ML23291A389
| ML23291A389 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/12/2023 |
| From: | Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23291A105 | List:
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| References | |
| PLA-8081 | |
| Download: ML23291A389 (207) | |
Text
SSES-FSAR Text Rev. 74 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 3.9.1.1 Design Transients This section shows the transients which are used in the design of the ASME Boiler and Pressure Vessel Code (ASME Code) Class 1 core support, reactor internals, and control rod drive (CRD) components. The number of cycles or events for each transient are included.
The design transients shown in this section are included in the design specifications for the components. Transients or combinations of transients are classified with respect to the component operating condition categories identified as "normal," "upset," "emergency,"
"faulted," or "testing" in the ASME Code if applicable. (The first four conditions correspond to Service Levels A, B, C, and D, respectively for those components constructed to the 1976 or later Edition of the ASME Code.)
3.9.1.1.1 Control Rod Drive Transients The normal and test service load cycles used for design purposes for the 40 year life of the Control Rod Drive (CRD) are as follows:
Transient Category Cycles
- 1. Reactor startup/shutdown Normal/upset 120
- 2. Vessel pressure tests Normal/upset 130
- 3. Vessel overpressure tests Normal/upset 10
- 5. Operational scrams Normal/upset 300
- 6. Jog cycles Normal/upset 30,000
- 7. Shim/drive cycles Normal/upset 1,000 FSAR Rev. 70 3.9-1
SSES-FSAR Text Rev. 74 In addition to the above cycles, the following have been considered in the design of the CRD.
Transient Category Cycles
- 8. Operating Basis Normal/upset 10 Earthquake* (OBE)
- 9. Safe Shutdown Faulted 1 Earthquake** (SSE)
- 10. Scram with inoperative buffer Normal/upset 10
- 11. Scram with stuck control blade Normal/upset 1 All ASME Code Class 1 components of the CRD have been analyzed according to the ASME Code.
The capability of the CRD's to withstand the emergency and faulted conditions is verified by test rather than analysis.
3.9.1.1.2 CRD Housing and Incore Housing Transients The number of transients, their cycles, and classification as considered in the design and fatigue analysis of the CRD housing and incore housing are as follows:
Transient Category Cycles
- 1. Normal startup & shutdown Normal/upset 120
- 2. Vessel pressure tests Normal/upset 130
- 3. Vessel overpressure tests Normal/upset 10
- 4. Interruption of feedwater flow Normal/upset 80
- 5. Scram Normal/upset 200
- 6. OBE Normal/upset 10
- 7. SSE Faulted 1
- 8. Stuck rod scram Normal/upset 1 The frequency of occurrence of this transient would indicate emergency category. However, for conservatism, this OBE condition was analyzed as an upset condition. Ten peak OBE cycles for each occurrence are postulated.
SSES is a faulted condition; however, in the stress analysis report, it was treated as emergency with lower stress limits.
FSAR Rev. 70 3.9-2
SSES-FSAR Text Rev. 74 Transient Category Cycles
- 9. Scram no buffer Normal/upset 10 3.9.1.1.3 Hydraulic Control Unit Transients The normal and test service load cycles used for the design and fatigue analysis for the 40 year life and the Hydraulic Control Unit (HCU) are as follows:
Transient Category Cycles
- 1. Normal startup and shutdown Normal/upset 120
- 2. Vessel pressure tests Normal/upset 130
- 3. Vessel overpressure tests Normal/upset 10
- 4. Scram tests (cold) Normal/upset 300
- 5. Operational scrams (hot) Normal/upset 300
- 6. Jog cycles Normal/upset 30,000
- 7. Drive cycles Normal/upset 1,000
- 10. SSE Faulted 1 3.9.1.1.4 Core Support and Reactor Internals Transients The events and number of cycles used for the design and fatigue analysis for the 40-year life of the core support and internals are shown in Table 3.9-1.
3.9.1.1.5 Main Steam System Transients The transients considered in the stress analysis of the main steam piping are included in Table 3.9-4.
3.9.1.1.6 Recirculation System Transients The transients considered in the stress analysis of the recirculation piping are included in Table 3.9-4.
3.9.1.1.7 Reactor Assembly Transients The reactor assembly includes the reactor pressure vessel, support skirt, shroud support, and shroud plate. The cycles listed in Table 3.9-1 were specified in the reactor assembly design and fatigue analysis.
FSAR Rev. 70 3.9-3
SSES-FSAR Text Rev. 74 Reactor design cycle or transient limits are as follows:
Transient Design Cycle Heatup and Cooldown 70°F to 551°F to 70°F (Envelopes items 3, 10, 13, 16b, 16d from Table 3.9-1)
Reactor Trip cycle 100% to 0% Rated Thermal Power (Item 9 from Table 3.9-1)
Hydrostatic Pressure and Leak Tests Pressurized 930 psig and 1250 psig (Item 2 from Table 3.9-1) 3.9.1.1.8 Main Steam Isolation Valve Transients The main steam isolation valves are designed for the following service conditions and thermal cycles:
Transient Category Cycles
- 1. Pre-op @100°F/hr Normal/upset 150
- 2. Startup (heating 100°F/hr) Normal/upset 120
- 3. Shutdown
- a. cooling cycles @100°F/hr Normal/upset 120 540°F to 375°F
- b. cooling cycles @270°F/hr Normal/upset 120 375°F to 330°F
- c. cooling cycles @100°F/hr Normal/upset 120 330°F to 100°F
- 4. Scram cooling cycles @100°F/hr Normal/upset 180
- 5. Emergency and faulted transients
- a. 546°F to 281°F in 15 sec Emergency/faulted 1
- b. 546°F to 375°F in 3.3 min Emergency/faulted 1 375°F to 281°F @300°F/hr Emergency/faulted 1 c1. 546°F to 375°F in 10 min Emergency/faulted 8 FSAR Rev. 70 3.9-4
SSES-FSAR Text Rev. 74 Transient Category Cycles c2. 375°F to 281°F @100°F/hr Emergency/faulted 8 d1. 546°F to 583°F in 2 sec Emergency/faulted 1 d2. 583°F to 538°F in 30 sec Emergency/faulted 1 d3. 538°F to 400°F @100°F/hr Emergency/faulted 1 d4. 400°F to 546°F @100°F/hr Emergency/faulted 1 e1. 561°F to 500°F in 7 min Emergency/faulted 10 e2. 500°F to 400°F @100°F/hr Emergency/faulted 10 e3. 400°F to 546°F @100°F/hr Emergency/faulted 10 3.9.1.1.9 Safety/Relief Valves Transients The transients used in the analysis of the safety/relief valves are as follows:
Transient Category Cycles
- 1. Pre-op and in-service testing (100°F/hr). Normal/upset 150
- 2. Startup (100°F/hr) and pressure increase Normal/upset 120 (0 psig to 1000 psig).
- 3. Shutdown (100°F/hr, pressure decrease Normal/upset 120 to 0 psig).
- 4. Scram. Normal/upset 180
- 5. System pressure and temperature decay from Emergency/faulted 1 1000 psig and 546°F to 35 psig and 281°F within 15 sec.
- 6. System temperature change from 546°F to Emergency/faulted 1 375°F within 3.3 mins and from 375°F to 281°F at a rate of 300°F/hr. Pressure change from 1000 psig to 35 psig
- 7. System temperature change from 546°F to Emergency/faulted 8 375°F within 10 min. and from 375°F to 281°F at a rate of 100°F/hr. Pressure change from 1000 psig to 35 psig.
FSAR Rev. 70 3.9-5
SSES-FSAR Text Rev. 74 Transient Category Cycles
- 8. System temperature change from 546°F to Emergency/faulted 1 583°F within 2 sec, from 583°F to 538°F within 30 sec, and from 538°F to 400°F and return to 546°F at a rate of 100°F/hr. Pressure change from 1000 psig thence to 1350 psig , thence to 240 psig and return to 1000 psig.
- 9. System temperature changes, greater than Emergency/faulted 10 30°F, from 561°F to 500°F within 7 min. and from 500°F to 400°F and return to normal operating temperature at 546°F at a rate of 100°F/hr. Pressure change from 1000 psig to 1180 psig to 240 psig and return to normal operating of 1000 psig.
Paragraph NB3552 of ASME III code excludes various transients and provides means for combining those which are not excluded. Review and approval of the equipment supplier's certified calculation provides assurance of proper accounting of the specified transients.
3.9.1.1.10 Recirculation Flow Control Valve Transients Not applicable since Susquehanna SES has no flow control valve.
3.9.1.1.11 Recirculation Pump Transients The following transients are listed in the design specification as a requirement for design considerations. However, a submitted certified analysis considering thermal stresses was not required. The vendor was required to submit a certification of compliance. The submitted certified design calculations only considered pressure transient. Nozzle piping loads were considered in accordance with the following paragraph:
"The pump case was designed to withstand secondary stresses due to piping reactions in accordance with Paragraph 452.4b of the ASME Standard Code for Pumps and Valves for Nuclear Power (1968 Draft)."
Transient Category Cycles
- 1. Heatup and cooldown at 100°F/hr Normal/upset 300
- 2. +/-29°F temperature changes Normal/upset 600
- 3. +/-50°F temperature changes Normal/upset 200
- 4. RPV pressure transients to 110% Normal/upset 1 design pressure FSAR Rev. 70 3.9-6
SSES-FSAR Text Rev. 74
- 5. SRV blowdowns Emergency 8 Emergency 1
- 6. Improper pump startup, 100°F to 552°F in 15 sec
- 7. Cooling transient 552°F to 281°F in 15 Faulted 1 sec
- 8. Hydrotest to 1300 psig Testing 130
- 9. Hydrotest to 1670 psig Testing 3 3.9.1.1.12 Recirculation Gate Valve Transients The following transients are considered in the design of the recirculation gate valves.
Transient Cycles
- 1. 50°F - 575°F - 50°F at 100°F/hr 300
- 2. +/-29°F between limits of 50°F and 575°F, instantaneous 600
- 3. +50°F between limits of 50°F and 546°F, instantaneous 200
- 4. 546°F to 375°F, instantaneous 30
- 5. 546°F to 281°F, instantaneous 2
- 6. 130°F to 546°F, instantaneous 1
- 7. 110% design pressure at 575°F 1
- 8. 1300 psi at 100°F installed hydrostatic test 130
- 9. 1670 psi at 100°F installed hydrostatic test 3 3.9.1.2 Computer Programs Used in Analysis The following sections discuss computer programs used in the analysis of specific NSSS components. (Computer programs were not used in the analysis of all components, thus, not all components are listed.) The NSSS programs can be divided into two categories.
The computer programs discussed in this section are those programs used for the original plant design. Changes to later versions of these programs or the addition of entirely new computer programs for safety related applications is controlled by procedures under our Operational Quality Assurance Program.
FSAR Rev. 70 3.9-7
SSES-FSAR Text Rev. 74 GE Programs The verification of the following GE programs has been performed in accordance with the requirements of 10CFR50, Appendix B. Evidence of the verification of input, output, and methodology is documented in GE Design Record Files.
(a) MASS (i) TSFOR (q) DYSEA (b) SNAP (MULTISHELL) (j) EZPYP (r) SPECA (c) HEATER (k) PDA (s) SEISM (d) SAP4 (l) SAP4G (e) ANSI7 (m) FTFLG01 (f) LUGSTR (n) ANSYS (g) PISYS (o) POSUM (h) RVFOR (p) BILRD Vendor Programs The verification of the following CB&I programs is assured by contractual requirements between GE and the vendor. Per the requirements, the quality assurance procedure of these proprietary programs used in the design of N-stamped equipment is in full compliance with 10CFR50, Appendix B.
(a) 711 GENOZZ (i) 928 TGRV (b) 948 NAPALM (j) 962 E0962A (c) 1027 (k) 984 (d) 846 (l) 992 GASP (e) 781 KALNINS (m) 1037 DUNHAM'S (f) 979 ASFAST (n) 1335 (g) 766 TEMAPR (o) 1606 & 1647 HAP (h) 767 PRINCESS (p) 1634 N A list of computer programs used in the BOP system components is provided in Table 3.9-5.
This list consists of computer programs that are developed and/or owned by Bechtel Power Corporation, and computer programs that are recognized and widely used in industry.
FSAR Rev. 70 3.9-8
SSES-FSAR Text Rev. 74 The Bechtel developed and/or owned computer programs are documented, verified, and maintained by Bechtel and meet the requirements of 10CFR50, Appendix B. A brief description of each of these programs is provided in Appendix 3.9A.
3.9.1.2.1 Reactor Vessel and Internals The computer programs used in the preparation of the reactor vessel stress report are identified and their use summarized in the following paragraphs.
3.9.1.2.1.1 Reactor Vessel 3.9.1.2.1.1.1 CB&I Program 711 "GENOZZ" The GENOZZ computer program is used to proportion barrel and double taper type nozzle configurations to comply with the specifications of the ASME Code,Section III and contract documents. The program will either design such a configuration or analyze the configuration input into it. If the input configuration will not comply with the specifications, the program will modify the design and redesign it to yield an acceptable result.
3.9.1.2.1.1.2 CB&I Program 948 - "NAPALM" The basis for the program NAPALM, Nozzle Analysis Program-All Loads Mechanical, is to analyze nozzles for mechanical loads and find the maximum stress intensity and location. The program analyzes at specified locations from the point of application of the mechanical loads.
At each location the maximum stress intensity is calculated for both the inside and outside surfaces of the nozzle. The program gives the maximum stress intensity for both the inside and outside surfaces of the nozzle as well as its angular location around the circumference of the nozzle from the 0° reference location. The principle stresses are also printed. The stresses resulting from each component of loading (bending, axial, shear, and torsion) are printed, as well as the loadings which caused these stresses.
3.9.1.2.1.1.3 CB&I Program 1027 This program is a computerized version of the analysis method contained in the "Welding Research Council Bulletin F107," December, 1965.
Part of this program provides for the determination of the shell stress intensities (S) at each of four cardinal points at both the upper and lower shell plate surfaces (ordinarily considered outside and inside surfaces) around the perimeter of a loaded attachment on a cylindrical or spherical vessel. With the determination of each S, there is also determined the components of that S (2 normal stresses, and one shear stress). This program provides the same information as the manual computation and the input data is essentially the input of the geometry of the vessel and attachment.
3.9.1.2.1.1.4 CB&I Program 846 This program computes the required thickness of a hemispherical head with a large number of circular parallel penetrations by means of the area replacement method in accordance with the ASME Code,Section III.
FSAR Rev. 70 3.9-9
SSES-FSAR Text Rev. 74 In cases where the penetration has a counterbore, the thickness is determined so that the counterbore does not penetrate the outside surface of the head.
3.9.1.2.1.1.5 CB&I Program 781 - "KALNINS" This program is a thin elastic shell program for shells of revolution. This program was developed by Dr. A. Kalnins of Lehigh University. Extensive revisions and improvements have been made by Dr. J. Endicott to yield the CB&I version of this program.
The basic method of analysis was published by Professor Kalnins in the Journal of Applied Mechanics, Volume 31, September, 1964, pages 467 through 476. The KALNINS thin shell program (Program 781) is used to establish the shell influence coefficient and to perform detail stress analysis of the vessel. The stresses and the deformations of the vessel can be computed for any combination of the following axi-symmetric loading:
a) Preload condition b) Internal pressure c) Thermal load 3.9.1.2.1.1.6 CB&I Program 979 - "ASFAST" ASFAST Program (Program 979) performs the stress analysis of axi-symmetric, bolted closure flanges between head and cylindrical shell.
3.9.1.2.1.1.7 CB&I Program 766 - "TEMAPR" This program will reduce any arbitrary temperature gradient through the wall thickness to an equivalent linear gradient. The resulting equivalent gradient will have the same average temperature and the same temperature-moment as the given temperature distribution. Input consists of plate thickness and actual temperature distribution. The output contains the average temperature and total gradient through the wall thickness. The program is written in FORTRAN IV language.
3.9.1.2.1.1.8 CB&I Program 767 - "PRINCESS" The PRINCESS computer program calculates the maximum alternating stress amplitudes from a series of stress values by the method in Section III of the ASME Pressure Vessel Code.
3.9.1.2.1.1.9 CB&I Program 928 - "TGRV" The TGRV program is used to calculate temperature distributions in structures or vessels.
Although it is primarily a program for solving the heat conduction equations, some provisions have been made for including radiation and convection effects at the surfaces of the vessel.
The TGRV program is a greatly modified version of the TIGER heat transfer program written about 1958 at Knolls Atomic Power Laboratory, by A. P. Bray. There have been many versions of TIGER in existence including TIGER II, TIGER II B, TIGER IV, and TIGER V, in addition to TGRV. This manual has been written for use with CB&I's version of TGRV.
FSAR Rev. 70 3.9-10
SSES-FSAR Text Rev. 74 This program utilizes an electrical network analogy to obtain the temperature distribution of any given system as a function of time. The finite difference representation of the three-dimensional equations of heat transfer are repeatedly solved for small time increments and continually summed. Linear mathematics are used to solve the mesh network for every time interval.
Included in the analysis are the three basic forms of heat transfer, conduction, radiation and convection, as well as internal heat generation.
Given any odd-shaped structure, which can be represented by a three dimensional field, its geometry and physical properties, boundary conditions, and internal heat generation rates, TGRV will calculate and give as output the steady state or transient temperature distributions in the structure as a function of time.
3.9.1.2.1.1.10 CB&I Program 962 - "EO962A" Program E0962A is one of a group of programs (E0953A, E1606A, E0962A, E0992N, E1037N, and E0984N) which are used together to determine the temperature distribution and stresses in pressure vessel components by the finite element method.
Program E0962A is primarily a plotting program. Using the nodal temperatures calculated by program E1606A or Program E0928A, and the node and element cards for the finite element model, it calculates and plots lines of constant temperature (isotherms). These isotherm plots are used as part of the stress report to present the results of the thermal analysis. They are also very useful in determining at which points in time the thermal stresses should be determined.
In addition to its plotting capability the program can also determine the temperatures of some of the nodal points by interpolation. This feature of the program is intended primarily for use with the compatible TGRV and finite element models that are generated by program E0953A.
3.9.1.2.1.1.11 CB&I Program 984 Program 984 is used to calculate the stress intensity of the stress differences, on a component level, between two different stress conditions. The calculation of the stress intensity of stress component differences (the range of stress intensity) is required by Section III of the ASME Code.
3.9.1.2.1.1.12 CB&I Program 992 - GASP The GASP computer program, originated by Prof. E. L. Wilson of the University of California at Berkeley, uses the finite element method to determine the stresses and displacements of plane or axi-symmetric structures of arbitrary geometry and is written in FORTRAN IV. For a detailed account, see the following reference document.
Wilson, E. L.; "A Digital Computer Program for the Finite Element Analysis of Solids with Non-Linear Material Properties," Aerojet General Corporation, Sacramento, California. Technical Memorandum No. 23, July 1965.
As mentioned above, the program determines the stresses and displacements of plane or axi-symmetric structures using the finite element method. The structures may have arbitrary geometry and have linear or non-linear material properties. The loadings may be thermal, mechanical, accelerational, or a combination of these.
FSAR Rev. 70 3.9-11
SSES-FSAR Text Rev. 74 The structure to be analyzed is broken up into a finite number of discrete elements or "finite-elements" which are interconnected at finite number of "nodal-points" or "nodes." The actual loads on the structure are simulated by statically equivalent loads acting at the appropriate nodes. The basic input to the program consists of the geometry of the stress-model and the boundary conditions. The program then gives the stress components at the center of each element and the displacements at the nodes, consistent with the prescribed boundary conditions.
3.9.1.2.1.1.13 CB&I Program 1037 - "DUNHAM'S" DUNHAM'S program is a finite ring element stress analysis program. It will determine the stresses and displacements of axi-symmetric structures of arbitrary geometry subjected to either axi-symmetric loads or non-axi-symmetric loads represented by a Fourier series.
This program is similar to the GASP program (CB&I 992). The major differences are that DUNHAM'S can handle non-axi-symmetric loads (which requires that each node have three degrees of freedom) and the material properties for DUNHAM'S must be constant. As in GASP, the loadings may be thermal, mechanical, and accelerational.
3.9.1.2.1.1.14 CB&I Program 1335 To obtain stresses in the shroud support, the baffle plate must be made a continuous circular plate. The program makes this modification and allows the baffle plate to be included in CB&I program 781 as two isotropic parts and an orthotropic portion at the middle (where the diffuser holes are located).
3.9.1.2.1.1.15 CB&I Programs 1606 and 1657 - "HAP" The HAP program is an axi-symmetric nonlinear heat analysis program. It is a finite element program and is used to determine nodal temperatures in a two-dimensional or axi-symmetric body subjected to transient disturbances. Programs 1606 and 1657 are identical except that 1606 has a larger storage area allocated and can thus be used to solve larger problems. The model for program 1606 is compatible with CB&I stress programs 992 and 1037.
3.9.1.2.1.1.16 CB&I Program 1634N This program is used to analyze thin cylindrical shells subjected to local loading beyond the range where Bijloard's curves are directly applicable, i.e., R/t >300.
This program computes stresses and displacements in thin walled elastic cylindrical shells subjected to mechanical loading such as radial loads, longitudinal and circumferential moments.
3.9.1.2.1.2 Reactor Internals The following programs are used in the analysis of core support structures and other safety related reactor internals: MASS, SNAP (MULTISHELL), and HEATER. These programs are described in detail in Section 4.1.
FSAR Rev. 70 3.9-12
SSES-FSAR Text Rev. 74 3.9.1.2.2 Piping The computer programs used in the analysis of NSSS piping systems are identified and summarized below:
3.9.1.2.2.1 Structural Analysis Program SAP 4 The Structural Analysis Program SAP 4, for the static and dynamic analysis of linear structural system is the result of several years research and development experience. The program has proven to be a very flexible and efficient analysis tool. The first version of the SAP Program was published in September, 1970. An improved static analysis program, namely SOLID SAP, or SAP 2, was presented in 1971. Work was then started on a new static and dynamic analysis program. The program SAP 3 was released towards the end of 1972. SAP 4 has the additional analysis capability of out-of-core direct integration for the time history analysis.
The structural systems to be analyzed may be composed of combinations of a number of different structural elements. The program presently contains the following element types:
a) three-dimensional truss element b) three-dimensional beam element c) plane stress and plane strain element, d) two-dimensional axisymetric solid, e) three-dimensional solid, f) thick shell element, g) thin plate or thin shell element, h) boundary element, i) pipe element (tangent and bend).
These structural elements can be used in a static or dynamic analysis. The capacity of the program depends mainly on the total number of nodal points in the system, the number of eigenvalues needed in the dynamic analysis and the computer used. There is practically no restriction on the number of elements used, the number of load cases or the order and bandwidth of the stiffness matrix. Each nodal point in the system can have from zero to six displacement degrees of freedom. The element stiffness and mass matrices are assembled in condensed form; therefore, the program is equally efficient in the analysis of one-, two-, or three-dimensional systems.
FSAR Rev. 70 3.9-13
SSES-FSAR Text Rev. 74 The formation of the structure matrices is carried out in the same way in a static or dynamic analysis. The static analysis is continued by solving the equations of equilibrium followed by the computation of element stresses. In a dynamic analysis the choice is between:
a) frequency calculations only, b) frequency calculations followed by response history analysis, c) frequency calculations followed by response spectrum analysis, d) response history analysis by direct integration.
