, 2051, 2052, 2053, 2054, 2056, 2059, 2060, 2061, 2062, 2063, 2064, 2065, 2066, 2067, 2068, and 2069. The structural analysis is performed using various calculational methods and techniques. Much of the structural design is performed utilizing the "Working Stress Design" method as defined in ACI 318-63 and in the AISC 6 th Edition.
Some portions of the Class I structures are designed by the "Ultimate Strength Design" method described in Chapters 15 through 19 of ACI 318-63.
Load combinations and allowable limits on stresses are as shown in the USAR Section C-2.3. Structural ductility of concrete members is provided by proportioning the structure so that a calculated value of 0.9 fy for the main tension reinforcement is the limiting stress value at the critical cross section in flexure.
No methods of inelastic analysis have been used on any of the components discussed in the appendix.[7l Only uni-dimensional stress conditions are considered to exist in all Class I steel structures other than the steel containment vessel.[BJ The allowable stresses appropriate to these uni-dimensiona l analyses are as generalized in USAR Table XII-2-1.
USAR Section II-5.2 describes the bases for the selection of maximum horizontal and vertical ground accelerations associated with the Safe Shutdown Earthquake and the Operating Basis Earthquake. The original vertical seismic acceleration component was specified as one-half of the horizontal ground acceleration. This design requirement was increased to two-thirds of the horizontal ground acceleration, which is the current Licensing/Desig n Basis. [ll Evaluations were performed to confirm that the increase in vertical seismic acceleration component was accommodated by existing margins and conservatisms in the original design analyses of the Principal Class I Structures. [67 l As described in USAR Section XII-2.3.5.2.1, the dynamic analysis consisted of 1) developing a mathematical model, 2) performing the analysis, C-2-5 01/02/15
- 3) obtaining a structural response, and 4) plotting the response spectra. For example, mathematical models of the Reactor and Control Buildings are shown in Figures C-2-1 and C-2-2, respectively.
2.5 Implementation of Criteria 2.5.1 Reactor Building Foundation (See Table C-2-1)
The soil bearing stress under the Reactor Building is derived from the application of all dead and uniformly distributed live loads on each main floor to allow for vertical loads during station operation and refueling. Several load cases are examined depending on how much live load and water is considered to be present. Seismic loads for the OBE and SSE are also considered.
The primary objective is to establish whether soil bearing stresses will be within the allowable stresses, that there is no possibility of excessive uneven settlement, and that there will be no possibility of soil liquefaction when subject to seismic vibrations.
The analysis shows that ( 1) maximum stresses are below allowable values, (2) relative settlement will be small, and (3) that liquefaction will not occur.
2.5.2 Reactor Building Floor System (See Table C-2-2)
Reactor Building floor systems are designed for the loadings tabulated in Subsection XII-2. 3. 2. The conservative assumption is made that the Operating Basis Earthquake (E) or the Safe Shutdown Earthquake (E')
vertical load could occur simultaneously with other vertical live loads.
Wind (W) and tornado (T) loads do not apply. Accident loads (R) are included where applicable.
The resulting stresses are within the Code allowable values, therefore, the floor systems are structurally adequate without excessive deflections or cracking of concrete.
2.5.3 Reactor Building Concrete Walls (See Table C-2-3)
The concrete walls of the Reactor Building are analyzed for the design loads (D) combined with seismic loads due to the Operating Basis Earthquake (E) or Safe Shutdown Earthquake (E'). Accident loads (R) are included where applicable. Tornado loads (T) were checked and found governing for the tornado generated missile criteria. Stresses due to seismic moments and shears were calculated, and combined together with the gravity stresses.
Jet forces are postulated to occur only in the Main Steam Line Tunnel ( see USAR Section C-2. 5. 8) * [ 121 Equipment reactions are taken as forces applied to the supporting structure and the structure is designed accordingly. The Reactor Building exterior walls were investigated for penetration by the postulated tornado generated missiles as described in USAR Section XII-2.3.3.2.
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USAR TABLE C-2-1 REACTOR BUILDING FOUNDATION Max.
Load Allowable Description/Criteria Method of Analysis Combination* Stress-ks£ Design load (D) includes all Max. net soil stress D+E 1. Sqa = 18 dead and equipment loads plus 25% of the full live loads. D + ~p + M (KSF) - qa net avg.
A s contact pressure (aDa+ Ysat+Db)
Vertical seismic load (~P) is where: D+E' 1. Sqa 18 assumed to be produced by a D Design load, kips vertical acceleration equal to ~p vert seismic load, kips 2/3 of the ground horizontal M horiz. seismic seismic acceleration, both overturning moment acting simultaneously** kip-ft A foundation mat area, E Loads due to OBE Overturning moment (M) is sq-ft E' Loads due to SSE produced by the horizontal s section modulus of seismic acceleration and all foundation mat ft 3 design load eccentricities y unit weight of soil above water table Ysat saturated unit wt. of soil Da depth of soil above water table Db depth of saturated soil above bottom of foundation
- Floor design load (D) includes dead loads, equipment loads, plus 25% of the full live loads (appropriate for operating conditions) for combination with seismic or tornado loads.
Floor design is also checked for maximum live loads (100 to 1,000 psf) not in combination with seismic or tornado loads.
- Vertical= 1/2 horizontal per PSAR; 2/3 was used for conservatism.
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USAR TABLE C-2-2 REACTOR BUILDING FLOOR SYSTEM Max.
Load Allowable Description/Criteria Method of Analysis Combination* Stress-ksi Floor system is designed for Working Stress Design Method D + E Fe 1. 35 the loadings tabulated in Fv 0.060 Subsection XII-2.3.2. Ft 20.0 Materials conform as follows: Same as for D + E' D + E' Fe 2.55 Concrete f'e = 3,000 psi at Fv 0.093 28 days max. strength per Ft 36.0 ACI 318-63; Reinforcing ASTM Designation: A615 Grade 40 per ACI 318-63; Structural Steel ASTM Designation: A36-67 per AISC Manual, 1963 Vertical seismic load is assumed to be produced by a vertical acceleration equal to 2/3 of horizontal ground acceleration**
Max. allowable stresses for D+E' load combination are: Concrete Fe= 0.85 f'e Reinforcing Ft= 0.90 fy Structural Steel= 0.90 fy
- Maximum allowable Stresses for D + E combination are not increased above code allowable values, i.e.,
customary 1/3 increase not used.
- Vertical= 1/2 horizontal per PSAR; 2/3 was used for conservatism.
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USAR TABLE C-2-3 REACTOR BUILDING CONCRETE WALLS Max.
Load Allowable Description/Criteria Method of Analysis Combination Stress-ksi Design Load (D) includes all Working Stress Design of D + E Ft 20.0 dead and equipment loads, plus ACI 318-63 Fe 1. 35 25% of the full live loads. Fy 0.060 Materials: Concrete Working Stress Design Method of D + E' Ft 36.0 f'e = 3,000 psi at 28 days max. ACI 318-63 Fe 2.55 strength per ACI 318-63 Fv 0.093 Reinforcing ASTM Designation:
A615 Grade 40 per ACI 318-63 Maximum allowable stresses for D+E' load Concrete Fe = 0. 8Sf e 1
Reinforcing Ft= 0.90 fy Maximum allowable stresses for D+E load combination are not increased above code allowable values, i.e., customary 1/3 increase not used.
Note: Ft and ft are tension in reinforcing Fe and f'e are compression in concrete Fv and fv are shear in concrete C-2-9 03/08/01
USAR The resulting stresses are within the Code allowable values, therefore, the Reactor Building walls are structurally adequate without excessive deflections or cracking of concrete.
As described in USAR Section XII-2.3.5.1.5, removable block walls are provided in the Reactor Building to satisfy equipment access requirements. These walls are shown in Burns and Roe Drawing 4215, and their horizontal seismic acceleration curve is shown in Figure C-2-12.
2.5.4 Reactor Building Steel Superstructure (See Tables C-2-4 and C-2-5)
The Reactor Building structural steel superstructure (truss frame and columns supporting roof and siding) are evaluated for various combinations of design loads (D), earthquake loads (E and E'), wind loads (W), and tornado loads (T). Design loads include dead loads, live loads and crane loads. While crane live loads are not considered as a controlling load combination with lateral loading in the Burns and Roe analysis, subsequent evaluation has demonstrated the 108 ton live load combined with earthquake loading is still bounded by the tornado load without crane live loads. r74 l Two cases of tornado loading are considered. Case A considers the siding to be intact and subject to a 75 psf pressure, and Case B assumes the siding to have blown off and the remaining superstructure to be subjected to full force tornado loads (270 psf due to 300 mph winds). For Case A, wind loads are taken as 45 psf on the walls, and as a 34 psf uplift on the roof due to 100 mph winds. For Case B, the roof is considered to remain intact and it is assumed that no pressure differential exists between the top and bottom of the roof. Tornado loads were found to govern over Safe Shutdown Earthquake (SSE) loads. D + W was considered and found not to be governing when compared to D + E or D + T. The analysis method is the Allowable Stress Design Method of the AISC 6th Edition considering unsymmetrical beam loading where applicable. For design loads in combination with wind or Operating Basis Earthquake (OBE) loads, the resultant stresses calculated are below the Code allowable stresses; therefore, the integrity of secondary containment will be assured. For design loads combined with SSE loads, it was obvious by inspection that the increase in stresses caused by the SSE would be small because of the relatively small masses involved; therefore, secondary containment integrity will be assured.
The siding will blow off when the design wind velocity of 100 mph is appreciably exceeded ( 7 5 psf in or out) . r 121 This is assured by the use of necked-down steel "control release fasteners" between the girt system and the supporting columns. A rigorous test and quality assurance program was instituted to provide assurance that these fasteners will fail at a maximum 75 psf. In addition, the barometric pressure drop that precedes the tornado need only be about1/2 psi before the siding blows out rather than inward.
The exposed structural steel frame of the Reactor Building superstructure has been designed to withstand winds of 300 mph intensity with stresses limited to 90% of yield. The 300 mph tornado forces have also been applied to the surfaces of the Reactor Building crane. Reactor Building metal siding is designed for normal 100 mph wind loading. The steel superstructure is not designed for tornado missiles.
C-2-10 11/22/10
USAR TABLE C-2-4 REACTOR BUILDING STRUCTURAL STEEL COLUMNS Max.
Load Allowable Description/Criteria Method of Analysis Combination Stress-ksi Material: ASTM Designation: Working Stress Design D + LL FA= 18.3 per AISC A36 per AISC Manual, 1963 Method considering + Crane Load FB = 22.0 unsymmetrical beam (Upper section crane columns) loading where applicable Structural Steel Column FA= 17.4 designed for all dead and FB=22.0 live loads as follows: (Lower section crane columns)
Wheel reactions from 108 ton crane in combination FA=24.3 with roof DL plus 20 psf FB = 29.2 LL plus wind@ 45 psf on (Upper section crane columns) building FA= 23.2 Vertical seismic load is FB = 29.2 assumed to be produced by a (Lower section crane columns) vertical acceleration equal to1/2 of horizontal ground FA= 17.4 acceleration plus normal DL FB = 22.0 partial LL crane DL (Lower section crane columns)
Tornado load - building Same as Above DL + Tornado FA = 27 siding remains intact until on structural FB=32.4 l
75 psf loading is reached at steel only (Upper section crane columns) which time it is considered removed and full tornado load FA = 27 is applied to the structural FB = 32.4 steel members (Lower section crane columns)
C-2-11 11/22/10
USAR TABLE C-2-5 REACTOR BUILDING STEEL ROOF TRUSS-FRAME ABOVE EL. 117' Max.
Method of Load Allowable Description/Criteria Analysis Combination* Stress-ksi Material: Structural steel Working D + E FA = 1 7. 62 comp.
ASTM Designation: A36 per Stress (Top chord)
AISC Manual, 1963 Design D + T FA = 2 6 . 6 comp .
(Top chord)
Design wind loading: 45 psf on walls due to 100 mph wind: D + E FA= 22.0 ten.
uplift 34 psf on roof due to (Bottom chord) 100 mph wind: dead weight of D + T FA= 32.4 ten.
roof deck and built-up roofing (Bottom chord) 8 psf D + E FA= 22.0 ten.
Tornado wind load (T) on walls (Diagonal member) averages a pressure of 270 psf D + T FA= 32.4 ten.
on exposed surfaces after (Diagonal member) siding has blown off at
+/- 75 psf D + E FA = 15. 4 comp.
(Truss vertical) 108 ton capacity crane on D + T FA = 22. 6 comp.
column bracket (Truss vertical)
D + E FA= 22.0 ten.
(Lower chord bracing)
D + T FA= 32.4 ten.
(Lower chord bracing)
E' load disregarded because tornado is governing at similar allowable stresses of
- 0. 9 Fy
- D+W was considered and found not governing; for D+W the margins between the maximum allowable stresses and the calculated stresses were greater than for D+E or D+Tornado.
C-2-12 11/22/10
USAR 2.5.5 Drywell Biological Shield Wall (See Table C-2-6)
The comments regarding the Reactor Building concrete walls ( see USAR Section C-2. 5. 3) generally apply also to the Drywell Biological Shield Wall. Drywell Safe Shutdown Earthquake loads control over tornado loads. The concrete surface is protected by the Drywell vessel which acts as a thermal shield, and by a 2 inch air gap which impedes the flow of heat. Thus, the transient itself has little effect on the concrete. ll 3 J The Drywell Biological Shield Wall structure is designed considering a bi-axial stress condition. There are vertical compressive stresses and tensile hoop stresses. The plane of shear resistance is a horizontal plane upon which compressive stresses act and is unaffected by the tensile hoop stresses. For the longitudinal or seismic shear the allowable stress is considered equal to the allowable shear stress, as a measure of diagonal tension, and as given in the ACI 318-63 Code. (The 1971 edition of the same code specifies a nominal permissible shear stress for nonprestressed concrete members equal in value with that prescribed for the nominal permissible shear stress in the plane of the shear walls: Ve = 2 ( f c) 0
- 5
- The 1
radial shear stress caused by discontinuities is considered to be diagonal tension and the allowable stress is given in the ACI 318-63 Code.
For the splices in the tensile hoop reinforcement in the Drywell Biological Shield Wall structure the allowable stress for bond is as given in the ACI 318-63 Code. These Code allowables are for a uni-axial stress condition. They should be conservative for this structure since the bi-axial compression which exists on the splices should improve the bond capability of the concrete.
The analysis shows that the stresses in the Drywell Biological Shield Wall will not exceed the allowable values.
No equipment or piping inside the drywell is anchored to the Drywell Biological Shield Wall. Drywell penetrations passing through the biological shield wall are not anchored to the biological shield wall annular sleeves. However, when required by a rupture analysis of the Drywell penetration nozzles, and/ or when required to provide positive bellows seal protection, limit stops were utilized. For penetrations which have expansion joint bellows seals, limit stops were provided to assure that the bellows seals would not fail due to excessive torsional rotation, or due to axial collapse of the joint at the time of pipe rupture. Stress analysis indicated that some penetrations require only lateral limit stops and, therefore, no axial or torsional limit stops are required for this type of penetration. The limit stops are normally disengaged. The limit stops consist of sleeves, embedded in the biological shield wall and extend to the flued head fitting outside the shield wall, coaxially with the containment penetration. In the vicinity of the flued head fitting, a flange is attached to the sleeve extension. A gap is maintained between the sleeve and the penetration assembly so that no contact is made due to thermal differential movements.
When rupture loading is applied to the containment penetration, the assembly deflects and contact is made between the flued head fitting and the limit stops at the flange. The flange, which is provided at this location, serves as a stiffening ring which transmits the ensuing rupture loading to the embedded sleeve, which in turn transmits the load to the biological shield wall. [13J [23J C-2-13 03/28/01
USAR TABLE C-2-6 DRYWELL SHIELDING CONCRETE Max.
Allowable Description/Criteria Method of Analysis Load Combination Stress-ksi Drywell shield acts as a Working Stress Design. Thermal D + E + thermal Ft 20 structural wall carrying stresses are included in the gradient floors, Design load (D) analysis. Seismic forces are Fe 1. 35 consists of all dead loads, super- imposed on the results.
equipment loads, plus 25% of Fv 0.060 full live load Seismic loads (E&E') are D + E' + R (Jet Ft 36.0 according to the response due to any pipe spectra for the reactor break within the Fe 2.55 building spherical area of the drywell Fv 0 .113 Accident load (R) includes section+ thermal 60°F normal operational gradient) thermal gradient in winter and jet forces due to ruptured pipe Materials conform as follows:
Concrete f'e = 3,000 psi at 28 days minimum strength per ACI 318-63 Reinforcing ASTM Designation:
A615 Grade 40 per ACI 318-63 Structural Steel ASTM Designation: A36-67 per AISC Manual, 1963 Maximum allowable stresses for D+E' load combination are:
Concrete Fe= 0.85 F e 1
Reinforcing Ft= 0.90 fy C-2-14 03/28/01
USAR The loading combination which is the controlling design case for these limit stops is D+R. To provide assurance that these limit stops can withstand this loading condition, the limit stops are designed for those forces which will produce a plastic hinge in the pipe at the flued head fitting (which is in the vicinity of the limit stop). In this manner an upper limit of capability is provided for by the limit stop. The allowable reinforcing steel stress for this condition is 0.90fy. The allowable concrete stress in the Drywell Biological Shield Wall for this condition is 0.85f'c*
For the loading condition when a jet reaction does not occur, the limit stops are disengaged and transmit no piping loadings to the Drywell Biological Shield Wall.
The design of the Drywell Biological Shield Wall in the vicinity of large openings was performed in accordance with the allowable stress criteria specified in USAR Table XII-2-1. The design considered stress concentration effects assuming that the Drywell Biological Shield Wall in the vicinity of the openings was a two dimensional system contained in the plane tangent to the mid-surface of the Drywell Biological Shield Wall in the area under consideration. Out-of-plane effects were considered separately.
Stresses resulting from both effects were accounted for in the design.
2.5.6 Reactor Pedestal (See Table C-2-7)
The Reactor Pedestal is designed using the Working Stress Design Method of ACI 318-63. The thermal gradient through the shell is considered.
The forces required to restrain a ruptured Reactor Recirculation pipe are also considered. The calculated maximum stresses are close to the normal Code allowable values indicating that excessive cracking will not occur.
The controlling load combinations include design loads (D) + temperature gradient loads (R) + Operating Basis Earthquake (OBE) loads (E) evaluated against normal Code allowables, and design loads (D) + temperature gradient loads (R) + Safe Shutdown Earthquake (SSE) loads (E') + jet loads (R) evaluated against 90% of ultimate allowables.
Since the stress criteria (i.e., 90% of ultimate) is the same for the loading combination of design+ accident (jet) + OBE + temperature effects and the loading combination of design+ accident (jet) +SSE+ temperature effects, the SSE case controls. [161 Pedestal Temperature Differential [161 During the steady state, the temperature differential between the inside and outside of the pedestal is very small and its effect may be neglected. This temperature condition is maintained through the use of the Drywell fan coil units which circulate air from the outside to the inside of the Reactor Pedestal.
During the transient state, or accident state, in which temperature builds up on the outside of the pedestal but in which the heat may pass through the openings in the pedestal to the inside of the pedestal, a temperature gradient may be built up across the wall. For accident conditions which include jet loads, higher stresses are permitted over the allowable ACI stresses. Since these stresses are below the ultimate capacity of the concrete and still within its elastic range the structures will function during and after an accident condition.
C-2-15 03/28/01
USAR TABLE C-2-7 REACTOR CONCRETE PEDESTAL Max.
Method of Allowable Description/Criteria Analysis Load Combination Stress-ksi The Reactor Vessel Pedestal consists Working Stress Design D + R + E Fe 1. 35 of a 3'-3" thick x 26'-7" high Method. For seismic loads Ft 20.0 cylindrical wall rising from a response spectra are concrete base, flaring out to 5'-10 3/4" used. Circumferential D + R + E Fe 1. 35 thick at the top. The base has an stresses due to (Horiz. Hoop Ft 20.0 average thickness of 7' at wall and a temperature are in stress) spherical shaped bottom matching the accordance with ACI 307.
sphere of drywell. The shears and D + R + E' + Jet Fe 2.55 moments are transferred to the drywell at pump restraint Ft 36.0 through a welded steel shear ring and the drywell and the concrete.
Materials: Concrete f'e = 3,000 psi at 28 days max. strength per ACI 318-63 Reinforcing ASTM Designation: A615 Grade 40 per ACI 318-63 Maximum allowable stresses for D+R+E' load:
Concrete Fe= 0.85 f e 1
Reinforcing Ft= 0.90 fy Maximum allowable stresses for D+R+E load combination are not increased above code allowable values, i.e., customary 1/3 increase not used Accident load (R) includes jet forces D + R + E' + Jet Fe 2.55 (where noted) due to ruptured pipe and (at reactor Ft 36.0 temperature gradient in concrete nozzle) Fv 0.16 C-2-16 03/08/01
USAR The method used for determining the temperature stresses in the circumferential and vertical directions were those set forth in the ACI Standard Specification for the Design and Construction of Reinforced Concrete Chimneys (ACI 307-69).
Ring Girder /RPV Support Loading [l 6 l The ring girder is designed to transfer the vertical and horizontal loads of the Reactor Pressure Vessel (RPV) skirt flange to the top of the Reactor Pedestal.
The horizontal shears on the RPV skirt flange are transferred to the top flange of the ring girder by 60-A490 high strength bolts in the same friction-type connection as is described in the AISC Code.
The amount of frictional force available to resist horizontal shear is directly proportional to the normal pressure (proof load) between the RPV skirt flange and top flange of the ring girder. The total frictional force and the coefficient of sliding friction are independent of the areas in contact. The friction-type connection of the RPV skirt flange to the ring girder, in which some of the bolts lose a part of their clamping force (proof load) due to applied tension during an earthquake, suffers no overall loss of frictional shear resistance during an earthquake. The bolt tension produced by the moment is coupled with a compensating compressive force on the other side of the axis of bending. The total frictional force remains constant so long as the total pressure remains the same.
The total frictional force due to a coefficient of friction of
. 15 and a proof load of 313 kips per bolt is 2,820 kips or 6. 9 times the Operating Basis Earthquake (OBE) shear load of 403 kips or 2.1 times the Safe Shutdown Earthquake (SSE) shear plus jet load of 1,307 kips. However, if the coefficient of friction is assumed zero, the bolts acting as bearing-type connections could resist a total horizontal shear of 15.0 (at AISC Code stresses) times the OBE shear load of 403 kips or 6.85 (at 90% yield stresses) times the SSE shear load plus jet of 1, 307 kips. Therefore, the high strength bolt connections of the RPV skirt flange to the top flange of the ring girder, with or without friction, are more than adequate for the respective design loads.
The vertical loads on the RPV skirt flange are transferred to the top of the Reactor Pedestal by the ring girder acting as a bearing plate. The ring girder is designed according to AISC 6th Edition.
The load from the ring girder is transmitted to the Reactor Pedestal by direct bearing of the ring girder on the Reactor Pedestal. For the seismic condition, and the seismic plus jet condition, the overturning moment and the shear force are resisted by the anchor bolts connecting the ring girder to the Reactor Pedestal. Without considering friction the maximum tension of 14. 0 ksi and the maximum shear of 3. 6 ksi occur in the anchor bolts for the conditions of SSE plus jet force plus operating load. These stresses are below the working allowable stresses. The shear force for the condition of operating plus OBE is resisted by the anchor bolts with a concrete bearing factor of safety of 2.8. The shear force for the condition of operating plus SSE plus jet condition is resisted by the anchor bolts with a concrete bearing factor of safety 1.76.
C-2-17 03/28/01
USAR The concrete shear which resists the anchor bolt shear force without considering friction is a peripheral or punching type shear. The ACI 318-63 allowable ultimate stress for peripheral shear when shear reinforcement is provided is 6¢ ( f 'C) 0. 5
- The resulting factor of safety for OBE shear load equals 3.5 and for SSE shear plus jet load equals 1.1.
For the case of OBE, the allowable shear stress at the base of the pedestal is 1.1 (f c) 0
- 5
- The concrete takes most of this shear (60 psi) 1 and the horizontal reinforcing in the Reactor Pedestal takes the excess shear, (2.3 psi) acting as stirrups. This stress represents a very low stress for the stirrup reinforcing.
For the case of SSE+ jet+ temperature the allowable shear stress is 3. 5 ¢ ( f' cl o.s. The concrete has the capacity to take this shear load and no shear reinforcing is required since the total shear stress is 154 psi which is less than the allowable.
