ML23129A268
| ML23129A268 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/20/2023 |
| From: | Nebraska Public Power District (NPPD) |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23129A261 | List:
|
| References | |
| NLS2023022 | |
| Download: ML23129A268 (1) | |
Text
1.0 2.0 3.0 SCOPE USAR APPENDIX C STRUCTURAL LOADING CRITERIA CONCRETE AND STEEL STRUCTURES 2.1 Description 2.2 Loads 2.3 2.4 2.5 Load Combinations and Allowable Limits Method of Analysis Implementation of Criteria 2.5.1 Reactor Building Foundation 2.5.2 Reactor Building Floor System 2.5.3 Reactor Building Concrete Walls 2.5.4 Reactor Building Steel Superstructure 2.5.5 Drywell Biological Shield Wall 2.5.6 Reactor Pedestal 2.5.7 2.5.8 2.5.9 2.5.10
- 2. 5.11 2.5.12 Primary Containment 2.5.7.1 Mark I Containment Program 2.5.7.1.1 Program History 2.5.7.1.2 Short-Term Program 2.5.7.1.3 Long-Term Program 2.5.7.1.4 Loads
- 2. 5. 7. 2 2.5.7.1.5 Load Combination & Allowable Limits 2.5.7.1.6 2.5.7.1.7 Pipe Whip 2.5.7.2.1 2.5.7.2.2 2.5.7.2.3 2.5.7.2.4 Method of Analysis Modifications Completed and Jet Force Analysis General Jet Force Analysis Pipe Whip Analysis Pipe Whip Energy Absorbing Material Main Steam Tunnel Sacrificial Shield Wall Concrete Attachments Control Building Elevated Release Point COMPONENTS 3.1 Intent and Scope 3.1.1 Components Designed By Rational Stress Analysis 3.1.2 Components Designed Primarily By Empirical Methods 3.2 Loading Conditions and Allowable Limits 3.3 3.2.1 Loading Conditions 3.2.1.1 Normal Conditions 3.2.1.2 Upset Conditions 3.2.1.3 Emergency Conditions 3.2.1.4 Faulted Conditions 3.2.2 Allowable Limits Method of Analysis and Implementation of Criteria 3.3.1 Reactor Pressure Vessel 3.3.1.1 Reactor Pressure Vessel Fatigue Analysis 3.3.1.2 Reactor Pressure Vessel Seismic Analysis c-1-1 PAGE C-1-1 C-2-1 C-2-1 C-2-1 C-2-2 C-2-5 C-2-6 C-2-6 C-2-6 C-2-6 C-2-10 C-2-13 C-2-15 C-2-18 C-2-24 C-2-24 C-2-25 C-2-25 C-2-26 C-2-26 C-2-27 C-2-27 C-2-29 C-2-29 C-2-29 C-2-30 C-2-33 C-2-34 C-2-35 C-2-36 C-2-42 C-2-42 C-3-1 C-3-1 C-3-1 C-3-1 C-3-1 C-3-1 C-3-2 C-3-2 C-3-2 C-3-2 C-3-2 C-3-74 C-3-74 C-3-74 C-3-74 01/02/15
4.0 3.3.2 3.3.3 3.3.4 USAR APPENDIX C STRUCTURAL LOADING CRITERIA (CONT'D) 3.3.1.3 3.3.1.4 Reactor 3.3.2.1 3.3.2.2 3.3.2.3 Piping 3.3.3.1 3.3.3.2 3.3.3.3 3.3.3.4 3.3.3.5 Reactor Pressure Vessel Support Structure Miscellaneous Associated Components and Supports Pressure Vessel Internals Internals Deformation Analysis Internals Fatigue Analysis Internals Seismic Analysis Piping Flexibility Analysis Piping Seismic Analysis Pipe Rupture Loading Drywell Penetration Limit Stops Other Piping Analyses 3.3.3.5.1 Main Steam Piping 3.3.3.5.2 SRV Discharge and Torus Attached Piping 3.3.3.5.3 Replacement Piping 3.3.3.5.4 Intake Structure Piping Equipment REFERENCES FOR APPENDIX C c-1-2 PAGE C-3-74 C-3-74 C-3-75 C-3-75 C-3-75 C-3-76 C-3-77 C-3-77 C-3-77 C-3-78 C-3-78 C-3-79 C-3-79 C-3-80 C-3-82 C-3-82 C-3-82 C-4-1 02/18/05
Figure No.
C-2-1 C-2-2 C-2-12 USAR LIST OF FIGURES (At End of Appendix C)
Title Reactor Bldg. Dynamic Analysis Model Control Bldg. Dynamic Model Reactor Bldg. Total Acceleration c-2-1 03/08/01
Table No.
C-2-1 C-2-2 C-2-3 C-2-4 C-2-5 C-2-6 C-2-7 C-2-8 C-2-9 C-2-10 C-2-11 C-2-12 C-2-13 C-2-14 C-2-15 C-2-16 C-3-1 C-3-2 C-3-3 C-3-4 C-3-5 C-3-6 C-3-7 USAR LIST OF TABLES Title Reactor Building Foundation Reactor Building Floor System Reactor Building Concrete Walls Reactor Building Structural Steel Columns Reactor Building Steel Roof Truss-Frame Above Drywell Shielding Concrete Reactor Concrete Pedestal Drywell Membrane Stresses Jet Impingement Force Stresses Drywell Stabilizer Shear Lugs El. 117' Summary of Mark I Containment and Piping Modifications Original Tabulation of Critical Pipe Stresses Due to Dead Weight Plus Pressure, Plus Operating Basis Earthquake, Plus Thermal Loads Control Building Foundation Control Building Floor System Control Building Concrete Walls Elevated Release Point Foundations and Tower Loading Condition Probabilities Deformation Limit Primary Stress Limit Buckling Stability Limit Fatigue Limit Minimum Safety Factors Loading Criteria Reactor Vessel Internals and Associated Equipment Main Steam Piping Recirculation Loop Piping Class IN/IS -
Core Spray Discharge Piping CS-ID Class IN/IS -
Clean-Up Recirculation Pump Suction Piping CU-IS Class IN/IS -
Main Steam Piping to HPCI Turbine and Residual Heat Exchangers lA and lB Class IN/IS - Reactor Feed Piping Class IN/IS - Residual Heat Removal Pump Suction Piping Class IN/IS - Residual Heat Removal Pump Discharge Piping c-3-1 Page C-2-7 C-2-8 C-2-9 C-2-11 C-2-12 C-2-14 C-2-16 C-2-19 C-2-20 C-2-22 C-2-28 C-2-32 C-2-39 C-2-40 C-2-41 C-2-43 C-3-4 C-3-5 C-3-6 C-3-8 C-3-9 C-3-10 C-3-11 C-3-11 C-3-24 C-3-26 C-3-27 C-3-28 C-3-29 C-3-31 C-3-33 C-3-35 01/02/15
Table No.
USAR LIST OF TABLES Title Recirculation Pumps RHR Pump Core Spray Pump HPCI Pump RCIC Pump Standby Liquid Control Pump HPCI Turbine RCIC Turbine Main Steam Isolation Valves Main Steam Safety Valves Main Steam Relief Valves Recirculation Valves Standby Liquid Control Tank c-3-2 Page C-3-37 C-3-41 C-3-43 C-3-45 C-3-47 C-3-49 C-3-51 C-3-55 C-3-59 C-3-62 C-3-67 C-3-70 C-3-73 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.
C-1-1 08/08/01
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 R
R' E
E' 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.
Loads resulting from jet forces and pressure and temperature transients associated with rupture of a single pipe within the primary containment.
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. [25l 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.
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
Flood w
T USAR for vertical seismic acceleration).
also known as Hypothetical Maximum Earthquake loads.
These loads are Possible Design Loads due to flooding the Drywell up to 23 feet above the Drywell flange level.
Design wind loading Section XII-2.3.3.1.
conditions.
Refer to USAR 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 D+E D+W D+R+E D+E+Flood Limits Stresses remain within normal code allowable stresses (AISC 6th Edition for structural steel, ACI 318-63 for reinforced
[Class BJ, S67 for Primary Containment). The customary increase in design stress for earthquake loadings is not permitted.
Maximum allowable stresses may be increased one-third above normal code allowable stresses.
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.
Local exceed membrane stresses the yield point, rupture.
C-2-2 in Primary but with Containment may a
margin against 03/08/01
Load Combination D+T D+R+E' D+R'+E' USAR Limits Maximum allowable stresses are as follows:
Steel -
AISC 6th 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.
Maximum allowable stresses are as follows:
Steel -
AISC 6th 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.
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 requirements for the design and the governing codes and supplementary 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 Institute of Steel Construction Specification for and Erection of Structural Steel for Buildings, (referred to herein as AISC 6th Edition).
code is the American the Design, Fabrication, adopted April 17, 1963 For concrete structures, the Concrete Institute Standard ACI 318-63, Reinforced
- Concrete, as amended March 6, ACI 318-63).
governing code is the American Building Code Requirements for 1963 (referred to herein as 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. £4l}
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 buildings has a
minimum 28 day strength of 3,000 psi:
Reactor superstructure, Reactor Pedestal, Control Building superstructure, Building superstructure, superstructure of the Intake Structure, all the Diesel Generator Building and Control Corridor.
parts of Building Radwaste parts of 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. E 45 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.
C-2-4 01/02/15
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 1c (2,550 psi).
In
- addition, concrete gains additional strength with age.
Figure 22 from the "Concrete Manual" [3l 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.[ 4l 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 6th 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-dimensional 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/Design 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 2.5.1 Implementation of Criteria 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 therefore, the floor systems are deflections or cracking of concrete.
are within the Code structurally adequate allowable values, without excessive 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.
C-2-6 03/08/01
Description/Criteria Design load (D) includes all dead and equipment loads plus 25% of the full live loads.
Vertical seismic load (~P) is assumed to be produced by a vertical acceleration equal to 2/3 of the ground horizontal seismic acceleration, both acting simultaneously**
Overturning moment (M) is produced by the horizontal seismic acceleration and all design load eccentricities USAR TABLE C-2-1 REACTOR BUILDING FOUNDATION Method of Analysis Max. net soil stress D + ~p
+
M (KSF) -
A s
(aDa+ Ysat+Db) where:
D Design load, kips
~p vert seismic load, kips M
horiz. seismic overturning moment kip-ft A
foundation mat area, sq-ft s
section modulus of foundation mat ft 3 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 Load Combination*
D+E D+E' Max.
Allowable Stress-ks£
- 1. Sqa = 18 qa net avg.
contact pressure
- 1. Sqa 18 E
E' Loads due to OBE Loads due to SSE 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.
C-2-7 03/08/01
Description/Criteria Floor system is designed for the loadings tabulated in Subsection XII-2.3.2.
Materials conform as follows:
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; 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 USAR TABLE C-2-2 REACTOR BUILDING FLOOR SYSTEM Method of Analysis Working Stress Design Method Same as for D + E' Load Combination*
D + E D + E' Max.
Allowable Stress-ksi Fe Fv Ft Fe Fv Ft
- 1. 35 0.060 20.0 2.55 0.093 36.0 Maximum allowable Stresses for D + E combination are not increased above code allowable values, customary 1/3 increase not used.
i.e.,
- Vertical= 1/2 horizontal per PSAR; 2/3 was used for conservatism.
C-2-8 03/08/01
Description/Criteria Design Load (D) includes all dead and equipment loads, plus 25% of the full live loads.
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+E' load Concrete Fe = 0. 8Sf 1 e 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.
USAR TABLE C-2-3 REACTOR BUILDING CONCRETE WALLS Method of Analysis Working Stress Design of ACI 318-63 Working Stress Design Method of ACI 318-63 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 Load Combination D + E D + E' Max.
Allowable Stress-ksi Ft Fe Fy Ft Fe Fv 20.0
- 1. 35 0.060 36.0 2.55 0.093 03/08/01
USAR The resulting stresses are within the Code allowable therefore, the Reactor Building walls are structurally adequate excessive deflections or cracking of concrete.