To obtain the frequencies and vibration mode shapes, solution routines are used which calculate the required eigenvalues and eigenvectors directly without a transformation of the structure stiffness matrix and mass matrix. This way the program operation and necessary input data for a dynamic analysis is a simple addition to what is needed for a static analysis.
3.9.1.2.2.2 Component Analysis/ANSI 7 Application. The ANSI 7 Computer Program determines stress and accumulative usage factors for thermal weight, seismic relief valve lift and turbine stop valve closure (as applicable) conditions of loadings derived from the Structural System Analysis in accordance with NB-3600 of ASME Code Section III.
Program Organization. For Class 1 ASME Code stress analysis, the program generates and prints hoop, bending, thermal discontinuity, linear temperature gradient and nonlinear temperature gradient components of stress for each equation of subarticle NB-3600 of Section III. Load combination results from possible load sets for Class 1 stress equations. The total stress (sum of component stresses) and the stress ratio (total stress divided by appropriate stress intensity limit) is printed for each Class 1 equation. The total stress (sum of the component stresses) and the stress ratio (total stress divided by the appropriate stress intensity limit) is printed for each one of the equations 9, 10, 12, and 13 of NB-3600. The alternating stresses and usage factor are calculated per NB-3653.6.
3.9.1.2.2.3 Integral Attachment/LUGSTR The computer program "LUGSTR" was prepared to evaluate the stress in the pipe wall that are produced by loads applied to the integral attachments. The program was prepared based on the Welding Research Council Bulletin 198 including the evaluation due to stress range and fatigue analysis.
3.9.1.2.2.4 Piping Analysis Program/PISYS PISYS is a computer code specialized for piping load calculations. It utilizes selected stiffness matrices representing standard piping components, which are assembled to form a finite element model of a piping system. The technique relies on dividing the pipe model into several discrete substructures, called pipe elements, which are connected to each other via nodes called pipe joints. It is through these joints that the model interacts with the environment, and loading of the structure becomes possible. PISYS is based on the linear classical elasticity in which the resultant deformation and stresses are proportional to the loading, and the superposition of loading is valid.
FSAR Rev. 70 3.9-14
SSES-FSAR Text Rev. 74 PISYS has a full range of static and dynamic analysis options which include distributed weight, thermal expansion, differential support motion modal extraction, response spectra, and time history analysis by modal or direct integration. The PISYS program has been benchmarked against five Nuclear Regulatory Commission piping models for the option-of-response-spectrum analysis and the results are documented in a report to the Commission, "PISYS Analysis of NRC Problem," NEDO-24210, August, 1979.
3.9.1.2.2.5 Relief Valve Discharge Pipe Forces/RVFO R The relief valve discharge pipe connects the relief valve to the suppression pool. When the valve is opened, the transient fluid flow causes time dependent forces to develop in the pipe wall. This computer program computes the transient fluid mechanics and the resultant pipe forces using the method of characteristics.
3.9.1.2.2.6 Turbine Stop Valve Closure/TSFOR The TSFOR program computes the time history forcing function in the main steam piping due to turbine stop valve closure. The program utilizes the method of characteristics to compute fluid momentum and pressure loads at each change in pipe section or direction.
3.9.1.2.2.7 Piping Analysis Program/EZPYP EZPYP links the ANSI-7 and SAP program together. The EZPYP program can be used to run several SAP cases by making user specified changes to a biasic SAP pipe model. By controlling files and SAP runs the EZPYP program gives the analyst the capability to perform a complete piping analysis in one computer run.
3.9.1.2.2.8 Pipe Whip Analysis/PDA The pipe whip analysis was performed using the PDA computer program. PDA is a computer program used to determine the response of a pipe subjected to the thrust force occurring after a pipe break. The program treats the situation in terms of generic pipe break configuration, which involves a straight, uniform pipe fixed at one end and subjected to a time-dependent thrust-force at the other end. A typical restraint used to reduce the resulting deformation is also included at a location between the two ends. Nonlinear and time-independent stress-strain relations are used for the pipe and the restraint. Similar to the popular plastic-hinge concept, bending of the pipe is assumed to occur only at the fixed end and at the location supported by the restraint.
Shear deformation is also neglected. The pipe bending moment-deflection (or rotation) relation used for these locations is obtained from a static nonlinear cantilever beam analysis. Using the moment-rotation relation, nonlinear equations of motion of the pipe are formulated using an energy consideration and the equations are numerically integrated in small time steps to yield time-history information of the deformed pipe.
3.9.1.2.3 Recirculation and ECCS Pumps and Motors 3.9.1.2.3.1 Recirculation Pumps No computer programs were used in the design of the recirculation pumps.
FSAR Rev. 70 3.9-15
SSES-FSAR Text Rev. 74 3.9.1.2.3.2 ECCS Pumps and Motors 3.9.1.2.3.2.1 Structural Analysis Program/SAP4G SAP4G is used to analyze the structural and functional integrity of the ECCS pump/motor systems. This is a general structural analysis program for static and dynamic analysis of linear elastic complex structures. The finite element displacement method is used to solve the displacements and stresses of each element of the structure. The structure can be composed of unlimited number of three-dimensional truss, beam, plate, shell, solid, plate strain-plane stress and spring elements that are axisymmetric. The program can treat thermal and various forms of mechanical loading. The dynamic analysis includes mode superposition, time history, and response spectrum analysis. Seismic loading and time-dependent pressure can be treated.
The program is versatile and efficient in analyzing large and complex structural systems. The output contains displacements of each nodal point as well as stresses at the surface of each element.
3.9.1.2.3.2.2 Effects of Flange Joint Connections/FTFLG01 The flange joints connecting the pump bowl castings are analyzed using FTFLG01. This program uses the local forces and moments determined by SAP4G to perform flat flange calculations in accordance with the rules set forth in Appendix II and Section III of the ASME Code.
3.9.1.2.3.2.3 Structural Analysis of Discharge Head/ANSYS ANSYS is used to analyze the pump discharge head flange and bolting taking into account of the prying action developed by the flag face contact surface. The program is described in detail in 3.12.
3.9.1.2.3.2.4 Beam Element Data Processing/POSUM POSUM is a computer code designed to process SAP generated beam element data for pump or heat exchanger models. The purpose is to determine the load combination that would produce the maximum stress in a selected beam element. It is intended for use on RHR heat exchangers with four nozzles or ECCS pumps with two nozzles.
3.9.1.2.4 RHR Heat Exchangers 3.9.1.2.4.1 Structural Analysis Programs/SAP4G SAP4G is used to analyze the structural and functional integrity of the RHR heat exchangers.
The description of this program is provided in Subsection 3.9.1.2.3.2.1.
3.9.1.2.4.2 Shell Attachment Parameters and Coefficients/BILRD BILRD is used to calculate the shell attachment parameters and coefficients utilized in the stress analysis of the support to shell junction. The method, per Welding Research Council Bulletin No. 107, is implemented to calculate local membrane stress due to the support reaction loads on the heat exchanger shell.
FSAR Rev. 70 3.9-16
SSES-FSAR Text Rev. 74 3.9.1.2.4.3 Beam Element Data Processing/POSUM POSUM is used to process SAP generated beam element data. The description of this program is provided in Subsection 3.9.1.2.3.2.4.
3.9.1.2.5 Dynamic Loads Analysis 3.9.1.2.5.1 Dynamic Analysis Program/DYSEA DYSEA simulates a beam model in the annulus pressurization dynamic analysis. A detailed description of DYSEA is provided in Section 4.1. DYSEA employs a preprocessor program names GZAPL. GZAPL converts pressure time histories into time varying loads and forcing functions for DYSEA. The overall resultant forces and moments time histories at specified points of resolution can also be obtained from GZAPL.
3.9.1.2.5.2 Acceleration Response Spectrum Program/SPECA SPECA generates acceleration response spectrum for an arbitrary input time history of piece-wise linear accelerations, i.e., to compute maximum acceleration responses for a series of single-degree-of-freedom systems subjected to the same input. It can accept acceleration time histories from a random file. It also has the capability of generating the broadened/enveloped spectra when the special points are generated equally spaced on a logarithmic scale axis of period/frequency. This program is also used in seismic and SRV transient analyses.
3.9.1.2.5.3 Fuel Support Loads Program/SEISM SEISM02 computes the vertical fuel support loads using the component element methods in dynamics. The methodology is based on the publication "The Component Element Method in Dynamics," by S. Levy and J. P. D. Wilkinson, McGraw Hill Co., New York, 1976.
3.9.1.3 Experimental Stress Analysis When experimental stress analysis is used in lieu of analytical methods for Seismic Category I ASME Code items, the requirements for experimental testing enumerated in the ASME Code applicable for the specific components under test are applied. When testing is required for Seismic Category I non-ASME Code items, account is taken of the effects of differences in size, dimensional tolerances, and ultimate strength (or other governing material properties) between the actual and tested parts to assure that the loads obtained from tests are realistic or conservative representation of the capability of the actual structure.
The following subsections in this section list the components upon which experimental stress analysis was used.
3.9.1.3.1 Experimental Stress Analysis of NSSS Piping Components The following components have been tested to verify their design adequacy:
a) Snubbers b) Pipe whip restraints FSAR Rev. 70 3.9-17
SSES-FSAR Text Rev. 74 Descriptions of the snubber and whip restraint tests are contained in Subsections 3.9.3.4 and 3.6, respectively.
3.9.1.3.2 Seismic Category I Items Other Than NSSS Experimental testing is performed in the qualification and acceptance of snubbers, compensating struts, and honeycomb material used in energy absorbing components for pipe break.
3.9.1.4 Considerations for the Evaluation of Faulted Conditions All Seismic Category I equipment in the NSSS is evaluated for the faulted loading conditions.
However, emergency stress limits rather than faulted stress limits were used in many cases.
In essentially all cases, calculated stresses are within allowable limits. The following paragraphs in subsection 3.9.1.4 show examples of the treatment of faulted conditions for the major components on a component by component basis. Additional discussion of faulted analysis can be found in Subsections 3.9.3 and 3.9.5, and Table 3.9-2.
Subsection 3.9.2.2 and Section 3.7 discuss the treatment of dynamic loads resulting from the postulated seismic and hydrodynamic events. Section 3.9.2.5 discusses the dynamic analysis of loads affected on reactor internals equipment resulting from blowdown. Deformations under faulted conditions have been evaluated in critical areas and no cases are identified where design limits, such as clearance limits, are violated.
3.9.1.4.1 Control Rod Drive System Components 3.9.1.4.1.1 Control Rod Drives The ASME Code components of the CRD have been analyzed for abnormal conditions shown in Subsection 3.9.1.1.1.
The load criteria, calculated and allowable stresses for various operating conditions are summarized in Table 3.9-2(u).
The design adequacy of non ASME code components of the CRD has been verified by analysis and extensive testing programs on both component parts, specially instrumented prototype drives and production drives. The testing has included postulated abnormal events as well as the service life cycle listed in Subsection 3.9.1.1.1.
3.9.1.4.1.2 Hydraulic Control Unit The Hydraulic Control Unit (HCU) was analyzed for the faulted condition.
The analysis of the HCU under faulted condition loads establishes the structural integrity of the system.
Section 3.9.2.2a.2.4 discusses the dynamic qualification of the HCU.
FSAR Rev. 70 3.9-18
SSES-FSAR Text Rev. 74 3.9.1.4.2 Standard Reactor Internal Components 3.9.1.4.2.1 CR Guide Tube The maximum calculated stress on the CR Guide Tube occurs in the base during the faulted condition. The loading criteria and calculated and allowable stresses are summarized in Table 3.9-2aa.
3.9.1.4.2.2 Incore Housing The maximum calculated stress on the Incore Housing occurs at the outer surface of the vessel penetration during the faulted condition. The loading criteria and calculated and allowable stresses are shown in Table 3.9-2ab.
3.9.1.4.2.3 Jet Pump The maximum stress in the jet pump occurs in the faulted condition due to impulse loading of the diffuser during a pipe rupture and blowdown. Table 3.9-2w summarizes loading criteria, and calculated and allowable stresses.
3.9.1.4.2.4 Orificed Fuel Support A series of vertical and horizontal load tests were performed on the orificed fuel support (OFS) in order to verify the design. Results from these tests indicate that the component and seismic loading of the OFS are well below the stress limit allowables with a safety margin of 1.26 for the normal and upset conditions, and 1.5 for the faulted condition. The allowable stress limits were arrived at by applying a .65 quality factor to the ASME Code allowables of 1.5 S for the upset condition, and 1.5 x .7 Su for the faulted condition.
3.9.1.4.2.5 Control Rod Drive Housing The CRD Housing is analyzed for the faulted condition considering SSE and hydrodynamic loads. Table 3.9-2v shows that the calculated stresses are within the allowable stresses.
3.9.1.4.3 Reactor Pressure Vessel Assembly The reactor pressure vessel, support skirt, and the shroud support were evaluated using elastic analysis methods for the faulted conditions. For the support skirt and shroud support an elastic analysis was performed and buckling was evaluated for the compressive load. Table 3.9-2a lists the loading criteria and calculated and allowable stresses for the various loading combinations.
3.9.1.4.4 Core Support Structure The evaluations for faulted conditions for the core support structure are discussed in Subsection 3.9.5. The loading criteria and calculated and allowable stresses are summarized in Table 3.9-2b.
FSAR Rev. 70 3.9-19
SSES-FSAR Text Rev. 74 3.9.1.4.5 Main Steam Isolation, Recirculation Gate, and Safety/Relief Valves Tables 3.9-2g, 3.9-2h, and 3.9-2j provide a summary of the analyses of the safety/relief, main steam isolation, and recirculation gate valves, respectively.
Standard design rules, as defined in the ASME Code,Section III, are utilized in the analysis of pressure boundary components of Seismic Category I valves. Conventional elastic stress analysis is used to evaluate components not defined in the ASME code. The code allowable stresses are applied to determine acceptability of structure under applicable loading conditions including faulted condition.
3.9.1.4.6 Main Steam and Recirculation Piping For Main Steam and Recirculation System piping, elastic analysis methods are used for evaluating faulted loading conditions. The equivalent allowable stresses using elastic techniques are obtained from ASME Code Section III, Appendix F, "Rules for Evaluation of Faulted Conditions," and these are above elastic limits. Additional information on the main steam and recirculation piping and pipe-mounted equipment is in Table 3.9-2d and 3.9-2e.
3.9.1.4.7 Nuclear Steam Supply System Pumps, Heat Exchanger, and Turbines The recirculation, ECCS, RCIC, and SLC pumps, RHR heat exchangers and RCIC turbine have been analyzed for the faulted loading conditions identified in Subsection 3.9.1.1. In all cases, stresses were within the elastic limits. The analytical methods, stress limits, and allowable stresses are discussed in Subsections 3.9.2.2 and 3.9.3.1.
3.9.1.4.8 Control Rod Drive Housing Supports Design adequacy of the CRD Housing Supports is shown by comparing the total static and dynamic loads to the original design loads. The comparison shows that the hydrodynamic loads and other dynamic loads combined by SRSS are less than the original design loads.
3.9.1.4.9 Fuel Storage Racks The stress criteria, loadings, calculated stresses, and stress limits for the faulted conditions for the new fuel storage racks are shown in Table 3.9-2s. No inelastic stress analyses were used on these components.
Similar information for the spent fuel storage racks was provided by the rack vendor.
3.9.1.4.10 Fuel Assembly (Including Channels)
Seismic/ LOCA loading evaluations for channeled ATRIUM-10 fuel, ATRIUM-11 fuel, and channeled Lead Use Assemblies are located in Section 4.2. (Fuel System) 3.9.1.4.11 Refueling Equipment Refueling and servicing equipment that is important to safety is classified as essential equipment per the requirements of 10 CFR 50, Appendix A. This equipment and other equipment whose failure would degrade an essential component is defined in Section 9.1 and is classified as Seismic Category I. These components are subjected to an elastic dynamic finite FSAR Rev. 70 3.9-20
SSES-FSAR Text Rev. 74 element analysis to generate loadings. This analysis utilizes appropriate seismic floor response spectra and combines loads at frequencies up to 33 Hz for seismic and up to 60 Hz for the hydrodynamic loads in three directions. Imposed stresses are generated and combined for normal, upset, and faulted conditions. Stresses are compared, depending on the specific safety class of the equipment, to Industrial Codes, ASME, ANSI or Industrial Standards, AISC, allowables. The calculated stresses and allowable limits for the faulted loads for the fuel preparation machine are provided in Table 3.9-2s. The refueling platform has also been examined; it can withstand the faulted loads due to seismic hydrodynamic events.
3.9.1.4.12 Seismic Category I Items Other than NSSS For statically applied loads, the stress allowables of Appendix F of the ASME Code,Section III, Winter 1972 were used for code components. For non-code components, allowables were based on tests or accepted standards consistent with those in Appendix F of the code.
Dynamic loads for components loaded in the elastic range were calculated using dynamic load factors, time history analysis, or any other method that assumes elastic behavior of the component.
The limits of the elastic range are defined in Paragraph 1323 of Appendix F for the code components. The local yielding due to stress concentration is assumed not to affect the validity of the assumptions of elastic behavior. The stress allowables of Appendix F for elastically analyzed components were used for code components. For non-code components, allowables were based on tests or accepted material standards consistent with those in Appendix F for elastically analyzed components.
The methods used in evaluating the pipe break effects are discussed in Section 3.6.
3.9.2 DYNAMIC TESTING AND ANALYSIS 3.9.2.1a Preoperational Vibration and Dynamic Effects Testing on NSSS Piping The test program is divided into three phases: preoperational vibration, startup vibration, and operational transients.
3.9.2.1a.1 Preoperational Vibration Testing The purpose of the preoperational vibration test phase is to verify that operating vibrations in the recirculation piping are acceptable. This phase of the test uses visual observation.
3.9.2.1a.2 Small Attached Piping There is no small attached piping in the NSSS scope of supply.
FSAR Rev. 70 3.9-21
SSES-FSAR Text Rev. 74 3.9.2.1a.3 Startup Vibration The purpose of this phase of the program is to verify that the main steam and recirculation piping vibration are within acceptable limits. Because of limited access due to high radiation levels, no visual observation is made during this phase of the test. Remote measurements were made during the following steady state conditions:
a) Main steam flow at 25% of rated; b) Main steam flow at 50% of rated; c) Main steam flow at 75% of rated; d) Main steam flow at 100% of rated.
3.9.2.1a.4 Operating Transient Loads The purpose of the operating transient test phase is to verify that pipe stresses are within Code Limits. The amplitude of displacements and number of cycles per transient of the main steam and recirculation piping were measured and the displacements compared with acceptance criteria. The deflections are correlated with the calculated deflections to assure that the stresses remain within Code limits. Remote vibration and deflection measurements were taken during the following transients:
a) Recirculation pump starts; b) Recirculation pump trip at 100% of rated flow; c) Turbine stop valve closure at 100% power; d) Manual discharge of representative S/R valves at 1,000 psig and at planned transient tests that result in S/R valve discharge.
3.9.2.1a.5 Test Evaluation and Acceptance Criteria The piping response to test conditions shall be considered acceptable if the organization responsible for the stress report reviews the test results and determines that the tests verify that the piping responded in a manner consistent with the predictions of the stress report and/or that the tests verify that piping stresses are within Code limits. To insure test data integrity and test safety, criteria have been established to facilitate assessment of the test while it is in progress.
These criteria, designated Level 1 and 2, are described in the following paragraphs.
3.9.2.1a.5.1 Level 1 Criterion Level 1 establishes the maximum limits for the level of pipe motion which, if exceeded, makes a test hold or termination mandatory.
If the Level 1 limit is exceeded, the plant will be placed in a satisfactory hold condition, and the responsible piping design engineer will be advised. Following resolution, applicable tests must be repeated to verify that the requirements of the Level 1 limits are satisfied.
FSAR Rev. 70 3.9-22
SSES-FSAR Text Rev. 74 3.9.2.1a.5.2 Level 2 Criterion Level 2 specifies the level of pipe motion which, if exceeded, requires that the responsible piping design engineer be advised. If the Level 2 limit is not satisfied, plant operating and startup testing plans would not necessarily be altered. Investigations of the measurements, criteria, and calculations used to generate the pipe motion limits would be initiated. An acceptable resolution must be reached by all appropriate and involved parties, including the responsible piping design engineer. Depending upon the nature of such resolution, the applicable tests may or may not have to be repeated.
3.9.2.1a.6 Corrective Actions During the course of the tests, the remote measurements shall be regularly checked to determine compliance with the Level 1 criterion. If trends indicate that the Level 1 criterion may be violated, the measurements shall be monitored at more frequent intervals. The test will be held or terminated as soon as the criterion is violated. As soon as possible after the test hold or termination, the following corrective actions will be taken:
a) Installation Inspection. A walkdown of the piping and suspension will be made to identify any obstruction or improperly operating suspension components. If vibration exceeds criteria, the source of the excitation must be identified to determine if it is related to equipment failure. Action will be taken to correct any discrepancies before repeating the test.
b) Instrumentation Inspection. The instrumentation installation and calibration will be checked and any discrepancies corrected. Additional instrumentation will be added, if necessary.
c) Repeat Test. If actions (a) and (b) identify discrepancies that could account for failure to meet the Level 1 criterion, the test will be repeated.
d) Resolution of Findings. If the Level 1 criterion is violated on the repeat test or no relevant discrepancies are identified in (a) and (b), the organization responsible for the stress report shall review the test results and criteria to determine if the test can be safely continued.
If the test measurements indicate failure to meet the Level 2 criterion, the following corrective actions will be taken after completion of the test:
a) Installation Inspection. A walkdown of the piping and suspension shall be made to identify any obstruction or improperly operating suspension components. If vibration exceeds limits, the source of the vibration must be identified. Action, such as suspension adjustment, will be taken to correct any discrepancies.
b) Instrumentation Inspection. The instrumentation installation and calibration will be checked and any discrepancies corrected.
c) Repeat Test. If (a) and (b) above identify a malfunction or discrepancy that could account for failure to comply with Level 2 criterion and appropriate corrective action has been taken, the test may be repeated.
FSAR Rev. 70 3.9-23
SSES-FSAR Text Rev. 74 d) Documentation of Discrepancies. If the test is not repeated, the discrepancies found under actions (a) and (b) above shall be documented in the test evaluation report and correlated with the test condition. The test will not be considered complete until the test results are reconciled with the acceptance criteria.
3.9.2.1a.7 Measurement Locations Remote shock and vibration measurements were made in the three orthogonal directions near the first downstream S/R valve on each steam line; and in the three orthogonal directions on the piping between the recirculation pump discharge and the first downstream valve. During preoperational testing prior to fuel load, visual inspection of the piping was made, and any visible vibration measured with a handheld instrument.
For each of the selected remote measurement locations, Levels 1 and 2 deflection and acceleration limits were prescribed in the startup test specification. Level 2 limits were based on the results of the stress report adjusted for operating mode and instrument accuracy; Level 1 limits were based on maximum allowable Code stress limits.
3.9.2.1b Preoperational Thermal Vibration and Dynamic Effects of Testing of Piping other than NSSS The dynamic effects on all safety-related ASME Class 1, 2, and 3 piping systems, including their supports and restraints, are considered as required by NB-3622.3, NC-3622, and ND-3622 of Section III of ASME B&PV Code. The structural and functional integrity of the piping system is ensured under a postulated seismic event by dynamic analysis only. Piping systems having significant anticipated transients loads, e.g., main stop valve closure or relief valve blow for example, are analyzed for the time-dependent forces. In addition, piping steady state vibration and dynamic transient tests were performed as summarized below, to ensure that a) Excessive steady state vibration is not present in the piping that would result in piping stresses and restraint loads above the allowables.
b) The piping is adequately restrained to withstand the dynamic transient loads.