Drywell Shear Ring/Reactor Pedestal Loading[ 161 The shear ring connected to the Drywell has the capacity to resist the shear delivered to it from the Reactor Pedestal. For the OBE condition, the bearing pressure of the concrete against the ring is 7 50 psi and the shear across the ring its elf is 1, 930 psi. For SSE + accident + temperature, the bearing pressure of the concrete against the ring is 2,300 psi and the shear across the ring itself is 4,670 psi. The stresses are below the allowables for concrete of 750 psi for OBE and 2,700 psi for SSE; and 8,750 psi for the ring material as per the ASME Code.
There are two safety factors relating to the action between the concrete foundation within the Drywell, and the Drywell shell. The factor of safety against overturning and the factor of safety against sliding, neglecting friction.
a - The maximum overturning moment for DBE+ accident 100,000 K-ft The stabilizing moment for this condition= 185,000 K-ft Factor of safety (overturning) = 185,000 = 1.85 100,000 b - The forces resisting sliding neglecting friction are discussed in the paragraph above, where the sliding forces resulted in stresses in the concrete which were below the allowable for OBE and for SSE+ accident.
2.5.7 Primary Containment (See Tables C-2-8 through C-2-10)
The Primary Containment is a radioactive material barrier consisting of the Drywell, in which the Reactor Pressure Vessel is located, the Suppression Chamber, and process lines out to the first isolation valve outside the containment wall.
C-2-18 03/28/01
USAR TABLE C-2-8 DRYWELL MEMBRANE STRESSES Method of Analysis Load Max. Allowable Description/Criteria Combination* Stress-ksi The vessel is bulb shaped and houses the ASME Section III including Code D + R + E Primary Reactor Pressure Vessel, the coolant Cases 1330-1 and 1177-5 and General recirculation lines, pumps, etc. In case addenda as of June 9, 1967, for Membrane of an operating accident the vessel must Vessel Class "B". Sm= 17.5 contain the steam released within the @ 281 °F Drywell, and conduct this steam to the Stress intensities are defined Suppression Chamber. per Code para. N-414 and their limits are per Code Structural steel plate material is para. N-414.
ASME SA516 fabricated to ASTM Designation A300. Minimum service End conditions are found with temperature 30°F with Charpy impact methods described in the book D + R + E Primary Local requirements at maximum 0°F. Theory of Plates and Shells by Membrane Timoshenko. PL = 1. 5 Sm Seismic design load includes load due to = 26.25 vertical acceleration equal to 2/3 of the @ 281 °F horizontal ground acceleration.
After an accident the Drywell may be D + R + E Primary+
flooded up to el. 1,000', stresses shall Secondary+
be below yield point (without seismic Bending load), or may exceed yield but with a Q = 3 Sm margin against rupture if seismic is = 52.50 considered. @ 281 °F Accident load (R) includes pressure and D + E + Flood Yield 38.0 temperature in the primary containment. @ ambient Ultimate 70.0 Critical Buckling 24.21 (meridional)
- The seismic load due to the OBE (E') was found by inspection to be not governing; for the OBE the margins between the maximum allowable stresses and the calculated stresses would be greater than for the SSE (E).
C-2-19 03/08/01
USAR TABLE C-2-9 JET IMPINGEMENT FORCE STRESSES Load Max. Allowable Description/Criteria Method of Analysis Combination* Stress-ksi A jet force is assumed to Find maximum load on shell D + R 30.33 (Equipment door) occur in any direction prior to breaking within the drywell. D + R 26.84 (Jet deflector support at vent)
Compare with given jet load The force is calculated D + R 90 (90% of yield of T-1 Steel at 300°F) as 1,250 psi pressure Apply the smaller load of (Jet deflector baffle plate) acting on the area equal either of the above:
to the cross section of calculate deformation, ruptured pipe. limiting stresses to prevent D + R 30.33(Top closure head) a progressive deformation or The jet impingement force strain as follows: D + R 30.33 (Cylinder above flanges) is considered to act coincidentally with the (a) Pm + Pb ~o. 9 yield Experimental CB&I experimental investigation proved design internal pressure (b) Pm + Pb + Q <2 yield test in 1964 that 3/4" thick plate can deform 3" and 150°F shell without failure (Spherical shell) temperature.
Pm general primary Temperature of the shell membrane stress and welds are assumed to Pb primary bending stress D + R 30.33 (Cylindrical shell) be 300°F if hit directly Q secondary membrane by jet. +bending stress D + R 27.90 (Upper spray header pipe)
Yield taken from ASME Local thermal effects and Sec. III Table N-424 at dynamic jet effects are pertinent temperature D + R 21.20 (Upper spray header weld) disregarded.
Assume shear type failure of There is 2" air gap weld, and its stress equal D + R 30.33 (Upper spray header support) between drywell shell and to 7/10 of parent material.
backup concrete.
D + R 27.90 (Upper spray header inlet pipe)
Material is ASME SA516 Grade 70 fabricated to ASTM Designation: A300 D + R 27.90 (Lower spray header pipe) or Tl by USS where noted.
C-2-20 03/08/01
USAR TABLE C-2-9 (Continued)
JET IMPINGEMENT FORCE STRESSES Load Max. Allowable Description/Cr iteria Method of Analysis Combination* Stress-ksi The load combination D+R D + R 21.20 (Lower spray header weld) is a lesser case of D+R+E or D+R+E'. The effect of 30.33 (Lower spray header support)
E or E' is insignificant D + R when compared to the effect of the jet impingement force.
- The seismic load due to the SSE (E') was found by inspection to be not governing; for the SSE, the margins between the maximum allowable stresses and the calculated stresses would be greater than for the OBE (E).
References:
Formulas for Stress & Strain by Roark (4 th Edition)
"Analysis of Shells of Revolution" by A. Kalnin, Journal of Applied Mechanics, September, 1964 "Stresses from Radial Loads" by P. P. Bijlaard, Welding Journal, December, 1954 "Theory of Plates" by Timoshenko C-2-21 03/08/01
USAR TABLE C-2-10 DRYWELL STABILIZER SHEAR LUGS Load Max. Allowable Description/Criteria Method of Analysis Combination* Stresses-ksi The stabilizer mechanism ASME Code Section III including Male Lug transfers into building the addenda as of June 9, 1967, for D + E GB= (plate) 22. 8 reaction due to seismic loads vessels Class "B" Gs= (plate) 13. 8 or seismic plus jet loads Ge= (weld) 15.8 acting on the Drywell, reactor Formulas for Stress and Strain and shield, or seismic, plus by Roark, Case 22 for plate. D + R + E flooding of the Drywell. GB=30. 32 Ge combined stress Gs=l8.35 Ge=21. 01 The geometry of the stabilizer 2 D + E + Flood GB=30. 32 allows for radial and vertical .J(aB + O'T) + as Gs=l8. 35 movements due to pressure and GB bending stress Ge=21. 01 temperature. GT tensile stress Gs shear stress Female Lug Materials: Components attached D + E Ge Fb (AISC) GB=22. 8 to the drywell are ASME SA516 Grade 70 fabricated to ASTM Gs=l3. 8 Designation: A300, per ASME CTe=l5.8 Code Section III: Components D + R + E outside the drywell are ASTM CTB=30.32 Designation: A36 per AISC-1963 Gs=l8.35 Ge=21.0l D + E + Flood Stress increase by 1/3 is allowed for jet loading or GB=30. 32 flooding. Gs=l8.35 Ge=2l.01 Accident loads (R) considered include jet loads, temperature and pressure.
- The seismic load due to the SSE (E') was found by inspection to be not governing; for the SSE, the margins between the maximum allowable stresses and the calculated stresses would be greater than for the OBE (E).
C-2-22 03/08/01
USAR The original design, fabrication, erection, inspection, and testing of Primary Containment conforms to ASME Boiler & Pressure Vessel Code,Section III (for Class B Vessel) using addenda up to and including the Summer, 1967 Addenda, and Code Cases 1177-5 and 1330-1. Elements such as platforms and accessories conform to AISC 6th Edition. Piping conforms to USAS Pressure Piping Code B. 31. 1. 0, 1967. Safety and construction requirements conform to the codes and regulations of the State of Nebraska.
The Drywell is analyzed for load combinations corresponding to several different plant conditions including testing, normal operating, refueling, accident, and flooding conditions. Accident load (R) includes pressure and temperature in the primary containment. Jet loads due to a pipe break inside containment are evaluated separately as discussed in the paragraph below. Stresses are calculated for Primary General Membrane, Primary Local Membrane, and Primary + Secondary + Bending stress conditions and are all within the appropriate Code allowables.
The Drywell shell, and the miscellaneous appurtenances (e.g.,
doors, jet deflectors, spray header, etc.) are evaluated for jet impingement stresses as a result of a pipe break inside containment. These components are evaluated for the load combination of design loads ( D) + accident loads (R)
(includes pressure and temperature transients as well as jet loads). The load combination D+R is a lesser case of D+R+E or D+R+E'. The effect of E or E' is insignificant when compared to the effect of the jet impingement force, therefore was not typically evaluated. Jet force analysis is described in more detail in USAR Section C-2.5.7.2.
The Drywell stabilizer shear lugs are analyzed for load combinations corresponding to several different plant conditions. The calculated stresses are all within the appropriate Code allowables.
The elements of the pressure suppression support system designed by ASME Band PV Code,Section III, and the structural elements designed by AISC specifications are listed below along with the sections of each code that applies. [l 7J Torus Columns, Ties, Bracing Struts, and Seismic Ties Designed per AISC Specification. Materials and inspection of welds to the torus shell all in accordance with ASME Section III. Stress analyses are described in the CNS Plant Unique Analysis Report.
Drywell Stabilizer Supports Designed per ASME Section III with allowable stresses, material, and inspections in accordance with ASME Section III.
Torus Ring Designed per ASME Section III with stress analysis as described in the CNS Plant Unique Analysis Report, material, and inspections in accordance with ASME Section III.
C-2-23 03/28/01
USAR The original design analysis for Primary Containment was based on the following general references:
Formulas for Stress and Strain by Roark (4th Edition)
Theory of Plates and Shells by Timoshenko.
Woinowsky-Krieger (2nd Edition)
Beams on Elastic Foundations by Hetenyi (7th Printing)
University of Michigan Press Welding Research Council Bulletin 107 Process Equipment Design by Brownell & Young (1959)
AISC Manual of Steel Construction (6th Edition)
Some problems were solved using other references.
As part of the Mark I Containment Program, the Primary Containment System was reevaluated to include hydrodynamic loads resulting from SRV discharges and postulated LOCA events. As a result of these evaluations, the Suppression Chamber support system was reinforced to enhance the response characteristics of the Suppression Chamber (also referred to as the torus) and to increase its load carrying capacity. The Mark I containment program is described in more detail below.
2.5.7.1 Mark I Containment Program The Mark I Containment Program involved a complete reevaluation of the CNS Mark I Primary Containment to include hydrodynamic loads that had not been included in the original design. This reevaluation included extensive testing and analysis, and resulted in a series of modifications which restored the originally intended design safety margins. This section briefly describes the history of this program, these loads, and the loading combinations, allowable limits, analytical procedures used for structural evaluations, and the resulting Long Term Program modifications. Mark I piping (Torus Attached Piping) evaluations are described in USAR Section C-3.3.3.5.3.
2.5.7.1.1 Program History The Mark I containment reevaluation program began in late 1974 with the discovery of internal damage in the suppression chambers of several foreign and domestic power plants. [47 l In early 197 5, the NRC transmitted letters to NPPD[ 52 , 53 l relating to hydrodynamic loadings associated with SRV discharges and LOCA events which were not explicitly considered in the original design of the containment system. They also requested that these loads be quantified and an assessment be perf armed of the effects of these loads on the CNS containment components.
Recognizing that these evaluation efforts would be similar for all Mark I BWR plants, NPPD joined an ad hoc Mark I Owners Group with G.E. as the lead technical organization. The objectives of the Owners Group were to determine the magnitude and significance of these dynamic loads and to identify courses of action needed to resolve outstanding safety concerns. The Mark I Owners Group divided this task into two programs: a Short-Term Program (STP) for early assessment of critical components, and a Long-Term Program (LTP) for final resolution of the issues.
General studies on the new postulated loads were conducted by G.E., Bechtel Power Corporation, and others. NPPD retained Kaiser Engineers, Inc., and EDS Nuclear, Inc., to determine the impact of program conclusions C-2-24 03/28/01
USAR on CNS and perform the plant-unique load definitions and structural evaluations to restore the originally intended design safety margins at CNS.
2.5.7.1.2 Short-Term Program The objectives of the Short-Term Program ( STP) were to verify that the Primary Containment System would maintain its integrity and functional capability when subjected to the most probable loads induced by a postulated design-basis LOCA, and to verify that continued plant operation was not inimical to the heal th and safety of the public. The STP justified interim plant operation while further tests and evaluations were conducted during the comprehensive Long term Program (LTP). During the STP review, structural safety margins were increased by the implementation of procedures to maintain a differential pressure of at least one pound per square inch between the drywell and the torus during reactor operation. Upon completion of the LTP, the requirement to maintain a differential pressure was recinded. [8 ll In addition, during the course of the STP review, NPPD per£ ormed modification to the Suppression Chamber support system to provide additional design safety margins.
2.5.7.1.3 Long-Term Program The Long-Term Program (LTP) activities were initiated in June, 197 6. The objectives of the LTP were to establish design basis loads that are appropriate for the life of each Mark I BWR facility, and to restore the originally intended design safety margins for each Mark I containment system. These objectives were satisfied through extensive testing and analytical programs that led to the development of generic methods for the definition of suppression pool hydrodynamic loading events and the associated structural assessment techniques. The program also included establishment of structural acceptance criteria, and evaluations of both load mitigation devices and system modifications to improve margins of safety. The results of these generic studies are available in numerous Topical Reports.
The results of the plant-unique analyses demonstrating compliance with Mark I Containment Program requirements for the CNS Primary Containment System and associated piping are documented in the CNS Plant Unique Analysis Report.
During Mark I LOCA tests at the full-scale test facility, unanticipated cycling of the vacuum breaker valve occurred which resulted in significant damage to the valve. As a result, plant-specific vacuum breaker structural evaluations were required to ensure that valve disc closing impacts, caused by Drywell vent pressure oscillations during the chugging phase of a LOCA, would not result in damage sufficient to prevent the vacuum breakers from per£ orming their intended safety function. Maximum expected impact velocity for this event was determined by Continuum Dynamics, Inc. For the CNS-specific vacuum breaker evaluation, the maximum impact velocity is 5.821 radians/second at a 0.0 psid vacuum breaker setpoint. This is the bounding impact velocity since impact velocity decreases with increasing setpoint. The vacuum breaker structural evaluation determined that component stresses due to hydrodynamic impact loads are all within ASME Code allowable stress levels. l 58 l l 59 l C-2-25 03/28/01
USAR 2.5.7.1.4 Loads The design loads used in performing the Mark I containment reevaluation were determined using the criteria established in the NRC Safety Evaluation Report.[ 32 l The load definition procedures for hydrodynamic loads in the CNS containment reevaluations were taken from the Load Definition Report (LOR) [54 l, as modified by the NRC Acceptance Criteria. [55 l The additional loads which were not considered in the original design basis are briefly described below.
LOCA-Related Loads In the event of a postulated LOCA, reactor steam and water would expand into the Drywell atmosphere. Depending upon the size of a postulated pipe break inside the Drywell, three LOCA categories are considered. These categories are the Design Basis Accident, Intermediate Break Accident, and Small Break Accident. The discharge of an air-steam mixture into the Suppression Chamber during a LOCA results in Suppression Chamber pressurization and heat-up, and hydrodynamic loads. These hydrodynamic loads are identified as follows:
- 1. Pool Swell - results from the air in the vent system being forced into the suppression pool at a sufficiently high rate that the upper water volume of the pool is displaced upward, later falling back to its original position;
- 2. Condensation Oscillation results from steam or a steam-and-air mixture flowing through the vent system at a high rate, and forming discharge bubbles at the end of the downcomers which oscillate in size and pressure;
- 3. Chugging - is a result of intermittent flow of nearly pure steam through the downcomer exits and into the suppression pool, forming large bubbles which expand and then rapidly collapse.
SRV Discharge-Related Loads CNS is equipped with SRVs to control primary system pressure transients. For these transients, the SRVs actuate to divert part or all of the generated steam to the suppression pool. Following an SRV actuation, steam enters the SRV discharge lines, compressing the air and expelling the water slug in the submerged portions into the suppression pool. Following water clearing, the compressed air is accelerated into the suppression pool and forms high-pressure bubbles. These bubbles expand and contract a number of times before they rise to the suppression pool surface. This is followed by injection of essentially pure steam into the pool. The loads on the Suppression Chamber, Suppression Chamber internal structures, and attached piping as a result of this discharge are referred to as SRV discharge-related loads.
2.5.7.1.5 Load Combinations and Allowable Limits The structural and mechanical acceptance criteria and the general analysis techniques were obtained from the Mark I LTP Structural Acceptance Criteria Plant Unique Analysis Application Guide. [33 l This Application Guide also defines the general categories of structures, the design load C-2-26 03/28/01
USAR combinations, and the corresponding service level limits for all the structural components. Potentially bounding load combinations were identified for evaluations of the structural components.
In general, requirements of the ASME B&PV Code,Section III, S77 and Code Case N-197 were used to determine the allowable stress limits. [5 ll 2.5.7.1.6 Method of Analysis Static and dynamic analysis procedures were used to evaluate the Suppression Chamber shell and supports, vent system, and structural components internal to the Suppression Chamber. Three-dimensional finite element models of segments of the Suppression Chamber and vent system were developed, taking advantage of symmetry conditions. The enclosed fluid in the Suppression Chamber was modeled using the added-mass formulation to consider fluid-structure interaction.
An additional stress analysis of the torus was performed to establish a general corrosion allowance. The model used in this analysis explicitly modeled the enclosed fluid using acoustic fluid elements. The details of this additional analysis are documented in the CNS Plant Unique Analysis Report.
Elastic analysis procedures were used for all components except the platform system components internal to the Suppression Chamber. Since this structure is classified as non-safety-related, its design was in accordance with the Limit Analysis Design rules of the ASME Code.
The ASME B&PV Code, Section III, S77, was generally used in demonstrating the margins of safety required for steel structures and piping.
The combined state of stress for each of the structural components meets the allowable values from the ASME Code for all design load combinations. Fatigue usage was found to be within allowables at all critical locations. Stability against buckling was also verified for the Suppression Chamber shell and the vent pipes. In several cases, direct application of the LTP design requirements resulted in unusual hardship without a compensating increase in plant safety margins. Alternate analytical approaches or interpretations were used in these cases. These exceptions to design requirements are discussed in the CNS Plant Unique Analysis Report.
2.5.7.1.7 Modifications Completed The Mark I containment reevaluation program resulted in numerous modifications to the containment system components which restored the originally intended design safety margins. Table C-2-11 lists the modifications that have been completed. Modifications to existing containment components and supports were designed, fabricated, and installed to the requirements of the ASME B&PV Code,Section III, S77. Modifications involving new structural components were also designed, fabricated, and installed to the requirements of ASME B&PV Code, Section III, S77. Modifications to existing structural components were designed, fabricated and installed to the requirements of the original code of record. This code of record was typically the latest edition of the AISC Code.
G.E. performed an evaluation of the LDR SRV Load Cases for CNS.
Plant changes were implemented which mitigates SRV subsequent actuation-induced loads during postulated LOCA events.[ 55 J The evaluation performed was NEDE-22197 (December, 1982) and installed under DC 83-001, Low-Low Set. The primary concerns were the potential high thrust loads on the discharge piping, and the high frequency pressure loading on the containment.
G. E. concluded that delayed isolation achieved by means of a PCIS Group 1 (MSIV closure) low reactor water level trip set point at Level 1, combined with a 90 psi minimum low-low set relief logic, produced the maximum potential benefit. These design changes are discussed in USAR Section IV-4. 5. [57 J C-2-27 05/22/02
USAR TABLE C-2-11
SUMMARY
OF MARK I CONTAINMENT AND PIPING MODIFICATIONS COMPONENT NAME NATURE OF MODIFICATION STRUCTURAL COMPONENTS Suppression Chamber Shell and Supports Suppression Chamber Support Column Plate reinforcement to column web and flanges Column Anchorage Installed anchor bolts, brackets, and box beam assemblies Column-to-Supp. Chamber Connection Additional full penetration weldment Suppression Chamber Saddle Full saddles connecting suppression chamber support columns Ring Girder Web stiffeners; local reinforcement of weld to shell Vent System Vent Header/Downcomer Intersection Reinforced 80 penetrations with stiffener plates and pads Down comers Reduced downcomer submergence by truncation Downcomer Ties Installed tie bar and ring assembly at each downcomer pair Vent Header Deflector Installed deflector assembly in all suppression chamber bays Vent Header Supports Removed existing supports; resupported from girder above DW/WW Vacuum Breakers Reinforced 12 vacuum breaker penetrations Miscellaneous Supp. Chamber Internals Monorail Installed midbay supports in all suppression chamber bays Service Platform Replaced existing supports; added new supports, bracing and grating tie-down Drywell Steel Framing Reinforcement of beam set connections and framing members MISCELLANEOUS SYSTEM MODIFICATIONS Drywell/Wetwell Pressure Differential Installed Pump Around System Supp. Chamber Temp. Monitoring System Installed monitoring system and instrumentation PIPING SYSTEMS S/RV Discharge Piping Wetwell Piping Rerouted with stronger pipe; added 12 new supports T-Quencher Discharge Device Installed T-quencher device on each S/RV line T-Quencher Support Installed quencher support assembly in 8 bays Quencher Support Bracing Installed quencher support bracing in 8 bays Vacuum Breakers Installed two, 10-inch vacuum breakers on each line Pipe Supports and Restraints Installed 89 new or modified supports in Drywell Torus Attached Piping Large Bore Supports Installed 151 new or modified supports Small Bore Supports Installed 54 new supports Small Bore Rerouting Rerouted 5 lines Branch Line Supports Installed 25 new or modified supports Suppression Chamber Penetrations Reinforced three large bore penetrations Valve Operator Supports Reinforced 13 valve yolks Pump Anchors Modified anchorage of 4 RHR pumps Torus Internal Piping HPCI Turbine Exhaust Rerouted and resupported HPCI sparger RCIC Turbine Exhaust Rerouted and resupported RCIC sparger Core Spray Return Test Line Truncated test lines RHR Return Test Line Installed reducer, discharge elbow, and new supports Spray Header Reinforced existing supports Vent Drain Line Rerouted lines and installed supports C-2-28 03/08/01
USAR 2.5.7.2 Pipe Whip and Jet Force Analysis 2.5.7.2.1 General The measures used to assure that the Drywell shell and all essential systems and components (as defined in Sections 3. 6 .1 and 3. 6. 2 of NUREG-0800, "Standard Review Plan") within the containment (components of the primary and secondary coolant systems, engineered safety features, and equipment supports) have been adequately protected against blowdown jet forces are similar to measures used on other plants of a design similar to CNS. A brief discussion is presented here, but a more detailed description of these protection measures was provided in Quad Cities Units 1 and 2, Docket 50-254 and 50-265, FSAR Amendments 23 and 25; Vermont Yankee, Docket 50-271, FSAR Amendment 26; and Pilgrim, Docket 50-293.
As noted in USAR Section XII-2. 3. 6 and USAR Table XI I-2-6, the Primary Containment System is designed to withstand all forces associated with a postulated Loss-Of-Coolant Accident, including forces resulting from impingement of steam and/or water from a pipe which has been postulated to be ruptured. [631 2.5.7.2.2 Jet Force Analysis 1491 For the analysis of the Drywell shell in the vicinity of penetration nozzles, the loading combination of pressure plus dead and live load plus Operating Basis Earthquake (OBE) plus thermal expansion load plus jet force from a pipe break is categorized as an Emergency Condition. When Safe Shutdown Earthquake (SSE) load is substituted for OBE load, the loading combination is categorized as a Faulted Condition. The allowable membrane stresses are as specified in the ASME B&PV Code,Section III [Class BJ The classification of stresses are in accordance with Table N-413 of Section III.
The limits of stress intensities are in accordance with Figure N-414 of Section III (Summer 1970 Addenda) for the applicable condition. This includes the effect of longitudinal or circumferential type rupture of the penetrating pipe, as well as pressure impingement from an adjacent pipe. Al though the "Maximum Seismic" combination is categorized as a Faulted Condition loading, containment vessel components were analyzed to Emergency Condition stress criteria, utilizing elastic analysis. No containment vessel components were designed to the faulted condition. [21 , 22 , 49 J The Main Steam and Feedwater piping penetrations are provided with expansion bellows seals between the flued head fittings and the penetration nozzles. As shown in Figure V-2-3, these seals are protected from pressure impingement by means of coaxial guardpipes and continuous rings at the inside end of the guardpipe, which serve the function of jet deflectors.