- values, without 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 C-2-5)
(See Tables C-2-4 and 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). r121 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
Description/Criteria Material: ASTM Designation:
A36 per AISC Manual, 1963 Structural Steel Column designed for all dead and live loads as follows:
Wheel reactions from 108 ton crane in combination with roof DL plus 20 psf LL plus wind@ 45 psf on building Vertical seismic load is assumed to be produced by a vertical acceleration equal to1/2 of horizontal ground acceleration plus normal DL partial LL crane DL Tornado load - building siding remains intact until 75 psf loading is reached at which time it is considered removed and full tornado load is applied to the structural steel members USAR TABLE C-2-4 REACTOR BUILDING STRUCTURAL STEEL COLUMNS Method of Analysis Working Stress Design Method considering unsymmetrical beam loading where applicable Same as Above C-2-11 Load Combination D + LL
+ Crane Load DL + Tornado on structural steel only l
Max.
Allowable Stress-ksi FA= 18.3 per AISC FB = 22.0 (Upper section crane columns)
FA= 17.4 FB=22.0 (Lower section crane columns)
FA=24.3 FB = 29.2 (Upper section crane columns)
FA= 23.2 FB = 29.2 (Lower section crane columns)
FA= 17.4 FB = 22.0 (Lower section crane columns)
FA = 27 FB=32.4 (Upper section crane columns)
FA = 27 FB = 32.4 (Lower section crane columns) 11/22/10
USAR TABLE C-2-5 REACTOR BUILDING STEEL ROOF TRUSS-FRAME ABOVE EL. 117' Description/Criteria Material: Structural steel ASTM Designation: A36 per AISC Manual, 1963 Design wind loading: 45 psf on walls due to 100 mph wind:
uplift 34 psf on roof due to 100 mph wind: dead weight of roof deck and built-up roofing 8 psf Tornado wind load (T) on walls averages a pressure of 270 psf on exposed surfaces after siding has blown off at
+/- 75 psf 108 ton capacity crane on column bracket E' load disregarded because tornado is governing at similar allowable stresses of
- 0. 9 Fy Method of Analysis Working Stress Design Load Combination*
D + E D + T D + E D + T D + E D + T D + E D + T D + E D + T Max.
Allowable Stress-ksi FA = 1 7. 62 comp.
(Top chord)
FA = 2 6. 6 comp.
(Top chord)
FA= 22.0 ten.
(Bottom chord)
FA= 32.4 ten.
(Bottom chord)
FA= 22.0 ten.
(Diagonal member)
FA= 32.4 ten.
(Diagonal member)
FA = 15. 4 comp.
(Truss vertical)
FA = 22. 6 comp.
(Truss vertical)
FA= 22.0 ten.
(Lower chord bracing)
FA= 32.4 ten.
(Lower chord bracing)
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 1 c) 0 *5
- The 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
Description/Criteria Drywell shield acts as a structural wall carrying floors, Design load (D) consists of all dead loads, equipment loads, plus 25% of full live load Seismic loads (E&E') are according to the response spectra for the reactor building Accident load (R) includes 60°F normal operational 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 1 e Reinforcing Ft= 0.90 fy USAR TABLE C-2-6 DRYWELL SHIELDING CONCRETE Method of Analysis Working Stress Design. Thermal stresses are included in the analysis. Seismic forces are super-imposed on the results.
C-2-14 Load Combination D + E + thermal gradient D + E' + R (Jet due to any pipe break within the spherical area of the drywell section+ thermal gradient)
Ft Fe Fv Ft Fe Fv Max.
Allowable Stress-ksi 20
- 1. 35 0.060 36.0 2.55 0.113 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
Description/Criteria The Reactor Vessel Pedestal consists of a 3'-3" thick x 26'-7" high cylindrical wall rising from a concrete base, flaring out to 5'-10 3/4" thick at the top. The base has an average thickness of 7' at wall and a spherical shaped bottom matching the sphere of drywell. The shears and moments are transferred to the drywell 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 1 e 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 (where noted) due to ruptured pipe and temperature gradient in concrete USAR TABLE C-2-7 REACTOR CONCRETE PEDESTAL Method of Analysis Working Stress Design Method. For seismic loads response spectra are used. Circumferential stresses due to temperature are in accordance with ACI 307.
C-2-16 Load Combination D + R + E D + R + E (Horiz. Hoop stress)
D + R + E' + Jet at pump restraint D + R + E' + Jet (at reactor nozzle)
Max.
Allowable Stress-ksi Fe Ft Fe Ft Fe Ft Fe Ft Fv
- 1. 35 20.0
- 1. 35 20.0 2.55 36.0 2.55 36.0 0.16 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 1 c) 0
- 5
- The concrete takes most of this shear (60 psi) 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.
2.5.7 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.
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
Description/Criteria The vessel is bulb shaped and houses the Reactor Pressure Vessel, the coolant recirculation lines, pumps, etc. In case of an operating accident the vessel must contain the steam released within the Drywell, and conduct this steam to the Suppression Chamber.
Structural steel plate material is ASME SA516 fabricated to ASTM Designation A300. Minimum service temperature 30°F with Charpy impact requirements at maximum 0°F.
Seismic design load includes load due to vertical acceleration equal to 2/3 of the horizontal ground acceleration.
After an accident the Drywell may be flooded up to el. 1,000', stresses shall be below yield point (without seismic load), or may exceed yield but with a margin against rupture if seismic is considered.
Accident load (R) includes pressure and temperature in the primary containment.
USAR TABLE C-2-8 DRYWELL MEMBRANE STRESSES Method of Analysis ASME Section III including Code Cases 1330-1 and 1177-5 and addenda as of June 9, 1967, for Vessel Class "B".
Stress intensities are defined per Code para. N-414 and their limits are per Code para. N-414.
End conditions are found with methods described in the book Theory of Plates and Shells by Timoshenko.
Load Combination*
D + R + E D + R + E D + R + E D + E + Flood Max. Allowable Stress-ksi Primary General Membrane Sm= 17.5
@ 281 °F Primary Local Membrane PL = 1. 5 Sm
= 26.25
@ 281 °F Primary+
Secondary+
Bending Q = 3 Sm
= 52.50
@ 281 °F Yield 38.0
@ 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
Description/Criteria A jet force is assumed to occur in any direction within the drywell.
The force is calculated as 1,250 psi pressure acting on the area equal to the cross section of ruptured pipe.
The jet impingement force is considered to act coincidentally with the design internal pressure and 150°F shell temperature.
Temperature of the shell and welds are assumed to be 300°F if hit directly by jet.
Local thermal effects and dynamic jet effects are disregarded.
There is 2" air gap between drywell shell and backup concrete.
Material is ASME SA516 Grade 70 fabricated to ASTM Designation:
A300 or Tl by USS where noted.
USAR TABLE C-2-9 JET IMPINGEMENT FORCE STRESSES Load Method of Analysis Find maximum load on shell prior to breaking Compare with given jet load Apply the smaller load of either of the above:
calculate deformation, limiting stresses to prevent a progressive deformation or strain as follows:
(a)
Pm + Pb ~o. 9 yield (b)
Pm + Pb + Q <2 yield Pm general primary membrane stress Pb primary bending stress Q
secondary membrane
+bending stress Yield taken from ASME Sec. III Table N-424 at pertinent temperature Assume shear type failure of weld, and its stress equal to 7/10 of parent material.
C-2-20 Combination*
D + R D + R D + R D + R D + R Experimental test in 1964 D + R D + R D + R D + R D + R D + R Max. Allowable Stress-ksi 30.33 (Equipment door) 26.84 (Jet deflector support at vent) 90 (90% of yield of T-1 Steel at 300°F)
(Jet deflector baffle plate) 30.33(Top closure head) 30.33 (Cylinder above flanges)
CB&I experimental investigation proved that 3/4" thick plate can deform 3" without failure (Spherical shell) 30.33 (Cylindrical shell) 27.90 (Upper spray header pipe) 21.20 (Upper spray header weld) 30.33 (Upper spray header support) 27.90 (Upper spray header inlet pipe) 27.90 (Lower spray header pipe) 03/08/01
Description/Criteria 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.
USAR TABLE C-2-9 (Continued)
JET IMPINGEMENT FORCE STRESSES Method of Analysis Load Combination*
D + R D + R Max. Allowable Stress-ksi 21.20 (Lower spray header weld) 30.33 (Lower spray header support)
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
Description/Criteria The stabilizer mechanism transfers into building the reaction due to seismic loads or seismic plus jet loads acting on the Drywell, reactor and shield, or seismic, plus flooding of the Drywell.
The geometry of the stabilizer allows for radial and vertical movements due to pressure and temperature.
Materials: Components attached to the drywell are ASME SA516 Grade 70 fabricated to ASTM Designation: A300, per ASME Code Section III: Components outside the drywell are ASTM Designation: A36 per AISC-1963 Stress increase by 1/3 is allowed for jet loading or flooding.
Accident loads (R) considered include jet loads, temperature and pressure.
USAR TABLE C-2-10 DRYWELL STABILIZER SHEAR LUGS Method of Analysis ASME Code Section III including addenda as of June 9, 1967, for vessels Class "B" Formulas for Stress and Strain by Roark, Case 22 for plate.
Ge combined stress
.J(aB +
O'T) 2 + as GB bending stress GT tensile stress Gs Ge shear stress Fb (AISC)
Load Combination*
D + E D + R + E D + E + Flood D + E D + R + E D + E + Flood Max. Allowable Stresses-ksi Male Lug GB= (plate) 22. 8 Gs= (plate) 13. 8 Ge= (weld)
GB=30. 32 Gs=l8.35 Ge=21. 01 GB=30. 32 Gs=l8. 35 Ge=21. 01 Female Lug GB=22. 8 Gs=l3. 8 CTe=l5.8 CTB=30.32 Gs=l8.35 Ge=21.0l GB=30. 32 Gs=l8.35 Ge=2l.01 15.8 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 combinations corresponding to several different plant conditions.
calculated stresses are all within the appropriate Code allowables.
load The 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
- material, and Section III.
Torus Ring Designed per described in
- material, and Section III.
ASME Section III inspections in ASME Section III the CNS Plant inspections in C-2-23 with allowable accordance
- stresses, with ASME with stress analysis as Unique Analysis
- Report, accordance with ASME 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. [5ll 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 COMPONENT NAME STRUCTURAL COMPONENTS
SUMMARY
OF MARK I CONTAINMENT AND PIPING MODIFICATIONS NATURE OF MODIFICATION Suppression Chamber Shell and Supports Suppression Chamber Support Column Column Anchorage Column-to-Supp. Chamber Connection Suppression Chamber Saddle Ring Girder Vent System Vent Header/Downcomer Intersection Down comers Downcomer Ties Vent Header Deflector Vent Header Supports DW/WW Vacuum Breakers Miscellaneous Supp. Chamber Internals Monorail Service Platform Drywell Steel Framing MISCELLANEOUS SYSTEM MODIFICATIONS Drywell/Wetwell Pressure Differential Supp. Chamber Temp. Monitoring System PIPING SYSTEMS S/RV Discharge Piping Wetwell Piping T-Quencher Discharge Device T-Quencher Support Quencher Support Bracing Vacuum Breakers Pipe Supports and Restraints Torus Attached Piping Large Bore Supports Small Bore Supports Small Bore Rerouting Branch Line Supports Suppression Chamber Penetrations Valve Operator Supports Pump Anchors Torus Internal Piping HPCI Turbine Exhaust RCIC Turbine Exhaust Core Spray Return Test Line RHR Return Test Line Spray Header Vent Drain Line Plate reinforcement to column web and flanges Installed anchor bolts, brackets, and box beam assemblies Additional full penetration weldment Full saddles connecting suppression chamber support columns Web stiffeners; local reinforcement of weld to shell Reinforced 80 penetrations with stiffener plates and pads Reduced downcomer submergence by truncation Installed tie bar and ring assembly at each downcomer pair Installed deflector assembly in all suppression chamber bays Removed existing supports; resupported from girder above Reinforced 12 vacuum breaker penetrations Installed midbay supports in all suppression chamber bays Replaced existing supports; added new supports, bracing and grating tie-down Reinforcement of beam set connections and framing members Installed Pump Around System Installed monitoring system and instrumentation Rerouted with stronger pipe; added 12 new supports Installed T-quencher device on each S/RV line Installed quencher support assembly in 8 bays Installed quencher support bracing in 8 bays Installed two, 10-inch vacuum breakers on each line Installed 89 new or modified supports in Drywell Installed 151 new or modified supports Installed 54 new supports Rerouted 5 lines Installed 25 new or modified supports Reinforced three large bore penetrations Reinforced 13 valve yolks Modified anchorage of 4 RHR pumps Rerouted and resupported HPCI sparger Rerouted and resupported RCIC sparger Truncated test lines Installed reducer, discharge elbow, and new supports Reinforced existing supports Rerouted lines and installed supports C-2-28 03/08/01
2.5.7.2 2.5.7.2.1 USAR Pipe Whip and Jet Force Analysis 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, 5o, 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 clearance so as not to interfere with normal system operation, Operating Basis Earthquake Loads.
sufficient including
- 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.
redundant concurrent postulated
- 5.