Cognizant design personnel familiar with the systems to be tested developed the test plans, and evaluate the test results. Also the cognizant design personnel witnessed the test. The data acquired from the tests was compared with the expected results to determine the acceptability of the total system response.
A list of all piping systems in the BOP is provided in Table 3.9-20. ASME Section III Class 1, 2 and 3 piping systems, high energy piping systems, moderate energy piping systems, seismic Category I and seismic Category II systems are identified in the Table. The Table also identifies the tests to be performed for each system.
Piping thermal expansion tests are performed for the safety-related piping systems with normal operating temperature exceeding 300°F. Safety-related piping systems with normal operating temperature less than 300°F do not have enough significant thermal expansion to warrant thermal expansion tests. Engineering review of all seismic Category I piping systems, including their supports, restraints or snubbers, is performed after completion of construction and prior to fuel load to ensure that no restraint of normal thermal movement occurs due to interferences and obstructions, and that the support and restraints are in accordance with the design intent.
FSAR Rev. 70 3.9-24
SSES-FSAR Text Rev. 74 For the systems receiving thermal expansion tests, the pipe movements are monitored to ensure that no restraint of normal thermal movement occurs at locations other than at the designed restraint locations.
The thermal expansion test program verifies that the free thermal expansion of piping systems takes place at the snubbers by monitoring the thermal movement. Performance of the snubbers designed for transient loads such as due to Main Stop Valve Closure or Main Steam Relief Valve Discharge are verified by measuring the load in the snubber during the dynamic transient tests. The snubbers are qualified by dynamic testing for cyclic loading as described in Subsection 3.9.3.4.1.
The acceptance criterion for thermal expansion tests and dynamic transient tests is that the measured pipe displacements or restraint loads shall be below the calculated or design values.
The acceptance criterion for the steady-state vibration tests is:
Either The maximum measured amplitude of the piping vibration shall not induce a stress in the pipe more than half the endurance limit of the material for B31.3 piping. The maximum stress in the pipe due to steady state vibration for class 1, 2 & 3 piping is limited to one-half of the endurance limit (allowable stress corresponding to 106 cycles in Appendix I of ASME Section III), the steady-state vibration-induced stress will not contribute to the reduction of fatigue life of piping.
Or Acceptance criterion are divided into two categories, i.e., Level 1 and Level 2. If the Level 1 criterion is violated, the test must be placed on hold. If the Level 2 criterion is violated, the test can continue, but the measurements must be evaluated to verify that continued test operation will not result in exceeding piping fatigue requirements.
For steady state vibration the piping peak stress zero to peak due to vibration only (neglecting pressure) will not exceed 10,000 psi for the Level 1 criterion and 5,000 psi for the Level 2 criterion. These limits are below the piping material fatigue endurance limits as defined for 106 cycles in Appendix I of the ASME Code,Section III.
The Table 3.9-20 provides cross reference between the FSAR Section 3.9 and the appropriate test description in FSAR Chapter 14.
3.9.2.1b.1 Piping Dynamic Transient Tests During the preoperational and/or startup testing, dynamic transient tests will be performed on the following piping for the indicated modes of operation.
a) Main steam piping outside the containment for main steam turbine stop valve closure at 30 percent +/-10%, 75 percent +/-10%, and 100 percent +0 -10% power.
b) Main steam bypass piping to the anchor near the bypass valves for the turbine stop valve closure.
FSAR Rev. 70 3.9-25
SSES-FSAR Text Rev. 74 c) Selected main steam safety/relief valve discharge piping for the main steam relief valve opening. The selected SRV discharge piping brackets all the SRV discharge piping.
d) HPCI turbine steam supply piping for HPCI turbine trip.
e) Feedwater system reactor feed pump trip/coastdown under operating conditions; Pump A only, B only, and C only and at normal pump flowrate.
From past experience, the dynamic transients in other piping systems are not significant.
Dynamic transient analysis of the subject lines has been performed to determine the response of the piping system and the restraint loads. During the test the displacement of the pipe, loads in the snubbers and restraints and pressure at representative locations will be measured.
Acceptance criteria for this test are that the measured loads in the snubbers and restraints shall be below the design values of the snubbers and restraints. In the case (e) the acceptance criteria is that the measured response shall be less than the acceptable response determined by analysis.
3.9.2.1b.2 Piping Steady State Vibration Testing The piping system associated with the following components' operation will be observed for steady state vibration during preoperational test programs or power ascension:
a) RHR pump b) HPCI pump and turbine c) RCIC pump and turbine d) Core spray pump e) Main Steam f) Feedwater g) Reactor Water Cleanup From experience on other nuclear power plants, the steady state vibration in other piping systems is not critical. However, abnormal vibrations of other systems during system walkdown on initial startup or power escalation will be noted and instrumented if necessary to determine the acceptability of such vibration.
Steady-state vibration in the subject piping systems is primarily induced by the flow in the pipe and the equipment motion. In general, the specific causes of the steady-state vibration is not known beforehand; therefore, design engineers with stress analysis experience and familiarity with the subject piping system will visually observe the lines or monitor inaccessible lines with suitable instrumentation during all significant modes of system operation and classify each line as acceptable if the vibration is not significant, or questionable if vibration is significant. The lines with questionable steady-state vibration will be monitored by suitable instrumentation to determine the system response.
FSAR Rev. 70 3.9-26
SSES-FSAR Text Rev. 74 The type of the instrumentation, if necessary, will be determined by the design engineer so that the maximum amplitude and frequency response of the piping system can be determined. The instrumentation will not screen out the significant frequencies.
For lines with questionable steady-state vibration, the acceptance criterion is as discussed in Subsection 3.9.2.1b.
When required, additional restraints will be provided to reduce the steady-state vibration and to keep the stresses below the acceptance criterion levels.
3.9.2.2a Seismic and Hydrodynamic Qualification of Safety-Related NSSS Mechanical Equipment This subsection describes the criteria for dynamic qualification of mechanical safety-related equipment and also describes the qualification testing and/or analysis applicable to this plant for all the major components on a component-by-component basis. In some cases, a module or assembly consisting of mechanical and electrical equipment is qualified as a unit, for example, motor powered pumps. These modules are generally discussed in this paragraph rather than providing discussion of the separate electrical parts in Sections 3.10 and 3.11. Dynamic qualification testing for pumps and valves is also discussed in Subsection 3.9.3.2. Electrical supporting equipment such as control consoles, cabinets, and panels which are part of the NSSS are discussed in Subsection 3.10.
All safety related NSSS mechanical equipment located in the Containment and the Reactor and Control Buildings are qualified for the combined seismic and hydrodynamic vibratory loadings.
Procedures for the assessment and requalification of safety-related NSSS mechanical equipment for the additional hydrodynamic loads are described in Section 7.1.6 of the Design Assessment Report (DAR.)
3.9.2.2a.1 Tests and Analysis Criteria and Methods The ability of equipment to perform its safety-related function during and after the application of dynamic loads is demonstrated by tests and/or analysis. Selection of testing, or analysis, or a combination of the two is determined by the type, size, shape, and complexity of the equipment being considered. When practical, equipment operability is demonstrated by testing. Otherwise, operability is demonstrated by mathematical analysis.
Equipment which is large, simple, and/or consumes large amounts of power is usually qualified by analysis or test to show that the loads, stresses and deflections are less than the allowable maximum. Analysis and/or testing is also used to show there are no natural frequencies below 33 Hz for seismic loads and 60 Hz for hydrodynamic loads. If a natural frequency is discovered, dynamic tests may be conducted and in conjunction with mathematical analysis used to verify operability and structural integrity at the required dynamic input conditions. When the equipment is qualified by dynamic test, either the response spectrum or time history of the attachment point is used in determining input motion.
Natural frequency may be determined by running a continuous sweep frequency search using a sinusoidal steady-state input of low magnitude. Dynamic loading conditions are simulated by testing using random vibration input or single frequency input (within equipment capability) at frequencies through 35 Hz. Whichever method is used, the input amplitude during testing envelopes the actual input amplitude expected during dynamic loading conditions.
FSAR Rev. 70 3.9-27
SSES-FSAR Text Rev. 74 The equipment being dynamically tested is mounted on a fixture which simulates the intended service mounting and causes no dynamic coupling to the equipment.
Equipment having an extended structure, such as a valve operator, is analyzed by applying static equivalent dynamic loads at the center of gravity of the extended structure. In cases where the equipment structural complexity makes mathematical analysis impractical, a test is used to determine spring constant and operational capability at maximum equivalent dynamic loading conditions. Pipe-mounted equipment is analyzed in the piping system dynamic analysis.
3.9.2.2a.1.1 Test Input Motion When random vibration input is used, the actual input motion envelopes the appropriate floor input motion at the individual modes. However, single frequency input, such as sine beats, can be used provided one of the following conditions are met:
a) The characteristics of the required input motion are dominated by one frequency.
b) The anticipated response of the equipment is adequately represented by one mode.
c) The input has sufficient intensity and duration to excite all modes to the required magnitude, such that the testing response spectra will envelope the corresponding response spectra of the individual modes.
3.9.2.2a.1.2 Application of Input Motion When dynamic tests are performed, the input motion is applied to one vertical and one horizontal axis simultaneously. However, if the equipment response along the vertical direction is not sensitive to the vibratory motion along the horizontal direction, and vice versa, then the input motion is applied to one direction at a time. In the case of single frequency input, the time phasing of the inputs in the vertical and horizontal directions are such that a purely rectilinear resultant input is avoided.
3.9.2.2a.1.3 Fixture Design The fixture design will simulate the actual service mounting and cause no dynamic coupling to the equipment.
3.9.2.2a.1.4 Prototype Testing Equipment testing has been conducted on prototypes of the equipment installed in this plant.
3.9.2.2a.2 Seismic and Hydrodynamic Qualification of Specific Mechanical Components The following sections discuss the testing or analytical qualification of NSSS equipment.
Seismic and hydrodynamic qualification is also described in Subsections 3.9.1.4, 3.9.3.1, and 3.9.3.2.
FSAR Rev. 70 3.9-28
SSES-FSAR Text Rev. 74 3.9.2.2a.2.1 Jet Pumps A dynamic analysis of the jet pumps was performed. The stresses resulting from the analysis are below the design allowables.
3.9.2.2a.2.2 CRD and CRD Housing The dynamic qualification of the CRD housing (with enclosed CRD) was done analytically. The results of this analysis established the structural integrity of these components. Preliminary dynamic tests have been conducted to verify the operability of the Control Rod Drive during a dynamic event. A test was performed in which the CRD was shown to function satisfactorily, while a static bow in the fuel channels was used to simulate dynamic loading.
3.9.2.2a.2.3 Core Support (Orificed Fuel Support and CR Guide Tube)
A detailed analysis imposing dynamic effects due to seismic and hydrodynamic events has shown that the maximum stresses developed during these events are much lower than the maximum allowed for the component material.
3.9.2.2a.2.4 Hydraulic Control Unit (HCU)
The seismic and hydrodynamic load adequacy of the HCU for the faulted condition is demonstrated by test and analysis. With the HCU's mounted on a seismic support structure, the dynamic loading results from application of 3.0g vertical at the natural frequency of 7 to 30 Hz, and 1.0g horizontal at 2 to 6 Hz, and 5.0g horizontal at 10 Hz. At these frequencies, the maximum HCU capability demonstrated for dynamic loading is 20g vertical at 7 to 30 Hz, and greater than 4g horizontal at 2 to 6 Hz, and 8g horizontal at 10 Hz.
3.9.2.2a.2.5 Fuel Channels Fuel channel loading is discussed in Section 3.9.1.4.10.
3.9.2.2a.2.6 Recirculation Pump and Motor Assembly Calculations are made to determine that the recirculation pump and motor assembly are designed to withstand the specific static to seismic and hydrodynamic forces. The flooded assembly was analyzed as a free body supported by constant support hangers from the brackets on the motor mounting member with mechanical snubbers attached to brackets located on the pump case and the top of the motor frame.
Primary stresses due to horizontal and vertical dynamic forces were considered to act simultaneously and are conservatively added directly. Horizontal and vertical dynamic forces were applied at mass centers and equilibrium reactions determined for motor and pump brackets.
3.9.2.2a.2.7 ECCS Pump and Motor Assembly Pump/motor assemblies were analyzed with static loading equivalent to seismic acceleration under faulted conditions since the natural frequencies are above 33 Hz. The maximum specified vertical and horizontal accelerations were applied simultaneously in the worst case FSAR Rev. 70 3.9-29
SSES-FSAR Text Rev. 74 combination and the results of the analysis indicate that the pump is capable of sustaining the loadings without overstressing the pump components.
A motor of similar design has been dynamically qualified by a combination of static analysis and dynamic testing. The motor has been seismically qualified via dynamic testing in accordance with IEEE 344, 1975. The qualification test program included demonstration of startup and shutdown capabilities, as well as no load operability during seismic and hydrodynamic loading conditions.
3.9.2.2a.2.8 RCIC Pump Assembly The barrel type RCIC pump is mounted on a large cross-section pedestal.
The RCIC pump assembly is qualified by analysis using static loading equivalent to seismic and hydrodynamic loading with the design operating loads and temperature. The results of this analysis confirm that the calculated stresses are substantially less than the allowable stresses.
3.9.2.2a.2.9 RCIC Turbine Assembly The RCIC turbine is qualified by analysis using static loading equivalent to dynamic loading.
The turbine assembly and its components were considered to be supported as designed.
Horizontal/vertical accelerations were applied to the mass centers of gravity. The magnitude of the acceleration coefficients was 3.0g horizontal and 1.0g vertical. The results of the analysis indicate that the turbine assembly is capable of sustaining the above loadings without overstressing any components.
The turbine assembly is qualified by dynamic testing, in accordance with IEEE 344-1975. The qualification test program demonstrated start-up, steady-state operability, and shutdown capabilities.
3.9.2.2a.2.10 Standby Liquid Control Pump and Motor Assembly The SLC positive displacement pump and motor are mounted on a common base plate which is qualified by static analysis using static loading equivalent to the dynamic loading with the design operating loads and temperature.
The results of this analysis confirm that the calculated stresses are substantially less than the allowable stresses.
3.9.2.2a.2.11 RHR Heat Exchangers A three-dimensional finite element model of the RHR heat exchanger and its support was developed and analyzed using the response spectrum method to verify that the heat exchanger can withstand seismic and hydrodynamic loads. The same model was statically analyzed to evaluate the effects of the external piping loads and dead weight to ensure that the nozzle load criteria and stress limits are met. Critical location stresses were evaluated and found to be lower than the corresponding allowable values.
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SSES-FSAR Text Rev. 74 3.9.2.2a.2.12 Safety Relief Valves (SRV)
Three SRVs of the Susquehanna design were subjected to the following qualification test programs in order to demonstrate compliance with the performance requirements under the specified conditions.
- 1. Life Cycle Tests - These tests consisted of subjecting each of the prequalification production units to approximately 300 safety and relief actuations in order to verify acceptability of the design to meet the requirements for (a) set pressure, (b) opening and closing response time, (c) blowdown, (d) seat tightness, (e) achievement of rated-capacity flow lift (ASME) during each actuation, (f) proper reclosure after each actuation without a tendency to stick open, chatter, or resulting in disc oscillation, and (g) opening without any inlet pressure applied which requirement (h) simulates an emergency operability condition.
Conditions such as operating temperature, pressure ramp rate, dynamic and static back-pressures, pneumatic operating pressure and solenoid voltage were varied to assure valve operability under normal and transient operating conditions. Upon completion of the tests, test units were disassembled and inspected. This test program established the qualified service life of the safety relief valve.
- 2. Seismic Tests - The test units were subjected to seismic tests to simulate the normal, upset, emergency, and faulted conditions.
Post-OBE and post-SSE reference frame tests were performed to determine the operability effects due to repeated combinations of seismic simulations, nozzle loadings, temperature and pressure. These reference frame tests consisted of set pressure determination during safety actuation, response time determination during relief actuation, valve leakage, and an emergency operability test. These reference frame tests were performed with induced nozzle loads applied.
In order to evaluate the design capability of the test unit, the OBE and SSE tests were repeated using a higher input level. The test conditions during these tests are shown in Table 5.2-3.
After the seismic tests, the electro-pneumatic actuator assembly was removed from the test unit and subjected to post seismic reference frame tests, a negative pressure test, post-negative pressure reference frame tests, a postulated Loss of Coolant Accident (LOCA) environmental test, and a post-LOCA reference frame test and inspection.
3.9.2.2a.2.13 Standby Liquid Control Tank The standby liquid control tank is a cylindrical tank 9 feet in diameter and 12 feet high bolted to the concrete floor. The Standby Liquid Control Tank is qualified by analysis for:
a) Stresses in the tank bearing plate b) Belt stresses c) Sloshing loads imposed at the natural frequency of sloshing, which is 0.58 Hz FSAR Rev. 70 3.9-31
SSES-FSAR Text Rev. 74 d) Minimum wall thickness e) Buckling The results of the analysis confirm that the calculated stresses at the investigated locations are below the allowable stresses.
3.9.2.2a.2.14 Main Steam Isolation Valves The main steam isolation valves were analyzed; representative models were statically tested to demonstrate operability at the specified faulted condition. Static testing consisted of loading the valve actuator mechanically to equivalent specified dynamic loading while valve closure was performed. Operation of the valve under simulated faulted conditions was demonstrated by this test.
3.9.2.2a.2.15 Main Steam Safety/Relief Valves Due to the complexity of the SRVs and the performance requirements, the total assembly of the safety/relief valve (including electrical, pneumatic devices) was dynamically tested at accelerations equal to or greater than the combined specified SSE and hydrodynamic loading.
Satisfactory operation of the valves was demonstrated during and after the test.
3.9.2.2a.2.16 HPCI Turbine The HPCI turbine was qualified by static analysis equivalent seismic acceleration. The turbine assembly and its components were considered to be supported as designed, and loading equivalent to horizontal/vertical accelerations was applied to the center of mass. The results of the analysis indicate that the turbine assembly is capable of sustaining the loadings without overstressing any components. The turbine electronic governor assembly has been seismically qualified via dynamic testing, in accordance with IEEE 344-1975. The qualification test program demonstrated startup, steady state operability, and shutdown capabilities.
3.9.2.2a.2.17 HPCI Pump The HPCI pump is a split body type comprising a booster pump and a main pump mounted on a common base plate. The pump assembly behaves as a rigid body; therefore, qualification by analysis was performed. Results are obtained by using acceleration forces acting simultaneously in two directions, one vertical and one horizontal and calculated using the square root of the sum of the squares method. The pump mass, support system, and accessory piping are shown by analysis to have a natural frequency less than 33 Hz.
3.9.2.2b Seismic Qualification Testing of Safety Related Non-NSSS Mechanical Equipment All Non-NSSS Seismic Category I equipment has been designed to withstand simultaneously the horizontal and vertical accelerations caused by the OBE and the SSE, in conjunction with other applicable loads. All equipment classified as active have demonstrated through qualification that they will perform their design function before, during and after a design basis accident.
The criteria for the seismic qualification of non-NSSS mechanical and electrical equipment, with the exception of valves, valve operators other than relief valves and the equipment found in the FSAR Rev. 70 3.9-32
SSES-FSAR Text Rev. 74 Diesel Generator 'E' Building, is contained in project specification No. 8856-G-10 for a seismic environment complemented by No. 8856-G-22 for a combined seismic and hydrodynamic environment. For the Diesel Generator 'E' facility, the criteria for seismic qualification of mechanical and electrical equipment is contained in project specification No. C-1041 and Cooper Energy Services Standard No. SD-140. The standard IEEE-344, "Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations", is referenced in the G-10 and C-1041 Specifications and is being used as a supplement to the G-10, G-22, and C-1041 Specifications in the individual equipment procurement documentation package. Specifications G-10, G-22, and C-1041 and Standard IEEE-344 address the requirements of the demonstration of the seismic adequacy of equipment by analysis and/or tests. NRC Regulatory Guide 1.100 Revision 1, August 1977 accepts the use of standard IEEE-344 with a few modifications. Table 3.9-18 shows the comparison of the specification G-10 with IEEE-344-1975.
Non-NSSS motor-and air-operated valves are addressed in Subsection 3.9.3.2b.2. Control valves are addressed in Section 3.10b.
The assessment and requalification of safety-related non-NSSS mechanical equipment for the additional hydrodynamic loads are described in Section 7.1.7 of the Design Assessment Report (DAR).
3.9.2.2b.1 Safety-Related and Safety-Impacted Mechanical Equipment Other than for the NSSS 3.9.2.2b.1.1 Dynamic Analysis Without Testing Structural analysis without testing was used if structural integrity alone could ensure the intended design function. Equipment that falls into this category includes:
Safety-Related a) Diesel oil storage tanks b) Containment instrument gas accumulators c) Suppression pool suction strainers d) Nuclear safety/relief valves e) Vacuum breakers Safety-Impacted f) Supports for air handling units g) Diesel building supports for cranes h) Reactor building supports for cranes i) Fuel pool skimmer surge tanks FSAR Rev. 70 3.9-33
SSES-FSAR Text Rev. 74 Rotational analysis without testing was used to qualify heavy rotating machinery where it had to be verified that deformations from seismic loading would not bind the rotating element so that the component could not perform its intended design function. Components that fall into this category include:
a) Diesel generators b) Diesel oil transfer pumps c) RHR service water pumps d) Emergency service water pumps e) Control room centrifugal water chiller pumps Refer to Tables 3.9-16 and 3.9-17 for listings of dynamically qualified equipment.
3.9.2.2b.1.2 Dynamic Testing The equipment subjected to dynamic testing are the hydrogen recombiners (NSD-E-JFW 1003 March 4, 1975) and rupture discs (Black Sivalls Bryson, January 3, 1977). The rupture discs are installed in the exhaust of the HPCI and RCIC turbines.
3.9.2.2b.2 Criteria For dynamic analysis without testing the equipment listed under Subsection 3.9.2.2b.1.1, and for dynamically testing the rupture discs under Subsection 3.9.2.2b.1.2, the criteria are as follows.
Response Spectrum Curves The appropriate response spectrum curves for the equipment in question were issued with the material requisition or the equipment specification, for OBE, SSE, LOCA and SRV (LOCA &
SRV only when applicable). Response spectrum curves are based upon the seismic analysis of the supporting structure and represent the maximum seismic response, as a function of oscillator frequency, of an array of single degree of freedom damped oscillators at a particular location within the structure. Response spectrum curves, plotted in terms of acceleration versus frequency, correspond to various locations within the buildings and are identified with respect to the points noted on the mathematical model for each direction of vibration to be considered.
This may include the vertical as well as both the north-south and the east-west horizontal directions. In addition, each response spectrum curve corresponds to a particular damping ratio, i.e., the ratio of damping of the single degree of freedom oscillators to critical damping. See Section 3.7 for the appropriate response spectrum curves.
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SSES-FSAR Text Rev. 74 Load Combinations and Allowable Stress Limits Seismic Category I equipment has been designed to withstand the more severe of the following load combinations:
a) OBE Conditions Gravity loads and operating loads (or Design Basis Accident loads, if applicable) including associated temperature and pressures combined by absolute sums with the dynamic seismic loading of the OBE. Allowable stresses in the structural steel portions may be increased to 125 percent of the allowable working stress limits as set forth in ASME Boiler and Pressure Vessel Code Section III, or other applicable industrial codes.