These penetration assemblies are also furnished with axial, torsional and lateral limit stops at the flued head fittings, as well as pipe anchors on the piping outside of the outer isolation valves for protection of the penetration assemblies due to the effects of jet force loading. r49 l For protection of other penetration assemblies, when required to resist the effects of jet force loading, lateral limit stops have been furnished in the vicinity of the applicable f lued head fittings. [49 l C-2-29 03/28/01
USAR 2.5.7.2.3 Pipe Whip Analysis [49 , 5 o, 511 The Reactor Recirculation System piping within the Primary Containment has been provided with a system of pipe restraints designed to protect the Drywell shell and adjacent essential piping systems and components from the effects of pipe whip. The maximum distance between restraint brackets is no greater than that distance between any one restraint and the Drywell shell plate.
The design criteria for the pipe restraints are as follows:
- 1. The pipe restraints are arranged and provide sufficient clearance so as not to interfere with normal system operation, including Operating Basis Earthquake Loads.
- 2. The magnitude of the pipe restraint design loads is established as the product of system operating pressure and pipe flow area.
Fluid dynamics are not considered as the ruptured pipe is assumed to load the restraints in a slow even manner before fluid dynamics occur.
- 3. The allowable stresses for the restraint brackets is established as 150 percent of AISC Code allowable stresses for the materials used. The pump restraint cables are limited to 90% of the catalog advertised breaking strength.
- 4. The restraint brackets are fabricated from ASTM - A36 cold rolled steel and the cables are constructed from extra strength improved plow steel wire rope.
- 5. The physical separation provided in the arrangement of redundant engineered safety features provides spatial separation to preclude concurrent damage to more than one redundant safety feature by a single postulated pipe failure.
- 6. The pipe ruptures are postulated to occur anywhere in the Reactor Recirculation System and are assumed as either circumferential guillotine type breaks or longitudinal pipe splits. For piping systems other than the Reactor Recirculation piping system, the following criteria of postulated pipe rupture location are applied: [5 0J
- a. Stress Criteria Any intermediate points between terminal points, where the primary plus secondary stresses derived on an elastically calculated basis under loadings associated with normal and upset operating conditions (i.e., dead weight plus pressure plus Operating Basis Earthquake plus thermal loads) exceeds 2 SH, where SH is the hot allowable stress specified in USAS B31.l.0-1967.
- b. Criteria for Pipe Size and Type of Break
( 1) Longitudinal split in piping exceeding one ( 1) inch and up to four (4) inches nominal size.
(2) Circumferential breaks in piping exceeding one (1) inch nominal size.
C-2-30 03/28/01
- c. Pipe Pressure Criteria Only that piping, which is directly exposed to RPV pressure during normal reactor operation, is postulated to rupture.
- d. Terminal Point Criteria [5 lJ Piping breaks were postulated at all terminal points in each piping run or branch run in all piping within Primary Containment, except where:
( 1) both of the following pipe system conditions are met:
- a. the service temperature is less than 200°F, and
- b. the design pressure is 275 psig or less, or (2) the piping is physically separated (or isolated) from other piping or components by protective barriers, or restrained from whipping by plant design features, such as concrete encasement, or (3) following a single break, the unrestrained pipe movement of either end of the ruptured pipe in any possible direction about a plastic hinge formed at the nearest pipe whip restraint cannot impact any structure, system or component important to safety, or
( 4) the internal energy level associated with the whipping pipe can be demonstrated to be insufficient to impair the safety function of any structure, system, or component to an unacceptable level.
The critical piping systems were investigated originally for location and magnitude of maximum stress, as well as the stresses at terminal points, such as RPV nozzle attachments and branch connections to major piping runs. [5 0J The stresses are due to dead weight plus pressure plus Operating Basis Earthquake plus thermal loads.
The original summary of stresses is shown on Table C-2-12.
The original analyses as well as subsequent reanalyses show that all intermediate points between terminal points have primary plus secondary stresses less than the critical value 2 SH. Summaries of the stresses associated with the subsequent analyses are documented in the existing calculations of record.
In order to provide additional assurance of protection against violation of containment, CNS installed a redundant safety system comprised of energy absorbing tornado siding to line the reactor containment in critical areas as described in USAR Section C-2.5.7.2.4.
C-2-31 03/28/01
USAR TABLE C-2-12 ORIGINAL TABULATION OF CRITICAL PIPE STRESSES DUE TO DEAD WEIGHT PLUS PRESSURE, PLUS OPERATING BASIS EARTHQUAKE, PLUS THERMAL LOADS Maximum Terminal Point Allowable Calculated System 2 Sh Stress 1. 2 Sh Stress Description (PSI) (PSI) (PSI) (PSI) Location Main Steam 35,000 18,710 21,000 7,395 R.P.V. Nozzle Core Spray Disch. 8,276 Pump lB 27,400 11,060 16,440 R.P.V. Nozzle Core Spray Disch. 8,013 Pump lA 27,400 10,892 16,440 R.P.V. Nozzle Clean-Up Recirc. 17,217 Pump Suet. 28,900 24,360 17,350 20" RH-lD HDR.
10" MS-1 to HPCI Turb. 30,000 22,120 18,000 7,690 24" MS-1 HDR.
3" MS-1 to RCIC Turb. 30,000 10,957 18,000 6,128 24" MS-1 HDR.
Reactor Feed Water 27,400 12,300 16,440 11,340 N.E. & S.E. R.P.V. Nozzle Reactor Feed Water 16,440 8,810 N.W. & S.W. R.P.V. Nozzle RHR Pump Suet. 27,400 17,200 16,440 6,418 Conn. to Recirc. HDR RHR Pump Disch. 30,000 13,528 18,000 10,718 South Conn. to Recirc. HDR RHR Pump Disch. 30,000 14,371 18,000 11,111 North Conn. to Recirc. HDR RHR Head Spray* 27,400 9,449 16,440 7,650 R.P.V. Nozzle
NOTE: See existing calculations of record for updated stress values.
C-2-32 03/08/01
USAR It is noted that no dynamic analysis for pipe whip was performed. [5 0J Piping within the containment, such as the Core Spray and RHR systems has been designed such that the physical separation provided from one another will preclude concurrent damage and thus not affect the core cooling capacity.
A comparison of the analysis performed in Amendment 3 to the Fitzpatrick FSAR, Docket No. 50-333, with the operating characteristics of CNS has shown that the core heatup resulting from an additional single active failure, beyond the initial pipe rupture, would produce a peak clad temperature that does not exceed the 1,370°F noted in the Fitzpatrick response. This is based on a Main Steam line rupture with the failure of an adjacent Core Spray and Reactor Recirculation riser line.
The difference in design of support structures within the Drywell, between the Fitzpatrick and CNS facilities, preclude the addition of piping restraints on CNS such as those which were committed to in Amendment 3 of the Fitzpatrick FSAR. However, the special containment impact protection discussed in the next section provides the desired redundance in pipe whip protection.
2.5.7.2.4 Pipe Whip Energy Absorbing Material [5 0J An investigation was conducted to determine the feasibility of selectively installing energy absorbing material inside the Drywell to reduce the potential mechanical effects to the interior of the Primary Containment resulting from the failure of a large, unrestrained pipe in the primary pressure boundary.
The piping in question is large pipe which is normally pressurized to reactor pressure (Main Steam, HPCI Steam Supply, Feedwater, and RHR). For this study, large pipes are assumed to rupture by instantaneous and complete severance at circumferential butt welds with the jet reaction force acting normal to the rupture surf ace and resulting in pipe movement around a plastic hinge.
In developing a solution to protect the reactor containment from pipe whip resulting from such a pipe rupture, the following guidelines were utilized:
- 1. The equipment or material substantially reduces the potential for reactor containment rupture.
- 2. The equipment or material installed does not weaken or endanger either structures, piping or components within the reactor containment during normal or accident conditions.
- 3. The equipment or material is adaptable for use with existing structures.
- 4. The equipment or material does not hinder access for inservice inspection.
Based on the above investigation, CNS installed tornado siding as manufactured by H. H. Robertson Company, to line the Drywell shell in C-2-33 03/28/01
USAR critical areas, where impact from a whipping pipe in question might penetrate the containment (see Burns and Roe Drawing 4286). This tornado siding is made of corrugated high-strength steel with a steel plate backing. All material used in construction is 12 gage carbon steel with a yield point of 60,000 psi. The siding is capable of absorbing . 93 x 10 6 ft-lbs of kinetic energy per square foot. The material is compatible with the containment environment under all normal or accident conditions. The siding is typically nine inches thick and is fabricated out of rectangular sections varying in size from 2' x 3' to 2' x 9' . The sections are connected to the Drywell shell. The use of small sections of siding permits the liner to follow the contour of the Drywell shell. The liner does not restrict access to piping, welds or components for inservice inspection. Special care has been taken to assure that proper and adequate Containment Spray distribution is not affected. The installation of the liner has a negligible affect on the free Drywell volume and has no affect on the accident or post-accident environment.
The use of this liner provides a high degree of confidence that the integrity of the containment barrier will be assured, to prevent an uncontrolled release of fission products.
2.5.8 Main Steam Tunnel The Main Steam Tunnel is a biological enclosure, with 5'-0" thick walls, floors, and ceilings ( except the floor of the main section is only 2'-0" thick), for the four Main Steam and the two Feedwater lines going from the Reactor Building to the Turbine Building. [l 5 J The inside dimensions are 28'-0" wide by 48'-3" long by 25'-0" high in the main section, and 11'-0" high in the shallow section. The enclosure houses a structural steel anchor for resisting operating and jet loads from the Main Steam and Feedwater lines. This support is anchored to the sides and bottom of the tunnel. Also located within the enclosure are the second (outer) isolation valves for two systems. Blowout panels consisting primarily of light-weight cellular concrete are located in the pipe chase blockouts of the wall separating the Main Steam Tunnel and the Turbine Building. [4 ll These blowout panels are designed to rupture[ 421 in postulated events of a High Energy Line Break (HELB) to relieve the resulting pressure in the Main Steam Tunnel. These blowout panels also function as a secondary containment boundary.
The analysis of the Main Steam Tunnel considered, in addition to the loadings associated with a postulated pipe break accident (pressure buildup and impact forces due to pipe whip), temperature effects, reactions due to pipe anchors and seismic effects. The seismic effects were defined as equivalent static forces corresponding to a horizontal component of earthquake equal to 30% of gravity. This value is consistent with the seismic criteria established for the Reactor Building for the Safe Shutdown Earthquake. The allowable stresses for this combination of loads are described in USAR Section App.C-2.3. The analysis concludes that the structural integrity of the Main Steam Tunnel is maintained under the load conditions expected to occur at the time of the hypothesized pipe break accident. [251 Local penetration, due to a whipping pipe, into the 5' -0" thick reinforced concrete wall is evaluated to be a maximum of 10.3", using analytical procedures and formula contained in: "Missile Generation and Protection in Light-Water-Cooled Power Reactor Plants," by Richard C. Gwaltney, Oak Ridge National Laboratory, Report #ORNL-NSIC-22. The C-2-34 03/28/01
USAR computed penetration depth is less than 1/3 the total wall thickness. It is concluded, therefore, that the structural integrity of the wall will not be locally impaired. [25 J The time variation of steam pressure and temperature following a steam line break in the Main Steam Tunnel was initially calculated using the CONTEMPT/ESCAPE computer program, was recalculated using the EDSFLOW (a version of RELAP4/MOD5) computer program (EDS calculation 0840-002-5.0) [39 l for evaluation of environmental effects, and was again recalculated using the GOTHIC computer program (NEDC 96-00 6) [4 0l to validate the results of the EDS calculation using a newer and more accurate analysis methodology. The pressure and temperature distribution curves from the EDS calculation indicate a peak temperature of 287°F and a peak pressure of approximately 15 psig after the pipe break. An investigation was made of vapor temperatures in the Main Steam Tunnel, and in the Reactor and Turbine buildings. The temperature differences in the two atmospheres do not create a temperature gradient in the tunnel wall because their durations are not long enough to allow a gradient to penetrate the full thickness of the wall. Hence; the temperature condition in the concrete is only a "skin" effect. Even though the pressure rise is a dynamic type of loading condition, since it occurs for a very short duration, the average pressure of 15 psig was conservatively used as a static load. The combination of these loadings will result in calculated stresses that will not exceed the yield stress in the reinforcing steel and 85% of the conservative ultimate strength (3000 psi) of the concrete as indicated in Amendment 25 of the FSAR.
The Main Steam line anchor (also referred to as the pipe tunnel anchor or piping anchor) is constructed of ASTM A-36 structural steel anchored to the wall by embedded anchor bolts. The structural steel under normal operating loads, is designed using the allowable stresses permitted by the AISC 6 th Edition. In addition to the normally encountered piping loads, the anchor is designed for the jet loading resulting from the break of one Main Steam or one Feedwater line. The jet loading includes the effects of all feasible combinations of pipe rupture loads. Forces from this anchor are resisted by the 5' -0" thick Main Steam Tunnel wall. The forces encountered are transmitted from the pipe to the wall by caged trunnions which are attached to the pipe. These are in turn attached to a steel truss framework structure which transmits all postulated loadings to the tunnel walls by means of bearing plates and embedded anchor bolts. [l 4 J For the loading condition of normal operating loads plus jet forces the stresses in the structural steel were permitted to reach O. 9 fy. The allowable loads on the anchor bolts under normal operating loads were those recommended by the Uniform Building Code. Under normal operating plus jet loads the anchor bolt loads were kept considerably under the ultimate capacity in bearing. The concrete tunnel has been designed to withstand the jet force plus a postulated temperature and pressure rise within the confined area so that the allowable stresses in the reinforcing steel are not greater than 0.9 fy, and the concrete stresses are limited to 0.85 f'c*
2.5.9 Sacrificial Shield Wall [l 5 J The Sacrificial Shield Wall is a 4 7' -33/4" high cylindrical shell with an inside diameter of 21' -8". Its base sets on top of the Reactor Pedestal. The shell consists of 12 structural steel columns equally spaced around the circumference. The outside surface of the wall is formed by a 5/16 inch steel liner plate welded to the outboard flange on the columns, C-2-35 03/28/01
USAR while the inside surface is formed by a1/4 inch steel liner plate welded to the outboard face of the column inboard flange.
A concrete fill is placed between these liner plates to act only as a radiological shield. The concrete is assumed not to have any structural significance. A heavier density shielding concrete fill, having a density of 210 lbs/ft 3 is used around the active core area. There is a removable section of steel shielding at each Reactor Pressure Vessel (RPV) penetration to facilitate inspection of the penetration and two removable sections near the base of the wall for inspection access at the base of the RPV. There are steel beams framing into the column webs to support platform floor beams and pipe hangers. The Sacrificial Shield Wall is anchored and transfers loading acting on it to the Reactor Pedestal through the column base anchor bolts and shear lugs. The RPV is laterally supported by stabilizers between the RPV and the Sacrificial Shield Wall at the top. There is a pipe truss between the Sacrificial Shield Wall and the Reactor Building to laterally support the wall.
The Sacrificial Shield Wall is designed, without considering the concrete for any structural purpose, to withstand seismic forces and to support normal pipe loading and pipe restraint loads in the event of a pipe rupture in the Drywell. Also considered in the design of the shield wall is the jet load of either a circumferential break of the Main Stearn reactor vessel penetration or the Reactor Recirculation outlet penetration.
The AISC 6 th Edition was used in the design of the steel in the Sacrificial Shield Wall for all the forces on the structure including the Operating Basis Earthquake without any increase in the allowable stresses.
However, an analysis was made of the steel under all loading conditions including jet loads and the Safe Shutdown Earthquake loading using the allowable tension and bending stresses as 90% of the yield stress. This stress criterion is considered acceptable and adequate for the one time loading condition.
The stresses due to the loads described above do not exceed the applicable allowable stresses, previously outlined, for ASTM A-36 steel having a yield stress of 36,000 psi.
2.5.10 Concrete Attachments A grid system of Richmond Screw Anchor Co. cast in place concrete inserts is provided on the underside of the Reactor Building floors and in walls for pipe supports and anchors. Shear and tension allowables for these inserts are the working load values recommended by Richmond Screw Anchor Co.
The safe working load values for self-drilling anchors is based on the application of a factor of safety to the ultimate tension and shear values recommended by Phillips Drill Co., Inc. The safety factor used in tension is 6 to 1, and in shear is 5 to 1 except as utilized under the IE Bulletin 79-02 pipe support base plate design review. Where location permits, pipe supports and anchors are connected to supplementary structural steel which in turn is connected to the wall or ceiling with expansion bolts. For installations prior to approximately 1982, these expansion bolts are Phillips Red Head self-drilling anchors, snap-off type ranging from 1/4 11 diameter to 7 /8" diameter and Phillips Red Head wedge anchors ranging from 3/8 11 diameter to 1-1/4" diameter. Each self-drilling anchor was installed with epoxy under rigidly controlled procedures. [l 4 J For installations after approximately 1982, Hilti concrete anchors and Drillco Maxi-bolts have been used. r 46 l Installation of other types or brands of anchors is acceptable when approved under an Engineering Change.
C-2-36 03/19/18
USAR In response to NRC IE Bulletin No. 7 9-02, an extensive concrete expansion anchor bolt verification program was undertaken in 1979. As a result, each Seismic Class IS pipe support secured by concrete expansion anchor bolts was analyzed taking into account base plate flexibility in the calculation of the anchor loads. Flexibility was also considered in cases where expansion anchor bolts were installed directly on structural members (i.e. angles, channels, etc.) instead of using baseplates. The allowable anchor loads are based on the ultimate anchor capacity based on vendor testing, using a factor of safety equal to 5 for shell type expansion anchors, and 4 for wedge and sleeve type expansion anchors. In addition, the bolt capacity takes into account the effects of shear-tension interaction (using a 4/3 elliptical interaction equation), minimum edge distance and proper bolt spacing. It was also noted that there were no cases of expansion anchor bolts being installed in concrete block (masonry) walls. Expansion anchor bolts installed after 1979 are installed in accordance with rigorous quality control procedures to ensure proper installation. l 63 ' 651 Pipe support loads from thermal constraint of piping and from seismic excitation (including SSE), as well as dead weight loads were considered in the design of all piping systems. Anchor bolt design loads were chosen to accommodate all of these loads utilizing the appropriate factor of safety described above under static primary load conditions. All bolts covered under the IEB 7 9-02 program have been properly torqued to their required loads, as described above, and as such are designed to account for the effects of cyclic loads such that the state of stress on the bolts does not change throughout the life of the plant. Therefore, fatigue action and loosening of the expansion anchor bolt assembly are prevented, and the stress range above the torque values are kept to a minimum. The prescribed torque loadings have accounted for the design requirement to permit the expansion anchor assemblies to successfully support loads which are repetitive. l 64 l The original construction test data indicated a maximum concrete design strength of 3,500 psi. Burns and Roe Report on Base Plates Design, dated October 7, 1980, drafted in response to NRC Bulletin 79-02, states that tests carried out on concrete samples taken from Class 1 areas shows the ultimate concrete strength over 5,000 psi. Statements from ITT Philips Drill Division Red Head Concrete Anchoring Handbook and Specifiers Guide, 1973, together with statements in the PCI Design, Precast and Prestressed Concrete, Prestressed Concrete Institute, 1971, indicate that anchor capacity increases as the concrete strength increases.
Concrete expansion anchors use the following ultimate concrete strengths based on installed location in Class 1 structures. These values are based on ACI 318-63 28 day concrete strength.
Building Compressive Concrete Strength (fc')
Intake Structure 5,000 psi Control Corridor 5,000 psi Diesel Generator Building 5,000 psi Control Building Substructure 5,000 psi Control Building 903'-6" 4,000 psi Control Building 918'-0" 5,000 psi Control Building 932'-6" 5,000 psi Control Building Roof 4,750 psi Reactor Building SW Quad Slab 4,200 psi Reactor Building SE Quad Slab 5,000 psi Reactor Building NW Quad Slab 5,000 psi C-2-37 01/02/15
USAR Reactor Building NE Quad Slab 5,000 psi Reactor Building SW Quad Wall 5,000 psi Reactor Building SE Quad Wall 5,000 psi Reactor Building NW Quad Wall 4,690 psi Reactor Building NE Quad Wall 5,000 psi Reactor Building Closure Walls 5,000 psi HPCI Room and Reactor Bents 5,000 psi Containment Walls to 903'-6" 5,000 psi Drywell, Quad Equipment Pads,
& Stairwells to 903'-6" 3,890 psi Reactor Building 903'-6" 4,420 psi Reactor Building 931'-6" 4,900 psi Reactor Building 958'-3" 5,000 psi Reactor Building 976'-0" 4,780 psi Reactor Building 1001'-0" 5,000 psi Reactor Sacrificial Shield Wall 5,000 psi The original design criteria for Philips Redhead-S Type anchor bolts associated with conduit/cable tray hangers was based upon the ICBO Codes. Using ultimate values from the manufacturer's test data the maximum allowable loading values were determined.
Event Factor of Safety OBE 4 SSE 3 NOTE: The factors of safety given in this paragraph only apply to original conduit/cable tray hangers. New hangers installed after January 1, 198 6, are to use factors of safety as laid down in IE Bulletin 79-02 (see above paragraph). The reason for upgrading the factors of safety with regard to conduit/ cable tray hanger anchor bolts is to ensure uniformity of anchor bolt design.
C-2-38 01/02/15
USAR TABLE C-2-13 CONTROL BUILDING FOUNDATION Maximum Load Allowable Description/Criteria Method of Analysis Combination Stress Design load (D) includes all Maximum soil stress= D + E 1. 5 qa = 18 dead and equipment loads plus D + ~p = ~ - (YDa + YsatDb) 25% of the full live loads A S Vertical seismic load (~P) is where:
assumed to be produced by a D Design load, kips D + E' 1. 5 qa 18 vertical acceleration equal to ~p Vert. seismic load, 1/2 of the ground horizontal 1/2 kips seismic acceleration, both M Horiz. seismic acting simultaneously overturning moment, kip-ft Overturning moment (M) is A Foundation mat area, produced by the horizontal sq-ft seismic acceleration and all s Section modulus of design load eccentricities foundation mat, ft 3 y Unit wt. of soil above water table Ysat = Saturated unit wt. of soil Da = Depth of soil above water table Db = Depth of saturated soil above bottom of foundation C-2-39 0110211s I
USAR TABLE C-2-14 CONTROL BUILDING FLOOR SYSTEM Max.
Load Allowable Description/Criteria Method of Analysis Combination Stress-ksi Floor system is designed for Working Stress Design Method - D + E Fe 1. 35 the loadings tabulated in Slabs, beams, and girders Fv 0.060 Subsection XII-2.3.2 Ft 20.0 Materials conform as follows:
Concrete f'c = 3,000 psi at 28 days max. strength per ACI 318-63. Reinforcing ASTM Designation: A615 Grade 40 per ACI 318-63 Vertical seismic load is D + E' Fe = 2.55 assumed to be produced by a Fv = 0.093 vertical acceleration equal to Ft= 36.0 1/2 of horizontal ground acceleration D + T Fe= 2.55 Maximum allowable stress for Fv = 0.093 D+E' and D+T load combinations Ft = 36.0 are: Concrete Fe= 0.85 f c 1
Reinforcing Ft= 0.90 fy Maximum allowable stresses for D+E combination are not increased above code allowable values, i.e., customary 1/3 increase not used Tornado wind load (T) on roof averages a pressure of 270 PSF or a 3 psi suction press.
C-2-40 0110211s I
USAR TABLE C-2-15 CONTROL BUILDING CONCRETE WALLS Max.
Load Allowable Description/Criteria Method of Analysis Combination Stress-ksi Design load (D) includes all Working Stress Design D + E Ft = 20.0 dead and equipment loads plus Bearing Walls Fe= 1.35 25% of the full live loads Fv=0.060 Materials: D + E' Ft= 36.0 Concrete f'e = 3,000 psi at Fe= 2.55 28 days max. strength per Fv=0.093 ACI 318. Reinforcing ASTM A615 Grade 40 per ACI 318 D + T Ft= 36.0 Fe = 2.55 Maximum allowable stresses for Fv = 0.093 D+E' and D+T loads:
Concrete Fe = 0. 85 f' e Reinforcing Ft - 0. 90 ft Maximum allowable stresses for D+E load combination are not increased above code allowable values, i.e., customary 1/3 increase not used Tornado wind load (T) on walls averages a pressure of 270 psf or a 3 psi suction pressure C-2-41 01102;15 I
- 2. 5 .11 Control Building (See Tables C-2-13 through C-2-15)
The Control Building has been reviewed in a manner similar to that described above for the Reactor Building for applicable combinations of design loads. The stresses in this structure have been found to be within the appropriate allowable limits.