The physical separation provided in the arrangement of engineered safety features provides spatial separation to preclude damage to more than one redundant safety feature by a single 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: [50J
- 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 one (1) inch nominal size.
C-2-30 piping exceeding 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 System Description Main Steam Core Spray Disch.
Pump lB Core Spray Disch.
Pump lA Clean-Up Recirc.
Pump Suet.
10" MS-1 to HPCI Turb.
3" MS-1 to RCIC Turb.
Reactor Feed Water Reactor Feed Water RHR Pump Suet.
RHR Pump Disch.
RHR Pump Disch.
RHR Head Spray*
2 Sh (PSI) 35,000 27,400 27,400 28,900 30,000 30,000 27,400 27,400 30,000 30,000 27,400 RHR Head Spray was removed per DC 86-78.
Maximum Allowable Stress (PSI) 18,710 11,060 10,892 24,360 22,120 10,957 12,300 17,200 13,528 14,371 9,449
- 1. 2 Sh (PSI) 21,000 16,440 16,440 17,350 18,000 18,000 16,440 16,440 16,440 18,000 18,000 16,440 Terminal Point Calculated Stress (PSI) 7,395 8,276 8,013 17,217 7,690 6,128 11,340 8,810 6,418 10,718 11,111 7,650 Location R.P.V. Nozzle R.P.V. Nozzle R.P.V. Nozzle 20" RH-lD HDR.
24" MS-1 HDR.
24" MS-1 HDR.
N.E. & S.E. R.P.V. Nozzle N.W. & S.W. R.P.V. Nozzle Conn. to Recirc. HDR South Conn. to Recirc. HDR North Conn. to Recirc. HDR 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. [50J 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 endanger either structures, piping or components containment during normal or accident conditions.
does not weaken or within the reactor
- 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. [4ll 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 reinforced analytical Protection Richard C.
Local penetration, due to a whipping pipe, into the 5' -0" thick concrete wall is evaluated to be a
maximum of 10.3",
using procedures and formula contained in: "Missile Generation and in Light-Water-Cooled Power Reactor Plants,"
by 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 6th 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 6th 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. r46 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. l64 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.
NOTE:
Event OBE SSE Factor of Safety 4
3 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
Description/Criteria Design load (D) includes all dead and equipment loads plus 25% of the full live loads Vertical seismic load (~P) is assumed to be produced by a vertical acceleration equal to 1/2 of the ground horizontal seismic acceleration, both acting simultaneously Overturning moment (M) is produced by the horizontal seismic acceleration and all design load eccentricities USAR TABLE C-2-13 CONTROL BUILDING FOUNDATION Method of Analysis Maximum soil stress=
D + ~p = ~ -
(YDa + YsatDb)
A S
where:
D
~p M
A s
y Design load, kips Vert. seismic load, 1/2 kips Horiz. seismic overturning moment, kip-ft Foundation mat area, sq-ft Section modulus of foundation mat, ft 3 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 Load Combination D + E D + E' Maximum Allowable Stress
- 1. 5 qa = 18
- 1. 5 qa 18 0110211s I
Description/Criteria Floor system is designed for the loadings tabulated in Subsection XII-2.3.2 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 assumed to be produced by a vertical acceleration equal to 1/2 of horizontal ground acceleration Maximum allowable stress for D+E' and D+T load combinations are: Concrete Fe= 0.85 f 1 c 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.
USAR TABLE C-2-14 CONTROL BUILDING FLOOR SYSTEM Method of Analysis Working Stress Design Method -
Slabs, beams, and girders C-2-40 Load Combination D + E D + E' D + T Max.
Allowable Stress-ksi Fe Fv Ft
- 1. 35 0.060 20.0 Fe = 2.55 Fv = 0.093 Ft= 36.0 Fe= 2.55 Fv = 0.093 Ft = 36.0 0110211s I
Description/Criteria Design load (D) includes all dead and equipment loads plus 25% of the full live loads Materials:
Concrete f'e = 3,000 psi at 28 days max. strength per ACI 318. Reinforcing ASTM A615 Grade 40 per ACI 318 Maximum allowable stresses for 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 USAR TABLE C-2-15 CONTROL BUILDING CONCRETE WALLS Method of Analysis Working Stress Design Bearing Walls C-2-41 Max.
Load Allowable Combination Stress-ksi D + E Ft = 20.0 Fe= 1.35 Fv=0.060 D + E' Ft= 36.0 Fe= 2.55 Fv=0.093 D + T Ft= 36.0 Fe = 2.55 Fv = 0.093 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
Description/Criteria Design load (D) including all dead and equipment loads plus 1/2" radial ice on all members Design wind loading (W) is 100 mph with wind pressures varying from 44 PSF at ground level to 73 PSF at top blowing against the projected area of members covered by 1/2" radial ice 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 USAR TABLE C-2-16 ELEVATED RELEASE POINT FOUNDATIONS AND TOWER Method of Analysis Free standing or rigid towers foundation Working Stress Method C-2-43 Load Combination D + W Ft Fe Max. Allowable Stress-ksi 20xl.33 26.7
- 1. 35xl. 33
- 1. 80 Fv 0.ll0xl.33 0.146 01102115 I
Description/Criteria Tower Steel Leg members A440 (50,000 psi yield) Other members A36 (36,000 psi yield)
Maximum allowable stresses for D+W load combinations are increased by 1/3 for foundation design. (Working stress method)
Concrete Ft = 1.33 fs Fe= 1.33 f 1 e Fv
- 1. 33 Ve USAR TABLE C-2-16 (Continued)
ELEVATED RELEASE POINT FOUNDATIONS AND TOWER Method of Analysis Tower Steel Ultimate design method based on formulae and provisions conforming to U.S. Bureau of Reclamation "Transmission Structures Design Standard No. 10" Load Combination D + W Max. actual stresses for load combination W+D or W-D (max uplift)
C-2-44 Max.
Allowable Stress-ksi Fa (leg) 35.77 for l = 57 r
Fa (brace) 10.60 for l = 174 r
Ft = 45.00 for A-440 Ft= 32.40 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 01102115 I
3.0 3.1 3.1.1 USAR COMPONENTS Intent and Scope 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
- Normal, Upset, Emergency, and Faulted generically described as follows:
C-3-1 be divided conditions.
into These four categories; categories are 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 Vessel Code, these conditions are designated as Service Level A Level B (Upset), Level C (Emergency), Level D (Faulted).
3.2.2 Allowable Limits Pressure (Normal),
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 PRIMARY STRESS LIMIT BUCKLING STABILITY LIMIT FATIGUE LIMIT Table C-3-2 Table C-3-3 Table C-3-4 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 SF min 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.
- b.
Where:
Permissible Deformation, DP Analyzed deformation causing loss of function, DL Permissible Deformation, DP Experimental deformation causing loss of function, DE 0.9 SF min
~
- 1. 0 SF min 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
- a.
- b.
- c.
- d.
USAR TABLE C-3-3 PRIMARY STRESS LIMIT Any One of (No More than One Required)
Elastic Evaluated Primary Stresses, PE Permissible Primary Stresses, PN Permissible Load, IP Largest Lower Bound Limit Load, CL Elastic Evaluated Primary Stress, PE Conventional Ultimate Strength at Temperature, US Elastic -
Plastic Evaluated Nominal Primary Stress, EP Conventional Ultimate Strength at Temperature, US
- e.
Permissible Load, LP Plastic Instability Load, PL
- f.
Permissible Load, LP Ultimate Load from Fracture Analysis, UF
- g.
Permissible Load, LP Ultimate Load or Loss of Function Load from Test, LE General Limit s;
s; s;
s; 2.25 SF min s; 1. 5 SF min s; 0.75 SF min s; 0. 9 SF min 0.9 SF min 0.9 SF min 1.0 SF min 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 us 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.
Conventional which would limiting.
ultimate strength at temperature or cause a system malfunction, whichever loading is more 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.
- b.
- c.
Permissible Load, LP Code normal event permissible load, PN Permissible Load, LP Stability Analysis Load, SL Permissible Load, LP Ultimate Buckling Collapse Load from Test, SE
- s; 2.25 SF min
- s; 0. 67 4 SF min
- s; 1.0 SF min Equation c was not used unless supporting data had been submitted for AEC Staff review. llBJ Where:
LP PN SL SE Permissible load under stated conditions of Normal, Upset, Emergency, or Faulted.
Applicable code normal event permissible load.
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.
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 usage including Emergency or Faulted events with design and operation loads following Miner's Hypotheses.... either one (not both)
- a.
- b.
Fatigue cycle usage
- . 05 (2l
- from analysis Fatigue cycle
- 0.33 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 SF min Upset N and E 10-1 2.25 or N and U 10-1 2.25 Emergency N and R 10-J
- 1. 5 Fault Where:
N N and E' Other combinations in this probability range N and E' and R Other combinations in this probability range Normal loads 10-J
<10- 1 to 10-J
- 1. 5 X 10-6 U
Upset loads (result in maximum system pressure)
E Maximum probable earthquake E'
Maximum possible earthquake
- 1. 5
<2.25 to 1. 5 1.125
<1.5 to 1.125 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: P40 < 10-6 ), then no safety factor is appropriate or required.
C-3-10 03/08/01
CRITERIA STABILIZER BRACKET AND ADJACENT SHELL Primary Stress Limit -
ASME Boiler and Pressure Vessel Code, Sect. III defines primary membrane plus primary bending stress intensity limit for SA302 -
Gr. B For normal and upset condition Stress limit= l.5x26,700 40,000 psi For emergency condition Stress limit= 1.5x40,000 60,000 psi For faulted condition Stress limit= 2.0x26,700 53,400 psi VESSEL SUPPORT SKIRT Primary Stress Limit -
ASME Boiler and Pressure Vessel Code, Sect. III, defines stress limit for SA516 GR70 For normal and upset condition B = 12,000 psi For emergency condition S1imit = 1.5 B = 18,000 psi For faulted condition S limit = 2. 0 B = 2 4, 0 0 0 p S i USAR TABLE C-3-7 LOADING CRITERIA Reactor Vessel Internals and Associated Equipment LOADING Normal and upset condition load
- 1. OBE (E)
- 2. Design pressure Emergency condition load
- 1. SSE (E')
- 2. Design pressure Faulted condition loads
- 1. SSE (E')
- 2. Jet reaction forces
- 3. Design pressure Normal and upset condition loads
- 1. Dead weight
- 2. OBE (E)
Emergency condition loads
- 1. Dead weight
- 2. SSE (E')
Faulted condition loads
- 1. Dead weight
- 2. SSE (E')
- 3. Jet reaction forces C-3-11 PRIMARY STRESS TYPE Membrane plus bending Membrane plus bending Membrane Compressive membrane Compressive membrane Compressive membrane ALLOWABLE STRESS 40,000 psi 60,000 psi 53,400 psi 12,000 psi 18,000 psi 24,000 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA SHROUD SUPPORT GUSSETS Primary Stress Limit -
ASME Boiler and Pressure Vessel Code, Sect. III defines allowable primary membrane stress plus bending stress for SB168 material.
For normal and upset condition SA = 1. 5 SM =
1.5 X 23.30 = 34.95 KSI For emergency condition S1imit =
1
- 5 SA=
1.5 X 34.95 = 52.43 KSI For faulted condition A limit =
2
- 0 SA=
2.0 X 34.95 = 69.90 KSI LOADING Normal and upset condition loads
- 1.
OBE (E)
- 2.
Pressure drop across shroud (normal)
- 3.
Subtract dead weight Emergency condition loads
- 1.
SSE (E')
- 2.
Pressure drop across shroud (normal)
- 3.
Subtract dead weight Faulted condition loads
- 1.
SSE (EI)
- 2.
Pressure drop across shroud during faulted condition
- 3.