The customary increase in normal allowable working stress due to an earthquake shall be used if, according to the appropriate code, it is less than 25 percent. Resulting deflections, misalignment or binding of parts, or effects on electrical performance (microphones, contact bounce, etc.) do not prevent operation of the equipment during or after the seismic disturbance.
b) SSE Conditions Gravity loads and operating loads (or Design Basis Accident loads, if applicable),
including associated temperatures and pressures combined by absolute sums with the dynamic seismic loading of the SSE. Allowable stresses in the structural portions may be increased to 150 percent of allowable working stress limits in accordance with the appropriate codes listed in (a); however, the stresses may not exceed 0.9 Fy in bending, 0.85 Fy for axial tension, and 0.5 Fy in shear, where Fy equals the material minimum yield stress at the design temperature. For equipment designed by the maximum shear stress theory, the difference between the maximum and minimum principal stresses will not exceed 0.9 Fy. The resulting deflections, misalignment, or binding of parts, or effects on electrical performance (microphones, contact bounce, etc.) will not prevent operation of the equipment during or after the seismic disturbance.
Prevention of Overturning and Sliding Stationary equipment is designed to prevent overturning or sliding by using anchor bolts or other suitable mechanical anchoring devices. The effect of friction on the ability to resist sliding is neglected. The effect of upward vertical seismic loads on reducing overturning resistance is considered. Anchoring devices are designed in accordance with the requirements of Items a) and b) and the AISC Manual of Steel Construction. The proposed anchoring system is shown on the Seller's drawings so that the Buyer can provide the proper foundation.
Dynamic Testing Seismic adequacy was established for the rupture discs by dynamically testing them to meet the criteria defined under a and b above. Actual testing of equipment was done with base connections simulating the actual installation in accordance with one of the following methods:
a) The equipment was subjected to an input excitation such that the measured response was equal to or greater than the specified design response.
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SSES-FSAR Text Rev. 74 b) The equipment was subjected to an input excitation whose response spectrum equaled or exceeded the specified response spectrum for that location.
Criteria for the Diesel Fuel Oil Storage Tanks These tanks are buried below grade under a cover of 16 ft (8 ft for diesel 'E' fuel tank) of earth.
Equivalent fluid pressure of soil is 110 lb/ft3 (100 lb/ft3 for diesel 'E' fuel tank).
Tanks and tank supports are designed to withstand an H-20 loading, according to AASHTO, applied above 16 ft (8 ft for diesel 'E' fuel tank) of saturated overburden. The H-20 loading acts simultaneously with normal fluid pressure. Tank walls and ends will not deflect more than 3 percent maximum under the most unfavorable loading conditions.
The diesel fuel oil storage tanks are designed, fabricated, tested, and stamped in accordance with the ASME Code,Section III, Class 3. The tanks, including vents and openings, are designed as underground atmospheric tanks in accordance with OSHA Section 1910.106.
Tanks and their supports are designated Seismic Category I, and are designed to resist the increased earth pressure from the OBE and the SSE. For the OBE, the lateral earth pressure is 90 psf (180 psf for diesel 'E' fuel tank), for the SSE, 180 psf (330 psf for diesel 'E' fuel tank).
When combined with other normal operating conditions, the stresses are limited to 125 percent of the ASME Code,Section III allowable stresses for the OBE condition, and are limited to 90 percent of the material's yield stress for the SSE condition.
Tanks are designed to withstand external pressure resulting from being buried in ground having a water table surface at ground level when the tanks are empty. Hydraulic uplift forces on buried tanks are resisted by the weight of the empty tank and the foundation mat plus 16 ft (8 ft for diesel 'E' fuel tank) of saturated overburden.
3.9.2.3 Dynamic Response of Reactor Internals under Operational Flow Transients and Steady State Conditions The major reactor internal components within the vessel are subjected to extensive testing.
In addition, dynamic system analyses are conducted to describe and evaluate the flow-induced vibration phenomena resulting from normal reactor operation and from anticipated operational transients.
In general, the vibration forcing functions for operational flow transients and steady state conditions are not predetermined by detailed analysis. Special analysis of the response signals measured from reactor internals of many similar designs are performed to obtain the parameters which determine the amplitude and modal contributions in the vibration responses.
These studies are useful for extrapolating the results from tests to components of similar design.
This vibration prediction method is appropriate where standard hydrodynamic theory cannot be applied due to complexity of the structure and flow conditions. Elements of the vibration prediction method are outlined as follows:
- 1) Dynamic analysis of major components and subassemblies is performed to identify natural vibration modes and frequencies. The analysis models used for Seismic Category 1 structures are similar to those outlined in subsection 3.7.2.
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SSES-FSAR Text Rev. 74
- 2) Data from previous plant vibration measurements is assembled and examined to identify predominant vibration response modes of major components. In general, response modes are similar but response amplitudes vary among BWRs of differing size and design.
- 3) Parameters are identified which are expected to influence vibration response amplitudes among the several reference plants. These include hydraulic parameters such as velocity and steam flow rates, and structural parameters such as natural frequency and significant dimensions.
- 4) Correlation functions of the variable parameters are developed which, multiplied by response amplitudes, tend to minimize the statistical variability between plants. A correlation function is obtained for each major component and response mode.
- 5) Predicted vibration amplitudes for components of the prototype plant are obtained from these correlation functions, based on applicable values of the parameters for the prototype plant. The predicted amplitude for each dominant response mode is stated in terms of a range, taking into account the degree of statistical variability in each of the correlations. The predicted mode and frequency are obtained from the dynamic analyses of paragraph 1 above.
The dynamic modal analysis also forms the basis for interpretation of the prototype plant preoperational and initial startup test results (Subsection 3.9.2.4). Modal stresses are calculated and relationships are obtained between sensor response amplitudes and peak component 3.9.2.4 Confirmatory Flow-Induced Vibration Testing of Reactor Internals Reactor internals were tested in accordance with provisions of Regulatory Guide 1.20, Revision 2, for Non-prototype Category I plants. The test procedure required operation of the recirculation system at rated flow with internals important to safety installed. Inspection for evidence of vibration, wear, or loose parts followed. Blade guides, incore instruments, neutron sources, dryer and fuel were not installed. Control rods were either not installed or fully withdrawn and prevented from inserting. The test duration was sufficient to subject critical components to at least 106 cycles of vibration during two-loop and single-loop operation of the recirculation system. At the completion of the flow test, the vessel head and shroud head were removed, the vessel was drained and major components will be inspected on a selected basis.
The inspection covered all components which were examined on the prototype design, including the shroud, shroud head, core support structures, jet pumps, peripheral control rod drive guide tubes and peripheral in-core guide tubes. Access will be provided to the reactor lower plenum.
Reactor internals for Susquehanna are substantially the same as the internals design configurations that have been tested in prototype BWR/4 plants. Results of the prototype tests are presented in a Licensing Topical Report (Ref. 3.9-7). This report also contains additional information on the confirmatory inspection program.
A labyrinth seal, consisting of five circumferential grooves on each jet pump mixer at the slip joint interface with the jet pump diffuser collar, reduces leakage at the slip joint. Tests performed by General Electric Company (Reference 3.9-10) demonstrated that the labyrinth seals reduce leakage through the slip joints. However, SSES no longer credits the labyrinth seals with this function. Jet pumps are equipped with either Slip Joint Diffuser Rings or Slip Joint Clamps. Slip Joint Diffuser Rings reduce leakage and leakage induced vibration, and are FSAR Rev. 70 3.9-37
SSES-FSAR Text Rev. 74 designed to clamp to the Diffuser Collar Ears holding the Ring in place. Slip Joint Clamps reduce Jet Pump vibration.
3.9.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions In order to assure that no significant dynamic amplification of load occurs as a result of the oscillatory nature of the blowdown forces, a comparison was made of the periods of the applied forces and the natural periods of the core support structures being acted upon by the applied forces. These periods were determined from a 12-node vertical dynamic model of the RPV and internals. Besides the real masses of the RPV and core support structures, account was made for the water inside the RPV.
The time-varying pressures are applied to the dynamic model of the reactor internals described above. Except for the dynamic model and the nature and locations of the forcing functions, the dynamic analysis method is identical to that described for seismic analysis and is detailed in Subsection 3.7.2.1.
Dynamic loads are combined by SRSS. The results are then combined with other static and steady state loads on an ABS basis to confirm the adequacy of design loads. The results of the dynamic analysis are summarized in Tables 3.9-2, 3.9-2b, 3.9-2w, and 3.9-2aa.
3.9.2.6 Correlations of Reactor Internals Vibration Test Results with the Analytical Results Prior to initiation of the instrumented vibration test program for the prototype plant, extensive dynamic analyses of the reactor and internals are performed. The results of these analyses are used to generate the allowable vibration levels during the vibration test. The vibration data obtained during the test are analyzed in detail. The results of the data analysis, vibration amplitudes, natural frequencies and mode shapes, are then compared to those obtained from the theoretical analysis.
Such comparisons provide the analysts with added insight into the dynamic behavior of the reactor internals. The additional knowledge gained is utilized in the generation of the dynamic models for seismic and LOCA analyses for this plant. The models used for this plant are the same as those used for the vibration analysis of the prototype plant.
The vibration test data are supplemented by data from forced oscillation tests of reactor internal components to provide the analysis with additional information concerning the dynamic behavior of the reactor internals.
3.9.3 ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTURES 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits This section delineates the criteria for selection and definition of design limits and loading combinations associated with normal operation, postulated accidents, and specified seismic and hydrodynamic events for the design of safety-related ASME code components (except containment components, which are discussed in Section 3.8.)
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SSES-FSAR Text Rev. 74 This section also lists the major ASME Class 1, 2, and 3 equipment and associated pressure retaining parts on a component by component basis and identifies the applicable loadings, calculation methods, calculated stresses, and allowable stresses. Design transients for ASME Class 2 equipment are not addressed in this section. They are covered in Subsection 3.9.1.1.
Seismic and hydrodynamic related loads are discussed in Subsections 3.9.2.2a, 3.9.2.2b and Section 3.7.
Table 3.9-2 is the major part of this section; it presents the loading combination, analytical methods (by reference or example) and also the calculated stress or other design values for the most critical areas of the ASME Code Class 1, 2 and 3 components, supports, and core support structures. These design values are also compared to applicable code allowables.
3.9.3.1.1 Plant Conditions All events that the plant might credibly experience during a reactor year are evaluated to establish a design basis for plant equipment. These events are divided into four plant conditions. The plant conditions described in the following paragraphs are based on event probability (i.e., frequency of occurrence) and correlated design conditions defined in the ASME Boiler and Pressure Vessel Code,Section III.
3.9.3.1.1.1 Normal Condition Normal conditions are any conditions in the course of System startup, operation in the design power range, normal hot standby (with main condenser available), and System shutdown other than Upset, Emergency, Faulted, or Testing.
3.9.3.1.1.2 Upset Condition Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The Upset Conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system, and transients due to loss of load or power, or an operating basis earthquake. Hot standby with the main condenser isolated is an Upset Condition.
3.9.3.1.1.3 Emergency Condition Those deviations from Normal Conditions which require shutdown for correction of the conditions or repair of damage in the RCPB. 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 any damage developed in the system. Emergency condition events include, but are not limited to, transients caused by one of the following: a multiple valve blowdown of the reactor vessel; loss of reactor coolant from a small break or crack which does not depressurize the reactor system nor result in leakage beyond normal makeup system capacity, but which requires the safety functions of isolation of containment, and reactor shutdown; improper assembly of the core during refueling; and seizure of one recirculation pump.
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SSES-FSAR Text Rev. 74 3.9.3.1.1.4 Faulted Condition Those combinations of conditions associated with extremely low probability, postulated events whose consequences are such that the integrity and operability of the system may be impaired to the extent that considerations of public health and safety are involved. Faulted conditions encompass events that are postulated because their consequences would include the potential for the release of significant amounts of radioactive material. These postulated events are the most drastic that must be designed against and thus represent limiting design bases. Faulted condition events include, but are not limited to, one of the following: a control rod drop accident, a fuel-handling accident, a main steam line break, a recirculation loop break, the combination of any pipe break plus the seismic motion associated with SSE and hydrodynamic loading plus a loss of offsite power, or the safe shutdown earthquake.
3.9.3.1.1.5 Correlation of Plant Conditions with Event Probability The probability of an event occurring per reactor year associated with the plant conditions is listed below. This correlation can be used to identify the appropriate plant condition for any hypothesized event or sequence of events.
Event Encountered Probability per Plant Conditions Reactor Year Normal (planned) 1.0 Upset (moderate probability) 1.0 > p > 10-2 Emergency (low probability) 10-2 > p > 10-4 Faulted (extremely low probability) 10-4 > p > 10-6 3.9.3.1.1.6 Safety Class Functional Criteria For any normal or upset design condition event, Safety Class 1, 2, and 3 equipment shall be capable of accomplishing its safety functions as required by the event and shall incur no permanent changes that could impair its ability to accomplish its safety functions as required by any subsequent design condition event.
For any emergency or faulted design condition event, Safety Class 1, 2, and 3 equipment shall be capable of accomplishing its safety functions as required by the event but repairs could be required to ensure its ability to accomplish its safety functions as required by any subsequent design condition event.
Functional capability of safety-related essential piping components will be assured using the criteria given in Enclosure 110-1 to NRC questions and the Rodabaugh criteria.
3.9.3.1.1.7 Compliance with Regulatory Guide 1.48 Regulatory Guide 1.48 was issued after the design of this plant was established. Compliance with this Regulatory Guide is addressed in Section 3.13.
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SSES-FSAR Text Rev. 74 3.9.3.1.2 Reactor Vessel Assembly The reactor vessel assembly consists of the reactor pressure vessel support skirt, shroud support and shroud plate.
The reactor pressure vessel, vessel support skirt, and shroud support are constructed in accordance with Section III of the ASME Code. The shroud support consists of the shroud support plate and the shroud support cylinder and its legs. The reactor pressure vessel is an ASME Code Class I component. Complete stress reports on these components have been prepared in accordance with ASME requirements. Table 3.9-2a provides a summary of the stress criteria, load combinations, calculated and allowable stresses. The stress analysis performed for the reactor vessel assembly, including the faulted condition, were completed using elastic methods. The stress Load combinations and stress analyses for the core support structures and other reactor internals are discussed in Subsection 3.9.5.
3.9.3.1.3 Main Steam Piping The main steam piping discussed in this paragraph includes that piping extending from the reactor pressure vessel to the outboard main steam isolation valve. This piping is designed in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB-3600.
The load combinations and stress criteria for the main steam piping and pipe-mounted equipment are shown in Table 3.9-2d.
The rules contained in Appendix F of ASME Code Section III will be used in evaluating faulted loading conditions independently of all other design and operating conditions. Stresses calculated on an elastic basis will be evaluated in accordance with F-1360.
3.9.3.1.4 Recirculation Loop Piping This piping is designed in accordance with the ASME for the recirculation piping and pipe-mounted equipment Code Section III, Subsection NB-3600. The load combinations and allowables are shown in Table 3.9-2e. The rules contained in Appendix F of ASME Code Section III are used in evaluating faulted loading conditions, independently of all other design and operating conditions. Stresses calculated on an elastic basis are evaluated in accordance with F-1360.
3.9.3.1.5 Recirculation System Valves The recirculation system valves are designed in accordance with the ASME Code,Section III, Class I, Subsection NB-3500. The discharge gate valves are required to close for LPCI flow injection. Loading combinations and other stress analysis information are presented in Table 3.9-2(j).
3.9.3.1.6 Recirculation Pump In the design of the recirculation pumps, the ASME Code,Section VIII, Division 1, 1971 Edition with latest addenda was used as a guide in calculations made for determining the thickness of pressure-retaining parts, and in sizing the pressure-retaining bolting.
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SSES-FSAR Text Rev. 74 The pump vendor made calculations for the design of the pressure-containing components to include the determination of minimum wall thickness, allowable stress and pressures. The design calculations are shown in Table 3.9-2i.
Load, shear, and moment diagrams were constructed to scale, using live loads, dead loads, and calculated snubber reactions. Combined bending, tension and shear stresses were determined for each major component of the assembly, including the pump driver mount, motor flange bolting, and pump case.
Replacement pump cover gaskets have been upgraded by the pump Original Equipment Manufacturer (OEM) to eliminate the use of asbestos and to improve reliability. The replacement pump cover gaskets require a higher bolt preload than the original gaskets. The OEM prepared a Gasket Upgrade Design Report that concludes the pump subcomponents are acceptable and meet the ASME Code requirements.
The maximum combined tensile stress in the cover bolting was calculated using tensile stress from design pressure.
Combined primary stresses did not exceed 150 percent of the code allowable stress shown in Section VIII of the ASME Code, 1971 Edition.
These methods and calculations demonstrate that the pump will maintain pressure integrity at all times.
3.9.3.1.7 Standby Liquid Control (SLC) Tank The SLC tank is designed in accordance with the ASME Code,Section III. A summary of the design calculations and stress criteria used are shown in Table 3.9-2m.
3.9.3.1.8 Residual Heat Removal Heat Exchangers The RHR heat exchanger is designed in accordance with the ASME B&PV Code,Section III.
The loading combinations considered and other stress analysis are presented in Table 3.9-2o.
3.9.3.1.9 RCIC Turbine Although not under the jurisdiction of the ASME Code, the RCIC turbine has been designed and fabricated following the basic guidelines for an ASME Code Section III, Class 2 component.
The RCIC Turbine is surveillance tested according to current Technical Specifications.
Design conditions for the RCIC turbine include:
a) Auto Quick start per Technical Specification surveillance requirements.
b) Turbine Inlet - 1250 psig at saturated temperature c) Turbine Exhaust - 165 psig at saturated temperature Table 3.9-9 contains a summary of the RCIC turbine components calculated and allowable loads.
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SSES-FSAR Text Rev. 74 3.9.3.1.10 RCIC Pump The RCIC pump has been designed and fabricated to the requirements for an ASME Code Section III Class 2 component.
The RCIC pump is surveillance tested in conjunction with the RCIC turbine. Surveillance testing is performed according to current Technical Specification surveillance requirements. Design conditions for the RCIC pump include:
a) Maximum NPSHR - 21.3 feet b) Total head, maximum High speed 3060 feet Low speed 525 feet c) Constant flow rate: 625 gpm d) Normal ambient operating temperature - 60°F to 100°F e) Normal/Upset conditions which control the pump design include:
Design pressure - 1500 psig Design temperature - 40°F - 140°F Seismic loads - 2/3 of SSE Table 3.9-2r contains a summary of the design calculation for the RCIC pump components.
3.9.3.1.11 ECCS Pumps The RHR, CS and HPCI pumps are designed in accordance with the ASME Code,Section III.
The stress criteria and calculated and allowable stresses are summarized in Table 3.9-2n.
Design condition for RHR and core spray pumps are as follows:
RHR CORE SPRAY Design pressure Suction 220 psig 125 psig Discharge 500 psig 500 psig Design Temperature 40-360°F 40-212°F 3.9.3.1.12 Standby Liquid Control Pump The standby liquid control pump has been designed and fabricated to the 1968 P&V Code for Class 2 component.
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SSES-FSAR Text Rev. 74 The SLC pumps and motors are functionally tested by pumping demineralized water through a closed test loop. The SLC pumps are capable of injecting the net contents of the storage tank into the reactor in less than an hour. The pumps are capable of injecting flow into the reactor against pressure up to the second lowest spring set pressure (1195) of the reactor safety relief valves.
Design conditions for the SLC pump include:
a) Flow rate 43 gpm b) Maximum operating discharge pressure 1250 psig c) Ambient conditions:
Temperature 70°F - 120°F Relative Humidity 20% - 95%
d) Normal/upset conditions which control the pump design include:
Design pressure 1500 psig Design temperature 150°F Seismic Loads 2/3 of SSE Stress limits for the pressure boundary are the ASME Code allowable stress (1.0S) for general membrane e) Faulted or emergency conditions include:
Design pressure 1500 psig Design temperature 150°F Safe shutdown earthquake Horizontal 1.5g Vertical 0.14g A summary of the design calculations for the SLC pump components is contained in Table 3.9-2l.
3.9.3.1.13 Main Steam Isolation Valves and Safety/Relief Valves The main steam isolation and safety relief valves are designed in accordance with the requirements of the ASME Code Section III, Subsection NB-3500 for Class 1 components.
Load combination, analytical methods, calculated stresses, and allowable limits are shown in Table 3.9-2g and 3.9-2(h), respectively.
3.9.3.1.14 Safety Relief Valve Piping See Subsection 3.9.3.1.19.
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SSES-FSAR Text Rev. 74 3.9.3.1.15 High Pressure Coolant Injection (HPCI) Turbine Although not under the jurisdiction of the ASME Code, the HPCI turbine has been designed and fabricated to the basic guidelines for an ASME Code Section III as a Class 2 component.
Surveillance testing is performed according to current Technical Specification surveillance requirements.
Design conditions for the HPCI turbine include:
a) Auto-Startup -- 30 cycles per year with reactor pressure at 1150 psig peak and saturated temperature, turbine exhaust pressure at 50 psig peak and saturated temperature.
b) Turbine Inlet - 1250 psig at saturated temperature.
c) Turbine Exhaust - 200 psig at saturated temperature d) Upset conditions, which control the turbine design include:
Design pressure Design temperature Operating basis earthquake Inlet and exhaust piping nozzle loads Stress limits for pressure boundary are ASME Code allowable stress (1.0S) for general membrane and 1.5S for bending (local membrane).
e) Faulted, or emergency conditions include:
Design pressure Design temperature Safe shutdown earthquake Inlet and exhaust piping nozzle loads Stress limits for pressure boundary are 120% of ASME Code Section III allowable stress (1.2S) for general membrane and 1.8S for bending (local membrane).
f) Nozzle loading definition includes:
Upset - Inlet F = (20,000 - M)/2.5, but < 5,000 lbs.
- Exhaust F = (20,000 - M)/0.8, but < 11,500 lbs.
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SSES-FSAR Text Rev. 74 Faulted (or - Emergency) Inlet F = (30,000 - M)/2.5, but < 7,500 lbs.
Exhaust F = (30,000 - M)/0.8, but < 17,250 lbs.
Where F (lbs) and M (ft-lb) are the resultant force and moment on the respective nozzle.
A summary of the design calculations for the HPCI turbine components are shown in Table 3.9-2(ae).
3.9.3.1.16 High Pressure Coolant Injection (HPCI) Pump The HPCI pump has been designed and fabricated to the requirements for an ASME Code Section III Class 2 component.
The HPCI pump is surveillance tested in conjunction with the HPCI turbine. The HPCI pump is surveillance tested according to the current Technical Specification surveillance requirements.
The HPCI pump takes condensate from the above-ground storage tank and at design flow discharges condensate back to the above-ground storage tank via a closed test loop.
Design conditions for HPCI pump include:
a) Total head , maximum- High speed: 3060 feet Low speed: 525 feet b) Constant flow rate - 5000 gpm c) Normal ambient operating temperature - 60°F to 100°F d) Normal plus Upset conditions which control the pump design include:
Design pressure - 1500 psig Design temperature - 40°F - 140°F Seismic Loads - 2/3 of SSE Suction nozzle loads - F = 1940 lb M = 2460 ft-lbs Discharge nozzle loads - F = 3715 lbs M = 4330 ft-lbs Stress limits for pressure boundary are ASME Code allowable stress (1.0S) for general membrane and 1.5S for bending (local membrane).