2.5.12 Elevated Release Point (See Table C-2-16)
The Elevated Release Point (ERP) is described in USAR Section XII-2.2.6. The ERP is designed for design loads (D), wind loads (W),
and seismic loads (E, E') . The ERP is not designed for tornado loads as explained in USAR Section XII-2.2.6. All the resulting stresses are within the Code allowable values.
C-2-42 01102115 I
USAR TABLE C-2-16 ELEVATED RELEASE POINT FOUNDATIONS AND TOWER Load Max. Allowable Description/Cr iteria Method of Analysis Combination Stress-ksi Design load (D) including all Free standing or rigid towers D + W Ft 20xl.33 dead and equipment loads plus foundation Working Stress Method 26.7 1/2" radial ice on all members Design wind loading (W) is Fe 1. 35xl. 33 100 mph with wind pressures varying from 44 PSF at ground 1. 80 level to 73 PSF at top blowing against the projected area of members covered by 1/2" radial Fv 0.ll0xl.33 ice 0.146 Live loads on platforms:
Platforms at top and 30' levels= 100 PSF Rest Platforms= 50 PSF Loading conditions D+E and D+E' were investigated and found not to be governing when compared with D+W Materials:
Concrete f'c = 3,000 psi at 28 days maximum strength per ACI 318-63. Reinforcing ASTM A615 Grade 40 per ACI 318-63 C-2-43 01102115 I
USAR TABLE C-2-16 (Continued)
ELEVATED RELEASE POINT FOUNDATIONS AND TOWER Max.
Load Allowable Description/Cr iteria Method of Analysis Combination Stress-ksi Tower Steel Tower Steel D + W Fa (leg)
Leg members A440 (50,000 psi Ultimate design method based on yield) Other members A36 formulae and provisions 35.77 (36,000 psi yield) conforming to U.S. Bureau of for l = 57 Reclamation "Transmission r Maximum allowable stresses for Structures Design Standard D+W load combinations are No. 10" Fa (brace) increased by 1/3 for 10.60 foundation design. (Working for l = 174 stress method) r Concrete Ft = 45.00 Ft = 1.33 fs for A-440 Fe= 1.33 f e 1
Ft= 32.40 Fv 1. 33 Ve for A-36 Fb = 45.00 for A-440 (not used in bending)
Fb = 32.40 for A-36 Soil pressure
= 12.0 ksf maximum
- Max. actual stresses for load combination W+D or W-D (max uplift)
C-2-44 01102115 I
USAR 3.0 COMPONENTS 3.1 Intent and Scope 3 .1.1 Components Designed By Rational Stress Analysis These general design criteria are intended to apply to those ductile metallic structures or components which are normally designed using rational stress analysis techniques such as pressure vessels, reactor internal components, etc. The criteria may also be applied to those components or structures whose ultimate loading capability is determined by tests. These criteria are intended to supplement applicable industry design codes where necessary. Compliance with these criteria is intended to provide design safety margins which are appropriate to extremely reliable structural components when account is taken of rare event potentialities such as might be associated with a Safe Shutdown Earthquake or primary pressure boundary coolant pipe rupture, or a combination of events.
3 .1. 2 Components Designed Primarily By Empirical Methods There are many important Class I components or equipment which are not normally designed or sized directly by stress analysis techniques. Simple stress analyses are sometimes used to augment the design of these components, but the primary design work does not depend upon detailed stress analysis.
These components are usually designed by tests and empirical experience, which may include earthquake experience for the seismic qualification of Class I components. Complete detailed stress analysis is currently not meaningful nor practical for these components. Examples of such components are valves, pumps, electrical equipment, and mechanisms. Field experience and testing are used to support the design. Alternatively, use of the SQUG Generic Implementation Procedure (GIP-3) may be used for seismic qualification of electrical and mechanical equipment within the scope of the GIP. Where the structural or mechanical integrity of components is essential to safety, the components referred to in these criteria must be designed to accommodate the events of the Safe Shutdown Earthquake (SSE) or Operating Basis Earthquake (OBE), or a design basis pipe rupture, or a combination where appropriate. The reliability requirements of such components cannot be quantitatively described in a general criterion because of the varied nature of each component and its specific function in the system.
3.2 Loading Conditions and Allowable Limits The loading conditions established herein are expressed in generic terms and are related in a probabilistic manner to the loads which are to be investigated for safety considerations. Related probabilistic definitions are used to determine an appropriate minimum safety factor which is used to establish structural design allowable limits and functional design allowable limits. Certain of the limits described in these criteria, i.e., deformation limit, and fatigue limit, are included for completeness, but do not necessarily require application to all components. Where it is clear to the designer that, based upon experience, fatigue or excess deformation are not of concern for a particular structure or component, a formal analysis with respect to that limit is not required.
3.2.1 Loading Conditions The loading conditions may be divided into four categories; Normal, Upset, Emergency, and Faulted conditions. These categories are generically described as follows:
C-3-1 08/08/01
USAR 3.2.1.1 Normal Conditions Any condition in the course of operation of the station under planned and anticipated conditions, in the absence of Upset, Emergency or Faulted Conditions.
3.2.1.2 Upset Conditions Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand those conditions.
The Upset Conditions include abnormal operational transients caused by a fault in a system component requiring its isolation from the system, transients due to loss of load or power, and any system upset not resulting in a forced outage. The Upset Conditions may include the effect of the Operating Basis Earthquake.
3.2.1.3 Emergency Conditions Any deviations from Normal Conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of specific damage developed in the system.
3.2.1.4 Faulted Conditions Those combinations of conditions associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the nuclear system may be impaired to the extent where considerations of public health and safety are involved. Such considerations require compliance with safety criteria.
Beginning with the 1977 Edition of the ASME Boiler and Pressure Vessel Code, these conditions are designated as Service Level A (Normal),
Level B (Upset), Level C (Emergency), Level D (Faulted).
3.2.2 Allowable Limits In addition to the generic definition of loading conditions in the preceding paragraphs, the meaning of these terms is expanded in quantitative probabilistic language. The purpose of this expansion is to clarify the classification of any hypothesized accident or sequence of loading events so that the appropriate limits or safety margins are applied. Knowledge of the event probability is necessary to establish meaningful and adequate safety factors for design. Table C-3-1 illustrates the quantitative event classifications. The probabilities of Table C-3-1 have been assigned to establish the appropriate structural design limits for the loading conditions in Subsection C-3.2.1. A summary of these limits is shown in the tables listed below.
DEFORMATION LIMIT Table C-3-2 PRIMARY STRESS LIMIT Table C-3-3 BUCKLING STABILITY LIMIT Table C-3-4 FATIGUE LIMIT Table C-3-5 There are many places where, through the exercise of designer judgment, it is unnecessary to actually carry out a formal analysis for each of these limits. A simple example consists of the case where two pieces of pipe of different wall thicknesses are joined at a butt weld. If they are both subjected to the same loading, only the thinner piece would require a formal analysis to demonstrate that the primary stress limit has been satisfied.
C-3-2 08/08/01
USAR The term SFmin is defined as the minimum safety factor on load or deflection and is related to the event probability by the following equation:
9 SFmin 3 - log10 P40 For event probabilities smaller than 10- 5 or greater than 10- 1, the following apply:
Event Probability Min. Safety Factor 1.125
- 1. o > P 40 ::::: 10- 1 2.25 These expressions show the probabilistic significance of the classical safety factor concept as applied to reactor safety. The SFmin values corresponding to the event probabilities are summarized in Table C-3-6.
The loadings which occur as a result of the conditions listed are factored into the design of the components in accordance with the requirements of the applicable design code, or to the requirements of these criteria. Where applicable design codes or acceptance criteria provide specific allowable limits for a given loading condition, these limits may be substituted for the criteria above on the basis that appropriate safety margins are provided in the limits. In particular, editions of the ASME Boiler and Pressure Vessel Code subsequent to 1977 specify allowable limits on primary stress for piping components under Emergency and Faulted Conditions. These limits are used whenever these ASME Code editions apply for design.
C-3-3 08/08/01
USAR TABLE C-3-1 LOADING CONDITION PROBABILITIES Upset (likely) 1.0 > p 40 ::.:: 10- 1 Emergency (low probability) 10- 1 > p 40 ::.:: 10- 3 Faulted (extremely low probability) 10- 3 > p 40 ::.:: 1 o- 6 Where P40 40 year event encounter probability C-3-4 03/08/01
USAR TABLE C-3-2 DEFORMATION LIMIT Either One of (Not Both) General Limit
- a. Permissible Deformation, DP 0.9 Analyzed deformation SFmin causing loss of function, DL
- b. Permissible Deformation, DP ~ 1. 0
- Experimental deformation SFmin causing loss of function, DE Where:
DP permissible deformation under stated conditions of normal, upset, emergency or fault DL analyzed deformation which would cause a system "loss of function"**
DE experimentally determined deformation which would cause a system "loss of function"**
- Equation b was not applied unless supporting data had been submitted for AEC staff evaluation. llBJ
- Note that "loss of function" can only be defined quite generally until attention is focused on the component of interest. In cases of interest, where deformation limits can affect the function of equipment and components, they will be specifically delineated. From a practical viewpoint, it is convenient to interchange, with the loss of function condition, some deformation condition at which function is assured if the required safety margins from the functioning condition can be achieved.
Therefore, it is often unnecessary to determine the actual loss of function condition because this interchange produces conservative and safe designs.
Examples where deformation limits apply are: control rod drive alignment and clearances for proper insertion, core support deformation causing fuel disarrangement or excess leakage of any component.
C-3-5 03/08/01
USAR TABLE C-3-3 PRIMARY STRESS LIMIT Any One of (No More than One Required) General Limit
- a. Elastic Evaluated Primary Stresses, PE s; 2.25 Permissible Primary Stresses, PN SFmin
- b. Permissible Load, IP s; 1. 5 Largest Lower Bound Limit Load, CL SFmin
- c. Elastic Evaluated Primary Stress, PE s; 0.75 Conventional Ultimate Strength at SFmin Temperature, US
- d. Elastic - Plastic Evaluated s; 0. 9 Nominal Primary Stress, EP SFmin Conventional Ultimate Strength at Temperature, US
- e. Permissible Load, LP s; 0.9
- Plastic Instability Load, PL SFmin
- f. Permissible Load, LP s; 0.9
- Ultimate Load from Fracture SFmin Analysis, UF
- g. Permissible Load, LP s; 1.0
- Ultimate Load or Loss of Function SFmin Load from Test, LE
- Equations e, f, g were not used unless supporting date had been submitted for NRC Staff evaluation. [lSJ Where:
PE Primary stresses evaluated on an elastic basis. The effective membrane stresses are to be averaged through the load carrying section of interest. The simplest average bending, shear or torsion stress distribution which will support the external loading will be added to the membrane stresses at the section of interest.
PM Permissible primary stress levels under normal or upset conditions under applicable industry code.
LP Permissible load under stated conditions of emergency or fault.
C-3-6 03/08/01
USAR TABLE C-3-3 (Continued)
PRIMARY STRESS LIMIT CL Lower bound limit load with yield point equal to 1. 5 Sm where Sm is the tabulated value of allowable stress at temperature as contained in the ASME III Code or its equivalent. The "lower bound limit load" is here defined as that produced from the analysis of an ideally plastic (nonstrain hardening) material where deformations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined material yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the unaxial case.
us Conventional ultimate strength at temperature or loading which would cause a system malfunction, whichever is more limiting.
EP Elastic-plastic evaluated nominal primary stress. Strain hardening of the material may be used for the actual monotonic stress strain curve at the temperature of loading or any approximation to the actual stress strain curve which everywhere has a lower stress for the same strain as the actual monotonic curve may be used. Either the shear or strain energy of distortion flow rule may be used.
PL Plastic instability load. The "plastic instability load" is defined here as the load at which any load bearing section begins to diminish its cross-sectional area at a faster rate than the strain hardening can accommodate the loss in area.
This type analysis requires a true stress-true strain curve or a close approximation based on monotonic loading at the temperature of loading.
UF Ultimate load from fracture analyses. For components which involve sharp discontinuities (local theoretical stress concentration > 3) the use of a "fracture mechanics" analysis where applicable, utilizing measurements of plane strain fracture toughness may be applied to compute fracture loads. Correction for finite plastic zones and thickness effects as well as gross yielding may be necessary. The methods of linear elastic stress analysis may be used in the fracture analysis where its use is clearly conservative or supported by experimental evidence. Examples where "fracture mechanics" may be applied are for fillet welds or end of fatigue life crack propagation.
LE Ultimate load or loss of function load as determined from experiment. In using this method, account shall be taken of the dimensional tolerances which may exist between the actual part and the tested part or parts as well as differences which may exist in the ultimate tensile strength of the actual part and the tested parts. The guide to be used in each of these areas is that the experimentally determined load shall use adjusted values to account for material properties and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part.
C-3-7 03/08/01
USAR TABLE C-3-4 BUCKLING STABILITY LIMIT Any One of (No More than One Required) General Limit
- a. Permissible Load, LP ::s; 2.25 Code normal event permissible load, PN SFmin
- c. Permissible Load, LP ::s; 1.0
- Ultimate Buckling Collapse Load from SFmin Test, SE
- Equation c was not used unless supporting data had been submitted for AEC Staff review. llBJ Where:
LP Permissible load under stated conditions of Normal, Upset, Emergency, or Faulted.
PN Applicable code normal event permissible load.
SL Stability analysis load. The ideal buckling analysis is often sensitive to otherwise minor deviations from ideal geometry and boundary conditions. These effects shall be accounted for in the analysis of the buckling stability loads. Examples of this are ovality in externally pressurized shells or eccentricity of column members.
SE Ultimate buckling collapse load as determined from experiment. In using this method, account shall be taken of the dimensional tolerances which may exist between the actual part and the tested part. The guide to be used in each of these areas is that the experimentally determined load shall be adjusted to account for material property and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part.
C-3-8 03/08/01
USAR TABLE C-3-5 FATIGUE LIMIT Summation of mean fatigue (lJ damage a. Fatigue cycle usage ::;; . 05 (2 l
- usage including Emergency or Faulted from analysis events with design and operation loads following Miner's b. Fatigue cycle ::;; 0.33
- Hypotheses .... either one (not both) usage from test
- Equations a and b were not used unless supporting data had been submitted for AEC Staff review. Fatigue analysis was performed per footnote 2. [lBJ (1) Fatigue failure is defined here as a 25% area reduction for a load carrying member which is required to function or excess leakage causing loss of function, whichever is more limiting. In the fatigue evaluation, the methods of linear elastic stress analysis may be used when the 3Sm range limit of ASME III has been met. If 3Sm is not met, account will be taken of (a) increases in local strain concentration, (b) strain ratcheting, (c) re-distribution of strain due to elastic-plastic effects. The January, 1969, draft of the USAS B31.7 Piping Code may be used where applicable or detailed elastic-plastic methods may be used. With elastic-plastic methods, strain hardening may be used not to exceed in stress for the same strain, the steady state cyclic strain hardening measured in a smooth low cycle fatigue specimen at the average temperature of interest.
(2) It is acceptable to use the ASME Section III Design Fatigue curves in conj unction with a cumulative usage factor of 1. 0 (using Miner's Hypothesis) in lieu of using the mean fatigue data curves with a limit on fatigue usage of 0.05, since the two methods are approximately equivalent.
C-3-9 03/08/01
USAR TABLE C-3-6 MINIMUM SAFETY FACTORS Loading Conditions Loads p 40 SFmin Upset N and E 10- 1 2.25 or N and U 10- 1 2.25 Emergency N and R 10-J 1. 5 N and E' 10-J 1. 5 Other combinations <10- 1 to 10-J <2.25 to 1. 5 in this probability range Fault N and E' and R 1. 5 X 10- 6 1.125 Other combinations <1.5 to 1.125 in this probability range Where: N Normal loads U Upset loads (result in maximum system pressure)
E Maximum probable earthquake E' Maximum possible earthquake R Loads resulting from jet forces and pressure and temperature transients associated with rupture of a single pipe within the primary containment. This load is considered as indicated in the tables.
The minimum safety factor decreases as the event probability diminishes and if the event is too improbable (incredible: P 40 < 10- 6 ) , then no safety factor is appropriate or required.
C-3-10 03/08/01
USAR TABLE C-3-7 LOADING CRITERIA Reactor Vessel Internals and Associated Equipment CRITERIA LOADING PRIMARY STRESS TYPE ALLOWABLE STRESS STABILIZER BRACKET AND ADJACENT SHELL Primary Stress Limit - ASME Normal and upset condition Membrane plus bending 40,000 psi Boiler and Pressure Vessel Code, load Sect. III defines primary 1. OBE (E) membrane plus primary bending 2. Design pressure Membrane plus bending 60,000 psi stress intensity limit for SA302 - Gr. B Emergency condition load
- 1. SSE (E')
For normal and upset condition 2. Design pressure Stress limit= l.5x26,700 Membrane 53,400 psi 40,000 psi Faulted condition loads For emergency condition 1. SSE (E')
Stress limit= 1.5x40,000 2. Jet reaction forces 60,000 psi 3. Design pressure For faulted condition Stress limit= 2.0x26,700 53,400 psi VESSEL SUPPORT SKIRT Primary Stress Limit - ASME Normal and upset condition Compressive membrane 12,000 psi Boiler and Pressure Vessel Code, loads Sect. III, defines stress limit 1. Dead weight for SA516 GR70 2. OBE (E)
Compressive membrane 18,000 psi Emergency condition loads For normal and upset condition 1. Dead weight B = 12,000 psi 2. SSE (E')
For emergency condition Faulted condition loads Compressive membrane 24,000 psi S1imit = 1.5 B = 18,000 psi 1. Dead weight
- 2. SSE (E')
For faulted condition 3. Jet reaction forces S limit = 2 . 0 B = 2 4 , 0 0 0 p S i C-3-11 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING PRIMARY STRESS TYPE ALLOWABLE STRESS SHROUD SUPPORT GUSSETS Primary Stress Limit - ASME Normal and upset condition Membrane plus bending 34,950 psi Boiler and Pressure Vessel Code, loads Sect. III defines allowable 1. OBE (E) primary membrane stress plus 2. Pressure drop across shroud bending stress for SB168 (normal) material. 3. Subtract dead weight For normal and upset condition Emergency condition loads Membrane plus bending 52,430 psi SA = 1. 5 SM = 1. SSE (E')
1.5 X 23.30 = 34.95 KSI 2. Pressure drop across shroud (normal)
- 3. Subtract dead weight For emergency condition Faulted condition loads Membrane plus bending S1imit = 1 *5 SA= 1. SSE (EI) 69,900 psi 1.5 X 34.95 = 52.43 KSI 2. Pressure drop across shroud during faulted condition For faulted condition 3. Subtract dead weight A limit = 2 *0 SA=
2.0 X 34.95 = 69.90 KSI C-3-12 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING PRIMARY STRESS TYPE ALLOWABLE STRESS TOP GUIDE-LONGEST BEAM Primary Stress Limit - The Normal and upset condition General membrane plus 25,388 psi allowable primary membrane stress loads bending plus bending stress is based on 1. OBE (E)
ASME Boiler and Pressure Vessel 2. Weight of structure Code, Sect. III for type 304 3. Weight of temporary control stainless steel plate. curtains.
For normal and upset condition Emergency condition loads General membrane plus 38,081 psi Stress Intensity SA= 1.5 SM= 1. SSE (E') bending l.5x16,925 psi= 25,388 psi. 2. Weight of structure
- 3. Weight of temporary control curtains.
For emergency condition S 1imi t = 1 . 5 SA= 1 . 5 X 2 5 , 3 8 8 38,081 psi For faulted condition Faulted condition loads General membrane plus 50,775 psi S1imit = 2 SA= 2x25, 388 50,775 psi (same as emergency condition) bending TOP GUIDE BEAM END CONNECTIONS Primary Stress Limit - ASME Boiler Normal and upset condition Pure shear 10,155 psi and Pressure Vessel Code, loads Sect. III, defines material stress 1. OBE (E) limit for type 304 stain-less 2. Weight of structure steel. 3. Weight of temporary control curtains.
For normal and upset condition Emergency condition loads Pure shear 15,232 psi Stress Intensity SA= 0.6 SM 1. SSE (E')
0.6x16,925 psi= 10,155 psi 2. Weight of structure
- 3. Weight of temporary control For emergency condition curtains.
S 1imit = 1 . 5 SA= 1 . 5 X 1 0 , 15 5 p Si 15,232 psi.
For faulted condition Faulted condition loads Pure shear 20,310 psi S 1imi t = 2 SA = 2 X 1 0 , 15 5 p Si (same as emergency condition) 20,310 psi C-3-13 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING PRIMARY STRESS TYPE ALLOWABLE STRESS TOP GUIDE ALIGNERS Primary Stress Limit - The Normal and upset condition General membrane plus 25,388 psi allowable primary membrane loads bending stress plus bending stress is 1. OBE (E) based on ASME Boiler and 2. Weight of structure Pressure Vessel Code, Sect. 3. Weight of temporary control III for type 304 stainless curtains.
steel plate.
Emergency condition loads General membrane plus 38,081 psi For normal and upset condition 1. SSE (E') bending Stress Intensity SA = 1. 5 SM = 2. Weight of structure l.5x16,925 psi 25,388 psi 3. Weight of temporary control curtains.
For emergency condition S limit = 1 . 5 SA = 1 . 5 X 2 5 , 3 8 8 38,081 psi Faulted condition loads General membrane plus 50,775 psi (same as emergency condition) bending For faulted condition S limit = 2 SA = 2 X 3 8 , 0 8 1 50,775 psi C-3-14 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING LOCATION ALLOWABLE STRESS RPV STABILIZER Primary Stress Limit Upset condition Rod 127,000 psi AISC 6th Edition specification 1. Spring preload for the construction, 2. OBE (E) Bracket 22,000 psi fabrication, 14,000 psi and erection of structural steel for buildings For normal & upset conditions Emergency condition Bracket 33,000 psi AISC allowable stresses, but 1. Spring preload without the usual increase for 2. SSE (E') 21,000 psi earthquake loads For emergency conditions Faulted condition Bracket 36,000 psi 1.5 x AISC allowable stresses 1. Spring preload
- 2. SSE (E') 21,500 psi For faulted conditions 3. Jet reaction load Material yield strength RPV SUPPORT (RING GIRDER)
Primary Stress Limit Normal and upset condition Top Flange 27,000 psi AISC 6 th Edition specification 1. Dead loads Bottom Flange 27,000 psi for the design, fabrication and 2. OBE (E) Vessel to girder bolts 60,000 psi erection of structural steel 3. Loads due to scram 18,000 psi for buildings For normal and upset conditions Emergency condition Top Flange 40,500 psi AISC allowable stresses, but 1. Dead loads Bottom Flange 40,500 psi without the usual increase for 2. SSE (E') Vessel to girder bolts 90,000 psi earthquake loads. 3. Loads due to scram 20,800 psi For faulted conditions 1.67 X AISC allowable stresses Faulted condition Top Flange 45,000 psi for structural steel members 1. Dead loads Bottom Flange 45,000 psi Yield strength for high 2. SSE (E') Vessel to girder bolts 125,000 psi strength bolts (vessel to ring 3. Jet reaction load 72,000 psi girder)
C-3-15 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING PRIMARY STRESS TYPE ALLOWABLE STRESS CORE SUPPORT Primary Stress Limit - The Normal and upset condition General membrane plus 25,388 psi allowable primary membrane loads bending stress plus bending stress 1. Normal operation pressure drop is based on ASME Boiler and 2. OBE (E)
Pressure Vessel Code, Sect.
III for type 304 stainless steel plate.
For allowable stresses see Emergency condition loads General membrane plus 38,081 psi top guide, longest beam, 1. Normal operation pressure drop bending above 2. SSE (E')
Faulted condition loads General membrane plus 50,775 psi
- 1. Pressure drop after bending recirculation line rupture
- 2. SSE (E')
CORE SUPPORT ALIGNERS Primary Stress Limit - ASME Normal and upset condition Pure shear 10,155 psi Boiler and Pressure Vessel Code, load Sect. III, defines material 1. OBE (E) stress limit for type 304 stainless steel.
For allowable shear stresses, Emergency condition load Pure shear 15,232 psi see top guide beam end connec- 1. SSE (EI) tions, above.
Faulted condition load Pure shear 20,310 psi
- 1. SSE (EI)
C-3-16 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
MOMENT LIMIT accounting for CRITERIA LOADING PRIMARY STRESS TYPE pressure loads FUEL CHANNELS Primary Stress Limit - Allowable Normal and upset condition Membrane and bending 28,230* in. lbs.
stress SM for Zircaloy or NSF load determined according to methods 1. OBE (E) recommended by ASME Boiler and 2. Normal pressure load Pressure Vessel Code, Sect. III.