Subtract dead weight C-3-12 PRIMARY STRESS TYPE Membrane plus bending Membrane plus bending Membrane plus bending ALLOWABLE STRESS 34,950 psi 52,430 psi 69,900 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA TOP GUIDE-LONGEST BEAM Primary Stress Limit -
The allowable primary membrane stress plus bending stress is based on ASME Boiler and Pressure Vessel Code, Sect. III for type 304 stainless steel plate.
For normal and upset condition Stress Intensity SA= 1.5 SM=
l.5x16,925 psi= 25,388 psi.
For emergency condition S 1 imi t = 1. 5 SA= 1. 5 X 2 5, 3 8 8 38,081 psi For faulted condition S1imit = 2 SA= 2x25, 388 50,775 psi TOP GUIDE BEAM END CONNECTIONS Primary Stress Limit -
ASME Boiler and Pressure Vessel Code, Sect. III, defines material stress limit for type 304 stain-less steel.
For normal and upset condition Stress Intensity SA= 0.6 SM 0.6x16,925 psi= 10,155 psi For emergency condition S 1 imit = 1. 5 SA= 1. 5 X 1 0, 15 5 p Si 15,232 psi.
For faulted condition S 1 imi t = 2 SA = 2 X 1 0, 15 5 p Si 20,310 psi LOADING Normal and upset condition loads
- 1.
OBE (E)
- 2.
Weight of structure
- 3.
Weight of temporary control curtains.
Emergency condition loads
- 1.
SSE (E')
- 2.
Weight of structure
- 3.
Weight of temporary control curtains.
Faulted condition loads (same as emergency condition)
Normal and upset condition loads
- 1.
OBE (E)
- 2.
Weight of structure
- 3.
Weight of temporary control curtains.
Emergency condition loads
- 1.
SSE (E')
- 2.
Weight of structure
- 3.
Weight of temporary control curtains.
Faulted condition loads (same as emergency condition)
C-3-13 PRIMARY STRESS TYPE General membrane plus bending General membrane plus bending General membrane plus bending Pure shear Pure shear Pure shear ALLOWABLE STRESS 25,388 psi 38,081 psi 50,775 psi 10,155 psi 15,232 psi 20,310 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA TOP GUIDE ALIGNERS Primary Stress Limit -
The allowable primary membrane stress plus bending stress is based on ASME Boiler and Pressure Vessel Code, Sect.
III for type 304 stainless steel plate.
For normal and upset condition Stress Intensity SA = 1. 5 SM =
l.5x16,925 psi 25,388 psi For emergency condition S limit = 1. 5 SA = 1. 5 X 2 5, 3 8 8 38,081 psi For faulted condition S limit = 2 SA = 2 X 3 8, 0 8 1 50,775 psi LOADING PRIMARY STRESS TYPE Normal and upset condition General membrane plus loads bending
- 1.
OBE (E)
- 2.
Weight of structure
- 3.
Weight of temporary control curtains.
Emergency condition loads General membrane plus
- 1.
SSE (E')
bending
- 2.
Weight of structure
- 3.
Weight of temporary control curtains.
Faulted condition loads General membrane plus (same as emergency condition) bending C-3-14 ALLOWABLE STRESS 25,388 psi 38,081 psi 50,775 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA RPV STABILIZER Primary Stress Limit AISC 6th Edition specification for the construction, fabrication, and erection of structural steel for buildings For normal & upset conditions AISC allowable stresses, but without the usual increase for earthquake loads For emergency conditions 1.5 x AISC allowable stresses For faulted conditions Material yield strength RPV SUPPORT (RING GIRDER)
Primary Stress Limit AISC 6th Edition specification for the design, fabrication and erection of structural steel for buildings For normal and upset conditions AISC allowable stresses, but without the usual increase for earthquake loads.
For faulted conditions 1.67 X AISC allowable stresses for structural steel members Yield strength for high strength bolts (vessel to ring girder)
LOADING Upset condition
- 1. Spring preload
- 2.
OBE (E)
Emergency condition
- 1. Spring preload
- 2.
SSE (E')
Faulted condition
- 1. Spring preload
- 2.
SSE (E')
- 3. Jet reaction load Normal and upset condition
- 1.
Dead loads
- 2.
OBE (E)
- 3.
Loads due to scram Emergency condition
- 1.
Dead loads
- 2.
SSE (E')
- 3.
Loads due to scram Faulted condition
- 1.
Dead loads
- 2.
SSE (E')
- 3. Jet reaction load C-3-15 LOCATION Rod Bracket Bracket Bracket Top Flange Bottom Flange Vessel to girder bolts Top Flange Bottom Flange Vessel to girder bolts Top Flange Bottom Flange Vessel to girder bolts ALLOWABLE STRESS 127,000 psi 22,000 psi 14,000 psi 33,000 psi 21,000 psi 36,000 psi 21,500 psi 27,000 psi 27,000 psi 60,000 psi 18,000 psi 40,500 psi 40,500 psi 90,000 psi 20,800 psi 45,000 psi 45,000 psi 125,000 psi 72,000 psi 03/08/01
CRITERIA CORE SUPPORT Primary Stress Limit -
The allowable primary membrane stress plus bending stress is based on ASME Boiler and Pressure Vessel Code, Sect.
III for type 304 stainless steel plate.
For allowable stresses see top guide, longest beam, above CORE SUPPORT ALIGNERS Primary Stress Limit -
ASME Boiler and Pressure Vessel Code, Sect. III, defines material stress limit for type 304 stainless steel.
For allowable shear stresses, see top guide beam end connec-tions, above.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
LOADING Normal and upset condition loads
- 1. Normal operation pressure drop
- 2.
OBE (E)
Emergency condition loads
- 1. Normal operation pressure drop
- 2.
SSE (E')
Faulted condition loads
- 1. Pressure drop after recirculation line rupture
- 2.
SSE (E')
Normal and upset condition load
- 1.
OBE (E)
Emergency condition load
- 1.
SSE (EI)
Faulted condition load
- 1.
SSE (EI)
C-3-16 PRIMARY STRESS TYPE General membrane plus bending General membrane plus bending General membrane plus bending Pure shear Pure shear Pure shear ALLOWABLE STRESS 25,388 psi 38,081 psi 50,775 psi 10,155 psi 15,232 psi 20,310 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA FUEL CHANNELS Primary Stress Limit - Allowable stress SM for Zircaloy or NSF determined according to methods recommended by ASME Boiler and Pressure Vessel Code, Sect. III.
Allowable moment determined by calculating limit moment using Table C-3-3, equation (b), then applying SF~n for applicable loading conditions.
( Sm = 9,270 psi; 1. 5 Sm 13,900 psi)
Emergency limit load= 1.5 x Normal limit load calculated using 1.5 Sm= yield.
LOADING Normal and upset condition load
- 1.
OBE (E)
- 2.
Normal pressure load Emergency condition load
- 1.
SSE (E')
- 2. Normal pressure load Faulted condition load
- 1.
SSE (E')
- 2.
Loss of cooling accident pressure PRIMARY STRESS TYPE Membrane and bending Membrane and bending Membrane and bending MOMENT LIMIT accounting for pressure loads 28,230* in. lbs.
42,350* in. lbs.
56,500* in. lbs.
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 CRD HOUSING SUPPORT Primary Stress Limit AISC specification for the design, fabrication and erection of structural steel for buildings For normal and upset condition Fa= 0.60 Fy (tension)
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 Structural Steel: AISC specification for the design, fabrication and erection of structural steel for buildings.
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 LOADING Faulted condition loads
- 1.
Dead weight
- 2.
Impact force from failure of a CRD housing (Dead weights and earthquake loads are very small as compared to jet force.)
Faulted condition loads
- 1.
Jet force from a complete circumferential failure (break) of recirculation line.
C-3-18 LOCATION Beams (top cord)
Beams (bottom cord)
Grid structure Brackets on 28" pipe Cable on pump restraints ALLOWABLE STRESS 33,000 psi 33,000 psi 33,000 psi 33,000 psi 41,500 psi 27,500 psi 33,000 psi 90% of listed breaking strength 03/29/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA CONTROL ROD DRIVE HOUSING Primary Stress Limit -
The allowable primary membrane stress is based on the ASME Boiler and Pressure Vessel Code, Sect. III, for Class A vessels, for type 304 stainless steel.
For normal and upset condition Sm= 15,800 psi @ 575°F For emergency conditions Siimit =
1.5 Sm= 1.5x15,800 23,700 psi CONTROL ROD DRIVE Primary Stress Limit -
The allowable primary membrane stress plus bending stress is based on ASME Boiler and Pressure Vessel Code, Sect. III for SA212 TP316 tubing.
For normal and upset condition SA
1. 5 Sm = 1. 5xl 7, 3 7 5
26,060 psi LOADING Normal and upset condition loads
- 1.
Design pressure
- 2.
Stuck rod scram loads
- 3.
OBE (E)
Emergency condition loads
- 1.
Design pressure
- 2.
Stuck rod scram loads
- 3.
SSE (E')
Normal and upset condition loads Maximum hydraulic pressure from the control rod drive supply pump.
NOTE: Accident conditions do not increase this loading.
Earthquake loads are neglig-ible.
C-3-19 LOCATION Maximum membrane stress intensity occurs at the tube to tube weld near the center of the housing for normal, upset and emergency conditions.
Maximum stress inten-sity occurs at a point on the Y-Y axis of the indicator tube.
ALLOWABLE STRESS 15,800 psi 23,700 psi 26,060 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA CONTROL ROD GUIDE TUBE Primary Stress Limit The allowable primary membrane stress plus bending stress is based on the ASME Boiler and Pressure Vessel Code,Section III for 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 allowable primary membrane stress is based on ASME Boiler and Pressure Vessel Code, Sect.
III, for Class A vessels for 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 LOADING Faulted condition loads
- 1.
Dead weight
- 2. Pressure drop across guide tube due to failure of recirculation line
- 3.
SSE (E')
Emergency condition loads
- 1. Design pressure
- 2.
SSE (E')
C-3-20 LOCATION The maximum bending stress under faulted loading conditions occurs at the center of the guide tube Maximum membrane stress intensity occurs at the outer surface of the vessel penetration ALLOWABLE STRESS 47,400 psi 23,700 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA HYDRAULIC CONTROL UNIT PIPING From USAS B31.1.0 -
1967 Code for power pressure piping For Normal Conditions:
sh= 15,000 psi For upset and emergency con-dition:
When upset or emergency condition exists for less than 1% of the time, the code allows 20% increase in stress.
Sa= 1.2 sh= 18,000 psi LOADING Normal Condition Load Maximum normal hydraulic system pump pressure Upset condition load
- 1. Shut off pump pressure
- 2.
OBE (E) (negligible load)
Emergency condition
- 1. Shut off pump pressure
- 2.
SSE (E') (negligible load)
C-3-21 LOCATION 3/4" drive withdraw piping 3/4" drive withdraw piping 3/4" drive withdraw piping ALLOWABLE STRESS 15,000 psi 18,000 psi 18,000 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Reactor Vessel Internals and Associated Equipment (Cont'd)
CRITERIA SPENT FUEL STORAGE RACKS Stresses due to normal, upset, or emergency loading shall not cause the racks to fail so as to result in a critical fuel array.
Primary Stress Limit -
Paper numbers 3341 and 3342, Pro-ceedings of the ASCE, Journal of the Structural Division, Dec., 1962 (task committee on light-weight alloys) (Aluminum)
Emergency Conditions Stress limit= yield strength at 0.2% offset.
LOADING Emergency condition "A" loads
- 1.
Dead loads
- 2. Full fuel load in rack
- 3.
SSE (E')
Emergency condition 11 B 11 loads (see below)
LOCATION At column to base welds At base hold down lug (casting)
ALLOWABLE STRESS 11,000 psi 20,000 psi
( 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 conservative.
(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, uncontrolled drop of the fuel grapple from its full retracted position into the weakest portion of the rack.
Method of Analysis The displacement 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 longitudinal 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
CRITERIA Stress due to all normal and upset loadings must not exceed the limits of USAS B31. 1. 0.
- 1. The sum of the longitudinal stresses due to pressure and dead weight must be less than the hot allowable stress.
- 2. The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of OBE(E) must be less than 1.2 times the hot allowable stress.
- 3. The sum of the longitudinal stresses due to pressure and dead weight plus the thermal expansion 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 stresses due to pressure, dead weight, and OBE (E) plus the thermal expansion stress intensity range must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Piping LOADING
- a.
- b.
- a.
- b.
- c.
- a.
- b.