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SSES-FSAR Text Rev. 74 e) Faulted, or Emergency conditions include:
Design pressure - 1500 psig Design temperature - 40°F - 140°F Safe shutdown earthquake - Horizontal - 1.50g Vertical - 0.14g Stress limits for pressure boundary are 120% of ASME Code allowable stress (1.2S) for general membrane and 1.8S for bending (local membrane).
f) Nozzle loading Pump nozzles are subject to loading from the connecting pipe. The method of analysis shows the maximum resultant moment is due to pipe reaction. The maximum resultant force shall not exceed the allowable. Allowable nozzle forces and moments are expressed as:
Normal plus Upset Suction - F = 33,000-0.79M Discharge - F = 32,000-1.54M Emergency:
Suction - F = 43,000-0.74M Discharge - F = 47,000-1.23M The calculated stress values are compared to allowable stresses for critical components in Table 3.9-2t.
3.9.3.1.17 Reactor Water Cleanup (RWCU) System The RWCU pump and heat exchangers are not part of a safety system and are not designed to Seismic Category I requirements.
However, the requirements for ASME Code Section III, Class 3 components are used as guidelines in evaluating the RWCU system components.
The design loading combinations and limits for the pump include the following:
a) Normal plus upset loads: This includes the simultaneous effect of normal operating loads, design pressure, temperature, nozzle loads from connected piping, dead weight loads, seismic loads, plus torsional loads due to rotating parts.
b) Seismic loading: This equipment and supports are designed to withstand the static seismic forces applied at the mass center, assuming that the pump is flooded.
FSAR Rev. 70 3.9-47
SSES-FSAR Text Rev. 74 c) Stresses in the supports and the anchor bolts due to seismic loads are combined with the stresses due to other live and dead loads and operating loads. The allowable stress for this combination of loads is based on the allowable stress as set forth in the ASME Code Section III.
d) Equipment operates between 70°F and 532.3°F. Transient analysis is not required for Class III components in this temperature range.
Tables 3.9-2(p) and 3.9-2(c) show the calculated stress values and allowable stress limits for the pump and heat exchangers, respectively.
3.9.3.1.18 This Section Has Been Intentionally Deleted 3.9.3.1.19 ASME Code Constructed Items Not Furnished with the NSSS The design loading combinations categorized with respect to plant operating conditions identified as normal, upset, emergency, and faulted for ASME code constructed items are presented in Table 3.9-6.
The method of combining the peak loads on components and supports resulting from different dynamic events was addressed by the Mark II Owners Subgroup on SRSS. The generic resolution has been reviewed and applies to Susquehanna SES.
The design criteria and stress limits associated with each of the plant operating conditions for each type of ASME code constructed item are presented in Tables 3.9-7, 3.9-8, 3.9-9, 3.9-10, 3.9-11 and 3.9-12.
The component operating condition will be the same as the plant operating condition, except for active pumps or valves for which, the emergency or faulted plant condition is considered a normal operating condition.
3.9.3.2a NSSS Pump and Valve Operability Assurance The active NSSS pumps and valves are listed in Table 3.9-3.
Active mechanical equipment classified as Seismic Category I are designed to perform their function during the life of the plant under postulated plant conditions. Equipment with faulted condition functional requirements include "active" (active equipment must perform a mechanical motion during the course of accomplishing a safety function) pumps and valves in fluid systems such as the emergency core cooling system. Operability is assured by satisfying the requirements of the following programs. Safety-related valves are qualified by prototype testing and analysis satisfying stress and deformation criteria at all critical locations and safety-related active pumps by analysis with suitable stress limits and nozzle loads. The content of these programs is detailed below.
3.9.3.2a.1 ECCS Pumps All active pumps are qualified for operability by first being subjected to rigid tests before installation in the plant. The in-shop tests include (1) hydrostatic tests of pressure-retaining parts to 125% of the design pressure times the ratio of material allowable stress at room temperature to the allowable stress value at the design temperature, (2) seal leakage tests, (3)
FSAR Rev. 70 3.9-48
SSES-FSAR Text Rev. 74 performance tests, while the pump is operated with flow, to determine total developed head, minimum and maximum head, Net Positive Suction Head (NPSH) requirements and other pump/motor parameters. Also monitored during these operating tests are bearing temperatures (except water cooled bearings) and vibration levels. Both are shown to be below specified limits. After the pump is installed in the plant, it undergoes the cold hydro tests, functional tests, and the required periodic in-service inspection and operation. These tests demonstrate reliability of the pump for the design life of the plant.
In addition to these tests, the safety-related active pumps are analyzed for operability during a faulted condition by assuring that (1) the pump will not be damaged during the seismic and hydrodynamic event, and (2) the pump will continue operating despite the faulted loads.
3.9.3.2a.1.1 Analysis of Loading, Stress, and Acceleration Conditions In order to avoid damage during the faulted plant condition, the stresses caused by the combination of normal operating loads, SSE, and dynamic system loads are limited to the material elastic limit, as indicated in Table 3.9-2. A three-dimensional finite element model of the pump/motor and its supports is developed using the response spectrum method of dynamic analysis. The same model is analyzed for static nozzle loads, pump thrust loads, and deadweight. Critical location stresses are evaluated and compared with the allowable criteria.
The average membrane stress m) for the faulted conditions loads are maintained at 1.2S, or approximately 0.75 y(y - yield stress), and the maximum stress in local fibers (m + bending stress o) is limited to 1.8S, or approximately 1.1 y. The maximum dynamic nozzle loads are considered in an analysis of the pump supports to assure that a system misalignment cannot occur.
Performing these analyses with the conservative loads stated and with the restrictive stress limits of Table 3.9-2 as allowables, will assure that critical parts of the pump will not be damaged during the faulted condition; therefore, the reliability of the pump for post-faulted condition operation will not be impaired by the seismic and hydrodynamic events.
A dynamic analysis is made to determine the seismic load from the applicable floor response spectra. Analysis is made to check that faulted condition nozzle loads and dynamic accelerations will not impair the operability of the pumps during or following the faulted condition. Components of the pump, when having a natural frequency above 33 Hz, are considered essentially rigid. This frequency is considered sufficiently high to avoid problems with amplification between the component and structure for all seismic areas.
If the natural frequency is found to be below 33 Hz, an analysis is performed to determine the amplified input accelerations necessary to perform the static analysis. The adjusted accelerations will be determined using the same conservatisms contained in the horizontal and vertical accelerations used for "rigid" structures. The static analysis is performed using the adjusted accelerations; the stress limits stated in Table-3.9-2 must still be satisfied.
3.9.3.2a.1.2 Pump Operation During and Following Faulted Condition Loading Active pump/motor rotor combinations are designed to rotate at a constant speed under all design conditions. Motors are designed to withstand short periods of severe overload. The high rotary inertia in the operating pump rotor, and the nature of the random, short duration loading characteristics of the seismic and hydrodynamic event, will prevent the rotor from becoming seized. In actuality, the dynamic loadings will cause only a slight increase, if any, in FSAR Rev. 70 3.9-49
SSES-FSAR Text Rev. 74 the torque (i.e., motor current) necessary to drive the pump at the constant design speed.
Therefore the pump will not shutdown during the faulted event and will operate at the design speed despite the faulted loads.
The functional ability of the active pumps after a faulted condition is assured since only normal operating loads and steady state nozzle loads exist. For the active pumps, the faulted condition is more severe than the normal condition only due to seismic and hydrodynamic loads on the equipment itself and the increase in nozzle loads due to the SSE on the connecting pipe. The SSE event is infrequent and of relatively short duration compared to the design life of the equipment. Since it is demonstrated that the pumps would not be damaged during the faulted condition, the post-faulted condition operating loads will be no worse then the normal plant operating limits. This is assured by requiring that the imposed nozzle loads (steady-state loads) for normal conditions and post-faulted conditions be limited by the magnitudes of the normal condition nozzle loads. The post-faulted condition ability of the pumps to function under these applied loads is proven during the normal operating plant conditions for active pumps.
3.9.3.2a.2 SLC Pump and Motor Assembly and RCIC Pump Assembly These equipment assemblies are small, compact, rigid assemblies, with natural frequencies well above 33 Hz. With this fact verified, each equipment assembly is qualified by static analysis only. This static qualification verifies operability under seismic and hydrodynamic conditions, and assures structural loading stresses within Code limitations.
3.9.3.2a.3 RCIC Turbine Analysis and testing done on the RCIC turbine is covered by Subsections 3.9.2.2, 3.9.3.1, and Table 3.9-2q.
3.9.3.2a.4 ECCS Motors Qualification of the Class 1E motors used for the ECCS motors is in compliance with IEEE Standard 323-1971. The qualification of motors of all sizes is based on completion of a type test, followed up with review and comparison of design and material details and seismic analysis of production units, ranging from 500 to 3500 Bhp. The motor is used in the type test.
All manufacturing, inspection, and routine tests performed by motor manufacturers on production units are performed on the test motor.
The type test has been performed on a 1250 HP vertical motor in accordance with IEEE Standard 323-1971. Normal operation during the design life is first simulated, then the motor is subjected to a number of seismic events. Then the abnormal environmental condition possible during and after a loss of coolant accident (LOCA) is simulated. The test plan for the type test was as follows:
a) Thermal aging of the motor electrical insulation system (which is a part of the stator only) was based on IEEE Standard 275-1966. The amount of aging equaled the total estimated operating days at maximum insulation surface temperature.
b) Radiation aging of the motor electrical insulation equals the maximum estimated integrated dose of gamma radiation during normal and abnormal conditions.
FSAR Rev. 70 3.9-50
SSES-FSAR Text Rev. 74 c) The normal operating inducted vibration effect on the insulation system has been simulated by 1.5g horizontal vibration acceleration of 60 Hz current frequency for one hour duration.
d) Motor bearings are selected and their operating life is established based on bearing manufacturer's test and operating data using the loads calculated to act on the bearing.
e) The dynamic load deflection analysis on the rotor shaft, performed to ensure adequate rotation clearance, has been verified by static loading and deflection of the rotor for the type-test motor.
f) Dynamic load aging and testing has been performed on a biaxial test table in accordance with IEEE 344-1975. During this type test the shake table was activated simulating the maximum design limit of the safe shutdown earthquake and hydrodynamic loads with motor starts and operation combination as may possibly occur during a plant life.
g) An environmental test simulating a LOCA condition with 100 days duration time has been performed with the test motor fully loaded to simulate pump operation. The test consisted of startup and six hours operation at 212°F ambient temperature and 100%
steam environment. Another startup and operation of the test motor after one hour stand-still in the same environment was followed by sufficient operation at high humidity and temperature. The operation was based on the temperature-life characteristic curve from IEEE 275-1966 for the insulation type used on the ECCS motors.
3.9.3.2a.5 NSSS Valves The Class 1 Active Valves are the Main Steam Isolation Valves, Safety/Relief Valves, Recirculation Discharge and Bypass Gate Valves, Standby Liquid Control Valves and Control Rod Drive Scram Discharge Volume Vent and Drain Valves. Each of these valves is dynamically qualified for operability in a manner unique to its design. Therefore, each method of qualification is detailed individually below.
3.9.3.2a.5.1 Main Steam Isolation Valve The MSIV's are evaluated for operability during dynamic acceleration by both analysis and test.
This analysis for MSIV operability is completed in two separate ways. First the valve body is designed in accordance with the ASME Code Section III Class 1 which limits deformation to be within the elastic limit of the material by limiting pressure and pipe reaction input loads (including seismic and hydrodynamic loads). This assures only small deformation in the operating area of the valve body, hence, there is no interference with valve operability.
A static deflection test was conducted on a MSIV of similar design to assure operability under maximum deformation from seismic loading. A maximum static load equivalent to 8 g's applied perpendicular to the actuator axis centerline resulted in no significant change in valve closure rate and no change in measured seat leakage following termination of the load.
To assure that design limits are not exceeded for both piping input loads and actuator dynamic loads, the MSIV is mathematically modeled in the Main Steam Line System Analysis. The valve input loads, amplified accelerations, and resonance frequencies are determined based on site excitation input to the system for MSIV as a part of the overall steamline analysis. Pipe anchors FSAR Rev. 70 3.9-51
SSES-FSAR Text Rev. 74 and restraints are applied as required to limit pipe system resonance frequencies and amplified accelerations to within acceptable limits for the MSIV's. Additional details on the analysis of these valves is shown on Table 3.9-2(h).
The main steam isolation valve operability during LOCA conditions was demonstrated as defined in the report APED-5750 (March 1969). The test specimen was a 20" valve of a design representative of the MSIV's.
3.9.3.2a.5.2 Main Steam Safety/Relief Valves The safety/relief valves are qualified for operability during seismic and hydrodynamic events by both design and test.
The valve is designed for the largest moments that can occur in service. These are 400,000 in-lbs and 300,000 in-lbs at the inlet and outlet, respectively. These moments are resultants due to dead weight plus dynamic seismic effects of 3 g's horizontal and 1 g vertical of both valve and connecting pipe, thermal expansion of the connecting pipe, and reaction forces from valve discharge. A production S/R valve demonstrated operability while being dynamically (shake table) tested at loads greater than equipment design limit loads.
A mathematical model of this valve is included in the main steam line system analysis. This analysis assures the design limits are not exceeded.
The safety relief valves are generically qualified by testing for seismic and hydrodynamic loads.
The natural frequencies are determined to be greater than 33 Hz for seismic and 60 Hz for hydrodynamic loads.
Additional details on the analysis of these valves are shown in Table 3.9-2g.
3.9.3.2a.5.3 Recirculation (Discharge and Bypass) Gate Valves Recirculation discharge and bypass gate valves are evaluated for operability during seismic and hydrodynamic events by both analysis and test.
Motor operators were generically qualified to IEEE 382-1980 which requires a dynamic test to verify the absence of any natural frequencies below 33 Hz and then a demonstration of operability during dynamic testing. The operators have been qualified to acceleration levels of 10 g from 2 Hz to 100 Hz.
The valve are designed in accordance with the ASME Code,Section III Class 1 design rules.
The discharge valves are designed to seismic accelerations of 9.8 gs horizontal and 2.188 gs vertical including gravity. Both valves extended structures were analyzed to show that they could withstand both compressive stresses and bending stresses imposed by the seismic accelerations. The valve fundamental frequencies were determined by frequency analysis to be less than the seismic cut-off frequency of 33 Hz. This required dynamic analysis considering multinode response. However, since the valves are pipe-mounted and the required response spectra at the valve location were not available, it was necessary to perform a dynamic analysis on the entire piping system. A simple lumped-mass model of the valve and its actuator was developed based on the valve fundamental frequency, and was used to represent the valve dynamic characteristics in the piping analysis.
FSAR Rev. 70 3.9-52
SSES-FSAR Text Rev. 74 Dynamic piping analysis indicated that the loads imposed on these valves are less than the design allowable loads. Additional details on analysis of the discharge valves are shown in Table 3.9-2(j).
3.9.3.2a.5.4 Explosive Valves The SLC explosive valve has been qualified to IEEE 344-1975. The qualification test included a demonstration of the absence of natural frequencies below 35 Hz and the ability to remain operable under a horizontal seismic coefficient of 6.5g and a vertical seismic coefficient of 4.5g at 33 Hz.
3.9.3.2a.5.5 CRD Scram Discharge Volume Vent and Drain Valves The CRD Scram Discharge Vent and Drain Valves are evaluated for operability during dynamic acceleration by test. The testing consisted of a combination of vibration aging testing. SRV cycling induced fatigue load testing and seismic testing at acceleration levels based upon both upset and faulted required response spectra (RRS). The valve successfully passed all qualification testing.
3.9.3.2a.6 HPCI Turbine The HPCI turbine is dynamically qualified by static analysis. The turbine assembly and its components were considered to be supported as designed, and horizontal/vertical accelerations were applied to the mass's center of gravity. The magnitude of the acceleration coefficients was 1.50 horizontal and 0.48 vertical. The results of the analysis indicate that the turbine assembly is capable of sustaining the above loadings without overstressing any component.
The turbine was dynamically qualified via dynamic testing by the 1st quarter 1983 in accordance with IEEE 344-1975. The qualification test program demonstrated start-up, steady state operability, and shutdown capabilities.
3.9.3.2b Non-NSSS Pump and Valve Operability Assurance The pumps under this category are:
a) Diesel oil transfer pumps b) RHR service water pumps c) Emergency service water pumps d) Control structure chiller - cooling water pumps.
All the above pumps are Class 3.
3.9.3.2b.1 Pumps The pumps listed above are subjected to testing both in the manufacturer's shop and following their installation to verify that they meet the criteria required by the respective specifications.
FSAR Rev. 70 3.9-53
SSES-FSAR Text Rev. 74 During manufacture, nondestructive test procedures including liquid penetrant examination, radiographic examination, magnetic particle inspection, and ultrasonic inspection are applied to the pumps. All these procedures are performed in accordance with ASME Code,Section III.
After the pumps have been assembled they are hydrostatically tested and performance tested in the manufacturer's shop in accordance with the Hydraulic Institute's standards.
After the pumps are installed they will undergo functional tests. Provisions will be made for in-service inspection and operational testing.
All these tests demonstrate that the pumps are reliable and will function as specified.
3.9.3.2b.1.1 Analysis of Loading, Stress, and Acceleration Conditions In addition to the tests and procedures referred to above, the pumps are seismically analyzed to ensure that they will be capable of operating both during and after the OBE and DBE.
In performing these analyses, conservative seismic accelerations and stress criteria were used; this ensures that critical parts of the pump are not damaged during a seismic event and that the pump can still operate following such an event.
3.9.3.2b.1.2 Pump Operation During and Following SSE Loading Each pump/motor combination is designed to rotate at a constant speed under all conditions unless the rotor becomes completely seized, i.e., with no rotation. Motors are designed to withstand short periods of severe overload and, typically, the rotor can be seized five full seconds before a circuit breaker shuts down the pumps. However, the high rotary inertia in the operating pump rotor, and the nature of the random, short duration loading characteristics of the seismic event, will prevent the rotor from becoming seized. In actuality, the seismic loadings will cause only a slight increase in the torque (i.e., motor current) necessary to drive the pump at the constant design speed. Therefore, the pump will not shut down during the event and will operate at the design speed despite the seismic loads.
From previous discussions, it is evident that the pump/motor units will withstand seismic loadings and, therefore, will perform their intended functions. These proposed requirements take into account the complex characteristics of the pump and are sufficient to demonstrate and ensure the seismic operability of these pumps. Post-seismic condition operating loads will be no worse than the normal plant operating limits.
3.9.3.2b.2 Valves Active ASME Class 1, 2, and 3 valves are identified in the Plants ISI program manual.
Safety related active valves are subjected to a series of tests prior to service and during the plant life. Prior to installation, the following tests are performed: shell hydrostatic test in accordance with ASME Code Section III requirements, backseat and main seat leakage tests, disc hydrostatic test, functional tests which verify that the valve will open and close within the specified time limits, and operability qualification of motor operators for the environmental conditions over the installed life (i.e., aging, radiation, accident environment simulation, etc). in accordance with IEEE 382-1972. After installation, cold hydrostatic construction tests, functional tests in accordance with the requirements of Chapter 14, and periodic in-service FSAR Rev. 70 3.9-54
SSES-FSAR Text Rev. 74 operation in accordance with the requirements of Chapter 16, are performed to verify and ensure the functional ability of the valve.
The valves are designed using either stress analyses or the pressure-containing minimum wall thickness requirements. On all active valves with extended topworks, an analysis is also performed for static equivalent SSE loads applied at the center of gravity of the extended structure. The maximum stress limits allowed in the analyses demonstrate structural integrity and are equal to the limits recommended by the ASME Code for the particular ASME class of valve analyzed as listed in Tables 3.9-7 and 3.9-12. Loading combinations are listed in Tables 3.9-6 and 3.9-14. In addition to these tests and analyses, a representative valve of each design type is tested for verification of operability during a simulated seismic event by demonstrating operational capabilities within the specified limits.
A. Selection of Representative Valve The valves requiring operability qualification are divided into different groups: by valve manufacturer, valve type, size, pressure class, material type (carbon steel, stainless steel, and alloy steel) and actuator type (AC electric, DC electric, air, hydraulic, etc.)
Valve sizes that cover the range of sizes in service are qualified as shown in Table 3.9-15 by tests, and the results are used to qualify all valves within the intermediate range of sizes. A tabulation is made of the weight of the valve actuator, the actuator thrust margin (a ratio of the maximum thrust available from the actuator divided by the design thrust required for the valve), and the yoke configuration (as related to stiffness) for each valve assembly. For a range of qualified valve sizes, as defined by the qualification table, the valve assembly with the heaviest actuator, lowest thrust margin, and least stiff yoke is picked as the test unit. In those cases where a test unit is not readily apparent, more than one unit is tested to provide a conservative test position. This procedure is repeated within each group until all listed units are represented by a test unit, and for each group until all the necessary valves are represented by a test unit.
In addition to the tests, the stress calculations for each valve assembly are reviewed.
A tabulation is made for all qualified valve assemblies comparing the yoke stress for all valve classes, the yoke-flange to body and the yoke-flange to actuator-bolting stresses, as applicable, for all classes of valves, and the body stress for Class 1 valves. This is done to provide further analytical justification for the qualification of non-tested valves by tested valves.
B. Qualification Testing Procedures The valve is mounted in a manner that will conservatively represent typical valve installations. The valve unit includes the actuator and all appurtenances usually attached to the valve in service. The operability of the valve during a SSE is demonstrated by satisfying the following criteria:
a) All the active valves with topworks are designed to have a first natural frequency greater than 33 Hz. This is shown by suitable test or analysis.
FSAR Rev. 70 3.9-55
SSES-FSAR Text Rev. 74 b) The extended topworks of the valve are subjected to a statically applied equivalent seismic load. Load is applied at the center of gravity of the topworks in the direction of the weakest axis of the yoke. The design pressure of the valve or the design pressure of the system is simultaneously applied to the valve during the static load tests.
c) The valve is then operated at minimum specified actuation supply voltage or pressure with equivalent seismic static load applied. The valve must perform its safety-related function within the specified operating time limits.
d) Motor operators and other electrical appurtenances necessary for operation are qualified as operable during the SSE in accordance IEEE-344-1975 prior to their installation on the valve.
For valves with and without the topworks supported, the statically applied load envelopes the specified G-force times the weight of the topworks. This load is generally greater than would result from 3.0 g horizontal and 3.0 g vertical. For valves with the topworks supported, additional loading from thermal and/or anchor movements may be imposed upon the valve through the support(s). If the loads due to thermal and/or anchor movements are greater than the statically applied load (which envelopes inertial forces), stress and critical deflection analyses are performed on the valve (considering the maximum applied loading) as an acceptable qualification alternative.
An exception to the above described seismic qualification approach is the RHR throttle valves (HV151F017A/B and HV251F017A/B) which were not tested with a static seismic load but instead were qualified by a combination of static seismic analysis and static deflection operability analysis. The static deflection operability analysis verified that adequate internal clearances exist to insure the binding does not occur within the valve during and after a design basis event.
The piping designer limits the valve accelerations and support loads to allowable values as determined by the qualification test and analysis.
The valve is leak tested following the test described above to show that the valve has not been damaged. The leak rates must not exceed the original allowable leakage rate specified for the valve.