Allowable moment determined by Emergency condition load Membrane and bending 42,350* in. lbs.
calculating limit moment using 1. SSE (E')
Table C-3-3, equation (b), then 2. Normal pressure load applying SF~n for applicable loading conditions.
Faulted condition load Membrane and bending 56,500* in. lbs.
( Sm = 9,270 psi; 1. 5 Sm 1. SSE (E')
13,900 psi) 2. Loss of cooling accident pressure Emergency limit load= 1.5 x Normal limit load calculated using 1.5 Sm= yield.
- The moments shown above are applicable to the previous analysis (e.g., prior to MELLL and ICF). The current fuel design is consistent with the MELLL and ICF evaluation and has been qualified based on the pressure drops associated with MELLL and ICF.
C-3-17 11/07/16
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING LOCATION ALLOWABLE STRESS CRD HOUSING SUPPORT Primary Stress Limit Faulted condition loads Beams (top cord) 33,000 psi AISC specification for the 1. Dead weight 33,000 psi design, fabrication and erection 2. Impact force from Beams (bottom cord) 33,000 psi of structural steel for buildings failure of a CRD housing 33,000 psi (Dead weights and earthquake Grid structure 41,500 psi For normal and upset condition loads are very small as 27,500 psi Fa= 0.60 Fy (tension) compared to jet force.)
Fb = 0.60 Fy (bending)
Fv = 0.40 Fy (shear)
For faulted conditions Fa limit= 1.5 Fa (tension)
Fb limit= 1.5 Fb (bending)
Fv limit= 1.5 Fv (shear)
Fy = Material yield strength RECIRCULATING PIPE AND PUMP WHIP RESTRAINTS Primary Stress Limit Faulted condition loads Brackets on 28" pipe 33,000 psi Structural Steel: AISC 1. Jet force from a specification for the design, complete circumferential Cable on pump restraints 90% of listed fabrication and erection of failure (break) of breaking strength structural steel for buildings. recirculation line.
For normal or upset conditions Fa= 0.60 Fy (tension)
For faulted conditions Fa limit= 1.5 Fa (tension)
Fy = yield strength Cable (wire rope):
For faulted conditions 90% of listed breaking strength C-3-18 03/29/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING LOCATION ALLOWABLE STRESS CONTROL ROD DRIVE HOUSING Primary Stress Limit - The Normal and upset condition Maximum membrane 15,800 psi allowable primary membrane loads stress intensity stress is based on the ASME 1. Design pressure occurs at the tube Boiler and Pressure Vessel Code, 2. Stuck rod scram loads to tube weld near the Sect. III, for Class A vessels, 3. OBE (E) center of the housing for type 304 stainless steel. for normal, upset and emergency conditions.
For normal and upset condition Sm= 15,800 psi @ 575°F For emergency conditions Siimit = Emergency condition loads 23,700 psi 1.5 Sm= 1.5x15,800 23,700 psi 1. Design pressure
- 2. Stuck rod scram loads
- 3. SSE (E')
CONTROL ROD DRIVE Primary Stress Limit - The Normal and upset condition Maximum stress inten- 26,060 psi allowable primary membrane loads sity occurs at a point stress plus bending stress is Maximum hydraulic pressure on the Y-Y axis of the based on ASME Boiler and from the control rod drive indicator tube.
Pressure Vessel Code, Sect. III supply pump.
for SA212 NOTE: Accident conditions TP316 tubing. do not increase this loading.
Earthquake loads are neglig-For normal and upset condition ible.
SA = 1. 5 Sm = 1. 5xl 7, 3 7 5 =
26,060 psi C-3-19 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING LOCATION ALLOWABLE STRESS CONTROL ROD GUIDE TUBE Primary Stress Limit Faulted condition loads The maximum bending 47,400 psi The allowable primary 1. Dead weight stress under faulted membrane stress plus 2. Pressure drop across guide tube loading conditions bending stress is based on due to failure of recirculation occurs at the center the ASME Boiler and Pressure line of the guide tube Vessel Code,Section III for 3. SSE (E')
Type 304 stainless steel tubing.
For normal and upset conditions SA 1.5 Sm= 1.5 X 15,800 =
23,700 psi For faulted condition S 1 imi t = 2 . 0 SA = 2 . 0 X 2 3 , 7 0 0 47,400 psi INCORE HOUSING Primary Stress Limit - The Emergency condition loads Maximum membrane 23,700 psi allowable primary membrane 1. Design pressure stress intensity stress is based on ASME Boiler 2. SSE (E') occurs at the outer and Pressure Vessel Code, Sect. surface of the III, for Class A vessels for vessel penetration Type 304 stainless steel.
For normal and upset conditions Sm= 15,800 psi@ 575°F For emergency condition (N+E')
S limit = 1 . 5 Sm = 1 . 5 X 15, 8 0 0 =
23,700 psi C-3-20 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING LOCATION ALLOWABLE STRESS HYDRAULIC CONTROL UNIT PIPING From USAS B31.1.0 - 1967 Code Normal Condition Load 3/4" drive 15,000 psi for power pressure piping Maximum normal hydraulic withdraw piping system pump pressure For Normal Conditions:
sh= 15,000 psi For upset and emergency con- Upset condition load 3/4" drive 18,000 psi dition: 1. Shut off pump pressure withdraw piping When upset or emergency 2. OBE (E) (negligible load) condition exists for less than 1% of the time, the code allows 20% increase in stress.
Sa= 1.2 sh= 18,000 psi Emergency condition 3/4" drive 18,000 psi
- 1. Shut off pump pressure withdraw piping
- 2. SSE (E') (negligible load)
C-3-21 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA LOADING LOCATION ALLOWABLE STRESS SPENT FUEL STORAGE RACKS Stresses due to normal, upset, Emergency condition At column to base 11,000 psi or emergency loading shall not "A" loads welds cause the racks to fail so as 1. Dead loads to result in a critical fuel 2. Full fuel load in rack At base hold down 20,000 psi array. 3. SSE (E') lug (casting)
Primary Stress Limit - Paper numbers 3341 and 3342, Pro- Emergency condition ceedings of the ASCE, Journal 11 B 11 loads (see below) of the Structural Division, Dec., 1962 (task committee on light-weigh t alloys) (Aluminum)
Emergency Conditions Stress limit= yield strength at 0.2% offset.
( 1) Load testing showed that the structure would not yield when subjected to simulated emergency condition "A"
loads. Strain gages mounted on the welds showed that calculated stresses are conservativ e.
(2) Calculated stresses compare very well with test results.
EMERGENCY CONDITION "B" Loading In addition to the loading conditions given above, the racks were tested and analyzed to determine their capability to safely withstand the accidental, uncontrolle d drop of the fuel grapple from its full retracted position into the weakest portion of the rack.
Method of Analysis The displacemen t of the vertical columns at the ends of the racks was determined by considering the effect of the grapple kinetic energy on the upper structure. The energy absorbed shearing the rack longitudina l structural member welds was determined.
C-3-22 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
Method of Analysis (continued)
The effect of the remaining energy on the vertical columns was analyzed. Equivalent static load tests were made on the structure to assure that the criteria were met.
Results of Analysis All criteria are met.
Analysis shows that the grapple would shear the welds in the area where the impact occurred. The longitudinal structural member bends but does not fail in shear. Grapple penetration into the rack is not sufficient to cause the vertical columns to deflect the fuel into a critical array. Static load testing showed that forces in excess of those resulting from a grapple drop are required to cause the columns to deflect to the extent that the criteria is violated.
C-3-23 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Piping CRITERIA LOADING Stress due to all normal and upset "Note: Analyses of this piping system demonstrate loadings must not exceed the limits of compliance with the applicable Code and are USAS B31. 1. 0. documented in the existing calculations of record."
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure and dead weight must b. Pressure be less than the hot allowable stress.
- 2. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of OBE(E) must be c. OBE (E) less than 1.2 times the hot allowable stress.
- 3. The sum of the longitudinal stresses a. Dead weight due to pressure and dead weight plus b. Pressure the thermal expansion stress C. Thermal intensity range must be less than the loads sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
- 4. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and OBE b. Pressure (E) plus the thermal expansion stress c. OBE (E) intensity range must be less than d. Thermal 1.2 times the sum of the allowable loads stress range for expansion stresses plus the hot allowable stress.
C-3-24 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Piping (Cont'd)
CRITERIA LOADING For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limit:
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of SSE (E') must be c. SSE (EI) less than 1.8 times the hot allowable stress. The probability of this load occurrence during the 40 year plant life is 10- 3 and SF= 1.5.
- 2. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead weight, b. Maximum pressure and inertia effects of OBE (E) must c. OBE (E) be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 40 year plant life is 10- 2 and SF= 1.8.
- 3. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead weight, b. Maximum pressure and inertia effects of SSE (E') must c. SSE (EI) be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 years plant life is
.25 x 10- 3 and SF= 1.36.
C-3-25 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Loop Piping "Note: Reanalysis of this piping system was performed as part of the piping replacement program in 1985. The basis for design criteria, loading conditions and stress allowables are provided in Impell Corp. Report No. 01-0840-1268; "Stress Analysis of Reactor Recirculation, Core Spray, and Reactor Water Cleanup Drywell Systems for Cooper Nuclear Station. [34 l Subsequent analyses based on the requirements of Reference 34 demonstrate compliance with the applicable Code and are documented in the existing calculations of records."
C-3-26 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Core Spray Discharge Piping CS-ID "Note: Reanalysis of this piping system was performed as part of the piping replacement program in 1985. The basis for design criteria, loading conditions and stress allowables are provided in Impell Corp. Report No. 01-0840-1268; "Stress Analysis of Reactor Recirculation, Core Spray, and Reactor Water Cleanup Drywell Systems for Cooper Nuclear Station. [34 l Subsequent analyses based on the requirements of Reference 34 demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
C-3-27 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Clean-Up Recirculation Pump Suction Piping CU-IS "Note: Reanalysis of this piping system was performed as part of the piping replacement program in 1985. The basis for design criteria, loading conditions and stress allowables are provided in Impell Corp. Report No. 01-0840-1268; "Stress Analysis of Reactor Recirculation, Core Spray, and Reactor Water Cleanup Drywell Systems for Cooper Nuclear Station. [34 l Subsequent analyses based on the requirements of Reference 34 demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
C-3-28 03/08/01 I
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Main Steam Piping to HPCI Turbine & Residual Heat Exchangers lA & 1B CRITERIA LOADING Stress due to all normal and upset "Note: Analyses of this piping system demonstrate loadings must not exceed the limits of compliance with the applicable Code and are USAS B31. 1. 0. documented in the existing calculations of record."
- 1. The sum of the longitudinal a. Dead weight stresses due to pressure and dead b. Pressure weight must be less than the hot allowable stress.
- 2. The sum of the longitudinal a. Dead weight stresses due to pressure, dead b. Pressure weight, and inertia effects of OBE c. OBE must be less than 1.2 times the hot allowable stress.
- 3. The sum of the longitudinal a. Dead weight stresses due to pressure and dead b. Pressure weight plus the thermal expansion c. Thermal loads stress intensity range must be less than the sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
- 4. The sum of the longitudinal a. Dead weight stresses due to pressure, dead b. Pressure weight, and OBE plus the thermal c. OBE expansion stress intensity range d. Thermal loads must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
C-3-29 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Main Steam Piping to HPCI Turbine & Residual Heat Exchangers lA & lB (Cont'd)
CRITERIA LOADING For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limits:
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of SSE must be less c. SSE than 1.8 times the hot allowable stress. The probability of this load occurrence during the 40 year plant life is 10- 3 and SF= 1.5.
- 2. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of OBE c. OBE must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 40 years plant life is 10- 2 and SF= 1.8.
- 3. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of SSE c. SSE must be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 years plant life is
.25 x 10- 3 and SF= 1.36.
C-3-30 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS Reactor Feed Piping CRITERIA LOADING Stress due to all normal and upset "Note: Analyses of this piping system demonstrate loadings must not exceed the limits of compliance with the applicable Code and are USAS B31. 1. 0. documented in the existing calculations of record."
- 1. The sum of the longitudinal a. Dead weight stresses due to pressure and dead b. Pressure weight must be less than the hot allowable stress.
- 2. The sum of the longitudinal a. Dead weight stresses due to pressure, dead b. Pressure weight, and inertia effects of OBE c. OBE must be less than 1.2 times the hot allowable stress.
- 3. The sum of the longitudinal a. Dead weight stresses due to pressure and dead b. Pressure weight plus the thermal expansion C. Thermal loads stress intensity range must be less than the sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
- 4. The sum of the longitudinal a. Dead weight stresses due to pressure, dead b. Pressure weight, and OBE plus the thermal C. OBE expansion stress intensity range d. Thermal loads must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
C-3-31 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Reactor Feed Piping (Cont'd)
CRITERIA LOADING For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limits:
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of SSE must be less c. SSE than 1.8 times the hot allowable stress. The probability of this load occurrence during the 40 year plant life is 10- 3 and SF= 1.5.
- 2. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of OBE c. OBE must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 40 years plant life is 10- 2 and SF= 1.8.
- 3. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of SSE c. SSE must be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 years plant life is
.25 x 10- 3 and SF= 1.36.
C-3-32 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS Residual Heat Removal Pump Suction Piping CRITERIA LOADING Stress due to all normal and upset "Note: Analyses of this piping system demonstrate loadings must not exceed the limits of compliance with the applicable Code and are USAS B31.1.0. documented in the existing calculations of record."
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure and dead weight must b. Pressure be less than the hot allowable stress.
- 2. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of OBE must be less c. OBE than 1.2 times the hot allowable stress.
- 3. The sum of the longitudinal stresses a. Dead weight due to pressure and dead weight plus b. Pressure the thermal expansion stress c. Thermal loads intensity range must be less than the sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
- 4. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure OBE plus the thermal expansion c. OBE stress intensity range must be less d. Thermal loads than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
C-3-33 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Residual Heat Removal Pump Suction Piping (Cont'd)
CRITERIA LOADING For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limits:
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of SSE must be less C. SSE than 1.8 times the hot allowable stress. The probability of this load occurrence during the 40 year plant life is 10- 3 and SF= 1.5.
- 2. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of OBE c. OBE must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 40 years plant life is 10- 2 and SF= 1.8.
- 3. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of SSE c. SSE must be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 years plant life is
.25 x 10- 3 and SF= 1.36.
C-3-34 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS Residual Heat Removal Pump Discharge Piping CRITERIA LOADING Stress due to all normal and upset "Note: Analyses of this piping system demonstrate loadings must not exceed the limits of compliance with the applicable Code and are USAS B31.1.0. documented in the existing calculations of record."
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure and dead weight must b. Pressure be less than the hot allowable stress.
- 2. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of OBE must be less c. OBE than 1.2 times the hot allowable stress.
- 3. The sum of the longitudinal stresses a. Dead weight due to pressure and dead weight plus b. Pressure the thermal expansion stress c. Thermal loads intensity range must be less than the sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
- 4. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure OBE plus the thermal expansion c. OBE stress intensity range must be less d. Thermal loads than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
C-3-35 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Residual Heat Removal Pump Discharge Piping (Cont'd)
CRITERIA LOADING For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limits:
- 1. The sum of the longitudinal stresses a. Dead weight due to pressure, dead weight, and b. Pressure inertia effects of SSE must be less c. SSE than 1.8 times the hot allowable stress. The probability of this load occurrence during the 40 year plant life is 10- 3 and SF= 1.5.
- 2. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of OBE c. OBE must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 40 years plant life is 10- 2 and SF= 1.8.
- 3. The sum of the longitudinal stresses a. Dead weight due to maximum pressure, dead b. Maximum pressure weight, and inertia effects of SSE c. SSE must be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 years plant life is
.25 x 10- 3 and SF= 1.36.
C-3-36 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Pumps CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR ACTUAL THICKNESS
- 1. Casing Minimum Wall Thickness t PR + C 3.00 in.
SE-06P Loads: Normal and upset condition where:
Design pressure & temperature t minimum required thickness, in.
p design pressure, psig Primary membrane stress R maximum internal radius, in.
limit: s allowable working stress, psi E joint efficiency Allowable working stress per ASME C corrosion allowance, in.
Section III, Class C
- 2. Casing Cover Minimum Thickness 3w [ 2 b4 (m-1)4b 4 (m+l)lna/b+a 2b 2(m+l)]
a - 2 b2 + - - - -
Sr --------
Loads: Normal and upset condition 4t2 a2(m-1) + b2(m+l)
Design pressure & temperature 2 2 3w- [ 12mb -2b (m+
+ - - -- -l)ln
- -a/-
b]
14,950 psi 2nt2 a (m-1)+ b2(m+l)
Primary bending stress limit:
2 4 4 2 2 3w(m -l) [a -b -4a b lna/b] +
1.5 Sm per ASME code for Pumps and St 2 2 Valves for Nuclear Power Class I 4mt a (m-1) + b 2(m+ 1)
~ [l + ma 2 (m-1) - mb 2(m+ 1) - 2(m2 - l)a 2 1n a/ b]
14,950 psi 2nmt 2 a2(m-1) + b 2(m+ 1) where: Sr radial stress at outer edge, psi t disc thickness, in.
St tangential stress at inner edge, psi m reciprocal of Poisson's ratio w pressure load, psi a radius of disc, in.
W uniform load along inner edge, lb. b radius of disc hole, in.
C-3-37 03/08/01
USAR TABLE C-3-7 (Cont'd) LOADING CRITERIA Recirculation Pumps (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR ACTUAL THICKNESS
- 3. Cover and seal flange bolt Bolting loads, areas and stresses are calculated in Cover Flange Bolts areas accordance with "Rules for Bolted Flange Connections"
- ASME Section VIII, Appendix II. 27,000 psi Loads: Normal and upset condition Seal Flange Bolts Design pressure & temperature. 20,000 psi Design gasket load Bolting Stress Limit:
Allowable working stress per ASME Section III, Class C
- 4. Cover Clamp Flange Thickness Flange thickness and stress are calculated in Flange Thickness accordance with "Rules for Bolted Flange Connections" Loads: Normal and upset - ASME Section VIII, Appendix II. 8. 25" condition Design pressure & temperature.
Design gasket loado Design bolting load.
Tangential Flange Stress Limit:
Allowable working stress per ASME Section III, Class C
- 5. Pump Nozzle Membrane and Bending Stress re D 2 P M F SL= -- + - + -
4 A Z A Loads: Normal and upset condition PD Sc=
Design pressure & temperature. 2t Piping reactions during normal Ss = TRo J
C-3-38 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Pumps CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR ACTUAL THICKNESS
- 5. Pump Nozzle Membrane and Bending Stress (Cont'd)
Combined Stress Limit:
1.5 Sm per ASME code for Pumps and Valves for Nuclear Power Class I.
+
2 where: Pump Nozzle Stresses SL longitudinal stress, psi Sc circimferential stress, psi Ss torsional stress, psi 28,650 psi D nozzle internal diameter, in.
p design pressure, psi 28,650 psi A nozzle cross section metal area, in 2 M maximum bending moment, in. lb. 28,650 psi F maximum longitudinal force, lb.
t nozzle wall thickness, in. 28,650 psi J polar moment of inertia, in 4 Ro nozzle outside radius, in.
T torsional moment
- 6. Mounting Bracket Combined Stress Loads: Bracket vertical loads are determined by Maximum Combined summing the equipment and fluid weights and Stresses Flooded weight vertical seismic forces. Bracket horizontal loads are determined by applying the specified SSE (EI) seismic force at mass center of pump-motor assembly (flooded). 15,600 psi Combined Stress Limit: Horizontal and vertical loads are applied simultaneously to determine tensile, shear and Yield Stress bending stresses in the brackets. Tensile, shear and bending stresses are combined to determine maximum combined stresses.
C-3-39 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Pumps (Cont'd)
ALLOWABLE STRESS OR CRITERIA METHOD OF ANALYSIS ACTUAL THICKNESS
- 7. Stresses Due to Seismic Loads The flooded pump-motor assembly is analyzed as a Motor Bolt Tensile free body supported by constant support hangers from Stress:
Loads: the pump brackets. Horizontal and vertical seismic forces are applied at mass center of assembly and 40,000 psi Operating pressure and temperature equilibrium reactions are determined for the motor SSE (E') and pump brackets. Load, shear, and moment diagrams Pump Cover Bolt Tensile are constructed using live loads, dead loads, and Stress:
calculated snubber reactions. Combined bending, tension and shear stresses are determined for each 43,200 psi Combined Stress Limit: major component of the assembly including motor, motor support barrel, bolting and pump casing. The Motor Support Barrel Yield stress maximum combined tensile stress in the cover bolting Combined Stresses:
is calculated using tensile stresses determined from loading diagram plus tensile stress from operating 22,400 psi pressure.
C-3-40 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RHR Pump OBE (E) Coefficients 0.23 g horizontal 0.07 g vertical SSE (E') Coefficients 0.46 g horizontal 0.14 g vertical The RHR pump is a verticaly mounted, single stage pump driven by direct-coupled motors. The motor is mounted to and above the pump, and the pump is mounted to the foundation.
Statement of Criteria Method of Analysis Results
- 1. Closure bolting is designed to contain 1. Bolting loads and stresses are 1. Allowable Stress 25,000 psi the internal design pressure of the calculated in accordance with the pump casing without exceeding the "Rules for Bolted Flange Connections", Pump Design Pressure 450 psig allowable stress of the bolting ASME Boiler and Pressure Vessel Code, Max. Design Temperature 350°F material. Allowable stresses at Section VIII, Appendix II.
design temperature are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the pump 2. Stress in the pump casing is 2. Allowable Stress 14,000 psi limits stress to the allowable stress calculated at the point of maximum when subjected to design pressure and internal pump diameter by the formula temperature. Allowable stresses are in accordance with ASME B&PV Code,Section VIII. P(D + .2 t)
Sc 2t where Sc calculated stress, psi p pump design pressure, psi D maximum pump internal diameter t actual min. metal thickness less corrosion allowance, 0.080 inches C-3-41 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RHR Pump (Cont'd)
Statement of Criteria Method of Analysis
- 3. Application of forces and moments by 3. Stresses will not be excessive if the attaching pipe on pump nozzles maximum force when taken with the under combined maximum thermal expansion maximum moment falls below the line.
and OBE loading (E) reaction plus load due to internal pressure does not produce an equivalent bending and torsional F~
stress in the nozzles in excess of the M
allowable stress as defined by the ASME B&PV Code,Section VIII. E E' suction Fintercept 77,007 124,988 lbs.
For SSE (E'), less than 1.5 of allowable (M=0) stress.
Mintercept 725,709 1,385,100 in.lbs.
(F=0)
Fintercept 65,696 116,898 lbs.
(M=0) discharge Mintercept 629,003 1,060,236 in.lbs.
(F=0)
C-3-42 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Core Spray Pump OBE (E) Coefficients 0.23 g horizontal 0.07 g vertical SSE (E') Coefficients 0.46 g horizontal 0.14 g vertical The Core Spray or RHR pump is a vertically mounted, single stage pump driven by direct coupled motors. The motor is mounted to and above the pump, and the pump is mounted to the foundation.
Statement of Criteria Method of Analysis Results
- 1. Closure bolting is designed to 1. Bolting loads and stresses are 1. Allowable Stress 20,000 psi contain the internal design calculated in accordance with the pressure of the pump casing "Rules for Bolted Flange without exceeding the allowable Connections", ASME Boiler and Pump Design Pressure 500 psi stress of the bolting material. Pressure Vessel Code, Section Max. Design Temperature 212°F Allowable stresses at VIII, Appendix II.
design temperature are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the 2. Stress in the pump casing is 2. Allowable Stress 14,000 psi pump limits stress to the calculated at the point of allowable stress when subjected to maximum internal pump diameter by design pressure and temperature. the formula Allowable stresses are in accordance with ASME B&PV Code, P(D + .2 t)
Section VIII. Sc 2t where Sc calculated stress, psi P pump design pressure, psi D maximum pump internal diameter t actual min. metal thickness less corrosion allowance, 0.080 inches C-3-43 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Core Spray Pump (Cont'd)
Statement of Criteria Method of Analysis
- 3. Application of forces and 3. Stresses will not be excessive if moments by attaching pipe on the maximum force when taken with pump nozzles under combined the maximum moment falls below the maximum thermal expansion and line.
OBE loading reaction (E) plus load due to internal pressure does not produce an F~
equivalent bending and torsional stress in the M nozzles in excess of the allowable stress as defined E E' by the ASME B&PV Code, Suction Fintercept 54,535 79,808 lbs.
Section VIII. (M=0)
For SSE (E') less than 1.5 of Mintercept 402,803 623,427 in.lbs.
allowable stress. (F=0)
Fintercept 30,624 59,668 lbs.
Discharge (M=0)
Mintercept 293,562 426,519 in.lbs.