C.
- a.
- b.
- c.
- d.
Dead weight Pressure Dead weight Pressure OBE (E)
Dead weight Pressure Thermal loads Dead weight Pressure OBE (E)
Thermal loads C-3-24 "Note: Analyses of this piping system demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
03/08/01
CRITERIA 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 due to pressure, dead weight, and inertia effects of SSE (E') must be 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 due to maximum pressure, dead weight, and inertia effects of OBE (E) must 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 due to maximum pressure, dead weight, and inertia effects of SSE (E') 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.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Piping (Cont'd)
LOADING
- a.
- b.
- c.
- a.
- b.
- c.
- a.
- b.
- c.
Dead weight Pressure SSE (EI)
Dead weight Maximum pressure OBE (E)
Dead weight Maximum pressure SSE (EI)
C-3-25 03/08/01
"Note:
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Loop Piping 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
"Note:
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS -
Core Spray Discharge Piping CS-ID 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
"Note:
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS -
Clean-Up Recirculation Pump Suction Piping CU-IS 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 Stress due to all normal and upset loadings must not exceed the limits of USAS B31. 1. 0.
- 1.
- 2.
- 3.
- 4.
The sum of the longitudinal stresses due to pressure and dead weight must be less than the hot allowable stress.
The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of OBE must be less than 1.2 times the hot allowable stress.
The sum of the longitudinal stresses due to pressure and dead weight plus the thermal expansion stress intensity range must be less than the sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
The sum of the longitudinal stresses due to pressure, dead weight, and OBE plus the thermal expansion stress intensity range must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
LOADING
- a.
Dead weight
- b. Pressure
- a.
Dead weight
- b.
Pressure
- c.
- a.
Dead weight
- b. Pressure
- c.
Thermal loads
- a.
Dead weight
- b. Pressure
- c.
- d.
Thermal loads C-3-29 "Note: Analyses of this piping system demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
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 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 due to pressure, dead weight, and inertia effects of SSE must be 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 due to maximum pressure, dead weight, and inertia effects of 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 due to maximum pressure, dead weight, and inertia effects of 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.
LOADING
- a.
Dead weight
- b. Pressure
- c.
- a.
Dead weight
- b.
Maximum pressure
- c.
- a.
Dead weight
- b.
Maximum pressure
- c.
SSE C-3-30 03/08/01
CRITERIA Stress due to all normal and upset loadings must not exceed the limits of USAS B31. 1. 0.
- 1.
- 2.
- 3.
- 4.
The sum of the longitudinal stresses due to pressure and dead weight must be less than the hot allowable stress.
The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of OBE must be less than 1.2 times the hot allowable stress.
The sum of the longitudinal stresses due to pressure and dead weight plus the thermal expansion stress intensity range must be less than the sum of the allowable stress range for expansion at stresses plus the hot allowable stress.
The sum of the longitudinal stresses due to pressure, dead weight, and OBE plus the thermal expansion stress intensity range must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS Reactor Feed Piping LOADING
- a.
- b.
- a.
- b.
- c.
- a.
- b.
C.
- a.
- b.
C.
- d.
Dead weight Pressure Dead weight Pressure OBE Dead weight Pressure Thermal loads Dead weight Pressure OBE Thermal loads C-3-31 "Note: Analyses of this piping system demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
03/08/01
CRITERIA 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 due to pressure, dead weight, and inertia effects of SSE must be 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 due to maximum pressure, dead weight, and inertia effects of 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 due to maximum pressure, dead weight, and inertia effects of 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.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS -
Reactor Feed Piping (Cont'd)
LOADING
- a.
Dead weight
- b. Pressure
- c.
- a.
Dead weight
- b.
Maximum pressure
- c.
- a.
Dead weight
- b. Maximum pressure
- c.
SSE 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 Stress due to all normal and upset loadings must not exceed the limits of USAS B31.1.0.
- 1. The sum of the longitudinal stresses due to pressure and dead weight must be less than the hot allowable stress.
- 2.
The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of OBE must be less than 1.2 times the hot allowable stress.
- 3.
The sum of the longitudinal stresses due to pressure and dead weight plus the thermal expansion 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 stresses due to pressure, dead weight, and OBE plus the thermal expansion stress intensity range must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
LOADING
- a.
Dead weight
- b. Pressure
- a.
Dead weight
- b. Pressure
- c.
- a.
Dead weight
- b. Pressure
- c.
Thermal loads
- a.
Dead weight
- b. Pressure
- c.
- d.
Thermal loads C-3-33 "Note: Analyses of this piping system demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS -
Residual Heat Removal Pump Suction Piping (Cont'd)
CRITERIA 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 due to pressure, dead weight, and inertia effects of SSE must be 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 due to maximum pressure, dead weight, and inertia effects of 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 due to maximum pressure, dead weight, and inertia effects of 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.
LOADING
- a.
Dead weight
- b. Pressure C.
- a.
Dead weight
- b.
Maximum pressure
- c.
- a.
Dead weight
- b.
Maximum pressure
- c.
SSE 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 Stress due to all normal and upset loadings must not exceed the limits of USAS B31.1.0.
- 1.
The sum of the longitudinal stresses due to pressure and dead weight must be less than the hot allowable stress.
- 2.
The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of OBE must be less than 1.2 times the hot allowable stress.
- 3.
The sum of the longitudinal stresses due to pressure and dead weight plus the thermal expansion 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 stresses due to pressure, dead weight, and OBE plus the thermal expansion stress intensity range must be less than 1.2 times the sum of the allowable stress range for expansion stresses plus the hot allowable stress.
LOADING
- a.
Dead weight
- b. Pressure
- a.
Dead weight
- b. Pressure
- c.
- a.
Dead weight
- b. Pressure
- c.
Thermal loads
- a.
Dead weight
- b.
Pressure
- c.
- d.
Thermal loads C-3-35 "Note: Analyses of this piping system demonstrate compliance with the applicable Code and are documented in the existing calculations of record."
03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Class IN/IS - Residual Heat Removal Pump Discharge Piping (Cont'd)
CRITERIA 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 due to pressure, dead weight, and inertia effects of SSE must be 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 due to maximum pressure, dead weight, and inertia effects of 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 due to maximum pressure, dead weight, and inertia effects of 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.
LOADING
- a.
Dead weight
- b. Pressure
- c.
- a.
Dead weight
- b.
Maximum pressure
- c.
- a.
Dead weight
- b.
Maximum pressure
- c.
SSE C-3-36 03/08/01
CRITERIA
- 1. Casing Minimum Wall Thickness Loads: Normal and upset condition Design pressure & temperature Primary membrane stress limit:
Allowable working stress per ASME Section III, Class C
- 2. Casing Cover Minimum Thickness Loads: Normal and upset condition Design pressure & temperature Primary bending stress limit:
1.5 Sm per ASME code for Pumps and Valves for Nuclear Power Class I USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Pumps METHOD OF ANALYSIS t
PR
+ C SE-06P ALLOWABLE STRESS OR ACTUAL THICKNESS 3.00 in.
where:
t p
R s
E C
minimum required thickness, in.
design pressure, psig maximum internal radius, in.
allowable working stress, psi joint efficiency corrosion allowance, in.
Sr 3w [ 2 2
2 b4 (m-1)4b4 (m+l)lna/b+a2b2(m+l)]
a -
b +------------
4t2 a2(m-1) + b2(m+l)
+ -- 1---------
3w [
2mb2 -2b2(m+ l)ln a/ b]
2nt2 a2(m-1)+ b2(m+l) 14,950 psi St 3w(m2-l) [a 4-b 4 -4a2b2lna/b] +
4mt2 a2(m-1) + b2(m+ 1)
~
[l + ma2 (m-1) - mb2(m+ 1) - 2(m2 - l)a 2 1n a/ b]
2nmt2 a2(m-1) + b2(m+ 1) 14,950 psi where:
Sr radial stress at outer edge, psi t
m a
b disc thickness, in.
reciprocal of Poisson's ratio radius of disc, in.
St tangential stress at inner edge, psi w
pressure load, psi W
uniform load along inner edge, lb.
radius of disc hole, in.
C-3-37 03/08/01
CRITERIA
- 3. Cover and seal flange bolt areas Loads: Normal and upset condition Design pressure & temperature.
Design gasket load Bolting Stress Limit:
Allowable working stress per ASME Section III, Class C
- 4.
Cover Clamp Flange Thickness Loads: Normal and upset 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 Loads: Normal and upset condition Design pressure & temperature.
Piping reactions during normal USAR TABLE C-3-7 (Cont'd) LOADING CRITERIA Recirculation Pumps (Cont'd)
METHOD OF ANALYSIS Bolting loads, areas and stresses are calculated in accordance with "Rules for Bolted Flange Connections" ASME Section VIII, Appendix II.
Flange thickness and stress are calculated in accordance with "Rules for Bolted Flange Connections" ASME Section VIII, Appendix II.
re D2 P M
F SL=
-- + -
+ -
4 A
Z A
PD Sc= 2t Ss = TRo J
C-3-38 ALLOWABLE STRESS OR ACTUAL THICKNESS Cover Flange Bolts 27,000 psi Seal Flange Bolts 20,000 psi Flange Thickness
- 8. 25" 03/08/01
CRITERIA
- 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.
- 6. Mounting Bracket Combined Stress Loads:
Flooded weight SSE (EI)
Combined Stress Limit:
Yield Stress USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Pumps where:
SL Sc Ss D
p A
M F
t J
Ro T
METHOD OF ANALYSIS s
+ s{
2 longitudinal stress, psi circimferential stress, psi torsional stress, psi nozzle internal diameter, in.
design pressure, psi nozzle cross section metal area, maximum bending moment, in. lb.
maximum longitudinal force, lb.
nozzle wall thickness, in.
polar moment of inertia, in4 nozzle outside radius, in.
torsional moment in2 Bracket vertical loads are determined by summing the equipment and fluid weights and vertical seismic forces. Bracket horizontal loads are determined by applying the specified seismic force at mass center of pump-motor assembly (flooded).
Horizontal and vertical loads are applied simultaneously to determine tensile, shear and bending stresses in the brackets. Tensile, shear and bending stresses are combined to determine maximum combined stresses.
C-3-39 ALLOWABLE STRESS OR ACTUAL THICKNESS Pump Nozzle Stresses 28,650 psi 28,650 psi 28,650 psi 28,650 psi Maximum Combined Stresses 15,600 psi 03/08/01
CRITERIA
- 7. Stresses Due to Seismic Loads Loads:
Operating pressure and temperature SSE (E')
Combined Stress Limit:
Yield stress USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Pumps (Cont'd)
METHOD OF ANALYSIS The flooded pump-motor assembly is analyzed as a free body supported by constant support hangers from the pump brackets. Horizontal and vertical seismic forces are applied at mass center of assembly and equilibrium reactions are determined for the motor and pump brackets. Load, shear, and moment diagrams are constructed using live loads, dead loads, and calculated snubber reactions. Combined bending, tension and shear stresses are determined for each major component of the assembly including motor, motor support barrel, bolting and pump casing. The maximum combined tensile stress in the cover bolting is calculated using tensile stresses determined from loading diagram plus tensile stress from operating pressure.
C-3-40 ALLOWABLE STRESS OR ACTUAL THICKNESS Motor Bolt Tensile Stress:
40,000 psi Pump Cover Bolt Tensile Stress:
43,200 psi Motor Support Barrel Combined Stresses:
22,400 psi 03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RHR Pump 0.23 g horizontal 0.46 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain 1.
the internal design pressure of the pump casing without exceeding the allowable stress of the bolting material. Allowable stresses at design temperature are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the pump 2.
limits stress to the allowable stress when subjected to design pressure and temperature. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
Sc Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections",
ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
Stress in the pump casing is calculated at the point of maximum internal pump diameter by the formula P(D +.2 t) 2t where Sc calculated stress, psi p
D t
pump design pressure, psi maximum pump internal diameter actual min. metal thickness less corrosion allowance, 0.080 inches C-3-41 Results
- 1. Allowable Stress 25,000 psi Pump Design Pressure 450 psig Max. Design Temperature 350°F
- 2. Allowable Stress 14,000 psi 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 attaching pipe on pump nozzles under combined maximum thermal expansion and OBE loading (E) reaction plus load due to internal pressure does not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME B&PV Code,Section VIII.