The above testing program applies only to valves with overhanging structures, i.e., the motor operator or air actuator assembly. Because of their simple characteristics, check and other compact valves are not affected by seismic acceleration. Check valves have no extended structures to distort the valve and cause a malfunction. Check valve discs are designed to allow sufficient clearance around the disc to prevent distortions from nozzle or other imposed loads.
Accordingly, check valves are qualified by a combination of the following tests and analysis:
a) The air-operated check valves are analyzed to ensure that the air cylinder cannot impair the ability of the valve to operate as a simple check valve during seismic loading. No functional test simulating seismic loading is performed. Air operators on check valves do not perform a safety function.
b) In-shop hydrostatic test FSAR Rev. 70 3.9-56
SSES-FSAR Text Rev. 74 c) In-shop seat leakage test d) Periodic valve exercise and inspection to ensure the functional ability of the valve in accordance with the requirements of Chapter 16.
Using the methods described, all the safety-related active valves in the systems are qualified for operability during a seismic event. These methods conservatively simulate the seismic event and ensure that the active valves will perform their safety-related functions when necessary.
3.9.3.3a Design and Installation of NSSS Supplied Pressure Relief Devices 3.9.3.3a.1 Main Steam Safety/Relief Valves Safety/relief valve lift results in a transient that produces momentary unbalanced forces acting on the discharge piping system for the period from opening of the safety/relief valve until a steady discharge flow from the reactor pressure vessel to the suppression pool is established.
This period includes clearing of the water slug from the end of the discharge piping submerged in the suppression pool. Pressure waves traveling through the discharge piping following the relatively rapid opening of the safety/relief valve cause the safety/relief valve discharge piping to vibrate. This in turn produces forces that act on the main steam piping.
The analysis of the relief valve discharge transient consists of a stepwise time history solution of the fluid flow equation to generate a time-history of the fluid properties at numerous locations along the pipe. Simultaneously, reaction loads on the pipe are determined at each location corresponding to the position of an elbow. These loads are composed of pressure-times-area, momentum change, and fluid friction terms. Figure 3.9-2 shows a set of fluid property and pipe section load transients typical of those produced by relief valve discharge.
The method of analysis applied to determine piping system response to relief valve operation is time history integration. The forces are applied at locations on the piping system where fluid flow changes direction, thus causing momentary reactions. The resulting loads on the safety/
relief valve, the main steam line, and the discharge piping are combined with loads due to other effects as specified in Subsection 3.9.3.1. The Code stress limits corresponding to load combinations classification as normal, upset, emergency and faulted, are applied to the steam and discharge pipe.
3.9.3.3b Design and Installation Details for Mounting of Pressure Relief Devices in ASME Code Class 1 and 2 Systems The design of pressure relieving devices can be grouped into two categories: open discharge and closed discharge.
a) Open Discharge There are no open discharge pressure relieving devices mounted on ASME Code Class 1 and 2 systems.
b) Closed Discharge A closed discharge system is characterized by piping between the valve and a tank, or some other terminal end. Under steady-state conditions, there are no net unbalanced FSAR Rev. 70 3.9-57
SSES-FSAR Text Rev. 74 forces. The initial transient response and resulting stresses are determined by using either a time-history computer solution or a conservative equivalent static solution. In calculating initial transient forces, pressure and momentum terms are included. Water slug effects are also considered.
Time history dynamic analysis is performed for the discharge piping and its supports.
The effect of the loading on the header is also considered. The design load combinations for a given transient are shown in Table 3.9-2, and the design criteria and stress limits for the relief valve are shown in Table 3.9-2g.
3.9.3.4 Component Supports 3.9.3.4.1 Recirculation Piping Supports The NSSS-designed recirculation piping supports are designed in accordance with Subsection NF of ASME Code Section III. (Non-NSSS designed pipe supports on recirculation piping are in accordance with Subsection 3.9.3.4.6.) Supports are either designed by load rating per paragraph NF-3260, or to the stress limits for linear supports per paragraph NF-3231. In general, the load combinations for the various operating conditions correspond to those used to design the supported pipe. Design transient cyclic data are not applicable to piping supports as no fatigue evaluation is necessary to meet the Code requirements.
The design criteria and dynamic testing requirements for component supports are as follows:
Component Supports All components supports are designed, fabricated, and assembled so that they cannot become disengaged by the movement of the supported pipe or equipment after they have been installed.
Hangers The design load on hangers is the load caused by dead weight. The hangers are calibrated to ensure that they support the design load at both their hot and cold load settings. Hangers provide a specified down travel and up travel in excess of the specified thermal movement.
Snubbers The design load on snubbers includes those loads caused by dynamic forces (OBE and SSE) and hydrodynamic loads, system anchor movements, and reaction forces caused by relief valve discharge, turbine stop valve closure, etc.
The snubbers are designed and load rated in accordance with NF-3000 to be capable of carrying the design load for all operating conditions. Faulted condition design uses the criteria outlined in Appendix F of the ASME code. They are designed to be able to carry the load under normal, upset, emergency, and faulted loading conditions.
The snubbers are also tested dynamically to ensure that they can perform as required in the following manner:
a) The snubber will be subjected to either force or displacement that varies approximately as the sine wave.
FSAR Rev. 70 3.9-58
SSES-FSAR Text Rev. 74 b) The frequency (Hz) of the input motion or force will be verified at small increments within the specified range.
c) The resulting relative displacements and corresponding loads across the working components, including end attachments, will be recorded.
d) The test will be conducted with the snubber at various temperatures representative of operating conditions.
e) The rated load in both tension and compression will be equal to or higher than the peak load.
f) The duration of the test at each frequency will be specified.
Dynamic Testing The criterion used to demonstrate the operability of the snubber under dynamic loading conditions is that the total travel of the unit, including lost motion and deflection during dynamic load cycling, shall not exceed +/-.060 inches (.120 inch total).
The dynamic testing on a prototype snubber consists of the following:
(a) The snubber is subjected to load cycling at 100%, 75%, 50% of the rated load and this load varies approximately as the sine wave function.
(b) The frequency of the input force is varied from 3 to 33 Hz in 3 Hz steps.
(c) The duration of the test (load cycles can be determined by this time interval) at each frequency is 10 seconds as a minimum.
Struts The design load on struts includes those loads caused by dead weight, thermal expansion, primary dynamic forces, i.e., (OBE) and SSE, system anchor displacements, and reaction forces caused by relief valve discharge, turbine stop valve closure, etc.
Struts are designed in accordance with NF-3000 to be capable of carrying the design load for all operating conditions.
3.9.3.4.2 Reactor Pressure Vessel Support Skirt The permissible compressive load on the reactor vessel support skirt cylinder (modeled as plate and shell type component support) was limited by the GE design specification to 90 percent of the load which produces yield stress, divided by the safety factor for the condition being evaluated. The effects of fabrication and operational eccentricity was included. The safety factor for faulted conditions was 1.125.
An analysis of reactor pressure vessel support skirt buckling for faulted conditions shows that the support skirt has the capability to meet ASME Code Section III, Paragraph F-1370(c) faulted condition limits of 0.67 times the critical buckling strength of the support at temperature. The FSAR Rev. 70 3.9-59
SSES-FSAR Text Rev. 74 faulted condition analyzed included the compressive loads due to the design basis maximum earthquake, the overturning moments and shears due to the jet reactor load resulting from a severed pipe, and the compressive effects on the support skirt due to the thermal and pressure expansion of the reactor vessel. The expected maximum earthquake loads for the reactor vessel support skirts are less than 60% of the maximum design basis loads used in the buckling analysis described; therefore, the expected faulted loads are well below the critical buckling limits of Paragraph F-1370(c) for the vessel support skirt. The expected earthquake loads were determined using the seismic dynamic analysis methods described in Section 3.7.
The loading condition, stress criterion, calculated and allowable stresses are summarized in Table 3.9-2a.
3.9.3.4.3 NSSS Floor-Mounted Equipment (Pumps, Heat Exchangers and RCIC Turbine)
The RHR pump, core spray pump, RHR heat exchangers, RCIC pump, SLC pump, and RCIC turbine are all analyzed to verify the adequacy of their support structures under various plant operating conditions. In all cases, the calculated loads in the critical support areas are within the ASME Code allowables.
3.9.3.4.4 Supports for ASME Code Class 1, 2 and 3 Active Components ASME Code Class 1, 2 and 3 active components are either pumps or valves. Since valves are supported by piping and not tied to building structures, pipe design criteria govern.
Seismic Category I active pumps supports are qualified for seismic and hydrodynamic loads by testing when the pump supports along with the pumps are fulfilling the following conditions:
- 1) Simulate actual mounting conditions.
- 2) Simulate all static and dynamic loadings on the pump.
- 3) Monitor pump operability during testing.
- 4) The normal operation of the pump during and after the test indicates that the supports are adequate. Any deflection or deformation of the pump supports which precludes the operability of the pump, is not accepted; and,
- 5) Supports are inspected for structural integrity after the test. Any cracking or permanent deformation is not accepted.
Seismic and hydrodynamic qualification of component supports by analysis is generally accomplished as follows:
- 1) Stresses at all support elements and parts such as pumps holddown, and baseplate holddown bolts, pump support pads, pump pedestal, and foundation are checked to be within the allowable limits as specified in ASME Subsection NF.
- 2) For Normal and Upset plant conditions, the deflections and deformations of the supports are assured to be within the elastic limits and not to exceed the values permitted by the designer based on his design verification tests to ensure the operability of the pumps.
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SSES-FSAR Text Rev. 74
- 3) For Emergency and Faulted plant conditions, the deformations must not exceed the values permitted by the designer to ensure the operability of the pumps. Elastic/plastic analysis will be performed if the deflections are above the elastic limits.
3.9.3.4.5 HPCI Turbine This section has been intentionally deleted.
3.9.3.4.6 Non-NSSS Supports The design loading combinations for supports for ASME Code Class 1, 2, and 3 components, categorized with respect to plant operating conditions identified as normal, upset, emergency, and faulted, are given in Table 3.9-14. This table also provides the stress limits for each plant operating condition.
The loads imposed on the ASME Class 1, 2 and 3 active valves and pumps are limited to values meeting both the manufacturer's and code allowables to insure operability of the active components by the design of the supports. The supports are designed to remain elastic under the maximum loads. The minor local deformations associated with the elastic deformation of the support will not impair operability of the active components.
3.9.3.4.6.1 Snubbers Snubbers will be used in seismic Category I systems. Snubbers will be purchased with load ratings appropriate for the design conditions and load combinations.
3.9.3.4.6.1.1 Snubber Design Specification The specification for the purchase of shock suppressors (snubbers) covers the following related to snubber design, supplier's performance qualification tests and load tests. Mechanical snubbers specified for Susquehanna SES are addressed below. Design specification information pertaining to hydraulic snubbers is also provided below.
Design Criteria Mechanical Snubbers a) The frictional resistance of purchased suppressors shall not exceed 2% of the rated load. However, for suppressors tested during in-service inspection, the frictional resistance may exceed 2% of the rated load provided that the effects of this increase on the Seismic Category I systems on which the suppressors are used have been evaluated and shown to be acceptable.
b) All purchased snubbers shall be designed such that they limit the acceleration of the pipe to a maximum of 0.02g when subjected to any load up to rated load. An evaluation of installed snubbers and the Seismic Category I systems on which they are used indicates that the permissible acceleration can be increased to 0.04g maximum, with some exceptions that are limited to 0.02g as defined in the snubber program.
c) The total pipe movement a long the axis of the suppressor shall not exceed +/-.060 inches due to any applied dynamic cycle load from 3 to 33 cps up to the rated load at the unit.
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SSES-FSAR Text Rev. 74 d) The suppressor shall be designed for an exposure to a temperature of 40°F prior to initial startup and 200°F during continuous operations and to a relative humidity of 55 percent normally and 90 percent during shutdown. The radiation exposure shall be 100 Roentgen/hour.
Performance Test: Two types of tests are required.
Production Test: This type of test is required to be performed on each unit.
a) Check unit to confirm acceleration level is less than specified maximum.
b) Check unit to confirm that it operates freely over the total stroke.
c) Measure and record the force required to initiate motion over the stroke in tension and compression.
d) Measure and record lost motion of the snubber mechanism.
Qualification Tests: These types of tests are to be performed on randomly selected production models. These tests are used to demonstrate the required load performance (load rating) and specified displacement when subjected to dynamic load cycling. Also included in these tests are low temperature, high temperature, humidity, salt/sand and dust spray test, life test and faulted load test.
Hydraulic Snubbers The following environmental conditions apply to each hydraulic snubber:
Normal Operation Emergency Faulted Temperature range 32°F to 176°F 32°F to 302°F 32°F to 302°F 0°C to 80°C 0°C to 150°C 0°C to 150°C Humidity Up to 100% Up to 100% Up to 100%
Pressure 1 bar 1 bar 5 bar Radiation (*total accum) 107 rad 107 rad 107 rad Max. no. cycles 20000 40 1 The following performance testing is required for each hydraulic snubber:
The frictional resistance, including breakaway, in compression and tension shall not exceed 1.5% of normal load for a normal load > 4500 lbs.
The standard lock-up velocity in compression and tension shall be between 4.7 ipm to 14.2 ipm.
The standard bleed velocity shall be between 0.47 imp to 4.7 ipm at normal load.
FSAR Rev. 70 3.9-62
SSES-FSAR Text Rev. 74 The piston rod travel during a dynamic functional test under rated load from 2 to 35 cps shall be kept at < +/- 0.16.
3.9.3.4.6.1.2 Snubber Analysis Model A piping system is idealized as a mathematical model consisting of lumped masses connected by massless elastic members. The elastic members are given the properties of the piping system being analyzed. The lumped masses are carefully located to adequately represent the dynamic and elastic properties of the piping system. A lumped mass is located at the beginning and end of every elbow, valve, at the extended valve operator, and at the intersection of every tee. On straight runs, lumped masses are located at spacings no greater than the span length corresponding to 33 cps. A mass point is located at every extended mass to account for torsional effects on the piping system. In addition, the increased stiffness and mass of valves is considered in the modeling of a piping system.
The three-dimensional stiffness matrix of the mathematical model is determined by the direct stiffness method. Axial, shear, flexural and torsional deformations of each member are included. For curved members and branch connectors a decreased stiffness is used in accordance with ASME Section III. The mass matrix is also calculated.
Snubbers are considered to be rigid members in the dynamic model. Differences in tension and compression spring rates will not effect design calculations; similarly entrapped air and temperature do not effect mechanical snubbers. Hydraulic snubbers manufactured by LISEGA have a pressurized reservoir that precludes air entrapment and environmental temperature range limits are provided in the design specification.
The load conditions and combinations are being addressed as a generic issue and are included in the SSES plant Design Assessment Report (DAR). The transients analyzed will include:
- 1. Seismic
- 3. Flow disruption transients (e.g., fast valve closure)
- 4. Normal Operating Loads.
Snubber locations and sizes are chosen to maintain the stresses due to the above listed loads to below the ASME code allowable stresses. Since all of the loads described above will be maintained below the code allowable stresses for the plant condition indicated in the DAR, the snubber with the appropriate load rating will be used.
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SSES-FSAR Text Rev. 74 3.9.4 CONTROL ROD DRIVE SYSTEM This plant is equipped with a hydraulic control rod drive system. The discussion in this section includes the Control Rod Drive Mechanism (CRDM), the Hydraulic Control Unit (HCU), the Condensate Supply System and the Scram discharge volume and extends to the coupling interface with the control rods.
3.9.4.1 Descriptive Information on CRDS Descriptive information on the control rod drives as well as the entire control and drive system is contained in Section 4.6.
3.9.4.2 Applicable CRDS Design Specifications The Control Rod Drive System (CRDS) is designed to meet the functional design criteria as outlined in Section 4.6 and consists of the following:
a) Locking piston control rod drive; b) Hydraulic control unit; c) Hydraulic power supply (pumps),
d) Interconnecting piping, e) Flow and pressure and isolation valves, f) Instrumentation and electrical controls.
Those CRD components forming part of the primary pressure boundary are designed according to ASME Code Section III.
The quality group classification of the CRD hydraulic system is outlined in Table 3.2-1, and the components are designed according to the codes and standards governing the individual quality groups.
Pertinent aspects of the design and qualification of the CRD components are discussed in the following locations: transients in Subsection 3.9.1.1, faulted conditions in Subsection 3.9.1.4, and dynamic testing in Subsection 3.9.2.2.
3.9.4.3 Design Loads, Stress Limits, and Allowable Deformation The ASME Code components of the CRDs and CRD Housings have been evaluated analytically and the design load combinations and stress limits for the CRD housing are listed in Table 3.9-2v. For the non-code components, experimental testing was used to assure the CRD performance under all possible conditions as described in Subsection 3.9.4.4.
Deformations are not a limiting factor in the analysis of the CRDs components based upon the results of numerous tests on the drive.
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SSES-FSAR Text Rev. 74 3.9.4.3.1 Control Rod Drive Housing Supports The Control Rod Drive (CRD) housing support system functions are described in Section 4.6.
The American Institute of Steel Construction (AISC) Manual of Steel Construction, "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," was used in designing the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses used were 90% of yield and the allowable shear stress used was 60% of yield. These design stresses are 1.5 times the AISC allowable stresses (60% and 40% of yield, respectively).
The CRD housing supports are designed as Seismic Category I equipment.
3.9.4.4 CRD Performance Assurance Program The CRD test program consists of the following tests:
a) Development tests b) Factory Quality Control Tests c) 5 year Maintenance Life tests d) 1.5X Design Life tests e) Operational tests f) Acceptance tests g) Surveillance tests All of the above tests except c) and d) are discussed in Subsections 4.6.3 through 4.6.3.1.1.5.
Tests c) and d) are discussed below:
c) "5 Year Maintenance Life" Tests Four Control Rod Drives are picked at random from the production stock each year and subjected to various tests under simulated reactor conditions and 1/8 of the cycles specified in Subsection 3.9.1.1. Upon completion of the test program, the control rod drive parts are checked to the drawings and all parts must meet or surpass the minimum specified requirements.
d) 1.5X Design Life Tests When a significant design change is made to the components of the drive, the drive is subjected to a series of tests equivalent to 1.5 times the life test cycles specified in Subsection 3.9.1.1.
Two CRDs were tested in 1976. Upon completion of the test program, the CRDs were disassembled and the parts checked to the drawing for wear and/or damage. All parts met or surpassed the minimum specified requirements.
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SSES-FSAR Text Rev. 74 3.9.5 REACTOR PRESSURE VESSEL INTERNALS This subsection identifies and discusses the structural and functional integrity of the major reactor pressure vessel internals.
3.9.5.1 Design Arrangements The core support structures and reactor vessel internals (exclusive of fuel, control rods, CRDs, and incore nuclear instrumentation) are identified below:
Core Support Structures Shroud Shroud support Core plate and holddown bolts Top guide (including bolts and keepers)
Fuel supports Control rod guide tubes Control rod drive housing Reactor Internals Jet Pump assemblies and instrumentation
- Feedwater spargers Vessel head spray nozzle Differential pressure and liquid control lines In-core flux monitor tubes
- Initial startup neutron sources
- Surveillance sample holders Core spray lines and spargers
- In-Core instrument housings
- Steam dryer
- Shroud head and steam separator assembly
- Guide rods FSAR Rev. 70 3.9-66
SSES-FSAR Text Rev. 74 CRD thermal sleeves
- Non-safety class components A general assembly drawing of the important reactor components is shown in Figure 3.9-3.
The floodable inner volume of the reactor pressure vessel can be seen in Figure 3.9-4. It is the volume inside the core shroud up to the level of the jet pump suction inlet.
The design arrangement of the reactor internals, such as the jet pumps, steam separators and guide tube, is such that one end is unrestricted and thus free to expand.
3.9.5.1.1 Core Support Structures The core support structures consist of those items listed in Subsection 3.9.5.1. These structures form partitions within the reactor vessel, to sustain pressure differentials across the partitions, direct the flow of the coolant water, and laterally locate and support the fuel assemblies. Figure 3.9-4 shows the reactor vessel internal flow paths.
3.9.5.1.1.1 Shroud The shroud support, shroud, and top guide make up a stainless steel cylindrical assembly that provides a partition to separate the upward flow of coolant through the core from the downward recirculation flow. This partition separates the core region from the Downcomer annulus, thus providing a floodable region following a recirculation line break. The volume enclosed by this assembly is characterized by three regions. The upper portion surrounds the core discharge plenum, which is bounded by the shroud head on top and the top guide's grid plate below. The central portion and the shroud surrounds the active fuel and forms the longest section of the assembly. This section is bounded at the bottom by the core support. The lower portion, surrounding part of the lower plenum, is welded to the reactor pressure vessel shroud support.
3.9.5.1.1.2 Shroud Head and Steam Separator Assembly The shroud head and steam separator assembly is bolted to the top of the top guide to form the top of the core discharge plenum. This plenum provides a mixing chamber for the steam-water mixture before it enters the steam separators. Individual stainless steel axial flow steam separators are attached to the top of standpipes that are welded into the shroud head. The steam separators have no moving parts. In each separator, the steam-water mixture rising through the standpipe passes vanes that impart a spin to establish a vortex separating the water from the steam. The separated water flows from the lower portion of the steam separator into the Downcomer annulus.
3.9.5.1.2 Core Plate The core plate consists of a circular stainless steel plate with bored holes stiffened with a rim and beam structure. The plate provides lateral support and guidance for the control rod guide tubes, in-core flux monitor guide tubes, peripheral fuel supports, and startup neutron sources.
The last two items are also supported vertically by the core support plate.
The entire assembly is bolted to a support ledge on the lower portions of the shroud.
FSAR Rev. 70 3.9-67
SSES-FSAR Text Rev. 74 3.9.5.1.3 Top Guide The top guide is formed by a series of stainless steel beams joined at right angles to form square openings and fastened to a peripheral rim. Each opening provides lateral support and guidance for four fuel assemblies or in the case of peripheral fuel, one or two fuel assemblies.
Sockets are provided in the bottom of the beam intersections to anchor the in-core flux monitors and startup neutron sources. The rim of the top guide rests on a ledge between the upper and central portions of the shroud. The top guide has alignment pins that engage and bear against slots in the shroud which are used to correctly position the assembly before it is secured.
Lateral restraint is provided by wedge blocks between the top guide and the shroud wall.
3.9.5.1.4 Fuel Support The fuel supports, shown in Figure 3.9-5 are of two basic types; namely, peripheral supports and four-lobed orificed fuel supports. The peripheral fuel support is located at the outer edge of the active core and is not adjacent to control rods. Each peripheral fuel support will support one fuel assembly and contains a single orifice assembly designed to assure proper coolant flow to the peripheral fuel assembly. Each four-lobed orificed fuel support will support four fuel assemblies and is provided with four orifice plates to assure proper coolant flow distribution to each rod-controlled fuel assembly. The four-lobed orificed fuel supports rest in the top of the control rod guide tubes which are supported laterally by the core support. The control rods pass through slots in the center of the four-lobed orificed fuel support. A control rod and the four adjacent fuel assemblies represent a core cell. (see Subsection 4.4.2).
3.9.5.1.5 Control Rod Guide Tubes The control rod guide tubes, located inside the vessel, extend from the top of the control rod drive housings up through holes in the core support plate. Each tube is designed as the guide for a control rod and as the vertical support for a four-lobed orificed fuel support piece and the four fuel assemblies surrounding the control rod. The bottom of the guide tube is supported by the control rod drive housing, which in turn transmits the weight of the guide tube, fuel support, and fuel assemblies to the reactor vessel bottom head. A thermal sleeve is inserted into the control rod drive housing from below and is rotated to lock the control rod guide tube in place.