(F=0)
C-3-44 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Pump OBE (E) Coefficients 0.23 g horizontal 0.07 g vertical SSE (E') Coefficients 0.46 g horizontal 0.14 g vertical The HPCI pumps are multi-stage, horizontally mounted, split case pumps driven by their respective steam turbines through couplings. The HPCI pump consists of a main pump and booster on a common baseplate separate from the turbine baseplate.
Statement of Criteria Method of Analysis Results Main Pump
- 1. Closure bolting is designed to 1. Bolting loads and stresses are 1. Design Pressure 1,500 psig contain the internal design pressure calculated in accordance with the Allowable Stress 20,000 psi of the pump casing without exceeding "Rules for Bolted Flange the allowable working stress of the Connections", ASME Boiler and bolting material. Allowable stresses Pressure Vessel Code, Results Boost Pump are in accordance with ASME B&PV Section VIII, Appendix II.
Code,Section VIII. Design Pressure 450 psi Allowable Stress 20,000 psi
- 2. The minimum wall thickness of the 2. Nozzle stress is calculated by 2. Main Pump pump limits stress to the allowable the following formula working stress when subjected to P(D + .2 t) Design Pressure 1,500 psig design pressure plus corrosion Sh = 2Et Allowable Stress 14,000 psi allowance. Allowable stresses are in accordance with ASME B&PV Code, Volute stress is calculated by Section III. Boost Pump the following formula The maximum stress in the pump a)
Design Pressure 450 psig casing when subjected to design Sv = Pb (R + Roark Allowable Stress 14,000 psi pressure does not exceed the 2t R allowable working stress of the p. 307 Case 26 material. The allowable stress are in accordance with ASME B&PV Code,Section III.
and R a - 0.5b C-3-45 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Pump (Cont'd)
Statement of Criteria Method of Analysis
- 3. Application of forces and moments 3. Stresses will not be excessive if by attaching pipe on pump nozzles the maximum force when taken with under combined maximum thermal the maximum moment falls below the expansion and OBE loading reaction line.
(E) plus load due to internal pressure does not produce an equivalent bending and torsional F~
stress in the nozzles in excess of the allowable stress as defined by M the ASME B&PV Code,Section VIII.
Suction E E' For SSE (E') less than 1.5 of rintercept 33,000 65,000 lbs.
allowable stress. (M=0)
Mintercept 430,000 700,000 in.lbs.
(F=0)
Discharge r intercept 31,000 63,000 lbs.
(M=0)
Mintercept 250,000 540,000 in.lbs.
C-3-46 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Pump OBE (E) Coefficients 0.23 g horizontal 0.07 g vertical SSE (E') Coefficients 0.46 g horizontal 0.14 g vertical The RCIC pump is a multi-stage, horizontally mounted, split case pump driven by their respective steam turbines through couplings. The RCIC pump and turbine are on a common baseplate.
Statement of Criteria Method of Analysis Results
- 1. Closure bolting is designed to contain 1. Bolting loads and stresses are 1. Design Pressure 1,500 psig the internal design pressure of the calculated in accordance with the Allowable Stress 20,000 psi pump casing without exceeding the "Rules for Bolted Flange Connections",
allowable working stress of the bolting ASME Boiler and Pressure Vessel Code, material. Allowable stresses are in Section VIII, Appendix II.
accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the pump 2. Stress in the pump nozzle is 2. Design pressure 1,500 psig limits stress to the allowable working calculated at the point of maximum Allowable Stress 14,000 psi stress when subjected to design internal pump diameter by the formula pressure plus corrosion allowance.
Allowable stresses are in accordance P(D + .2t) with ASME B&PV Code,Section III. Sc 2Et The maximum stress in the pump casing Sc= 0.8Sa when subjected to design pressure does not exceed the allowable working Design Pressure 1,500 psig volute stress is computed by the stress of the material. The allowable Allowable Stress 14,000 psi following formula:
stress is in accordance with ASME B&PV Code,Section III.
Sb=/JPb2 Roark p. 225 t2 Case# 36 factor from Roark volute length volute width C-3-47 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Pump (Cont'd)
Statement of Criteria Method of Analysis
- 3. Application of forces and moments 3. Stresses will not be excessive if the by attaching pipe on pump nozzles maximum force when taken with the under combined maximum thermal maximum moment falls below the line.
expansion and OBE loading reaction (E) plus load due to internal pressure does not F~
produce an equivalent bending and torsional stress in the nozzles M in excess of the allowable stress Suction E E' as defined by the ASME B&PV Code, Fintercept 6,200 11,800 lbs.
Section VIII. (M=0)
For SSE less than 1.5 of Mintercept 34,700 64,000 in.lbs.
allowable stress. (F=0)
Discharge Fintercept 2,800 7,200 lbs.
(M=0)
Mint~rcept 12,300 30,000 in.lbs.
(F=0)
C-3-48 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Standby Liquid Control Pump OBE (E) Coefficients 0.33 g horizontal 0.07 g vertical SSE (E') Coefficients 0.66 g horizontal 0.14 g vertical The standby liquid control pump is a horizontally mounted, multiple piston, positive displacement pump. It is driven by an ACgear motor through a coupling. The pump and motor are independently bolted to a common baseplate. The baseplate is anchored to its foundation.
Statement of Criteria Method of Analysis Results
- 1. Closure bolting is designed to 1. Bolting loads and stresses are 1. Stuffing Box Bolts contain the internal design calculated in accordance with the Allowable Stress 25,000 psi pressure of the pump without "Rules for Bolted Flange Connections",
exceeding the allowable working ASME Boiler and Pressure Vessel Code, Cylinder Head Bolts stress of the bolting material. Section VIII, Appendix II. Allowable Stress 25,000 psi Allowable stresses are in accordance with ASME B&PV Code.
- 2. The maximum stress in the pump 2. Stress in the pump fluid cylinder is 2. Design Pressure 1,400 psig fluid cylinder when subjected to calculated at the point of maximum Allowable Stress 16,500 psi design pressure does not exceed stress by the pressure area method.
the allowable working stress of the material. The allowable stress is in accordance with ASME B&PV Code,Section VIII.
- 3. The stresses in the motor 3. The seismic forces acting on the 3. The bolt stresses under (E')
mounting bolts when the motor is motor to subject the bolting to Tension subjected to the SSE does not shear or tension are considered. Allowable Stress exceed 0.9 of yield stress and The motor is isolated from the twice the allowable shear stress pump and nozzle forces by the Shear for bolting material in accordance flexible coupling. Allowable Stress with the ASME Boiler and Pressure Vessel Code,Section VIII.
C-3-49 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Standby Liquid Control Pump (Cont'd)
Statement of Criteria Method of Analysis
- 4. The stresses in the pump mounting 4. The maximum force taken with the bolts due to the combination of OBE maximum moment shall fall below acting on the flooded pump plus the the line on the force-moment attaching pipe reactions under diagram:
combined maximum thermal expansion plus OBE does not exceed the allowable shear and tensile stresses F for the bolting material in accordance with the ASME Boiler and M Pressure Vessel Code,Section VIII.
E The attaching pipe reaction plus the Discharge F 342 lbs load due to internal pressure does not produce an equivalent M .9450 in. lbs., but bending and torsional stress in not to exceed 3400 nozzles in excess of the allowable in. lbs.
stress.
Suction F 711 lbs.
The stresses in the pump mounting M = 39,100 in. lbs.,
bolts due to the combination of the but not to exceed SSE acting on the flooded pump plus 16,600 in. lbs.
the attaching pipe reaction under E'
combined maximum thermal expansion Discharge F 684 lbs plus SSE does not exceed 0.9 M = 18,900 in. lbs.,
times the yield stress in tension but not to exceed and twice the allowable shear stress 5,330 in. lbs.
for the bolting material in accord-ance with the ASME Boiler and Suction F 1,422 lbs Pressure Vessel Code,Section VIII.
M 78,200 in. lbs. ,
The attaching pipe reaction plus the load due to internal pressure does but not to exceed 24,700 in. lbs.
not produce an equivalent bending and torsional stress in nozzles in excess where Mis maximum moment (in. lbs.)
of 1.5 times allowable stress.
in any direction and Fis maximum force (lbs.) in any direction.
C-3-50 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine OBE (E) Coefficients 0.75 g horizontal 0.07 g vertical SSE (E') Coefficients 1.50 g horizontal 0.14 g vertical The HPCI turbine is a horizontally mounted solid wheel turbine, having two stages (or wheels) in tandem. The turbine drives its pump through a flexible coupling. The turbine to baseplate mounting is accomplished by bolting and doweling at the exhaust end, and by a sliding pedestal (axial movement only) at the inlet end. The HPCI turbine is mounted on its own baseplate, independent of the pump mounting.
Statement of Criteria Method of Analysis Results
- 1. Closure bolting is designed to contain 1. Bolting loads and stresses are 1. Allowable Stress 20,000 psi the internal design pressure of the calculated in accordance with the turbine casing without exceeding the "Rules for Bolted Flange Connections",
allowable working stress of the bolting ASME Boiler and Pressure Vessel Code, material. Allowable stresses are in Section VIII, Appendix II.
accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the 2. Stresses in the various pressure 2. Allowable Stress 17,500 psi turbine casing limits stress to the containing portions of the turbine allowable working stress when subjected to design pressure plus corrosion casing are calculated according to the Rules of Part UG,Section VIII, of the I
allowance. Allowable stresses are in ASME Boiler & Pressure Vessel Code.
accordance with ASME B&PV Code,Section VIII.
C-3-51 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine (Cont'd)
Statement of Criteria Method of Analysis
- 3. The forces and moments imposed by the 3. The total resultant of the forces and the total resultant of attached piping on the turbine inlet the moments on both the inlet and the exhaust connections of and exhaust connections satisfy the the turbine satisfies the following conditions:
following conditions:
For the combination of dead weight and maximum thermal The resultant force and moment expansion, from the combination of dead weight and thermal expansion is Inlet-----F 2,520 lbs.
less than that stipulated by the M 7,570 ft. lbs.
equipment vendor.
Exhaust---F 3,310 lbs.
The resultant force and moment M 9,930 ft. lbs.
from the combination of dead weight, thermal expansion and OBE For the combination of dead weight, maximum thermal (E) or SSE (E') is less than that expansion and OBE (E),
demonstrated acceptable by detailed seismic analysis of the Inlet-----F 8,000 lbs, but not to exceed 5,000 lbs.
equipment. M 20,000 ft. lbs.
Exhaust---F 25,000lbs, but not to exceed 11,500 lbs.
M 20,000ft. lbs.
For the combination of dead weight, maximum thermal expansion, and SSE (E'),
Inlet-----F 12,000 lbs, but not to exceed 7,500 lbs.
M 30,000 ft. lbs.
Exhaust---F 37,500 lbs, but not to exceed 17,250 lbs.
M 30,000 ft. lbs.
C-3-52 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine (Cont'd)
Statement of Criteria Method of Analysis
- 3. Continued where "F" is the resultant force in lbs, and "M" is the resultant moment in ft-lbs Typical acceptable area on the force-moment diagram is indicated below:
F M
- 4. The stresses in the turbine anchor 4. Vertical forces on the anchor bolts, subjecting them to bolts (turbine to baseplate, and tension, are the sum of the following:
baseplate to foundation) due to the combination of the OBE (E) acting on a) Weight of the turbine assembly times the vertical component of the turbine while running plus the acceleration, total piping loads [weight, thermal & b) The vertical pipe force reactions, OBE (E)] does not exceed the c) The pipe moment reactions tending to tip the turbine and allowable tensile stress nor subject the bolting to tension the allowable shear stress for the bolting materials as specified Horizontal forces on the anchor bolts, subjecting them to shear in the ASME Boiler & Pressure Vessel and tension, are the sum of the following:
Code,Section VIII.
a) Weight of the turbine assembly times the horizontal component of acceleration acting on the center of gravity, b) The horizontal pipe force reactions, c) The effect of pipe moment reactions causing horizontal loading at the anchor bolts.
NOTE: Friction shall not be considered to be restrictive C-3-53 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine (Cont'd)
Statement of Criteria Method of Analysis
- 5. The stresses in the turbine anchor 5. Same as analysis under (4), above.
bolts (turbine to baseplate and baseplate to foundation) due to the combination of SSE (E') acting* on the turbine in standby plus the total piping loads [weight, thermal and SSE (E')] does not exceed 0.9 times the yield stress in tension, nor twice the allowable shear stress for the bolting materials as specified in the ASME Boiler and Pressure Vessel Code,Section VIII.
C-3-54 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine OBE (E) Coefficients 0.75 g horizontal 0.07 g vertical SSE (E') Coefficients 1.50 g horizontal 0.14 g vertical The RCIC turbine is a horizontally mounted solid wheel turbine. The turbine drives its pump through a flexible coupling.
The turbine to baseplate mounting is accomplished by bolting and doweling at the exhaust end, and by a sliding pedestal (axial movement only) at the inlet end. The RCIC turbine is mounted on a common baseplate with its pump.
Statement of Criteria Method of Analysis Results
- 1. Closure bolting is designed to contain 1. Bolting loads and stresses are 1. Maximum Allowable Stress the internal design pressure of the calculated in accordance with the 20,000 psi turbine casing without exceeding the "Rules for Bolted Flange Connections",
allowable working stress of the bolting ASME Boiler and Pressure Vessel Code, material. Allowable stresses are in Section VIII, Appendix II.
accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the 2. Stresses in the various pressure 2. Maximum Allowable Stress turbine casing limits stress to the containing portions of the turbine 17,500 psi allowable working stress when subjected casing are calculated according to the to design pressure plus corrosion rules of Part UG,Section VIII, of the allowance. Allowable stresses are in ASME Boiler & Pressure Vessel Code.
accordance with ASME B&PV Code,Section VIII.
C-3-55 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine (Cont'd)
Statement of Criteria Method of Analysis
- 3. The forces and moments imposed by the 3. The total resultant of the forces and the total resultant of attached piping on the turbine inlet the moments on both the inlet and the exhaust connections of and exhaust connections satisfy the the turbine satisfy the following conditions:
following conditions:
For the combination of dead weight and maximum thermal The resultant force and expansion, moment from the combination of dead weight and thermal Inlet-----F 870 lbs.
expansion are less than M 2,620 ft. lbs.
that stipulated by the Exhaust---F 2,000 lbs.
equipment vendor. M 6,000 ft. lbs.
The resultant force and For the combination of dead weight, maximum thermal moment from the combination of expansion, and OBE (E),
dead weight, thermal expansion and OBE or SSE is less than that Inlet-----F 1,200 lbs demonstrated acceptable by M 3,000 detailed seismic analysis of the Exhaust---F 18,000 lbs, but not to exceed 8,370 lbs.
equipment. M 6,000 ft. lbs.
For the combination of dead weight, maximum thermal expansion, and SSE (E'),
Inlet-----F 1,800 lbs.
M 4,500 ft. lbs.
Exhaust---F 27,000 lbs, but not to exceed 12,555 lbs.
M 9,000 ft. lbs.
where "F" is the resultant force in lbs and "M" is the resultant moment in ft. lbs.
C-3-56 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine (Cont'd)
Statement of Criteria Method of Analysis Typical acceptable area on the force-moment diagram is indicated below:
max . ...,)
,____..,.-"-'-:...-;..--------------- M
- 4. The stresses in the turbine anchor 4. Vertical forces on the anchor bolts are the sum of the bolts (turbine to baseplate) due following:
to the combination of the OBE (E) acting on the turbine while running a) Weight of the turbine assembly times the vertical component plus the total piping loads [weight, of acceleration, thermal and SSE (E)] shall not exceed b) The vertical pipe force reactions, the allowable tensile stress nor the c) The pipe moment reactions tending to tip the turbine and allowable shear stress for the bolting subject the bolting to tension.
materials as specified in the ASME Boiler and Pressure Vessel Code, Horizontal forces on the anchor bolts, subjecting them to shear Section VIII. and tension, are the sum of the following:
a) Weight of the turbine assembly times the horizontal component of acceleration acting on the center of gravity, b) The horizontal pipe force reactions, c) The effect of pipe moment reactions causing horizontal loading at the anchor bolts.
NOTE: Friction shall not be considered to be restrictive C-3-57 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine (Cont'd)
Statement of Criteria Method of Analysis
- 5. The stresses in the turbine anchor 5. Same as analysis under (4), above.
bolts (turbine to baseplate) due to the combination of SSE (E')
acting on the turbine in standby plus the total piping loads [weight, thermal and SSE (E')] does not exceed 0.9 times the yield stress in tension nor twice the allowable shear stress for the bolting materials as specified in the ASME Boiler and Pressure Vessel Code,Section VIII.
C-3-58 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Isolation Valves CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 1. Body Minimum Wall Thickness Minimum wall thickness in the cylindrical portions of the valve is calculated using the Load: following formula: t 1. 74 in.
Design pressure and temperature 1.5 [ Pd + C]
2S-1.2P Primary membrane stress where:
limit: s allowable stress of 7000 psi p primary service pressure, 655 psi S = 7,000 psi per ASA d inside diameter of valve at section Bl6.5 being considered, inches C corrosion allowance of 0.12 inches
- 2. Cover Minimum Thickness 1/2 CP
-+
1.78 WhG + C1 Loads: t=d [ S ]
Sd 3 Design pressure and where:
temperature t minimum thickness, inches t 4.50 in Design bolting load d diameter or short span, inches Gasket load C attachment factor s allowable stress, psi Primary stress limit: w total, bolt load, pounds hC gasket moment arm, inches Allowable working stress C1 corrosion allowance, inches per ASME Section VIII.
C-3-59 03/08/01
USAR TABLE C-3-7 (Cont'd)
Main Steam Isolation Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 3. Cover Flange Bolt Area: Total, bolting loads and stresses are calculated in accordance with "Rules for Bolted Flange Loads: Connections" - ASME Boiler & Pressure Vessel Code, s 33,100 psi Section VIII, Appendix II, except that the stem @ 5 7 5°F Design pressure and temperature operational load and seismic loads are included in Gasket load the total load carried by bolts. The horizontal Stem operational load and vertical seismic forces are applied at the Seismic load - E' mass center of the valve operator assuming that the valve body is rigid and anchored Bolting stress limit:
Allowable working stress per ASME Nuclear Pump & Valve Code, Class I.
- 4. Body Flange Thickness & Flange thickness and stress are calculated in Stress accordance with "Rules for Bolted Flange Connections" - ASME Boiler & Pressure Vessel Code, Loads: Section VIII, Appendix II, except that the stem s 35,000 psi operational load and seismic loads are included in s 35,000 psi Design pressure & temperature the total load carried by the flange. The s 35,000 psi Gasket load horizontal and vertical seismic forces are applied Stem operational load at the mass center of the valve operator assuming Seismic load - E' that the valve body is rigid and anchored.
Flange Stress Limits:
fT I
fHr ~Rt 1.5 Sm per ASME Nuclear Pump
& Valve Code, Class I
- 5. Valve Disc Thickness 3W 2
2 a (m + 1) log ~+ a 2
(m - 1) - b 2
(M - 1)
Load: Max Sr s 23,300 psi 2 Jr t 2 a (rn + 1) +
2 b 2 (m - 1)
Design pressure & temperature Primary bending stress where:
limit: w total applied load, lb.
Allowable working stress per ASME t thickness of disc, in.
Section VIII. a outer radius of disc b inner radius (fixed) of disc m 3.85, reciprocal of Poisson's ratio C-3-60 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Isolation Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 6. Valve Operator Supports The valve assembly is analyzed assuming that the valve body is an anchored, rigid mass and s 20,000 psi Loads: that the specified vertical and horizontal seismic forces are applied at the mass center Design pressure & temperature of the operator assembly, simultaneously with Stem operational load operating pressure plus dead weight plus Equipment dead weight operational loads. Using these loads, stresses Seismic load - E' and defelections are determined for the operator support components.
Support Rod Stress Limit:
Allowable working stress per ASME Section VIII.
C-3-61 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Stearn Safety Valves CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 1. Inlet Nozzle Wall Thickness PR + C Load: SE-0.6P t 0.143 in.
where:
1.1 x Design Press. @ 600°F t min. required thickness, inches s allowable stress, psi Primary Membrane Stress p 1.1 x design press., psi Limit: R internal radius, inches E joint efficiency C corrosion allowable, inches Allowable stress intensity as defined by ASME Standard Code for Pumps & Valves for Nuclear Power
- 2. Valve Disc Thickness w PA1 Ss = A A
Load: where:
1.1 x Design press. @ 600°F w shear load, lb. Ss 12,780 psi A shear area, in 2 Diagonal Shear Stress Limit: p 1.1 x design press, psi A1 disc area, in 2 0.6 x allowable stress inten-sity as defined by ASME and:
(R + R )
1 Standard Code for Pumps & A TCS Valves for Nuclear Power. s slope of frustrurn of shear cone, in.
R radius at base of cone, in.
Rl radius at top of cone, in.
C-3-62 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 3. Inlet Flange Bolt Area Total bolting loads and stresses are Sb= 27,700 psi calculated in accordance with procedures of Loads: Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code.
Design pressure & temperature Gasket load Operational load SSE (E')
Bolting Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Stand-ard Code for Pumps & Valves for Nuclear Power
- 4. Inlet Flange Thickness Flange thickness and stresses are calculated SH 27,300 psi in accordance with procedures of Para. 1- SR 27,300 psi Loads: 704.5.1 Flanged Joints, of B31.7 Nuclear ST 27,300 psi Piping Code.
Design pressure & temperature Gasket load Operational load Seismic load - E' Flange Stress Limits,
.§.Hr _§_R, .§.T 1.5 Sm per ASME Nuclear Pump & Valve Code C-3-63 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 5. Valve Spring-Torsional Stress Set Point Smax 8PD [4C-1 + 0.615]
Loads: Hd 3 4C-4 C S = 62,500 psi W1 Set point load, lbs where: Max. Lift W2 Spring load at maximum Smax torsional stress, psi p W1 or W2 = spring load, lbs lift, lbs. S = 93,750 psi D mean diameter of coil, in.
Torsional Stress Limit: d diameter of wire, in.
C D = correction factor D
0.67 x torsional elastic limit when subjected to a load of W1
- 0.90 x torsional elastic limit when subjected to a F A
load of W2
- 2Sm
- 6. Yoke Rod Area Load: where:
A required area per rod, in 2 A 0.575 in 2 Spring load at maximum lift F total spring load, lbs Sm allowable stress, psi Primary Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Standard Code for Pumps & Valves for Nuclear Power C-3-64 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 7. Yoke Bending & Shear Stresses M V Sb z, Ss A Sb 18,200 psi Load:
Ss = 10,900 psi Spring load at maximum lift where:
Sb bending stress, psi Bending & Shear Stress Limits:
Ss shear stress, psi M bending moment, in. lbs.
Bending - allowable stress intensity, Sm, per ASME Nuclear z section modulus, in3 Pump & Valve Code Shear - 0.6 x V vertical shear, lb.
allowable stress intensity, 0.6 A shear area, in2 Sm, per ASME Nuclear Pump &
Valve Code.
Body Bowl
- 8. Body Minimum Wall Thickness t = 1.5 [ Pd ] + C t = 0.311 in.
2S 1.2P Load: where: Inlet Nozzle t required thickness, in.
Primary service pressure s allowable stress, 7,000 psi t = 0.218 in.
p primary service pressure, 150 psi Primary Stress Limit: d inside diameter of valve at sec- Outlet Nozzle tion being considered, in.
Allowable stress, 7,000 psi, in t = 0.250 in.
accordance with ASA Bl6.5.
s F1 + F2 + M1 + M2
- 9. Inlet Nozzle Combined Stress s 27,300 psi A Z Loads:
where:
Spring load at maximum lift Operational load s combined bending & tensile stress, F1 maximum spring load, lb.
Seismic load - E' F2 vertical component of reaction thrust, lb.
Combined Stress Limit:
A cross section area of nozzle, in 2 M1 moment resulting from horizontal 1.5 x allowable stress component of reaction, lb. in.
intensity, 1.5 Sm, per ASME M2 moment resulting from horizontal Code for Pumps & Valves for Nuclear Power.
C-3-65 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 10. Spindle Diameter Fe= 1/EI Load Limit (0.2Fc)
Load: L2 F = 85,900 lb.
Spring load at maximum lift where:
Fe critical buckling load, lb.
Spindle Column Load Limit: E modulus of elasticity, psi I moment of inertia, in 4 0.2 x critical buckling load L length of spindle in compression, in.
F
- 11. Spring Washer Shear Area Ss = A Ss 15,960 psi where:
Load: shear stress, psi Ss F spring load, lb.
Spring load at maximum lift A shear area, in 2 Shear Stress Limit:
0.6 x allowable stress in-tensity, 0.6 Sm, per ASME Nuclear Pump & Valve Code.