For SSE (E'), less than 1.5 of allowable stress.
suction discharge
- 3. Stresses will not be excessive if the maximum force when taken with the maximum moment falls below the line.
F~
E Fintercept 77,007 (M=0)
M E'
124,988 lbs.
Mintercept 725,709 1,385,100 in.lbs.
(F=0)
Fintercept 65,696 (M=0) 116,898 lbs.
Mintercept 629,003 1,060,236 in.lbs.
(F=0)
C-3-42 03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Core Spray Pump 0.23 g horizontal 0.46 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain the internal design pressure of the pump casing without exceeding the allowable stress of the bolting material.
Allowable stresses at design temperature are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the pump limits stress to the allowable stress when subjected to design pressure and temperature.
Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
Method of Analysis
- 1. Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections", ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
- 2. Stress in the pump casing is calculated at the point of maximum internal pump diameter by the formula Sc where P(D +.2 t) 2t 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 Results
- 1. Allowable Stress 20,000 psi Pump Design Pressure Max. Design Temperature 500 psi 212°F
- 2. Allowable Stress 14,000 psi 03/08/01
Statement of Criteria
- 3. Application of forces and moments by attaching pipe on pump nozzles under combined maximum thermal expansion and OBE loading reaction (E) plus load due to internal pressure does not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME B&PV Code,Section VIII.
For SSE (E') less than 1.5 of allowable stress.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Core Spray Pump (Cont'd)
Suction Method of Analysis
- 3. Stresses will not be excessive if the maximum force when taken with the maximum moment falls below the line.
F~
E Fintercept 54,535 (M=0)
M E'
79,808 lbs.
Mintercept 402,803 623,427 in.lbs.
(F=0)
Fintercept 30,624 Discharge (M=0) 59,668 lbs.
Mintercept 293,562 426,519 in.lbs.
(F=0)
C-3-44 03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Pump 0.23 g horizontal 0.46 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain the internal design pressure of the pump casing without exceeding the allowable working stress of the bolting material. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the pump limits stress to the allowable working stress when subjected to design pressure plus corrosion allowance. Allowable stresses are in accordance with ASME B&PV Code,Section III.
The maximum stress in the pump casing when subjected to design pressure does not exceed the allowable working stress of the material. The allowable stress are in accordance with ASME B&PV Code,Section III.
Method of Analysis
- 1. Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections", ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
- 2. Nozzle stress is calculated by the following formula P(D +.2 t)
Sh =
2Et Volute stress is calculated by the following formula Sv = Pb (R + a) 2t R
Roark
- p. 307 Case 26 and R a -
0.5b C-3-45 Results Main Pump
- 1. Design Pressure 1,500 psig Allowable Stress 20,000 psi Results Boost Pump Design Pressure 450 psi Allowable Stress 20,000 psi
- 2. Main Pump Design Pressure 1,500 psig Allowable Stress 14,000 psi Boost Pump Design Pressure 450 psig Allowable Stress 14,000 psi 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Pump (Cont'd)
Statement of Criteria
- 3. Application of forces and moments by attaching pipe on pump nozzles under combined maximum thermal expansion and OBE loading reaction (E) plus load due to internal pressure does not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME B&PV Code,Section VIII.
For SSE (E') less than 1.5 of allowable stress.
Suction Discharge C-3-46 Method of Analysis
- 3. Stresses will not be excessive if the maximum force when taken with the maximum moment falls below the line.
F~
M E
E' rintercept 33,000 65,000 lbs.
(M=0)
Mintercept 430,000 700,000 in.lbs.
(F=0) r intercept 31,000 63,000 lbs.
(M=0)
Mintercept 250,000 540,000 in.lbs.
03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Pump 0.23 g horizontal 0.46 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain the internal design pressure of the pump casing without exceeding the allowable working stress of the bolting material. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the pump limits stress to the allowable working stress when subjected to design pressure plus corrosion allowance.
Allowable stresses are in accordance with ASME B&PV Code,Section III.
The maximum stress in the pump casing when subjected to design pressure does not exceed the allowable working stress of the material. The allowable stress is in accordance with ASME B&PV Code,Section III.
Method of Analysis
- 1. Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections",
ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
- 2. Stress in the pump nozzle is calculated at the point of maximum internal pump diameter by the formula Sc P(D +.2t) 2Et Sc= 0.8Sa volute stress is computed by the following formula:
Sb=/JPb2 t2 Roark p. 225 Case# 36 factor from Roark volute length volute width C-3-47 Results
- 1.
- 2.
Design Pressure Allowable Stress Design pressure Allowable Stress Design Pressure Allowable Stress 1,500 psig 20,000 psi 1,500 psig 14,000 psi 1,500 psig 14,000 psi 03/08/01
Statement of Criteria
- 3. Application of forces and moments by attaching pipe on pump nozzles under combined maximum thermal expansion and OBE loading reaction (E) plus load due to internal pressure does not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME B&PV Code,Section VIII.
For SSE less than 1.5 of allowable stress.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Pump (Cont'd)
Method of Analysis
- 3. Stresses will not be excessive if the maximum force when taken with the maximum moment falls below the line.
F~
Suction Fintercept (M=0)
E 6,200 Discharge C-3-48 Mintercept 34,700 (F=0)
Fintercept 2,800 (M=0)
Mint~rcept 12,300 (F=0)
M E'
11,800 lbs.
64,000 in.lbs.
7,200 lbs.
30,000 in.lbs.
03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Standby Liquid Control Pump 0.33 g horizontal 0.66 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain the internal design pressure of the pump without exceeding the allowable working stress of the bolting material.
Allowable stresses are in accordance with ASME B&PV Code.
- 2. The maximum stress in the pump fluid cylinder when subjected to design pressure does not exceed 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 mounting bolts when the motor is subjected to the SSE does not exceed 0.9 of yield stress and twice the allowable shear stress for bolting material in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII.
Method of Analysis
- 1. Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections",
ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
- 2. Stress in the pump fluid cylinder is calculated at the point of maximum stress by the pressure area method.
- 3. The seismic forces acting on the motor to subject the bolting to shear or tension are considered.
The motor is isolated from the pump and nozzle forces by the flexible coupling.
C-3-49 Results
- 1. Stuffing Box Bolts Allowable Stress Cylinder Head Bolts Allowable Stress
- 2. Design Pressure Allowable Stress 25,000 psi 25,000 psi 1,400 psig 16,500 psi
- 3. The bolt stresses under (E')
Tension Allowable Stress Shear Allowable Stress 03/08/01
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Standby Liquid Control Pump (Cont'd)
Statement of Criteria
- 4.
The stresses in the pump mounting bolts due to the combination of OBE acting on the flooded pump plus the attaching pipe reactions under combined maximum thermal expansion plus OBE does not exceed the allowable shear and tensile stresses for the bolting material in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII.
The attaching pipe reaction plus the load due to internal pressure does not produce an equivalent bending and torsional stress in nozzles in excess of the allowable stress.
The stresses in the pump mounting bolts due to the combination of the SSE acting on the flooded pump plus the attaching pipe reaction under combined maximum thermal expansion plus SSE does not exceed 0.9 times the yield stress in tension and twice the allowable shear stress for the bolting material in accord-ance with the ASME Boiler and Pressure Vessel Code,Section VIII.
The attaching pipe reaction plus the load due to internal pressure does not produce an equivalent bending and torsional stress in nozzles in excess of 1.5 times allowable stress.
C-3-50 Method of Analysis
- 4.
The maximum force taken with the maximum moment shall fall below the line on the force-moment diagram:
E Discharge Suction E'
Discharge Suction F
F M
F M
342 lbs
.9450 in. lbs., but not to exceed 3400 in. lbs.
711 lbs.
M =
39,100 in. lbs.,
but not to exceed 16,600 in. lbs.
F 684 lbs M =
18,900 in. lbs.,
but not to exceed 5,330 in. lbs.
F M
1,422 lbs 78,200 in. lbs.,
but not to exceed 24,700 in. lbs.
where Mis maximum moment (in. lbs.)
in any direction and Fis maximum force (lbs.) in any direction.
03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine 0.75 g horizontal 1.50 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain the internal design pressure of the turbine casing without exceeding the allowable working stress of the bolting material. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the turbine casing limits stress to the allowable working stress when subjected to design pressure plus corrosion allowance. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
Method of Analysis
- 1. Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections",
ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
Results
- 1.
Allowable Stress
- 2. Stresses in the various pressure
- 2.
Allowable Stress containing portions of the turbine casing are calculated according to the Rules of Part UG,Section VIII, of the ASME Boiler & Pressure Vessel Code.
C-3-51 20,000 psi 17,500 psi 03/08/01 I
Statement of Criteria
- 3. The forces and moments imposed by the attached piping on the turbine inlet and exhaust connections satisfy the following conditions:
The resultant force and moment from the combination of dead weight and thermal expansion is less than that stipulated by the equipment vendor.
The resultant force and moment from the combination of dead weight, thermal expansion and OBE (E) or SSE (E') is less than that demonstrated acceptable by detailed seismic analysis of the equipment.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine (Cont'd)
Method of Analysis
- 3.
The total resultant of the forces and the total resultant of the moments on both the inlet and the exhaust connections of the turbine satisfies the following conditions:
For the combination of dead weight and maximum thermal expansion, Inlet-----F M
Exhaust---F M
2,520 7,570 3,310 9,930 lbs.
ft. lbs.
lbs.
ft. lbs.
For the combination of dead weight, maximum thermal expansion and OBE (E),
Inlet-----F M
Exhaust---F M
8,000 lbs, but not to exceed 5,000 lbs.
20,000 ft. lbs.
25,000lbs, but not to exceed 11,500 lbs.
20,000ft. lbs.
For the combination of dead weight, maximum thermal expansion, and SSE (E'),
Inlet-----F M
Exhaust---F M
C-3-52 12,000 lbs, but not to exceed 7,500 lbs.
30,000 ft. lbs.
37,500 lbs, but not to exceed 17,250 lbs.
30,000 ft. lbs.
03/08/01
Statement of Criteria
- 3.
Continued
- 4.
The stresses in the turbine anchor bolts (turbine to baseplate, and baseplate to foundation) due to the combination of the OBE (E) acting on the turbine while running plus the total piping loads [weight, thermal &
OBE (E)] does not exceed the allowable tensile stress nor the allowable shear stress for the bolting materials as specified in the ASME Boiler & Pressure Vessel Code,Section VIII.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine (Cont'd)
Method of Analysis 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. Vertical forces on the anchor bolts, subjecting them to tension, are the sum of the following:
a)
Weight of the turbine assembly times the vertical component of acceleration, b)
The vertical pipe force reactions, c)
The pipe moment reactions tending to tip the turbine and subject the bolting to tension Horizontal forces on the anchor bolts, subjecting them to shear 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-3-53 03/08/01
Statement of Criteria
- 5. The stresses in the turbine anchor 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.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA HPCI Turbine (Cont'd)
Method of Analysis
- 5. Same as analysis under (4), above.
C-3-54 03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine 0.75 g horizontal 1.50 g horizontal 0.07 g vertical 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
- 1. Closure bolting is designed to contain the internal design pressure of the turbine casing without exceeding the allowable working stress of the bolting material. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
- 2. The minimum wall thickness of the turbine casing limits stress to the allowable working stress when subjected to design pressure plus corrosion allowance. Allowable stresses are in accordance with ASME B&PV Code,Section VIII.
Method of Analysis
- 1. Bolting loads and stresses are calculated in accordance with the "Rules for Bolted Flange Connections",
ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II.
- 2. Stresses in the various pressure containing portions of the turbine casing are calculated according to the rules of Part UG,Section VIII, of the ASME Boiler & Pressure Vessel Code.
C-3-55 Results
- 1. Maximum Allowable Stress 20,000 psi
- 2. Maximum Allowable Stress 17,500 psi 03/08/01
Statement of Criteria
- 3. The forces and moments imposed by the attached piping on the turbine inlet and exhaust connections satisfy the following conditions:
The resultant force and moment from the combination of dead weight and thermal expansion are less than that stipulated by the equipment vendor.
The resultant force and moment from the combination of dead weight, thermal expansion and OBE or SSE is less than that demonstrated acceptable by detailed seismic analysis of the equipment.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine (Cont'd)
Method of Analysis
- 3. The total resultant of the forces and the total resultant of the moments on both the inlet and the exhaust connections of the turbine satisfy the following conditions:
For the combination of dead weight and maximum thermal expansion, Inlet-----F M
Exhaust---F M
870 2,620 2,000 6,000 lbs.
ft. lbs.
lbs.
ft. lbs.