A key is inserted into a locking slot in the bottom of the control rod drive housing to hold the thermal sleeve in position.
3.9.5.1.6 Jet Pump Assemblies The jet pump assemblies are located in two semi-circular groups in the downcomer annulus between the core shroud and the reactor vessel wall. The design and performance of the jet pump is covered in detail in References 3.9-1 and 3.9-2. Each stainless steel jet pump consists of driving nozzles, suction inlet, throat or mixing section, and diffuser (see Figure 3.9-6). The driving nozzle, suction inlet, and throat are joined together as a removable unit, and the diffuser is permanently installed. High pressure water from the recirculation pumps is supplied to each pair of jet pumps through a riser pipe welded to the recirculation inlet nozzle thermal sleeve. A riser brace consists of cantilever beams welded to a riser pipe and to pads on the reactor vessel wall.
The nozzle entry section is connected to the riser by a metal-to-metal, spherical-to-conical seal joint. Firm contact is maintained by a hold- down clamp. The throat section is supported laterally by a bracket attached to the riser. There is a slip-fit joint between the throat and FSAR Rev. 70 3.9-68
SSES-FSAR Text Rev. 74 diffuser. The diffuser is a gradual conical section changing to a straight cylindrical section at the lower end.
3.9.5.1.7 Steam Dryers The steam dryers remove moisture from the wet steam leaving the steam separators. The extracted moisture flows down the dryer vanes to the collecting troughs, then flows through tubes into the downcomer annulus. A skirt extends from the bottom of the dryer vane housing to the steam separator standpipe, below the water level. This skirt forms a seal between the wet steam plenum and the dry steam flowing from the top of the dryers to the steam outlet nozzles.
The steam dryer and shroud head are positioned in the vessel during installation with the aid of vertical guide rods. The dryer assembly rests on steam dryer support brackets attached to the reactor vessel wall. Upward movement of the dryer assembly, which would occur only under accident conditions, is restricted by steam dryer hold-down brackets attached to the reactor vessel top head.
3.9.5.1.8 Feedwater Spargers The feedwater spargers are stainless steel headers located in the mixing plenum above the downcomer annulus. A separate sparger is fitted to each feedwater nozzle and is shaped to conform to the curve of the vessel wall. Sparger end brackets are pinned to vessel brackets to support the spargers. Feedwater flow enters the center of the spargers and is discharged radially inward to mix the cooler feedwater with the downcomer flow from the steam separators and steam dryer before it contacts the vessel wall. The feedwater also serves to condense the steam in the region above the downcomer annulus and to subcool the water flowing to the jet pumps and recirculation pumps.
3.9.5.1.9 Core Spray Lines The core spray lines are the means for directing flow to the core spray nozzles which distribute coolant during accident conditions.
Two core spray lines enter the reactor vessel through the two core spray nozzles. (see Section 5.4). The lines divide immediately inside the reactor vessel. The two halves are routed to opposite sides of the reactor vessel and are supported by clamps attached to the vessel wall.
The lines are then routed downward into the downcomer annulus and pass through the upper shroud immediately below the flange. The flow divides again as it enters the center of the semicircular sparger, which is routed halfway around the inside of the upper shroud. The two spargers are supported by brackets designed to accommodate thermal expansion. The line routing and supports are designed to accommodate differential movement between the shroud and vessel. The other core spray line is identical except that it enters the opposite side of the vessel and the spargers are at a slightly different elevation inside the shroud. The correct spray distribution pattern is provided by a combination of distribution nozzles pointed radially inward and downward from the spargers (see Section 6.3).
3.9.5.1.10 Vessel Head Spray Nozzle When reactor coolant is returned to the reactor vessel part of the flow can be diverted to a spray nozzle in the reactor head. This spray maintains saturated conditions in the reactor vessel head FSAR Rev. 70 3.9-69
SSES-FSAR Text Rev. 74 volume by condensing steam being generated by the hot reactor vessel walls and internals.
The spray also decreases thermal stratification in the reactor vessel coolant. This ensures that the water level in the reactor vessel can rise. The higher water level provides conduction cooling to more of the mass of metal of the reactor vessel and, therefore, helps maintain the cooldown rate.
The vessel head spray nozzle is mounted to a short length of pipe and a flange, which is bolted to a mating flange on the reactor vessel head nozzle (see Subsection 5.4.7).
3.9.5.1.11 Differential Pressure and Liquid Control Line The differential pressure and liquid control line serves a dual function within the reactor vessel -
to provide a path for the injection of the liquid control solution into the coolant stream and to sense the differential pressure across the core support plate (described in Section 5.4). This line enters the reactor vessel at a point below the core shroud as two concentric pipes. In the lower plenum, the two pipes separate. The inner pipe terminates near the lower shroud with a perforated length below the core support plate. It is used to sense the pressure below the core support plate during normal operation and to inject liquid control solution if required. This location facilitates good mixing and dispersion. The inner pipe also reduces thermal shock to the vessel nozzle should the standby liquid control system be actuated. The outer pipe terminates immediately above the core support plate and senses the pressure in the region outside the fuel assemblies.
3.9.5.1.12 In-Core Flux Monitor Guide Tubes In-core flux monitor guide tubes provide a means of positioning fixed detectors in the core as well as provide a path for calibration monitors (TIP System).
The in-core flux monitor guide tubes extend from the top of the in-core flux monitor housing (see Section 5.4) in the lower plenum to the top of the core support plate. The power range detectors for the power range monitoring units and the dry tubes for the source range monitoring and intermediate range monitoring (SRM/IRM) detectors are inserted through the guide tubes. A lattice work of clamps, tie bars, and spacers give lateral support and rigidity to the guide tubes. The bolts and clamps are welded, after assembly, to prevent loosening during reactor operation.
3.9.5.1.13 Surveillance Sample Holders The surveillance sample holders are welded baskets containing impact and tensile specimen capsules (see Section 5.4). The baskets hang from the brackets that are attached to the inside wall of the reactor vessel and extend to mid-height of the active core. The radial positions are chosen to expose the specimens to the same environment and maximum neutron fluxes experienced by the reactor vessel itself while avoiding jet pump removal interference or damage.
3.9.5.2 Design Loading Conditions 3.9.5.2.1 Events to be Evaluated Examination of the spectrum of conditions for which the safety design basis must be satisfied reveals four significant faulted events:
FSAR Rev. 70 3.9-70
SSES-FSAR Text Rev. 74 a) Recirculation Line Break: a break in a recirculation line between the reactor vessel and the recirculation pump suction.
b) Steam line break accident: a break in one main steam line between the reactor vessel and the flow restrictor. The accident results in significant pressure differentials across some of the structures within the reactor.
c) Earthquake: subjects the core support structures and reactor internals to significant forces as a result of ground motion.
d) Safety relief valve discharge in combination with an SSE.
Analysis of other conditions existing during normal operation, abnormal operational transients, and accidents shows that the loads affecting the core support structures and other engineered safety feature reactor internals are less severe than these three postulated events.
The faulted conditions for the reactor pressure vessel internals are discussed in Subsection 3.9.1.4. Loading combination and analysis for the reactor pressure vessel internals are discussed in Subsection 3.9.3.l, Tables 3.9-1 and 3.9-2.
3.9.5.2.2 Pressure Differential During Rapid Depressurization A digital computer code is used to analyze the transient conditions within the reactor vessel following the recirculation line break accident and the steam line break accident. The analytical model of the vessel consists of nine nodes, which are connected to the necessary adjoining nodes by flow paths having the required resistance and inertial characteristics. The program solves the energy and mass conservation equations for each node to give the depressurization rates and pressure in the various regions of the reactor. Figure 3.9-7 shows the nine reactor nodes. The computer code used is the General Electric Short-Term Thermal-Hydraulic Model described in Reference 3.9-3. This model has been approved for use in ECCS conformance evaluation under 10CFR50, Appendix K. In order to adequately describe the blowdown pressure effect on the individual assembly components, three features are included in the model that are not applicable to the ECCS analysis and are, therefore, not described in Reference 3.9-3. These additional features are discussed below:
a) The liquid level in the steam separator region and in the annulus between the dryer skirt and the pressure vessel is tracked to more accurately determine the flow and mixture quality in the steam dryer and in the steamline.
b) The flow path between the bypass region and the shroud head is more accurately modeled since the fuel assembly pressure differential is influenced by flashing in the guide tubes and bypass region for a steamline break. In the ECCS analysis, the momentum equation is solved in this flow path, but its irreversible loss coefficient is conservatively set at an arbitrary low value.
c) The enthalpies in the guide tubes and the bypass are calculated separately, since the fuel assembly P is influenced by flashing in these regions. In the ECCS analysis, these regions are lumped.
FSAR Rev. 70 3.9-71
SSES-FSAR Text Rev. 74 3.9.5.2.3 Recirculation Line and Steam Line Break 3.9.5.2.3.1 Accident Definition Both a recirculation line break (the largest liquid break) and an inside steam line break (the largest steam break) are considered in determining the design basis accident for the engineered safety feature reactor internals. The recirculation line break is the same as the design basis loss-of-coolant accident described in Section 6.3. A sudden, complete circumferential break is assumed to occur in one recirculation loop. The pressure differentials on the reactor internals and core support structures are in all cases lower than for the main steam line break.
The analysis of the steam line break assumes a sudden, complete circumferential break of one main steam line between the reactor vessel and the main steam line restrictor. A steam line break upstream of the flow restrictors produces a larger blowdown area and thus a faster depressurization rate than a break downstream of the restrictors. The larger blowdown area results in greater pressure differentials across the reactor internal structures.
The steam line break accident produces significantly higher pressure differentials across the reactor internal structures than does the recirculation line break. This results from the higher reactor depressurization rate associated with the steam line break. Therefore, the steam line break is the design basis accident for internal pressure differentials.
3.9.5.2.3.2 Effects of Initial Reactor Power and Core Flow The maximum internal pressure loads can be considered to be composed of two parts: steady-state and transient pressure differentials. For a given plant the core flow and power are the two major factors which influence the reactor internal pressure differentials. The core flow essentially affects only the steady-state part. For a fixed power, the greater the core flow, the larger will be the steady-state pressure differentials. The core power affects both the steady-state and the transient parts. As the power is decreased, there is less voiding in the core and consequently the steady-state core pressure differential is less. However, less voiding in the core also means that less steam is generated in the reactor pressure vessel and thus the depressurization rate and the transient part of the maximum pressure load is increased. As a result, the total loads on some components are higher at low power.
To ensure that the calculated pressure differences bound those which could be expected if a steam line break should occur, an analysis is conducted at a low power-high recirculation flow condition in addition to the standard safety analysis condition at high power, rated recirculation flow. The power chosen for analysis is the minimum value permitted by the recirculation system controls at rated recirculation drive flow (that is, the drive flow necessary to achieve rated core flow at rated power.)
This condition maximizes those loads which are inversely proportional to power. It must be noted that this condition, while possible, is unlikely; first, because the reactor will generally operate at or near full power; second, because high core flow is neither required nor desirable at such a reduced power condition.
FSAR Rev. 70 3.9-72
SSES-FSAR Text Rev. 74 3.9.5.2.4 Seismic and Hydrodynamic Loads The seismic and hydrodynamic loads acting on the structures within the reactor vessel are based on a dynamic analysis as described in Section 3.7. Seismic analysis is performed by coupling the lumped mass model of the reactor vessel and internals, as described in Section 3.7, with the building model to determine the acceleration force and moment time histories in the reactor vessel and internals. This is accomplished by using the modal superpostion method.
Acceleration response spectra are also generated for subsystem analyses of selected components.
3.9.5.3 Design Loading Categories Loading combinations for the core support structures are shown in Table 3.9-2. The basis for determining faulted loads on the reactor internals is shown for dynamic loads in Section 3.7 and for pipe rupture loads in Subsection 3.9.5.2.3 and 3.9.5.4.3. Table 3.9-2b shows allowable and calculated stress values for highly stressed core support structures and selected reactor internal components. Table 3.9-2aa provides this same type of information for the CRD guide tubes.
Stress intensity and other design limits are discussed in Subsection 3.9.5.4.4. The core support structures which are fabricated as part of the reactor pressure vessel assembly are discussed in Subsection 3.9.1.3 in conjunction with the reactor pressure vessel.
The design requirements for equipment classified as "other" e.g., steam dryers and shroud heads, were specified by the designer with appropriate consideration of the intended service of the equipment and expected plant and environmental conditions under which it will operate.
Where possible, design requirements are based on applicable industry codes and standards.
If these are not available, the designer relies on accepted industry or engineering practices.
3.9.5.4 Design Bases 3.9.5.4.1 Safety Design Bases The reactor core support structures and internals shall meet the following safety design bases:
- 1) They shall be arranged to provide a floodable volume in which the core can be adequately cooled in the event of a breach in the nuclear system process barrier external to the reactor vessel.
- 2) Deformation shall be limited to assure that the control rods and core standby cooling systems can perform their safety functions.
- 3) Mechanical design of applicable structures shall assure that safety design bases (1) and (2), above, are satisfied so that the safe shutdown of the plant and removal of decay heat are not impaired.
3.9.5.4.2 Power Generation Design Bases The reactor core support structures and internals shall be designed to the following power generation design bases:
FSAR Rev. 70 3.9-73
SSES-FSAR Text Rev. 74
- 1) They shall provide the proper coolant distribution during all anticipated normal operating conditions to full power operation of the core without fuel damage.
- 2) They shall be arranged to facilitate refueling operations.
- 3) They shall be designed to facilitate inspection.
3.9.5.4.3 Response of Internals Due to Inside Steam Break Accident It is concluded that the maximum pressure loads acting on the reactor internal components result from an inside steam line break, and on some components the loads are greatest with operation at the minimum power associated with the maximum core flow. This has been substantiated by the analytical comparison of liquid versus steam breaks and by the investigation of the effects of core power and core flow.
It has also been pointed out that, although possible, it is not probable that the reactor would be operating at the rather abnormal condition of minimum power and maximum core flow. More realistically, the reactor would be at or near a full power condition and thus the maximum pressure loads acting on the internal components would be less.
3.9.5.4.4 Stress, Deformation, and Fatigue Limits for Reactor Internals (Except Core Support Structure)
These limits are summarized in Table 3.9-2(b).
Design Condition SF Normal 2.25 Upset 2.25 Emergency 1.5 Faulted 1.125 Components inside the reactor pressure vessel such as control rods which must move during accident condition have been examined to determine if adequate clearances exist during emergency and faulted conditions. No mechanical clearance problems have been identified.
No plastic deformation occurs in the reactor internal components during emergency or faulted conditions as shown in Subsections 3.9.4 and 3.9.3.1, and Table 3.9-2. This is used in demonstrating that no mechanical interferences exist. No fatigue analysis is required under the faulted conditions due to the low encounter frequency of faulted events and the low number of cycles. The forcing functions applicable to the reactor internals are discussed in Subsection 3.9.2.5.
3.9.5.4.5 Stress, Deformation, and Fatigue Limits for Core Support Structures These limits are summarized in Tables 3.9-2a, 3.9-2b, and 3.9-2v.
3.9.6 IN-SERVICE TESTING OF PUMPS AND VALVES The construction permits for the Susquehanna Steam Electric Station were issued in November 1973. Relating this date to the requirements of 10CFR50.55a(g), the preservice examination FSAR Rev. 70 3.9-74
SSES-FSAR Text Rev. 74 program, including provisions for design and access to enable operational readiness testing of pumps and valves, complied, as a minimum, with the 1971 Edition of the ASME B&PV Code Section XI including Addenda through Summer 1972.
This ASME Code Edition does not require preservice and in-service testing of pumps and valves to ensure operational readiness. The requirements for in-service testing of pumps and valves were added as Subsections IWV and IWP to ASME B&PV Code Section XI, Summer 1973 Addenda, effective December 30, 1973. By then, design and procurement for SSES was under way; however, the preservice testing program for pumps and valves for assessing operational readiness was conducted, to the extent practical, so that it complied with requirements of the 1974 Edition, ASME B&PV Code through Winter 1975 Addenda.
The first 120 months' in-service tests will assess operational readiness of pumps and valves.
These tests complied, to the extent practical within design limitations, with the requirements of 10CFR50.55a.
During successive 120-month periods, in-service tests of pumps and valves for assessing operational readiness will comply, to the extent practical within design limitations, with the requirements of 10CFR50.55a.
3.
9.7 REFERENCES
3.9-1 "Design and Performance of G.E. BWR Jet Pumps," General Electric Company, Atomic Power Equipment Department, APED-5460, July 1968.
3.9-2 Moen, H.H., "Testing of Improved Jet Pumps for the BWR/6 Nuclear System," General Electric Company, Atomic Power Equipment Department, NEDO-10602, June 1972.
3.9-3 General Electric Company, "Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K," Proprietary Document, General Electric Company, NEDE-20566.
3.9-4 Not Used 3.9-5 Not Used 3.9-6 Seismic Analysis of Piping Systems, BP-TOP-1, Bechtel Power Corporation, San Francisco, California, Rev. 2, January, 1975.
3.9-7 "Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants,"
NEDE-24057-P (Class III) and NEDO-24057 (Class I), November, 1977.
3.9-8 "Functional Capability Criteria for Essential Mark II Piping," NEDO-21985, 78 NED174 (Class I), September, 1978.
3.9-9 "Power Uprate Engineering Report for Susquehanna Steam Electric Station, Units 1 and 2," NEDC-32161P, As Revised by PP&L Calculation EC-PUPC-1001, Revision 0, March, 1994.
3.9-10 BWR Jet Pump Assembly Maintenance Issues, General Electric Company, San Jose, CA, June 2002 FSAR Rev. 70 3.9-75
TABLE 3.9-1 TRANSIENTS AND THE NUftBEB OF ASSOCIATED CYCLES CONSIDERED IN THE DESIGN AND FATIGUE ANALYSIS OF THE RPV ASSE~BLJ ANO INTERNAL TRANSIENTS
- 1. Bolt Up* 123
- 2. Desiqn Hydrostatic Test 130
- 3. Startup (100°F/hr Heatup Rate)** 117
- 4. Daily Reduction to 751 Power* 10,000
- 5. Weekly Reduction to 501 Power* 2,000
- 6. Control Rod Pattern Cbange* 400
- 7. Loss of Feeawater Heaters (80 Cycles Total): 80
- 8. 50% safe Shutdown Earthquake Event at Rated Operating Conditions 10****
- 9. Scram:
- a. Turbine Generator Trip, Feedvater On, Isolation Valves Stay Open 40
- b. Other Scrams 140
- 10. Reduction to 01 Power, Hot Standby vith main condenser available, Shutdown (100°F/hr Cooldovn Rate)** 111
- 11. Unbolt 123
- 12. Pre-op Blovdovn 10
- 13. Natural Circulation Startup 3
- 14. Loss of AC Power, Natural Circulation Restart 5
~mergencLConditions
- 15. Scram:
- a. Reactor Overpressure with Delayed Scram, Peedvater Stays on, Isolation Valves Stay Open 1***
Rev. 35, 07/84
SSES-PSAB Irsrn§.i~n.t§ E ergenc1_condit!2ns(Continued)
- 16. Scram a.
b.
c.
lutoaatic Blovdovn Loss of Feedvater Pulps, Isolation ValYes Closed 5
- d. Sinqle Safety or Belief Valve Blovdown 8
- 17. ! proper Start of Cold Becirculation Loop
- 18. Sudden Start of
- 19. ! proper Pu p I
in Cold Recirculation Loop Startup with Reactor Drain Shot Off Fault~~_condi~ion
- 20. Pipe Rupture and Blovdovn
- 21. Saf~ Sbutdovn Earthquake at Rated Operatinq Conditions 1***
- Applies to reactor pressure vessel only.
- Bulk averaqe vessel coolant teapera~ure change in aoy 1-hour period.
- The annual encounter probability for the one cycle events is< 10-z for emerqency and 10 for faulted events.
- Includes 10 axi u load cycles per event. Rot required to be considered in fatique analysis due to low encounter frequency (<10-2) and lov number of cycles.
Rev. 35, 07 /84 ,
SSES-FSAR This table lists the major echanical safety related aechanical components in the plant. Various parts of the table are referenced in Section 3.9. The foraat in the various parts of the table is consistent. since variations e1ist on analytical ethods and depth of detail necessar, to de onstrate the safety aspe=ts of various coap~nents.
Rev. 35, 07/84
SSES-fSAR
~QHTiH7:~ -
l.9-2 LOAD CO"BINATION AND ACCEPTANCE CBITEBIA POB 1S!E CODE CLASS 1~ 2, AND 3 PIPING AID CO"PONEMTS 3.9-2a REACTOB PRESSURE VESSEL IND SHROUD SUPPORT lSSEftBLf (i) Vessel Sapport Skirt (ii) Shroud Support (i ii l BPV Feedvater lozzle (i V) CBD Penetration - CRD Rousinq (V) CBD Penetration - stub tube 3.9-2b REACTOR INTERNALS & lSSOCilTBD BQUIP!EMT (i) Top Guide - Hiqbest stressed Beam (ii) Core Plate (Liqament in Top Plate)
(iii) Vessel Head Spray Nozzle 3.9-2c REACTOR VAT!R CLEANUP (REGENERATIVB & IOM-REGENEBlTIYE)
HEAT EXCHANGERS 3.9-2d lSftE CODE CLASS 1 ftAIN STEAN PIPIN~ AND PIPE BOOHTED EOOlPftENT 3.9-2e AS~E CODE CLASS 1 BECIRCULATIOI PIPING ARD PIPE "OONTED EQUIPMENT 3.9-2f NOT USED 3.9-2q "AIN STEA~ SlPETY/BELIEP VALVES 3.9-2h ftAIN STEAft ISOLATION VALVE 3.9-21 RECIRCOLATION PUftP REACTOR RECIRCULATION SfSTE" GlTB VALYBS
- 3. 9-2 lc CLASS III SAFETY RELIEF VALVE DISCHARGE PIPING 3.9-21 STANDBY LIQUID CONTROL PU"P
- 3. 9-2 m STANDBY LIQUID COtfTBOL TANK
- 3. 9-2 n !CCS PtJKPS (i) RRR Pumps (ii) Core Spray Pomps 3.9-20 RESIDUAL HEAT RE~OVlL .JBHBl HEAT EXCHANGER Rev. 35, 07/84
SS!S-FSAR 3.9-2p REACTOR WATER CLElNUP (BWcu, PO!IP
- 3. 9-2q RCIC TDBBINE 3.9-2r RCIC PUMP 3.9-2s REACTOR REFUELING & SERVICING EQOIPIIBVT (il Puel storaqe Racks (ii) Fuel Preparation ftachine
- 3. 9-2t HIGH PRESSORE COOLANT INJECTION (HPCI) PDlH, 3.9-2u CONTROL BOO DRIVE (INDEX TUBE) 3.9-2v CONTROL BOO DRIVE HOUSING
- 3. 9-2v JET POPIPS 3.CJ-2x NOT USED 3.9-2, NOT USED 3.9-2z NOT USED 3.9-2aa CONTROL BOD GOIDE TOBE 3.9-2ab IMCORE HOUSING 3.9-2ac NOT USED 3.9-2ad NOT USED 3.9-2ae RPCI TOBBIRE DESIGN CALCULATIONS Rev. 35, 07/84
TABLE 3.9-2 LOAD COMBINATION AND ACCEPTANCE CRITERIA FOR ASME CODE CLASS 1, 2, AND 3 NSSS PIPING AND EQUIPMENT Page 1 of 2 Design Evaluation Service Load Combination Basis Basis Level N + SRV (ALL) Upset Upset (B)
N + OBE Upset Upset (B)
N + OBE + SRV (ALL) Emergency Upset (B)
N + SSE+ SRV(ALL) Faulted Faulted* (0)
N + SBA+ SRV Emergency Emergency* (C)
N + IBA+ SRV Faulted Faulted* (D)
N + SBA+ SRV{ADS) Emergency Emergency* (C)
N + SBA+ OBE + SRV (ADS) Faulted Faulted* (D)
N + IBA+ OBE + SRV(ADS) Faulted Faulted* (D)
N + SBA/IBA+ SSE+ SRV(ADS) Faulted Faulted* (O)
N + LOCA** + SSE Faulted Faulted* (D)
NOTE; All dynamic loads are combined by SRSS LOAD DEFINITION LEGEND Norrnal(N) - Normal and/or abnormal loads depending on acceptance criteria.