C-3-66 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Relief Valves CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 1. Body Minimum Wall Thickness t = 1.5 [ Pd ] + C Main Body Load: 2S-1.2P t = 0.625 in.
where:
Design Pressure & Temperature t minimum required thickness, in.
s allowable stress, 7,000 psi Bonnet Primary Membrane Stress Limit: p primary service pressure, 655 t = 0.287 in.
d inside diameter of valve at sec-Allowable working stress as tion being considered, in.
defined by USAS B16.5 (7,000 C corrosion allowance, 0.12 in.
psi@ primary service pres-sure).
t = d [- + 3 ]
+ C1 Sm Smd Loads: Pilot Base where: t = 2.219 in.
Design pressure & temperature t minimum required thickness in.
Gasket load d diameter or short span, in.
C attachment factor, ASME Section Primary Stress Limit: VIII p design pressure, psi Allowable stress intensity, Sm, Sm allowable stress, psi as defined by ASME Standard w total bolt load, lb.
Code for Pumps & Valves for hG gasket moment arm, in.
Nuclear Power. C1 corrosion allowance, 0.12 in.
C-3-67 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Relief Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
- 3. Flange Bolt Area - Inlet Total bolting loads and stresses are Body to Base Flange, Outlet Flange, Body calculated in accordance with procedures of Ab = 10 . 2 6 in 2 to Bonnet, Bonnet to Base Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code.
Loads:
Bonnet to Cap Design pressure & temperature Ab = 1. 4 52 in 2 Gasket load Operational load Inlet Flange SSE (EI) Ab = 13. 9 in 2 Bolting Stress Limit:
Allowable stress intensity, Outlet Flange Sm, as defined by ASME Stand- Ab = 12. 2 in 2 ard Code for Pumps & Valves for Nuclear Power.
- 4. Flange Thickness - Inlet, Out- Flange thickness and stresses are calculated Body to Base let, Bonnet Flanges in accordance with procedures of Para. 1- SH= 26,250 psi 704.5.1 Flanged Joints, of B31.7 Nuclear SR = 26,250 psi Loads: Piping Code. ST = 26,250 psi Design pressure & temperature Gasket load Cap to Bonnet Operational load SH= 26,250 psi SSE (EI) SR = 26,250 psi ST = 26,250 psi Flange Stress Limits,
_§_H, _§_R, _§_T Inlet Flange 1.5 Sm per ASME Nuclear SH = 26,250 psi Pumps and Valve Code. SR = 26,250 psi ST = 26,250 psi C-3-68 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Relief Valves (Cont')
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Outlet Flange SH= 26,250 psi SR = 26,250 psi ST = 26,250 psi 3 (3 + u) PR 2
- 5. Valve Disc Thickness & Sr St Disc Stress Stress 8 t2 Sm = 15 , 8 0 0 psi Load: where:
Sr radial stress, psi Design pressure & temperature St tangential stress, psi V Poisson's ratio Primary Stress Limit: p design pressure, psi R radius of disc, in.
Allowable stress intensity, Sm, T thickness of disc, in.
as defined by ASME Standard Code for Pumps & Valve for Nuclear Power.
- 6. Inlet Nozzle Diameter Thick- F1 + F2 + M1 + M2 ness & Stress s Inlet Nozzle Stress A Z S = 26,250 psi Loads: where:
s combined bending and tensile stress, Design pressure & temperature psi Operational load vertical load due to design pressure, F1 SSE (EI) lb.
F2 vertical component of reaction Primary Stress Limit:
thrust, lb.
A cross section area of nozzle, in 2 1.5 x allowable stress in- moment resulting from horizontal M1 tensity, 1.5 Sm, as defined by reaction, in. lb.
ASME Standard Code for Pumps & moment resulting from horizontal M2 Valves for Nuclear Power. seismic force at mass center of valve, in. lb.
C-3-69 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Valves CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS
- 1. Body Minimum Wall Thickness Minimum wall thickness in the cylindrical portions of the valve are Load: calculated using the following formula:
Design pressure and tempera- t = 1.5 [ Pd ] + C ture 2S-1.2P Primary membrane stress limit: where:
s allowable stress of 7000 psi Body Wall Thickness S = 7,000psiperASAB16.5 p primary service pressure, 655 psi t = 1.875 in.
d inside diameter of valve at section being considered, inches C corrosion allowance of 0.12 inches 1/2
- 2. Cover Minimum Thickness CP d [-+
1.78 Wha + C1 s
3
]
Sd Loads: Valve Cover Thickness where: and Stress Design pressure and tempera- t minimum thickness, inches t = 5.469 in.
ture d diameter of short span, inches Sallow = 17,800 psi Design bolting load C attachment factor Gasket load s allowable stress, psi w total, bolt load, pounds Primary stress limit: he= gasket moment arm, inches C1 = corrosion allowance, inches Allowable working stress per ASME Section VIII.
C-3-70 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS
- 3. Cover Flange Bolt Area Total, bolting loads and stresses are calculated in accordance with "Rules for Loads: Bolted Flange Connections" - ASME Boiler &
Pressure Vessel Code,Section VIII, Appendix s 30,900 psi@ 575°F Design pressure and tempera- II, except that the stem operational load and ture seismic loads are included in the total load Gasket load carried by bolts. The horizontal and vertical Stem operational load seismic forces are applied at the mass center Seismic load - DBE of the valve operator assuming that the valve body is rigid and anchored.
Bolting stress limit:
Allowable working stress per ASME Nuclear Valve and Pump Code, Class I.
- 4. Body Flange Thickness & Flange thickness and stress are calculated in Stress accordance with "Rules for Bolted Flange Connections" - ASME Boiler & Pressure Vessel Loads: Code,Section VIII, Appendix II, except that s 26,700 psi the stem operational load and seismic loads Design pressure & temperature are included in the total load carried by the Gasket load flange. The horizontal and vertical seismic Stem operational load forces are applied at the mass center of the Seismic load - DBE valve operator assuming that the valve body is rigid and anchored.
Flange Stress Limits:
_§_Hr _§_R, _§_T 1.5 Sm per ASME Nuclear Pump
& Valve Code, Class I C-3-71 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Valves (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS OR MIN. THICKNESS 3 (3 + v) PR 2
- 5. Valve Disc Thickness Sr St 8 t2 Load:
s 17,800 psi Design pressure & temperature where:
Sr radial stress, psi Primary heading stress St tangential stress limit: V Poisson's ratio p design pressure, psi Allowable working stress per R radius of disc, inches ASME Section VIII. t thickness of disc, inches
- 6. Valve Operator Supports The valve assembly is analyzed assuming that the valve body is an anchored, rigid mass and Loads: that the specified vertical and horizontal seismic forces are applied at the mass center s 18,000 psi Design pressure & temperature of the operator assembly, simultaneously with Stem operational load operating pressure plus dead weight plus Equipment dead weight operational loads. Using these loads, stresses Seismic load - SSE and deflections are determined for the operator support components.
Support Rod Stress Limit:
Allowable working stress per ASME Section VIII.
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USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Standby Liquid Control Tank OBE (E) Coefficients 0.33 g horizontal 0.07 g vertical SSE (E') Coefficients 0.66 g horizontal 0.14 g vertical The standby liquid control tank is a vertical tank under atmospheric pressure. All vertical and horizontal shell joints are double welded, full penetration butt joints. The baseplate supports the tank on its foundation. The baseplate is fastened to the foundation by means of anchor bolts.
Statement of Criteria Method of Analysis Results
- 1. The minimum thickness of the shell 1. The following formula is used in 1. Minimum Thickness 0.0154" plates is computed from the stress on calculating the minimum thickness of the vertical joints using a joint the shell plate efficiency of 0.85. The 2.6(D) (H-1) (G) corrosion allowance is added t =
to the minimum thickness to obtain fo E design thickness. In no case is where:
the nominal thickness of shell plates t minimum thickness - inches less than 3/16".
D nominal inside diameter of tank-ft The allowable stress is in H height from bottom of tank to accordance with ASME Boiler and overflow level-ft Pressure Vessel Code,Section VIII.
G specific gravity of fluid fo= Allowable design stess-psi E = Joint efficiency dimensionless
- 2. The dead weight of the tank, the 2. The dead weight of shell, liquid and 2. Allowable Stress seismic acceleration [SSE (E')] on attachments, the seismic bending Tensile 17,500 psi its center of gravity and pipe moment, and the pipe reaction forces Compressive 10,000 psi reaction on the outlet nozzle from and moments does not produce tensile maximum thermal expansion and stresses in excess of allowable or SSE (E') does not produce compressive (buckling) stresses in excessive shell stress. excess of 1/3 of yield strength.
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USAR 3.3 Method of Analysis and Implementation of Criteria 3.3.1 Reactor Pressure Vessel The Reactor Pressure Vessel (RPV) has been designed, fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code, Section III, its interpretations, and applicable requirements for Class A vessels as defined therein, 1965 Edition, using addenda up to and including the Winter 1966 Addenda.
Stress analysis requirements and load combinations for the Reactor Pressure Vessel have been evaluated for the cyclic conditions expected throughout the 40 year life, with the conclusion that ASME Code limits are satisfied.
The vessel design report contains the results of the detailed design stress analyses performed for the Reactor Pressure Vessel to meet the Code requirements. Selected components, considered to possibly have higher than Code design primary stresses as a result of rare events or a combination of rare events, have been analyzed in accordance with the requirements of the loading criteria in this appendix. Descriptions of the most critical of those analyses are included in Table C-3-7, LOADING CRITERIA. The analyses show that the limits in the criteria have been met.
3.3.1.1 Reactor Pressure Vessel Fatigue Analysis An analysis of the Reactor Pressure Vessel shows that all components are adequate for cyclic operation by the rules of Section III of the ASME Code. The critical components of the vessel are evaluated on a fatigue basis, calculating cumulative usage factors (ratios of required cycles to allowed cycles-to-failure) for all operating cycle conditions. The cumulative usage factors for the critical components of the vessel are below the Code allowable of 1.0.
3.3.1.2 Reactor Pressure Vessel Seismic Analysis A seismic analysis was performed for a coupled system consisting of the Reactor Building, the Drywell, the Reactor Pressure Vessel, and the Reactor Pressure Vessel internals. The analysis is discussed below in USAR Section C-3.3.2.3.
3.3.1.3 Reactor Pressure Vessel Support Structure The Reactor Pressure Vessel (RPV) is supported on the bottom by a steel cylindrical skirt welded to the vessel. The skirt sets on top of a ring shaped girder, which in turn is bolted to the top of the Reactor Pedestal. The RPV is also supported, near the top, by lateral stabilizers which transfer the loads through the Sacrificial Shield Wall and the Drywell shell to the Drywell Biological Shield Wall. Some of the more critical components of the RPV support structure include the RPV stabilizer, the RPV stabilizer brackets and their connection to the RPV shell, the RPV vessel support skirt, and the RPV ring girder. The results of the analysis for the Reactor Pressure Vessel support structure are documented in the calculations of record and show that the acceptance criteria described above are met.
3.3.1.4 Miscellaneous Associated Components and Supports There are several miscellaneous components and supports associated with the Reactor Pressure Vessel which are evaluated for various loading conditions. These include the Control Rod Drive (CRD) housing support, the Reactor Recirculation ( RR) pipe and pump restraints, the hydraulic control C-3-74 03/08/01
USAR unit piping, and the spent fuel storage racks. The results of the analysis for these components show that the acceptance criteria described above are met.
3.3.2 Reactor Pressure Vessel Internals Al though not mandatory at the time, the design of the Reactor Pressure Vessel (RPV) internals is in accordance with the intent of Section III of the ASME B&PV Code. The Material used for fabrication of most of the materials is solution heat treated, unstabilized Type 304 austenitic stainless steel conforming to ASTM specifications. Allowable stresses for the internals materials under normal operating conditions are taken directly from Section III. Methods of analysis, use as a guide, the design procedures of Section III. For rare events or a combination of rare events, the internals have been analyzed in accordance with the requirements of the loading criteria in this appendix. Details of the most critical analyses for RPV internal components are included in Table C-3-7, LOADING CRITERIA. The analyses show that the limits in the criteria have been met.
3.3.2.1 Internals Deformation Analysis Control Rod System If there were excessive deformation of the control rod system, made up of the control rod drive, control rod drive housing, control rod, control rod guide tube and fuel channels and the core structural elements which support them (top guide, core support and shroud and shroud support) it could possibly impede control rod insertion. The maximum loading condition that would tend to deform these long, slender components is the Safe Shutdown Earthquake. Descriptions of the analyses of the internal components, which have the highest calculated stresses, are included in the following paragraphs. The highest calculated stresses occur where the Safe Shutdown Earthquake and loads resulting from the Design Basis Accident (OBA) line break are considered to occur simultaneously. Even in these cases, the general stress levels are relatively low. No significant deformation is associated with these calculated stresses; therefore, rod insertion would not be impeded after an assumed simultaneous maximum possible earthquake and line break accident.
Core Support The core support sustains the pressure drop across the fuel. This pressure drop is the only load which causes significant deflection of the core support. Excessive core support deflection could lift the control rod guide tubes off their seats on the control rod drive housings and thereby increase core bypass leakage. This upward deflection would have to be 1/2-inch to begin to lift guide tubes. The maximum deflections under normal operation conditions and pipe rupture differential pressures for the core support are calculated to be very small as compared to 1/2-inch. The guide tubes will, therefore, not be lifted off, although even if they were, this would not be of concern because bypass leakage at this time is not important.
3.3.2.2 Internals Fatigue Analysis Fatigue analysis is performed using ASME Section III as a guide.
The method of analysis used to determine the cumulative fatigue usage is described in APED - 5460, "Design and Performance of GE-BWR Jet Pumps",
September, 1968. The most significant fatigue loading occurs in the jet pump - shroud - shroud support area of the internals. The analysis was performed for a plant where the configuration (gusset type shroud support) was almost identical to the Cooper Station. Therefore, the calculated fatigue usage is expected to be a reasonable approximation for this station.
C-3-75 03/08/01
USAR Loading Combinations and Transients Considered
- 1. Normal startup and shutdown
- 2. Operating Basis and Safe Shutdown Earthquakes
- 3. Ten minute blowdown from a stuck relief valve
- 4. HPCI operation
- 6. Improper start of a Reactor Recirculation loop The allowable cumulative fatigue usage factor is 1.0. The analysis shows that the actual usage factor is less than the allowable.
3.3.2.3 Internals Seismic Analysis The seismic loads on the Reactor Pressure Vessel (RPV) and internals are based on a dynamic analysis of the coupled model consisting of the Reactor Building, the RPV and internals as described in the succeeding paragraphs. The natural frequencies and mode shapes for the system were determined. The relative displacement, acceleration and load response of the RPV and internals were then determined using the time history method of analysis. The dynamic response was determined for each mode of interest and added algebraically for each instant of time. Resulting response time-histories were then examined, and the maximum value of displacements, accelerations, shears and moments were used for design calculations. These results were combined with the results of other loads for the various loading conditions. The loading conditions for the critical components are also presented in Table C-3-7, LOADING CRITERIA. The dynamic model of the RPV and internals is briefly described below. [l 9 J The presence of fluid and other structural components, e.g., fuel within the RPV, introduces a dynamic coupling effect. Dynamic effects of water enclosed by the RPV are accounted for by introduction of a hydrodynamic mass matrix.
The seismic model of the RPV and internals analyzed has one horizontal translation coordinate for each node point considered in the analysis. Due to the approximate coincidence of the mass center and elastic axis of the building and symmetry of the RPV and internals, one horizontal coordinate was excluded, i.e., motion in the two horizontal directions do not interact and separate analysis can be performed. The remaining translational coordinate for the various node points was the vertical coordinate. This coordinate (vertical) was excluded because the frequency content of earthquakes is such that the vertical frequencies of the RPV and internals are well above those of earthquakes. Dynamic loads due to vertical motion were added to, or subtracted from, the static loads of the components, whichever is the more conservative. The two rotational coordinates about each node point were excluded because the moment contribution of rotary inertia is negligible.
The remaining rotational coordinate has been omitted since the seismic response of the RPV is negligible.
Seismic analyses were performed by coupling the lumped mass model of the RPV and internals with the building model to determine the system natural frequencies and mode shapes. The load response of the RPV and internals was then determined by the time history method. The maximum ground acceleration time histories for the design earthquakes are described in USAR l C-3-76 03/08/01
USAR Section II-5.2.3. The seismic response was determined by uncoupling the equations of motion with a coordinate transformation, dividing the ground acceleration time history into small time increments and then finding the modal response by way of duhamels integral. Applying the coordinate transformation by the modal responses, the time history displacements acceleration and loads were determined. Finally maximum responses were isolated from the time history response results and used to verify seismic design adequacy of the Reactor Pressure Vessel and internals.
3.3.3 Piping 3.3.3.1 Piping Flexibility Analysis The piping has been analyzed for the effects of dead loads, external loads, and thermal loads. Combined bending and torsional stresses were calculated in accordance with USAS B31.l.0-1967, Power Piping, including intensification factors. Several pressure temperature cycles were evaluated and the cycle representing the worst for thermal expansion stresses was selected for the design case. Critical points were evaluated to the stress limits of USAS B31. l. 0-1967 for the standard load combination events. In addition, for events with very low probability of occurrence, stresses at critical points were compared with the limits defined in this appendix.
Fatigue analyses were also performed for Class IN piping using methods and allowable limits of ANSI B31.7-1969. The loading criteria and allowable stresses are summarized in Table C-3-7, LOADING CRITERIA.
Analysis of the plant piping systems demonstrate compliance with the above criteria. Note- the load combinations involving Pmax (peak pressure) were evaluated on a generic basis and were shown not to control in comparison to other load combinations. Also, enveloping load combinations are used in some of the calculations of record to simplify the analysis.
Certain piping systems, such as those associated with the Piping Replacement Project and the Mark I Program, are designed and analyzed to a different set of criteria and/ or are subjected to additional dynamic loads.
These other piping analyses are described in more detail in USAR Section App.C-3.3.3.5.
3.3.3.2 Piping Seismic Analysis Seismic Class IS piping systems 21/2" and greater in diameter were dynamically analyzed using the "response spectrum method" of analysis. For each of the piping systems, a mathematical model consisting of lumped masses at discrete joints connected together by weightless elastic elements was constructed. Valves were also considered as lumped masses in the pipe, and valve operators as lumped masses acting through the operator center of gravity. Where practical, a support is located on the pipe at or near each valve. Stiffness matrix and mass matrix were generated and natural periods of vibration and corresponding mode shapes were determined. Input to the dynamic analyses were the 0.5% damped acceleration response spectra for the applicable floor elevation. The increased flexibility of the curved segments of the piping systems was also considered. The results for earthquakes acting in the X and Y (vertical) directions simultaneously (combined by absolute summation), and Z and Y directions simultaneously (combined by absolute summation) were computed separately. The maximum responses of each mode are calculated and combined by the root mean square method to give the maximum quantities resulting from all modes (the response of closely spaced modes was combined by absolute summation). The response thus obtained was combined with the results produced by other loading conditions to compute the resultant stresses. For Seismic Class IS piping systems less than 21/2" in diameter, piping and supports were field routed using span and load chart tables. [661 C-3-77 03/08/01
USAR The dynamic seismic stress analyses includes both bending and torsional effects of the eccentric masses on the piping. The torsional effect of eccentric masses (e.g., valve operators, etc.) having a significant effect on the results of the analysis have been included in the mathematical model.
However, if the pipe stress due to the torsional effect is expected to be less than 500 psi (based upon hand calculations and experience), the offset moment due to the eccentric mass may be neglected. [2 oi In addition to the piping tabulated in Table C-3-7, other Class IS piping systems were analyzed for stress due to normal and upset loadings and were shown not to exceed the limits of B31. 1. 0. [22 i These piping systems were also analyzed for the following load combinations that have a very low probability of occurrence:
- 1. The sum of the primary longitudinal stresses due to pressure, dead weight and inertia effects of maximum possible earthquake must be less than 1.8 times the hot allowable stress. The probability of this load occurrence during the 40-year plant life is l0E-3 and the safety factor (SF) = 1.5.
- 2. The sum of the primary longitudinal stresses due to maximum pressure, dead weight and inertia effects of maximum probable earthquake must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 40-year plant life is l0E-2 and SF= 1.8.
- 3. The sum of the primary longitudinal stresses due to maximum pressure, dead weight and inertia effects of maximum possible earthquake must be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40-year plant life is 0.25 x l0E-3 and SF= 1.36.
3.3.3.3 Pipe Rupture Loading [231 For pipe rupture loading (R), the maximum capacity of the process pipe to deliver load was utilized. This concept defines the maximum capability of piping to transmit load, as the load caused by an ultimate bending moment (Mu), and occurs when all the piping material is at the actual yield strength of the material. Any additional pipe loading would cause increase in strain without increase in stress. The actual yield strength of the material was derived from mill test reports for the process piping in question.
In addition to the moment (Mu), rupture loading (R) also considers a force (P) applied to the nozzle. This force was assumed to be the process pipe pressure multiplied by the cross-sectional area of the pipe.
When rupture loading is applied to a containment penetration, the assembly deflects and contact is made between the flued head fitting and the limit stops at the flange. The flange, which is provided at this location, serves as a stiffening ring which transmits the ensuing rupture loading to the embedded sleeve, which in turn transmits the load to the Drywell Biological Shield Wall.
3.3.3.4 Drywell Penetration Limit Stops [23 J When required by analysis of the Drywell penetration nozzles . I and/or when required to provide positive bellows seal protection, limit stops are utilized in the vicinity of the flued head fittings. There are three different types of Drywell penetrations: I C-3-78 03/08/01
USAR Type 1 Penetrations have expansion joint bellows seals to allow thermal expansion ( refer to USAR Figure V-2-3)
Type 1 limit stops are shown typically on Burns and Roe Drawing SKM200. Note that the axial limit stops provide assurance that the bellows seals will not fail due to excessive torsional rotation, or due to axial collapse of the joint at the time of pipe rupture.
Type 2 Penetrations have no provisions for thermal expansion (refer to USAR Figure V-2-4). Limit stops are not required for Type 2 penetrations.
Type 3 Penetrations have thermal sleeves for minor thermal expansion ( refer to USAR Figure V-2-5) . Type 3 limit stops are shown typically on Burns and Roe Drawing SK101670R. No axial or torsional stops are required, and only certain Type 3 penetrations require lateral stops.
The loading combinations which control the design of the limit stops are dead load plus pipe rupture.
As shown on Burns and Roe Drawings SKM200 and SK101670R, the limit stops consist of sleeves, embedded in the Drywell Biological Shield Wall and extend to the flued head fitting outside the shield wall, coaxially with the containment penetration. In the vicinity of the flued head fitting, a flange is attached to the sleeve extension. A gap is maintained between the sleeve and the penetration assembly, so that no contact is made due to thermal differential movements.
3.3.3.5 Other Piping Analyses 3.3.3.5.1 Main Steam Piping An analysis was performed [241 to verify that the Main Steam piping within the Reactor Coolant Pressure Boundary, the relief valve discharge piping, and the associated piping suspension systems are designed to withstand the dynamic effects produced by a turbine stop valve closure and by Safety Relief Valve ( SRV) discharge [44 l.
Two of the four Main Steam lines, with their associated SRV discharge piping, were analyzed for their dynamic response to a turbine stop valve closure and SRV operation using the SAP IV Computer Program. The maximum stresses from either of these two dynamic analyses were combined with the seismic inertia stresses by SRSS and included in the standard piping load combinations as discussed in USAR Section C-3. 3. 1. These resultant stresses were then compared to the corresponding allowable stresses. The results of these analyses are documented in the calculations of record, and conclude that the pipe stresses are within the appropriate allowable stresses. Also, the analysis shows that excessive deflections, which may cause interference problems, will not occur. The effects on the SRV discharge lines during an SRV discharge transient is evaluated in the CNS Plant Unique Analysis Report (see USAR Section App.C-3.3.3.5.3 ). The turbine stop valve closure transient was not considered a critical load case for the SRV discharge lines in that analysis.