For the combination of dead weight, maximum thermal expansion, and OBE (E),
Inlet-----F M
Exhaust---F M
1,200 3,000 18,000 6,000 lbs lbs, but not to exceed 8,370 lbs.
ft. lbs.
For the combination of dead weight, maximum thermal expansion, and SSE (E'),
lbs.
ft. lbs.
Inlet-----F M
Exhaust---F M
1,800 4,500 27,000 9,000 lbs, but not to exceed 12,555 lbs.
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
Statement of Criteria
- 4. The stresses in the turbine anchor bolts (turbine to baseplate) due to the combination of the OBE (E) acting on the turbine while running plus the total piping loads [weight, thermal and SSE (E)] shall not exceed the allowable tensile stress nor the allowable shear stress for the bolting materials as specified in the ASME Boiler and Pressure Vessel Code,Section VIII.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine (Cont'd)
Method of Analysis Typical acceptable area on the force-moment diagram is indicated below:
max....,)
,____..,.-"-'-:...-;..--------------- M
- 4. Vertical forces on the anchor bolts are the sum of the following:
a)
Weight of the turbine assembly times the vertical component of acceleration, b)
The vertical pipe force reactions, c)
The pipe moment reactions tending to tip the turbine and subject the bolting to tension.
Horizontal forces on the anchor bolts, subjecting them to shear 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
Statement of Criteria
- 5. The stresses in the turbine anchor 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.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA RCIC Turbine (Cont'd)
Method of Analysis
- 5. Same as analysis under (4), above.
C-3-58 03/08/01
CRITERIA
- 1.
Body Minimum Wall Thickness Load:
Design pressure and temperature Primary membrane stress limit:
S = 7,000 psi per ASA Bl6.5
- 2.
Cover Minimum Thickness Loads:
Design pressure and temperature Design bolting load Gasket load Primary stress limit:
Allowable working stress per ASME Section VIII.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Isolation Valves METHOD OF ANALYSIS Minimum wall thickness in the cylindrical portions of the valve is calculated using the following formula:
where:
s p
d C
where:
t d
C s
w hC C1 1.5 [
Pd
+ C]
2S-1.2P allowable stress of 7000 psi primary service pressure, 655 psi inside diameter of valve at section being considered, inches corrosion allowance of 0.12 inches
]
1/2
[
CP 1.78 WhG
+ C t=d-+
3 1
S Sd minimum thickness, inches diameter or short span, inches attachment factor allowable stress, psi total, bolt load, pounds gasket moment arm, inches corrosion allowance, inches C-3-59 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
t
- 1. 74 in.
t 4.50 in 03/08/01
CRITERIA
- 3.
Cover Flange Bolt Area:
Loads:
Design pressure and temperature Gasket load Stem operational load Seismic load -
E' Bolting stress limit:
Allowable working stress per ASME Nuclear Pump & Valve Code, Class I.
- 4.
Body Flange Thickness &
Stress Loads:
Design pressure & temperature Gasket load Stem operational load Seismic load -
E' Flange Stress Limits:
fHr
~Rt fT 1.5 Sm per ASME Nuclear Pump
& Valve Code, Class I
- 5.
Valve Disc Thickness Load:
Design pressure & temperature Primary bending stress limit:
Allowable working stress per ASME Section VIII.
USAR TABLE C-3-7 (Cont'd)
Main Steam Isolation Valves (Cont'd)
METHOD OF ANALYSIS Total, bolting loads and stresses are calculated in accordance with "Rules for Bolted Flange Connections" -
ASME Boiler & Pressure Vessel Code,Section VIII, Appendix II, except that the stem operational load and seismic loads are included in the total load carried by bolts. The horizontal and vertical seismic forces are applied at the mass center of the valve operator assuming that the valve body is rigid and anchored Flange thickness and stress are calculated in accordance with "Rules for Bolted Flange Connections" -
ASME Boiler & Pressure Vessel Code,Section VIII, Appendix II, except that the stem operational load and seismic loads are included in the total load carried by the flange. The horizontal and vertical seismic forces are applied at the mass center of the valve operator assuming that the valve body is rigid and anchored.
Max Sr where:
w t
a b
m 3W 2 Jr t 2 2 a 2 (m + 1) log ~
+
a 2 (m - 1) -
b 2 (M - 1) I a 2 (rn + 1) +
b 2 (m - 1) total applied load, lb.
thickness of disc, in.
outer radius of disc inner radius (fixed) of disc 3.85, reciprocal of Poisson's ratio C-3-60 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
s s
s s
s 33,100 psi
@ 5 7 5°F 35,000 psi 35,000 psi 35,000 psi 23,300 psi 03/08/01
CRITERIA
- 6.
Valve Operator Supports Loads:
Design pressure & temperature Stem operational load Equipment dead weight Seismic load -
E' Support Rod Stress Limit:
Allowable working stress per ASME Section VIII.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Isolation Valves (Cont'd)
METHOD OF ANALYSIS The valve assembly is analyzed assuming that the valve body is an anchored, rigid mass and that the specified vertical and horizontal seismic forces are applied at the mass center of the operator assembly, simultaneously with operating pressure plus dead weight plus operational loads. Using these loads, stresses and defelections are determined for the operator support components.
C-3-61 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
s 20,000 psi 03/08/01
CRITERIA
- 1.
Inlet Nozzle Wall Thickness Load:
1.1 x Design Press. @ 600°F Primary Membrane Stress Limit:
Allowable stress intensity as defined by ASME Standard Code for Pumps & Valves for Nuclear Power
- 2.
Valve Disc Thickness Load:
1.1 x Design press. @ 600°F Diagonal Shear Stress Limit:
0.6 x allowable stress inten-sity as defined by ASME Standard Code for Pumps &
Valves for Nuclear Power.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Stearn Safety Valves METHOD OF ANALYSIS PR
+ C SE-0.6P where:
t s
p R
E C
w Ss = A where:
and:
w A
p A1 A
s R
Rl min. required thickness, inches allowable stress, psi 1.1 x design press., psi internal radius, inches joint efficiency corrosion allowable, inches PA1 A
shear load, lb.
shear area, in2 1.1 x design press, psi disc area, in 2 TCS (R + R1 )
slope of frustrurn of shear cone, in.
radius at base of cone, in.
radius at top of cone, in.
C-3-62 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
t 0.143 in.
Ss 12,780 psi 03/08/01
CRITERIA
- 3.
Inlet Flange Bolt Area Loads:
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 Loads:
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 USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
METHOD OF ANALYSIS Total bolting loads and stresses are calculated in accordance with procedures of Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code.
Flange thickness and stresses are calculated in accordance with procedures of Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code.
C-3-63 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Sb= 27,700 psi SH SR ST 27,300 psi 27,300 psi 27,300 psi 03/08/01
CRITERIA
- 5.
Valve Spring-Torsional Stress Loads:
W1 Set point load, lbs W2 Spring load at maximum lift, lbs.
Torsional Stress Limit:
0.67 x torsional elastic limit when subjected to a load of W1
- 0.90 x torsional elastic limit when subjected to a load of W2 *
- 6.
Yoke Rod Area Load:
Spring load at maximum lift Primary Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Standard Code for Pumps & Valves for Nuclear Power USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
METHOD OF ANALYSIS Smax where:
Smax p
D d
C where:
A F
Sm 8PD [4C-1 + 0.615]
Hd3 4C-4 C
torsional stress, psi W1 or W2 = spring load, lbs mean diameter of coil, in.
diameter of wire, in.
D = correction factor D
A F
2Sm required area per rod, in2 total spring load, lbs allowable stress, psi C-3-64 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Set Point S = 62,500 psi Max. Lift S = 93,750 psi A
0.575 in2 03/08/01
CRITERIA
- 7.
Yoke Bending & Shear Stresses Load:
Spring load at maximum lift Bending & Shear Stress Limits:
Bending - allowable stress intensity, Sm, per ASME Nuclear Pump & Valve Code Shear -
0.6 x allowable stress intensity, 0.6 Sm, per ASME Nuclear Pump &
Valve Code.
- 8.
Body Minimum Wall Thickness Load:
Primary service pressure Primary Stress Limit:
Allowable stress, 7,000 psi, in accordance with ASA Bl6.5.
- 9.
Inlet Nozzle Combined Stress Loads:
Spring load at maximum lift Operational load Seismic load -
E' Combined Stress Limit:
1.5 x allowable stress intensity, 1.5 Sm, per ASME Code for Pumps & Valves for Nuclear Power.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
METHOD OF ANALYSIS M
V A
Sb z, Ss where:
Sb Ss M
z V
A bending stress, psi shear stress, psi bending moment, in. lbs.
section modulus, in3 vertical shear, lb.
shear area, in2 t = 1.5 [
Pd
] + C 2S 1.2P where:
t s
p d
required thickness, in.
allowable stress, 7,000 psi primary service pressure, 150 psi inside diameter of valve at sec-tion being considered, in.
s F1 + F2 + M1 + M2 where:
s F1 F2 A
M1 M2 A
Z combined bending & tensile stress, maximum spring load, lb.
vertical component of reaction thrust, lb.
cross section area of nozzle, in2 moment resulting from horizontal component of reaction, lb. in.
moment resulting from horizontal C-3-65 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Sb 18,200 psi Ss = 10,900 psi Body Bowl t
= 0.311 in.
Inlet Nozzle t = 0.218 in.
Outlet Nozzle t = 0.250 in.
s 27,300 psi 03/08/01
CRITERIA
- 10. Spindle Diameter Load:
Spring load at maximum lift Spindle Column Load Limit:
0.2 x critical buckling load
- 11. Spring Washer Shear Area Load:
Spring load at maximum lift Shear Stress Limit:
0.6 x allowable stress in-tensity, 0.6 Sm, per ASME Nuclear Pump & Valve Code.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Safety Valves (Cont'd)
METHOD OF ANALYSIS Fe= 1/EI L2 where:
Fe E
I L
F Ss = A where:
Ss F
A critical buckling load, lb.
modulus of elasticity, psi moment of inertia, in4 length of spindle in compression, in.
shear stress, psi spring load, lb.
shear area, in2 C-3-66 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Load Limit (0.2Fc)
F = 85,900 lb.
Ss 15,960 psi 03/08/01
CRITERIA
- 1.
Body Minimum Wall Thickness Load:
Design Pressure & Temperature Primary Membrane Stress Limit:
Allowable working stress as defined by USAS B16.5 (7,000 psi@ primary service pres-sure).
- 2.
Bonnet Cap & Pilot Base Minimum Thickness Loads:
Design pressure & temperature Gasket load Primary Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Standard Code for Pumps & Valves for Nuclear Power.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Relief Valves METHOD OF ANALYSIS t = 1.5 [
Pd
] + C 2S-1.2P where:
t s
p d
C minimum required thickness, in.
allowable stress, 7,000 psi primary service pressure, 655 inside diameter of valve at sec-tion being considered, in.
corrosion allowance, 0.12 in.
[
]
1/2 d
CP 1.78 WhG t =
+
3
+ C1 Sm Smd where:
t d
C p
Sm w
hG C1 minimum required thickness in.
diameter or short span, in.
attachment factor, ASME Section VIII design pressure, psi allowable stress, psi total bolt load, lb.
gasket moment arm, in.
corrosion allowance, 0.12 in.
C-3-67 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Main Body t = 0.625 in.
Bonnet t = 0.287 in.
Bonnet Cap t = 1.0 in.
Pilot Base t = 2.219 in.
03/08/01
CRITERIA
- 3.
Flange Bolt Area -
Inlet Flange, Outlet Flange, Body to Bonnet, Bonnet to Base Loads:
Design pressure & temperature Gasket load Operational load SSE (EI)
Bolting Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Stand-ard Code for Pumps & Valves for Nuclear Power.
- 4.
Flange Thickness -
Inlet, Out-let, Bonnet Flanges Loads:
Design pressure & temperature Gasket load Operational load SSE (EI)
Flange Stress Limits,
_§_H,
_§_R,
_§_T 1.5 Sm per ASME Nuclear Pumps and Valve Code.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Relief Valves (Cont'd)
METHOD OF ANALYSIS Total bolting loads and stresses are calculated in accordance with procedures of Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code.