Selected portions of the MSIV leakage pathway to the condenser, as shown in Drawing CNS-MS-43, were analyzed for seismic ruggedness as part of the licensing of the LOCA dose calculations. Walkdowns and associated analyses evaluated the seismic capacity of the MSIV leakage pathway piping. The Main Steam System (including the bypass piping), the primary leakage pathway, and C-3-79 02/18/05
USAR the alternate leakage pathway were selected for detailed computer analysis using the methods of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, 2001 Edition, Appendix N. Since the consideration of a Safe Shutdown Earthquake (SSE) was not in the original design basis for the piping systems under review, the capacity criteria given below was established for use in piping system limited analytical reviews and detailed analyses:
P + . 7 5 i [ ( MA/ Z) ] < 1 . 0 S i(Mc/Z)<SA + {S-(P + .75i[MA/Z])}
P + . 7 5 i [ ( MA/ Z ) + (M81 / Z ) ] < 2 . 4 S i [ (Mc/Z) + (MEsam/Z)] < 2. 5 SA i [ (2 X MEsam) / Z) ] < 2 . 5 SA Where:
p Pressure Loadings Applied Moments due to Deadweight Loadings Applied Moments due to SSE Seismic Intertial Loadings Mssam 1/2 the range of Applied SSE Moments due to Seismic Anchor Motion Loadings Range of Applied Moments due to Thermal Expansion and Thermal Anchor Motions z Piping Section Modulus s Allowable Primary Stress limit per the B31.l Code Allowable Expansion Stress range per B31.l Code i Stress intensification factor as defined in the B31.1 Code For detailed dynamic analysis of piping in both the Turbine Building and the Reactor Building, the horizontal seismic demand is the 5%
damped specific floor response spectra which have been calculated following the guidance in NUREG-0800 Sections 3.7.1 and 3.7.2, using a Regulatory Guide 1.60 spectral shape as the input ground response spectrum.
For static analysis, the methodology incorporated into Appendix N of ASME Section III was used. For localized evaluation of piping systems containing multiple changes of directions between lateral supports which provide the same direction of restraint, the piping can be projected into single beams in each of two orthogonal horizontal directions and the vertical direction. This approach called the Equivalent Static Load Method is used in conjunction with the methodology of NEDC-31858P and Regulatory Guide 1.61. For the CNS MSIV Leakage Pathway, the equivalent static seismic ruggedness evaluations of piping systems utilized a factor of 1. 0 applied to the peak acceleration of the amplified floor response spectra.
3.3.3.5.2 SRV Discharge and Torus Attached Piping As part of the Mark I Containment Program, dynamic analyses were performed to evaluate piping which would experience hydrodynamic loads C-3-80 02/18/05
USAR associated with a postulated Loss of Coolant Accident (LOCA) event or a Safety Relief Valve (SRV) discharge. The design bases for the SRV Discharge and the torus attached piping are described in detail in the CNS Plant Unique Analysis Report. A summary level discussion is provided below. In addition, a summary of the Mark I containment program and the modifications performed is provided in USAR Section C-2.5.7.1.
Evaluations were performed for the SRV discharge piping (extending from the Main Stearn connection through the SRVs to the T-quencher assemblies in the Suppression Chamber) and all piping systems which have an attachment point on the Suppression Chamber (torus) shell. Most of the torus attached piping systems have a segment both external to, and inside the Suppression Chamber. The piping external to the Suppression Chamber was modeled from the Suppression Chamber shell connection point to a point along the pipe beyond which the load effects from the Suppression Chamber excitation are no longer significant. Components included in these evaluations are the piping and piping supports, branch lines along the piping system, and all associated equipment. The piping inside the Suppression Chamber is modeled and analyzed separately, and generally consists of short partially submerged suction and discharge piping.
The torus attached piping systems were originally modeled and analyzed using the SUPERPIPE computer program developed by EDS Nuclear. The piping systems attached to the Suppression Chamber shell were analyzed for the hydrodynarnically induced Suppression Chamber vibrations using the response spectra method. The response spectra representing the Suppression Chamber vibrations were developed with consideration of the dynamic coupling between the piping and the Suppression Chamber shell. Reanalysis for selected piping systems was performed using the ADLPIPE, ANSYS, and PISTAR computer programs.
The Mark I hydrodynamic stress results were combined with stresses due to the originally defined loads (dead weight, thermal and seismic) in accordance with the CNS Plant Unique Analysis Report. The resultant stresses were then compared to the allowable stresses for ASME Code Class 2 or Class 3
( depending on the system) piping systems per the applicable subsections of ASME Section III, S77. The piping components satisfy the applicable stress requirements of the Code. Fatigue evaluations due to thermal loading were performed to meet the applicable requirements of the Code. An augmented fatigue evaluation method for ASME Code Class 2/3 piping subjected to Mark I hydrodynamic loads was developed in MPR Report-7 51, titled, "Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus Attached on SRV Piping System," dated November 1982. This report was reviewed by the NRC and it was concluded that all CNS torus attached piping systems have a fatigue usage of less than O. 5 during the plant life. [GlJ The original Mark I program methodology used for combining dynamic loads allowed up to two dynamic loads to be combined with a Square-Root-of-the-Sum-of-Squares (SRSS) methodology. The SRSS combination included an additional conservative increase factor of 1.1. The original methodology also required that if there were more than two independent dynamic loads, the additional loads had to be summed absolutely to the (1.1) SRSS combination. An alternate methodology for combining dynamic loads for torus attached piping ( for pipe stress combinations, pipe support load combinations, Suppression Chamber penetration piping reactions, and equipment loads) which removes the 1.1 factor from the SRSS dynamic load combination and allows all dynamic loads to be combined via SRSS is used on subsequent analyses. This alternate methodology is acceptable since it meets the same criteria (i.e., meeting an 8 4 % non-exceedance probability with more than a 90% confidence level) that was used to justify the original 1.1 SRSS method as described in Appendix D of the CNS Plant Unique Analysis Report. This C-3-81 01/26/07
USAR combination method was submitted to the NRC by the BWR Owners Group and was reviewed and found acceptable by the NRC. [35 , 36 l 3.3.3.5.3 Replacement Piping Portions of the Reactor Recirculation, Core Spray, Reactor Water Cleanup, and Residual Heat Removal systems were replaced in 1985 with piping made of new material and, in some cases, configuration changes were made.
Flexibility and seismic analyses of these systems have been performed which demonstrate compliance with the requirements of 1983 Edition of the ASME Boiler and Pressure Vessel Code. The load combinations and allowable stresses are provided in Impell Corp. Report No. 01-0840-1268[ 34 l_
3.3.3.5.4 Intake Structure Piping In addition to the standard piping loads, piping systems attached to the Intake Structure are also subjected to a dynamic transient associated with a barge impact with the Intake Structure. A response spectrum for this transient has been developed for the operating floor of the Intake Structure.
Piping attached to this structure is analyzed using this response spectrum.
This load is considered similar in probability to an SSE. Therefore for piping subjected to this transient, the maximum stresses due to the inertia effects of an SSE or the barge impact load is substituted for SSE in the applicable load combinations ( see USAR Section C-3. 3. 3. 1) The allowable stress is the same as for load combinations involving SSE.
3.3.4 Equipment The extent of stress analyses performed on equipment is dependent upon the type of equipment and the type of fabrication. Fabricated shapes are generally made from plate or rolled shapes with uniform thickness and shapes with regular geometric configurations. Cast shapes are generally made with non-uniform material thickness in complicated shapes that are not regular geometric configurations. Manufacturers have traditionally designed cast shapes conservatively since they do not lend themselves to rational analysis.
Typically a design is developed based on extensive test and experience.
For the seismic design of the equipment and components, the analytical results from both the horizontal and vertical earthquake loadings are considered to act concurrently and in the most disadvantageous directions.
The earthquake loads in the horizontal direction were obtained from the appropriate amplified floor response spectra plotted for the supporting structure. In lieu of determining the natural frequency of the equipment, the peak value of the applicable floor response spectrum was used in calculating the earthquake induced loads. Alternately, the natural frequency of the equipment was determined and corresponding input acceleration was obtained from the appropriate floor response spectra. Where equipment is represented by multi-degree of freedom systems, dynamic analyses which accounts for the contribution of all the significant modes of vibration are used. [27 l In the vertical direction, the buildings are considered to be very stiff as compared to the horizontal direction. The equipment and components (pumps, motors, tanks, valves, etc.) are also considered to be very stiff in the vertical direction. These factors tend to keep the vertical earthquake loads quite low. Since this equipment is designed to withstand vertical loads equal to its weight (lg) the effect of the vertical earthquake loads is, therefore, not as significant in the overall design of the equipment. The original vertical seismic acceleration component was specified as one-half of the horizontal ground acceleration. This design requirement was increased to two-thirds of the horizontal ground acceleration, which is the current Licensing/Design Basis. [2 si C-3-82 02118/05 I
USAR The original criteria, method of analysis and summary of critical stresses for various equipment are included in Table C-3-7, LOADING CRITERIA.
Dynamic testing is accepted as an alternate procedure to prove seismic adequacy of Class I equipment in cases where analytical procedures can not be used to evaluate equipment adequacy. r261 Use of the earthquake experience-base d method, as described in the Seismic Qualification Utility Group (SQUG) Generic Implementation Procedure (GIP) was accepted for seismic qualification of certain electrical and mechanical equipment in the scope of the GIP. r68 l For new and replacement electrical equipment, either IEEE 344-1971 or IEEE 344-1975 "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Genera ting Stations" is acceptable for seismic qualification of Class lE electrical equipment. CNS is designed to the IEEE 344-1971 standard, however the IEEE 344-1975 standard provides additional guidance and is commonly used in industry.
Depending on the design objective, the following two types of tests can be used in the dynamic testing of equipment. r261
- 1) Free Vibration Test This test is done on equipment whose response is dominated by the fundamental mode. The critical damping ratio and fundamental frequency can be determined from this test. In this test, an initial displacement or initial velocity is imparted to the equipment. The initial displacement is introduced by forcibly displacing the equipment and then suddenly releasing the force.
The initial velocity is obtained by impinging the equipment with an impulse.
Accelerometers or strain gauges are mounted on the equipment. Several readings are made first to assure that the equipment is vibrated in its primary mode.
The critical damping ratio is then calculated by the standard formula for determining damping from logarithmic decrement.
- 2) Forced Vibration Test The equipment is mounted on a shake table to which a prescribed acceleration is applied. By this test, both the critical damping ratio and the equipment's functional capability can be determined. The test can be performed also by means of eccentric shakers. In order to obtain the critical damping ratio of the equipment, a sinusoidal acceleration is applied by the shake table or eccentric shaker. The forced response curve (maximum amplitude vs.
forcing frequency curve) of the equipment is obtained first. The critical damping ratio is then calculated by using the half-power method of fitting a theoretical forced response curve through the data points at and near resonance.
In order to prove the equipment's functional capability, a prescribed acceleration, which can be sinusoidal, sine beat, random or shock type is applied at the shake table. The vibratory motion used is such that vibratory loads are equal to or more than the seismic loads represented by the applicable floor spectra. All loads normally acting on the equipment are also simulated. When seismic loading on mechanical equipment are considered analytically, the resultant stresses are added to those from normal and accident conditions in appropriate loading combinations to assure that the equipment will function as specified.
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USAR When dynamic testing is used to verify the seismic adequacy of Class I mechanical equipment, the equipment is operated during and after the test to ensure operability.
When testing is used to supplement analysis, the equipment is not in the operating mode but, the dynamic loads computed are added to those from normal and accident conditions in appropriate loading combinations to assure that the equipment will function as specified r261
- The components satisfy the applicable requirements of the ASME and ANSI Codes.
Revision 3 of the Generic Implementation Procedure (GIP-3) r 69 l, as modified and supplemented by the U.S. Nuclear Regulatory Commission Supplemental Safety Evaluation Report ( SSER) No. 2 r 7 oi and SSER No. 3 [7 ll, may be used as an alternative to other authorized methods for the seismic design and verification of modified, new and replacement equipment classified as Class I.
Only those portions of GIP-3 listed below, which apply to the seismic design and verification of mechanical and electrical equipment, electrical relays, tanks and heat exchangers, and cable and conduit raceway systems shall be used. The other portions of the GIP are not applicable since they contain administrative, licensing, and documentation information which is applicable to the USI A-46 program.
Part I, Section 2. 3. 4, Future Modifications and New and Replacement Equipment.
Part II, Sections 2.1.2 and 2.4, Seismic Capability Engineers.
Part II, Section 4, Screening Verification and Walkdown.
The following differences shall be used when applying this section of GIP-3 for modified, new, and replacement equipment.
Use of Method A for determining the seismic demand is directly applicable to equipment mounted in those buildings and at those elevations which are at or below the locations where safe shutdown equipment was mounted and evaluated as a part of the USI A-4 6 programl 72 l and accepted by the NRC in its review of that program. [681 Method A may be used in other buildings and elevations, but with appropriate justification consistent with the NRC' s acceptance of Method A for USI A-4 6 at CNS [681 and guidance provided to SQUG member utilities. r73 l For new anchorage installations and where previous bolt sizes and patterns are not used, the allowable anchorage capacities should be based on factors of safety specified in this Appendix (or, if not specified herein, as recommended by the anchorage manufacturer), instead of the GIP-3 factor of safety of 3.0.
Documentation of the results of the Screening Verification and Walkdown in Section 4. 6 may be limited to the use of walkdown checklists. It is not necessary to complete Screening Verification Data Sheets (SVDSs).
Part II, Section 6. 4, Comparison of Relay Seismic Capacity to Seismic Demand. It is not necessary to identify "essential relays" and perform functionality screening as defined in Section 6 of GIP-3. Only the seismic capacity compared to seismic demand evaluation in the GIP will be applied to relays designated as Class I.
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USAR Part II, Section 6.5, Relay Walkdown.
Part II, Section 7, Tanks and Heat Exchangers Review.
This section of GIP-3 may be used in its entirety for replacement of existing tanks and heat exchangers, as well as for the design and construction of new tanks and heat exchangers, except for new flat-bottom vertical tanks. For new flat-bottom vertical tanks, the following attributes, in addition to appropriate GIP-3 criteria shall be used:
The cast-in-place anchor bolts and associated hardware (chairs, transfer plates, etc.) in Subsection 7. 3. 3 shall be designed and installed in accordance with embedment depth, edge distance, anticipated concrete cracking, and corrosion allowance specified in GIP-3. The maximum strain in the anchor bolts shall not exceed that corresponding to the yield strength of the bolt material.
In Step 16 of Subsection 7.3.3.3, change the equation to:
CTc = 0. 6 [min. (ape , CTpctl] [psi]
In Subsection 7.3.7, the tank foundation (ring-type or otherwise) shall be designed to resist uplifting and the overturning moment.
Part II, Section 8, Cable and Conduit Raceway Review.
The additional evaluations and alternate methods for resolving raceway outliers in Sections 8.4.1 through 8.4.8 may be used. However, the generic methods for resolving outliers in Part II, Section 5 shall not be used.
Part II, Section 10, References.
Appendix B, Summary of Equipment Class Descriptions and Caveats.
Appendix C, Generic Equipment Characteristics for Anchorage Evaluations.
Appendix D, Seismic Interaction.
Appendix G, Screening Evaluation Work Sheets (SEWS).
Notes:
( 1) The GIP method should be used only for equipment located in mild environments and when concurrent vibratory loads, e.g., hydrodynamic loads from DBA, in combination with SSE loads are less than the GIP Reference Spectrum.
(2) The GIP method should not be used on equipment for which seismic qualifications have been imposed or committed to in connection with the resolution of other specific issues (e.g., Regulatory Guide 1.97, Three Mile Island (TMI) Action Item II.F.2, IPEEE) unless justified on a case-by-case basis.
Specific details of the criteria and method of analysis for various plant equipment is provided below:
Core Spray, RHR, HPCI, and RCIC Pumps and HPCI and RCIC Turbines The Core Spray, RHR, HPCI and RCIC pumps and the HPCI and RCIC turbines are evaluated for the expected loading conditions. The C-3-85 10/02/14
USAR structura l design criteria and method of analysis for these pumps and turbines are described in Table C-2-7. The Plant Unique Analysis Report describes additiona l analyses performed on this equipment . The results of the analysis for these pumps and turbines show that the applicabl e acceptanc e criteria have been met.
Reactor Recircula tion Pumps The Reactor Recircula tion pumps are evaluated for the expected loading condition s associate d with both normal operating and seismic condition s. The structura l design criteria and method of analysis for the key component s of these pumps are described in Table C-2-7. The results of the analysis for the Reactor Recircula tion pumps show that the applicabl e acceptanc e criteria have been met.
Standby Liquid Control Tank and Pumps The Standby Liquid Control tank and pumps are evaluated for the expected loading condition s. The structura l design criteria and method of analysis for the key component s of these pumps are described in Table C-2-7.
The results of the analysis for the Standby Liquid Control Pumps show that the applicabl e acceptanc e criteria have been met.
Main Steam Isolation Valves, Safety /Relief Valves, and Reactor Recircula tion Valves The Main Steam Isolation Valves, Safety /Relief Valves, and the Reactor Recircula tion Valves are evaluated for the expected loading condition s associate d with both normal operating and seismic condition s. The results of the analysis for these valves show that the applicabl e acceptanc e criteria have been met.
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4.0 REFERENCES
FOR APPENDIX C
- 1. Q/A 12.37; Amendment 9.
- 2. "Design of Concrete Structures," by Winter, Urguhart, O'Rourke, Nilson, 7th Edition, McGraw- Hill Book Company, New York, 1964.
- 3. "Concrete Manual," U.S. Department of the Interior, Bureau of Reclamation, 7th Edition, U.S. Government Printing Office, 1966, page 48.
- 4. Q/A 12.8; Amendment 9.
- 5. Q/A 12.10; Amendment 9.
- 6. Q/A 12.5; Amendment 9.
- 7. Q/A 12.51; Amendment 13.
- 8. Q/A 12.7; Amendment 9.
- 9. Deleted.
- 10. Deleted.
- 11. Deleted.
- 12. Q/A 12.17; Amendment 9.
- 13. Q/A 12.12; Amendment 9.
- 14. Q/A 12.18; Amendment 9.
- 15. Q/A 12.19; Amendment 9.
- 16. Q/A 12.20; Amendment 9.
- 17. Q/A 12.26; Amendment 9.
- 18. Q/A 12.50; Amendment 13.
- 19. Q/A 12.60; Amendment 13.
- 20. Q/A 12.59; Amendment 13.
- 21. Q/A 12.28; Amendment 13.
- 22. Q/A 12.52; Amendment 14.
- 23. Q/A 12.46; Amendment 13.
- 24. General Electric Report, "Dynamic Analysis of the Effects of Turbine Stop Valve Closure and Relief Valve Discharge on CNS Main Steam Lines," forwarded to NRC by Letter J. Pilant (NPPD)/J. Stolz (NRC) dated November 23, 1973, (17556 0527).
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- 25. CNS FSAR Amendments 20 and 25, forwarded by Letters R. Reder (NPPD)/L. M. Muntzing (NRC) dated April 6, 1973, and June 8, 19 7 3 .
- 26. Q/A 12.61; Amendment 14.
- 27. Q/A 12.38; Amendment 9.
- 28. Q/A 12.40; Amendment 9.
2 9. Deleted.
- 30. Q/A 12.45; Amendment 13.
- 31. Deleted.
- 32. "Safety Evaluation Report Mark I Containment Long-Term Program (Resolution of Generic Technical Activity A-7),"
NUREG-0661, U.S. Nuclear Regulatory Commission, July, 1980.
- 33. "Mark I Containment Program Structural Acceptance Criteria, Plant Unique Analysis Application Guide," NEDO-24583-1, Revision 1, October, 1979.
- 34. Impell Corp. Report No. 01-0840-1268, "Stress Analysis of Reactor Recirculation, Core Spray, and Reactor Water Cleanup Drywell Systems for Cooper Nuclear Station."
- 35. "Mark I Containment Program Cumulative Distribution Functions for Typical Dynamic Responses of a Mark I Torus and Attached Piping Systems," NEDE-24632, Revision 0, December 1980.
- 36. "Safety Evaluation by the Office of Nuclear Reactor Regulation for Acceptability of the SRSS Method for Combining Dynamic Responses in Mark I Piping Systems," forwarded by NRC letter D. Vassallo (NRC)/H. Pfefferlen (GE) dated March 10, 1983.
- 37. Deleted.
- 38. Deleted.
- 39. EDS Calculation 0840-002-5.0, "CNS Environmental Analysis Final Results."
- 41. NPPD Drawing CNS-BLDG-374, "Steam Tunnel Blowout Panels."
- 42. NPPD Calculation NEDC 95-205, "Review of VECTRA calculation FEMBOP-1, Finite Element Analysis of Steam Tunnel Blowout Panels."
- 43. Deleted.
- 44. Q/A 4.19; Amendment 30.
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- 45. Q/A 12.9; Amendment 9.
- 46. NPPD Calculation NEDC 87-140, "Anchor Bolt Load Cale. For 5000 psi Concrete."
- 47. Short-Term Program, Mark I Evaluation, Plant Unique Analysis, forwarded by letter from J. Pilant (NPPD) to V. Stello (NRC) dated 7-30-76 (17512 0924).
- 48. Kaiser Engineers Report, "Torus Support System Modifications," No. 77-14-R, March, 1977 (17512 0014).
4 9. Q/A 5. 12; Amend. 9.
- 50. Q/A 5.16; Amend. 14.
- 51. Q/A 5.20; Amend. 17.
- 52. U.S. Nuclear Regulatory Commission Letter (D. L. Ziemann) to NPPD (J. M. Pilant), February 14, 1975.
- 53. U.S. Nuclear Regulatory Commission Letter (D. L. Ziemann) to NPPD (J.M. Pilant), April 19, 1975.
- 54. "Mark I Containment Program Load Definition Report,"
NEDO-21888, General Electric Company, Revision 2, November, 1981.
- 55. U.S. Nuclear Regulatory Commission, "NRC Acceptance Criteria for the Mark I Containment Long Term Program,"
Revision 1, February, 1980.
- 56. "Evaluation of Mark I S/RV Load Cases C3.l, C3.2, and C3.3 for the Cooper Nuclear Station," NEDC-24359, General Electric Company, August, 1982.
- 58. Continuum Dynamics, Inc., Technical Note No. 92-28, Rev. 0, dated December 1992.
- 59. NPPD Calculation NEDC 93-007, "Torus to Drywell Vacuum Breaker Structural Evaluation for Various Setpoints."
- 60. NRC Safety Evaluation Report ( SER) for CNS Technical Specification Amendment 91, dated May 13, 1985.
- 61. NRC Safety Evaluation Report Related to the Mark I Containment Long-Term Program, Structural Review, NPPD Docket 50-298," dated January 20, 1984.
- 62. Q/A 12.31; Amend. 14.
- 63. Letter from Jay M. Pilant (NPPD) to Karl V. Seyfri t (NRC),
Subject "IE Bulletin No. 79-02 Pipe Support Base Plate C-4-3 03/08/01
USAR Designs Using Concrete Expansion Anchor Bolts," dated February 5, 1980.
- 64. Letter from Jay M. Pilant (NPPD) to Karl V. Seyfrit (NRC),
Subject "IE Bulletin No. 79-02 (Revision No. 1) Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts," dated July 6, 1979.
- 65. Letter from Jay M. Pilant (NPPD) to Karl V. Seyfrit (NRC),
Subject "IE Bulletin No. 79-02 (Revision No. 2) Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts," dated December 10, 1979.
- 66. Letter from Jay M. Pilant (NPPD) to Karl V. Seyfrit (NRC),
Subject "IE Bulletin No. 79-07 Seismic Stress Analysis of Safety Related Piping," dated June 29, 1979.
- 67. NEDC 01-035, Evaluation of Class I Structures for Change in Design Basis Vertical Seismic Factor from 1/2-horizontal g to 2/3.
- 68. U.S. NRC Letter to NPPD dated September 30, 1999, Plant-Specific Safety Evaluation Report for USI A-46 Program Implementation at Cooper Nuclear Station (TAC No. M69439).
- 69. Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 3, Updated 05/16/97 (GIP-3). Prepared by SQUG and sent to the NRC by letter dated May 16, 1997.
- 70. Supplement No. 1 to Generic Letter (GL) 87-02 that Transmits Supplemental Safety Evaluation Report No. 2 (SSER No. 2) on SQUG Generic Implementation Procedure Revision 2, as Corrected on February 14, 1992 (GIP-2), May 22, 1992.
- 71. NRC Letter to SQUG dated December 4, 1997, Supplemental Safety Evaluation Report No. 3 (SSER No. 3), on the Review of Revision 3 to the Generic Implementation Procedure for Seismic Verification of Nuclear Power Plant Equipment, Updated 05/16/97 (GIP-3).
7 2. NPPD Letter to U.S. NRC Document Control Desk, "Submittal of the Unresolved Safety Issue ( USI) A-4 6 Summary Report, Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46," June 13, 1996, and Attachment 2, "Cooper Nuclear Station, USI A-46 Seismic Evaluation Report" May 1996.
- 73. NRC Letter to SQUG dated 6/23/99, Review of Seismic Qualification Utility Group's Report on the Use of the Generic Implementation Procedure for New and Replacement Equipment and Parts.
- 74. Engineering Evaluation 10-024, "Reactor Building Up-Rate Evaluation."
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