Flange thickness and stresses are calculated in accordance with procedures of Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code.
C-3-68 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Body to Base Ab = 10. 2 6 in 2 Bonnet to Cap Ab = 1. 4 52 in2 Inlet Flange Ab = 13. 9 in2 Outlet Flange Ab = 12. 2 in2 Body to Base SH= 26,250 psi SR = 26,250 psi ST = 26,250 psi Cap to Bonnet SH= 26,250 psi SR = 26,250 psi ST = 26,250 psi Inlet Flange SH = 26,250 psi SR = 26,250 psi ST = 26,250 psi 03/08/01
CRITERIA
- 5.
Valve Disc Thickness &
Stress Load:
Design pressure & temperature Primary Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Standard Code for Pumps & Valve for Nuclear Power.
- 6.
Inlet Nozzle Diameter Thick-ness & Stress Loads:
Design pressure & temperature Operational load SSE (EI)
Primary Stress Limit:
1.5 x allowable stress in-tensity, 1.5 Sm, as defined by ASME Standard Code for Pumps &
Valves for Nuclear Power.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Main Steam Relief Valves (Cont')
METHOD OF ANALYSIS Sr St where:
Sr St V
p R
T 3 (3 + u) PR2 8 t2 radial stress, psi tangential stress, psi Poisson's ratio design pressure, psi radius of disc, in.
thickness of disc, in.
s F1 + F2 + M1 + M2 where:
s F1 F2 A
M1 M2 A
Z combined bending and tensile stress, psi vertical load due to design pressure, lb.
vertical component of reaction thrust, lb.
cross section area of nozzle, in2 moment resulting from horizontal reaction, in. lb.
moment resulting from horizontal seismic force at mass center of valve, in. lb.
C-3-69 ALLOWABLE STRESS OR MIN. THICKNESS REQD.
Outlet Flange SH= 26,250 psi SR = 26,250 psi ST = 26,250 psi Disc Stress Sm = 15, 8 0 0 psi Inlet Nozzle Stress S = 26,250 psi 03/08/01
CRITERIA
- 1.
Body Minimum Wall Thickness Load:
Design pressure and tempera-ture Primary membrane stress limit:
S
=
7,000psiperASAB16.5
- 2.
Cover Minimum Thickness Loads:
Design pressure and tempera-ture Design bolting load Gasket load Primary stress limit:
Allowable working stress per ASME Section VIII.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Valves METHOD OF ANALYSIS Minimum wall thickness in the cylindrical portions of the valve are calculated using the following formula:
t = 1.5 [
Pd
] + C 2S-1.2P where:
s p
d allowable stress of 7000 psi primary service pressure, 655 psi inside diameter of valve at section being considered, inches C
corrosion allowance of 0.12 inches
]
1/2
[
CP 1.78 Wha
+ C1 d -+
3 s
Sd where:
t d
C s
w he=
C1 =
minimum thickness, inches diameter of short span, inches attachment factor allowable stress, psi total, bolt load, pounds gasket moment arm, inches corrosion allowance, inches C-3-70 ALLOWABLE STRESS OR MIN. THICKNESS Body Wall Thickness t = 1.875 in.
Valve Cover Thickness and Stress t = 5.469 in.
Sallow = 17,800 psi 03/08/01
CRITERIA
- 3.
Cover Flange Bolt Area Loads:
Design pressure and tempera-ture Gasket load Stem operational load Seismic load -
DBE Bolting stress limit:
Allowable working stress per ASME Nuclear Valve and Pump Code, Class I.
- 4.
Body Flange Thickness &
Stress Loads:
Design pressure & temperature Gasket load Stem operational load Seismic load -
_§_Hr
_§_R,
_§_T 1.5 Sm per ASME Nuclear Pump
& Valve Code, Class I USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Valves (Cont'd)
METHOD OF ANALYSIS Total, bolting loads and stresses are calculated in accordance with "Rules for Bolted Flange Connections" -
ASME Boiler &
Pressure Vessel Code,Section VIII, Appendix II, except that the stem operational load and seismic loads are included in the total load carried by bolts. The horizontal and vertical seismic forces are applied at the mass center of the valve operator assuming that the valve body is rigid and anchored.
Flange thickness and stress are calculated in accordance with "Rules for Bolted Flange Connections" -
ASME Boiler & Pressure Vessel Code,Section VIII, Appendix II, except that the stem operational load and seismic loads are included in the total load carried by the flange. The horizontal and vertical seismic forces are applied at the mass center of the valve operator assuming that the valve body is rigid and anchored.
C-3-71 s
ALLOWABLE STRESS OR MIN. THICKNESS 30,900 psi@ 575°F s
26,700 psi 03/08/01
CRITERIA
- 5.
Valve Disc Thickness Load:
Design pressure & temperature Primary heading stress limit:
Allowable working stress per ASME Section VIII.
- 6.
Valve Operator Supports Loads:
Design pressure & temperature Stem operational load Equipment dead weight Seismic load -
SSE Support Rod Stress Limit:
Allowable working stress per ASME Section VIII.
USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Recirculation Valves (Cont'd)
METHOD OF ANALYSIS Sr St where:
Sr St V
p R
t 3 (3 + v) PR2 8 t2 radial stress, psi tangential stress Poisson's ratio design pressure, psi radius of disc, inches thickness of disc, inches The valve assembly is analyzed assuming that the valve body is an anchored, rigid mass and that the specified vertical and horizontal seismic forces are applied at the mass center of the operator assembly, simultaneously with operating pressure plus dead weight plus operational loads. Using these loads, stresses and deflections are determined for the operator support components.
C-3-72 ALLOWABLE STRESS OR MIN. THICKNESS s
17,800 psi s
18,000 psi 03/08/01
OBE (E) Coefficients SSE (E') Coefficients USAR TABLE C-3-7 (Cont'd)
LOADING CRITERIA Standby Liquid Control Tank 0.33 g horizontal 0.66 g horizontal 0.07 g vertical 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
- 1. The minimum thickness of the shell plates is computed from the stress on the vertical joints using a joint efficiency of 0.85. The corrosion allowance is added to the minimum thickness to obtain design thickness. In no case is the nominal thickness of shell plates less than 3/16".
The allowable stress is in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
- 2. The dead weight of the tank, the seismic acceleration [SSE (E')] on its center of gravity and pipe reaction on the outlet nozzle from maximum thermal expansion and SSE (E') does not produce excessive shell stress.
Method of Analysis
- 1. The following formula is used in calculating the minimum thickness of the shell plate t =
where:
t D
H G
fo=
E =
2.6(D) (H-1) (G) fo E minimum thickness -
inches nominal inside diameter of tank-ft height from bottom of tank to overflow level-ft specific gravity of fluid Allowable design stess-psi Joint efficiency dimensionless
- 2. The dead weight of shell, liquid and attachments, the seismic bending moment, and the pipe reaction forces and moments does not produce tensile stresses in excess of allowable or compressive (buckling) stresses in excess of 1/3 of yield strength.
C-3-73 Results
- 1.
Minimum Thickness 0.0154"
- 2.
Allowable Stress Tensile 17,500 psi Compressive 10,000 psi 03/08/01
3.3 3.3.1 USAR Method of Analysis and Implementation of Criteria 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
- 5.
- 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 3.3.3.1 Piping 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. [2oi 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 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:
C-3-78 03/08/01
. I I
Type 1 Type 2 Type 3 USAR 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.
Penetrations have no provisions for thermal expansion (refer to USAR Figure V-2-4).
Limit stops are not required for Type 2 penetrations.
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 3.3.3.5.1 Other Piping Analyses 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 Mssam z
s i
Pressure Loadings Applied Moments due to Deadweight Loadings Applied Moments due to SSE Seismic Intertial Loadings 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 Piping Section Modulus Allowable Primary Stress limit per the B31.l Code Allowable Expansion Stress range per B31.l Code 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-based 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.
C-3-83 10/02/14
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) r69l, as modified and supplemented by the U.S. Nuclear Regulatory Commission Supplemental Safety Evaluation Report
( SSER) No. 2 r7oi and SSER No. 3 [7ll, 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.
C-3-84 10/02/14
Notes:
( 1)
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).
The GIP method should be used environments and when concurrent loads from
only for equipment located in mild vibratory loads, e.g., hydrodynamic with SSE loads are less than the (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, RCIC turbines are evaluated
and The 10/02/14
USAR structural design criteria and method of analysis for these pumps and turbines are described in Table C-2-7.
The Plant Unique Analysis Report describes additional analyses performed on this equipment. The results of the analysis for these pumps and turbines show that the applicable acceptance criteria have been met.
Reactor Recirculation Pumps The Reactor Recirculation pumps are evaluated for the expected loading conditions associated with both normal operating and seismic conditions. The structural design criteria and method of analysis for the key components of these pumps are described in Table C-2-7. The results of the analysis for the Reactor Recirculation pumps show that the applicable acceptance criteria have been met.
Standby Liquid Control Tank and Pumps The Standby Liquid Control tank and pumps are evaluated for the expected loading conditions.
The structural design criteria and method of analysis for the key components of these pumps are described in Table C-2-7.
The results of the analysis for the Standby Liquid Control Pumps show that the applicable acceptance criteria have been met.
Main Steam Isolation Valves, Safety /Relief Valves, and Reactor Recirculation Valves The Main Steam Isolation Valves, Safety /Relief Valves, and the Reactor Recirculation Valves are evaluated for the expected loading conditions associated with both normal operating and seismic conditions. The results of the analysis for these valves show that the applicable acceptance criteria have been met.
C-3-86 10/02/14
4.0 USAR REFERENCES FOR APPENDIX C
- 1.
Q/A 12.37; Amendment 9.
- 2.
"Design of Concrete Structures,"
by
- Winter, O'Rourke,
- Nilson, 7th Edition, McGraw-Hill Book New York, 1964.
- Urguhart, Company,
- 3.
"Concrete Manual," U.S. Department of the Interior, Bureau of Reclamation, 7th Edition, U.S.
Government Printing Office, 1966, page 48.
- 4.
- 5.
- 6.
- 7.
- 8.
- 9.
- 10.
- 11.
- 12.
- 13.
- 14.
- 15.
- 16.
- 17.
- 18.
- 19.
- 20.
- 21.
- 22.
- 23.
Q/A 12.8; Q/A 12.10; Q/A 12.5; Q/A 12.51; Q/A 12.7; Deleted.
Deleted.
Deleted.
Q/A 12.17; Q/A 12.12; Q/A 12.18; Q/A 12.19; Q/A 12.20; Q/A 12.26; Q/A 12.50; Q/A 12.60; Q/A 12.59; Q/A 12.28; Q/A 12.52; Q/A 12.46; Amendment 9.
Amendment 9.
Amendment 9.
Amendment 13.
Amendment 9.
Amendment 9.
Amendment 9.
Amendment 9.
Amendment 9.
Amendment 9.
Amendment 9.
Amendment 13.
Amendment 13.
Amendment 13.
Amendment 13.
Amendment 14.
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).
C-4-1 03/08/01
- 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."
- 40.
NPPD Calculation NEDC 96-006, "Estimate of Steam Tunnel's HELB."
- 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.
C-4-2 03/08/01
- 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 Modifications," No.
- Report, "Torus Support System 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 NEDO-21888, General November, 1981.
Program Load Electric Definition Report,"
- Company, Revision 2,
- 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.
- 57.
DC 83-001, "Low-Low Set and Lower MSIV Water Level Trip."
- 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 Containment Long-Term
- Program, Structural Docket 50-298," dated January 20, 1984.
the Mark I
- Review, NPPD
- 62.
Q/A 12.31; Amend. 14.
- 63.
Letter from Jay M. Pilant (NPPD)
Subject "IE Bulletin No. 79-02 C-4-3 to Karl V. Seyfri t (NRC),
Pipe Support Base Plate 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)
Subject "IE Bulletin No. 79-02 Support Base Plate Designs Using Bolts," dated December 10, 1979.
to Karl V. Seyfrit (NRC),
(Revision No. 2)
Pipe Concrete Expansion Anchor
- 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 Plant-Specific Safety Program Implementation No. M69439).
NPPD dated Evaluation at Cooper September 30,
- 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,
"Cooper Nuclear
- Station, USI A-46 Seismic Evaluation Report" May 1996.
- 73.
- 74.
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.
Engineering Evaluation 10-024, Evaluation."
C-4-4 "Reactor Building Up-Rate 11/22/10