OBE - Operational basis earthquake loads.
SSE - Safe Shutdown earthquake loads.
SRVALL - The loads induced by the actuation of all safety/relief valves which activate within milliseconds of each other (e.g.,
turbine trip operational transient).
- The loads induced by the actuation of safety/relief valves associated with Automatic Depressurzation System which actuate within milliseconds of each other during the postulated small or intermediate size pipe rupture.
SRV - Safety/relief-valve-discharge-induced loads from two adjacent valves (one valve actuated when adjacent valve is cycling).
Rev. 40, 09/88
LOCI - The loss of coolant accident associate~ with tie postul4ted pipe rupture of larqe pipes (e.g., aain stea1, feedwater, recirculation pipinq).
LOCA l - Pool swell ~,tslf!ll2Yt-l2!~§ on piping and co1ponents located between tbe aain *ent discharge outlet sod the suppression pool water upper surface.
LOCA - Pool swell il~A~t-12A~§ on piping and co1ponents located 2 above the suppression pool water upper surface.
tOCA - oscillating pressure induced loads on sab erged piping ana 3
components durinq condensation oscillations.
LOCA - Buildinq motion induced loads fro* choqqinq.
4 LOCA S - B~ildinq otioo induced loads fro1 aain *ent air clearing.
LOCA 6 - Vertical and horizontal loa~s on ain vent piping.
LOCA 7 - Annulus pressurization loads.
SBA The abnor*~l transients associated vitb a S all Break lcci4ent.
IBA - The abnor~al transie~ts associated with sn Interaediate Break Accident.
All lS~E coae Class 16 2. and J pipinq systems that are required to function for safety shutdown under the postu13ted events shall eet the requirements of NRc*s "Interim Technical Position-Function capability of oassive components" - by "EB.
- The most severe load combination amonq LOCA.
Rev. 35. 07/84
SSES-FSAR Table Rev. 54 TABLE 3.9-2a REACTOR PRESSURE VESSEL AND SHROUD SUPPORT ASSEMBLY ASME B&PV CODE SEC. III LOADING PRIMARY STRESS ALLOWABLE MAXIMUM SUBSECTION NB PRIMARY TYPE STRESS CALCULATED STRESS LIMIT CRITERIA -(psi) STRESS (psi)
(i) VESSEL SUPPORT SKIRT MATERIAL: SA-516 GR-70 A. NORMAL & UPSET CONDITION:
Pm Sm 1. Deadweight PRIMARY MEMBRANE 19,150 14,566 U1 Sm = 19,150 @ 575°F 2. Pressure < 14,566 U2 Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 28,725 20,330 U1 1.5 Sm = 28,725 @ 575°F 4. SRV PLUS BENDING < 20,330 U2 B. EMERGENCY CONDITION:
Pm Sy 1. Deadweight PRIMARY MEMBRANE 29,425 20,565 U1 Sy = 29,425 @ 546°F 2. Pressure < 20,565 U2 Pm + Pb 1.5 Sy 3. DBA PRIMARY MEMBRANE 44,150 29,377 U1 1.5 Sy = 44,138 @ 546 °F 4. SRV PLUS BENDING < 29,377 U2 C. FAULTED CONDITION:
Pm 2.4 Sm 1. Deadweight PRIMARY MEMBRANE 45,960 36,410 2.4 Sm = 45,960 @ 546°F 2. Pressure Pm + Pb 3.6 Sm 3. Annulus PRIMARY MEMBRANE 68,940 60,570 3.6 Sm = 68,940 @ 546°F PLUS BENDING Pressurization
- 4. SSE
- 5. LOCA D. MAXIMUM CUMULATIVE FATIGUE USAGE FACTOR: 0.913 U1 and 0.888 U2 At Skirt Base Junction These CUFs account for the original 40-year plant design. During the period of extended operation, actual fatigue design margins are periodically evaluated in accordance with the Plant Transient and Fatigue Monitoring Program.
FSAR Rev. 65 Page 1 of 5
SSES-FSAR Table Rev. 54 TABLE 3.9-2a (Continued)
REACTOR PRESSURE VESSEL AND SHROUD SUPPORT ASSEMBLY ASME B&PV CODE SEC. III LOADING PRIMARY STRESS ALLOWABLE MAXIMUM SUBSECTION NB PRIMARY TYPE STRESS CALCULATED STRESS LIMIT CRITERIA -(psi) STRESS (psi)
(ii) SHROUD SUPPORT MATERIAL: INCONEL SB-168 A. NORMAL & UPSET CONDITION:
Pm Sm 1. Deadweight PRIMARY MEMBRANE 23,300 7,600 Sm = 23,300 @ 575°F 2. Pressure Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 35,000 11,600 1.5 Sm = 34,950 @ 575°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
Pm Sy 1. Deadweight PRIMARY MEMBRANE 35,000 7,600 Sy = 28,125 @ 575°F 2. Pressure Pm + Pb 1.5 Sy 3. SBA PRIMARY MEMBRANE 54,400 11,600 1.5 Sy = 42,188 @ 575°F 4. SRV PLUS BENDING C. FAULTED CONDITION:
(1)
Pm Sy 1. Deadweight PRIMARY MEMBRANE 35,000 15,300 Sy = 28,125 @ 575°F 2. Pressure Pm + Pb 1.5 Sy 3. Annulus PRIMARY MEMBRANE 52,400 23,400 1.5 Sy = 42,188 @ 575°F Pressurization PLUS BENDING
- 4. SSE D. MAXIMUM CUMULATIVE FATIGUE USAGE FACTOR: 0.41 At Vessel - Support Plate Junction. This CUF accounts for the original 40-year plant design. During the period of extended operation, actual fatigue design margins are periodically evaluated in accordance with the Plant Transient and Fatigue Monitoring Program.
(1)
Emergency allowable values used.
FSAR Rev. 65 Page 2 of 5
SSES-FSAR Table Rev. 54 TABLE 3.9-2a (Continued)
REACTOR PRESSURE VESSEL AND SHROUD SUPPORT ASSEMBLY ASME B&PV CODE SEC. III LOADING PRIMARY STRESS ALLOWABLE MAXIMUM SUBSECTION NB PRIMARY STRESS TYPE STRESS CALCULATED LIMIT CRITERIA (psi) STRESS (psi)
(iii) RPV FEEDWATER NOZZLE MATERIAL: SA 508 CL. I.
A. NORMAL & UPSET CONDITION:
Pm Sm 1. Deadweight PRIMARY MEMBRANE 17,700 14,200 Sm = 17,700 @ 525°F 2. Pressure Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 26,550 19,700 1.5 Sm = 26,550 @ 525°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
Pm Sy 1. Deadweight PRIMARY MEMBRANE 26,500 15,600 Sy = 25,900 @ 594°F 2. Pressure Pm + Pb 1.5 Sy 3. SBA PRIMARY MEMBRANE 39,800 20,800 1.5 Sy = 38,850 @ 594°F 4. SRV PLUS BENDING C. FAULTED CONDITION:
(2)
Pm 3.0 Sm 1. Deadweight PRIMARY MEMBRANE 53,100 28,300 3.0 Sm = 53,100 @ 575°F 2. Pressure Pm +Pb 1.5 Sy 3. LOCA PRIMARY MEMBRANE 39,800 20,800 1.5 Sy = 38,850 @ 594°F 4. SSE PLUS BENDING D. MAXIMUM CUMULATIVE FATIGUE USAGE FACTOR: 0.82 At Safe End. This CUF accounts for the original 40-year plant design. During the period of extended operation, actual fatigue design margins are periodically evaluated in accordance with the Plant Transient and Fatigue Monitoring Program.
(2)
Without Thermal Bending.
FSAR Rev. 65 Page 3 of 5
SSES-FSAR Table Rev. 54 TABLE 3.9-2a (Continued)
REACTOR PRESSURE VESSEL AND SHROUD SUPPORT ASSEMBLY ASME B&PV CODE SEC. III LOADING PRIMARY STRESS ALLOWABLE MAXIMUM SUBSECTION NB PRIMARY STRESS TYPE STRESS CALCULATED LIMIT CRITERIA (psi) STRESS (psi)
(iv) IV CRD PENETRATION - CRD HOUSING MATERIAL: TYPE 304SS A. NORMAL & UPSET CONDITION:
Pm Sm 1. Deadweight PRIMARY MEMBRANE 16,600 8,506 Sm = 16,600 @ 575°F 2. Pressure Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 24,900 8,480 1.5 Sm = 24,990 @ 575°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
Pm 1.2 Sm 1. Deadweight PRIMARY MEMBRANE 19,920 12,470 1.2 Sm = 20,000 @ 575°F 2. Pressure Pm + Pb 1.8 Sm 3. SRV PRIMARY MEMBRANE 29,880 12,710 1.8 Sm = 30,000 @ 575°F 4. SBA PLUS BENDING C. FAULTED CONDITION:
Pm 2.4 Sm 1. Deadweight PRIMARY MEMBRANE 39,840 12,470 2.4 Sm = 39,984 @ 575°F 2. Pressure Pm +Pb 3.6 Sm 3. LOCA PRIMARY MEMBRANE 59,760 12,710 3.6 Sm = 59,975 @ 575°F 4. SCRAM PLUS BENDING
- 5. SSE D. MAXIMUM CUMULATIVE FATIGUE USAGE FACTOR: 0.208 at Lower CRD Housing. This CUF accounts for the original 40-year plant design. During the period of extended operation, actual fatigue design margins are periodically evaluated in accordance with the Plant Transient and Fatigue Monitoring Program.
FSAR Rev. 65 Page 4 of 5
SSES-FSAR Table Rev. 54 TABLE 3.9-2a (Continued)
REACTOR PRESSURE VESSEL AND SHROUD SUPPORT ASSEMBLY ASME B&PV CODE SEC. III LOADING PRIMARY STRESS ALLOWABLE MAXIMUM SUBSECTION NB PRIMARY TYPE STRESS CALCULATED STRESS LIMIT CRITERIA -(psi) STRESS (psi)
(v) CRD PENETRATION - STUB TUBE MATERIAL: SB-167 INCONEL A. NORMAL & UPSET CONDITION:
Pm Sm 1. Deadweight PRIMARY MEMBRANE 20,000 5,005 Sm = 20,000 @ 575°F 2. Pressure Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 30,000 28,200 1.5 Sm = 30,000 @ 575°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
Pm Sy 1. Deadweight PRIMARY MEMBRANE 24,100 6,755 Sy = 24,100 @ 575°F 2. Pressure Pm + Pb 1.5 Sy 3. SBA PRIMARY MEMBRANE 36,150 30,260 1.5 Sy = 36,150 @ 575°F 4. SRV PLUS BENDING C. FAULTED CONDITION:
Pm 2.4 Sm 1. Deadweight PRIMARY MEMBRANE 48,000 6,755 2.4 Sm = 48,000 @ 575°F 2. Pressure Pm +Pb 3.6 Sm 3. LOCA PRIMARY MEMBRANE 72,000 30,260 3.6 Sm = 72,000 @ 575°F 4. SSE PLUS BENDING D. MAXIMUM CUMULATIVE FATIGUE USAGE FACTOR: 0.36 At Bottom of Weld. This CUF accounts for the original 40-year plant design. During the period of extended operation, actual fatigue design margins are periodically evaluated in accordance with the Plant Transient and Fatigue Monitoring Program.
FSAR Rev. 65 Page 5 of 5
SSES-FSAR TABLE 3.9-2aa CONTROL ROD GUIDE TUBE FLANGE PRIMARY ALLOWABLE CALCULATED CRITERIA LOADING STRESS TYPE STRESS (psi) STRESS {psi)
CONTROL ROD GUIDE TUBE Prima[:i Stress Limit - The allowable primary membrane stress plus bending stress is based on the ASME Boiler and Pressure Vessel Code, Secti0fl Ill for Type 304 S.S.
Sm= 16,000 psi @575°F For normal and upset 1. Dead weight Primary_ membrane and condition: 2. Pressure bending (The maximum drop across bending stress occurs at S1irnit = 1.5 Sm guide tube the guide tube base.)
24,000 11,563
- 5. Scram For emergency condition: 1. Dead weight Primary membrane and
- 2. Pressure bending .
Slimit = 2.25 Sm 3. . OBE
36,000 26,103
- 4. SRV Pm+ Pb 5. Chugging For faulted condition: 1. Dead weight Primary membrane and
- 2. Pressure bending S,i~it = 2.4 Sm 3. Chugging 38,400 27,010
SSES-FSAR TABLE 3.9-2ab INCORE HOUSING ALLOWABLE CALCULATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi)
. isallowable i Primary Stress Limrt-The primary membrane stress based on ASME Boller and I: for vessels, for Type 304 Pressure Vessel Code, Section Ill Class I
! stainless steel J .
- Pm== 16,660 psi@ 575"F For normal and upset conditions: 1. Design pressure Maximum membrane stress 6,000 111
. 2. OBE . occurs at the outer surface of i 16,660 15,548(2)
- 3. SRV I the vessel penetration I
, For faulted condition: 1. Design pressure Maximum membrane stress
- 2. SSE occurs at the outer surface of 6,4SO(II 39,840 27,830(2)
- Slim~=2.4 P,.. 3. LOCA the vessel penetration
- 4. Annulus pressurization NOTE: The emergency condition loads are less than the nonnal/upset loads.
(1) Unit No. 1 (2) Unit No. 2 Rev. 53, 04/99 Page 1 of 1
TABLE 3.9-2 (ac)
REACTOR VESSEL SUPPORT EQUIPMENT CRD HOUSING SUPPORT THIS TABLE HAS BEEN INTENTIONALLY LEFT BLANK Rev. 35. 07/84
SSES-FSAR I TABLE 3.9-2 (ae)
\
HPCI TURBINE DESIGN CALCULATIONS TURBINE PART CALCULATED ALLOWABLE .I
- Pressure Boundary Castings Stop Valve 8,~75 psi 14,000 psi Turbine Inlet {high pres.)
- 6,550 psi 14,000 psi Turbine Wheel Case (low pres.) 6,000 psi 14,000 psi Pressure Boundary Bolting I Stop Valve 17,700 psi 20,000 psi 20,000 psi Turbine 18,290 psi Non-Pressure Boundary Components Turbine Shaft . 5,000 psi 60,500 psi ;I Thrust Bearing 4,400 lbf 5,600 lbf
_ Journal Bearing 2,680 lbf 19,500 lbf i Stop Valve Yoke 13,500 psi 33,000 psi Pedestal Dowel Pins 29,800 psi 61,100 psi Pedestal Bolts 11 .400 lbf 28.300 lbf Rev. 53, 04/99 Page 1 of 1
SSES - FSAR TABLE 3.9 - 2 (af), page l of 2 HIGI NNSITY SPENI' FUEL RACKS
'r.iPES OF ANALYSIS PERFOR-!ED m'NAMIC ANALYSIS:
A dynamic JOOdal analysis using the seismic, SRV, aoo IJXA resp:,nse spectra was perforrre:I on a simplified mxlel consisting of 6 racks {l quadrant). 'lhe resulting loads on the corner roodule were extracted and a nore detailed analysis performed.
- Sl'ATIC ANALYSIS:
A detailed finite element (1364 elenents) nodel of the oorner nodule was developed arxl a static analysis perfornm using the 10:iding results of the dynamic analysis. 'lhe section descriptions, allo.rable stresses and stress ratios for the detailed nooel are given on page 2 of this table.
FUEL RA'ITLIM3 NmLYSIS:
A time history analysis was performed to detennine local inpact loads due to fuel rattling. A cx:m!8rison of the sll!l)Ort loads fran the fuel rattlin:J analysis with those of the response spectrtlll analysis sho,.red that the fuel rattling results are less than or equal to the response spectrwn results.
Analysis of the poison can was coopleted using the local inpact loads.
K>DEL IMPACT ANALYSIS:
An equivalent static load was determined for the following drop corrlitions:
. 1) 18" fuel drcp on oorner of top casting
- 2) 18" fuel drop on middle of top casting
- 3) fuel drq, full length through the cavity impacting oottom casting at the middle.
For the first 2 cases the equivalent static loads calculated were canbined with dead load and applied to the detailed m:>del. For the 3rd case, the.
ultinate load of the burrlle shearing out of the fuel seat was determinoo aoo coobined with dead load. '!his oonbined load was then awlied to the detailed nodel.
Rev. 35, 07/84
SSE5 - FSAR TABLE 3.9-2 (af), page 2 of 2 HIGH DENSITY SPENr FUEL RAO<
SlM1ARY CF WSUUIS FOR 1ffE DETA.IIED M:>DEL ~
OORMAL CESIGJ ACCIDENT AND
\ .ALI.CWABLE NJRMAL OPEFAT~ COIDITIOO EXTRfflE ENVIH'.NttmrAL I
STRESSFS COIDITICNS MAX MAX StL"r. fa fbx STR&SS fa fbx STRFSS NO. SECTIOO DFSCRIPTICN Fa Ft,y Fm Fa ~ FtJx: RM'IO (l)
Fa ~ Ft,y RATIO(!
1 Top Grid Outer Section 9941 15760 15760 .026 .009 .747 .78 .018 .006 )
- 715 .74 2 Tep Grid Inner Section 9420 15760 15760 .057 .055 .813 .93 .040 .039 .766 .85 3 Bottc,n Grid Outer Sect. 8830 15760 12120 .062 .248 .108 .42 .062 .248 , .108 .42 4 Botton Grid Inner Sect. 8550 15760 12120 .* oos .831 .013 .85 .005 .831 .013 .85 5 Bott.an Grid OJter 9650 15760 12120 .047 .249 .269 .57 .047 .249 .269 .57 I
Section Near Leg 6 Ihtt.an Grid Inner 9530 15760 l12120 .046 .508 .248 .80 .046 .508 .248 .80 Section Near Leg I
7 Bottan Grid Foot 10250 15760 12120 .132 0 .001 .13 .160 0 .003 .16 j.006 I l - I 8 Bottan Grid Foot 11020 14180114180 ( .161 0 .003 .16 .195 0 .20 1;2* Plate 3320 I - - - - - .,6(2) l -l l-9 Fv 99(2)
= 1390 10 7/8* Plate 17370 fa F., = 10970.
fbx
- - ~92(2) -
I - .92< 2 >
(1) Stress Ratio = Fa + ~ + F6i rUIB (2) Plate Stress Patio = ~ fx AllCMable stresses are a + Fa factored up per Table 9. l-7a oft.he SSES-FSAR.
Rev. 359 07 /84
SSES - FSAR Table Rev. 55 TABLE 3.9-2b REACTOR INTERNALS AND ASSOCIATED EQUIPMENT ASME B&PV CODE SEC. III LOADING PRIMARY STRESS TYPE ALLOWABLE MAXIMUM SUBSECTION NG PRIMARY STRESS CALCULATED STRESS LIMIT CRITERIA (psi) STRESS (psi)
(i) TOP GUIDE - BEAM WITH HIGHEST STRESS MATERIAL: SA-240 TYPE 304SS A. NORMAL & UPSET CONDITION:
Pm Sm 1. Normal PRIMARY MEMBRANE 16,950 2,770 Sm = 16,900 @ 550°F 2. Pressure Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 25,430 11,500 1.5 Sm = 25,350 @ 550°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
Pm 1.5 Sm 1. Normal PRIMARY MEMBRANE 25,430 730 1.5 Sm = 25,350 @ 550°F 2. Pressure Pm + Pb 2.25 Sm 3. Chugging PRIMARY MEMBRANE 38,140 9,600 PLUS BENDING 2.25 Sm = 38,025 @ 550°F 4. SRV C. FAULTED CONDITION:
Pm 2.4 Sm 1. Normal PRIMARY MEMBRANE 40,680 3,500 2.4 Sm = 40,560 @ 550°F 2. Pressure Pm + Pb 3 Sm 3. SRV PRIMARY MEMBRANE 61,020 11,800 3 Sm = 50,7000 @ 550°F 4. SSES PLUS BENDING
- 5. LOCA D. MAXIMUM CUMULATIVE USAGE FACTOR: 0.22 At Beam-Slot Location This CUF was evaluated for the period of extended operation and found to be acceptable.
FSAR Rev. 65 Page 1 of 3
SSES - FSAR Table Rev. 55 TABLE 3.9-2b (Continued)
REACTOR INTERNALS AND ASSOCIATED EQUIPMENT ASME B&PV CODE SEC. III LOADING PRIMARY STRESS TYPE ALLOWABLE MAXIMUM SUBSECTION NG PRIMARY STRESS CALCULATED STRESS LIMIT CRITERIA (psi) STRESS (psi)
(ii) CORE PLATE (LIGAMENT IN TOP PLATE)
MATERIAL: TYPE 304SS A. NORMAL & UPSET CONDITION:
Pm Sm 1. Normal PRIMARY MEMBRANE 16,900 783 Sm = 16,900 @ 550°F 2. Pressure Pm + Pb 1.5 Sm 3. OBE PRIMARY MEMBRANE 25,470 10,140 1.5 Sm = 25,350 @ 550°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
(1)
Pm 1.5 Sm 1. Normal PRIMARY MEMBRANE 25,350 647 1.5 Sm = 25,350 @ 550°F 2. Pressure Pm + Pb 2.25 Sm 3. Chugging PRIMARY MEMBRANE 38,200 12,460 2.25 Sm = 38,025 @ 550°F 4. SRV PLUS BENDING C. FAULTED CONDITION:
(1)
Pm 2.4 Sm 1. Normal PRIMARY MEMBRANE 40,560 9,570 2.4 Sm = 40,560 @ 550°F 2. Pressure PL + Pb 3 Sm 3. SRV PRIMARY MEMBRANE 61,020 15,710 3 Sm = 50,700 @ 550°F 4. SSE PLUS BENDING
- 5. LOCA D. MAXIMUM CUMULATIVE USAGE FACTOR: 0.005 At Beam-Rim Junction. This CUF was evaluated for the period of extended operation and found to be acceptable.
(1)
Value is not given in New Loads Report. However, Primary and Bending Stresses are, and these are bounding.
FSAR Rev. 65 Page 2 of 3
SSES - FSAR Table Rev. 55 TABLE 3.9-2b (Continued)
REACTOR INTERNALS AND ASSOCIATED EQUIPMENT ASME B&PV CODE SEC. III LOADING PRIMARY STRESS ALLOWABLE MAXIMUM SUBSECTION NG PRIMARY STRESS TYPE STRESS CALCULATED LIMIT CRITERIA (psi) STRESS (psi)
(ii) VESSEL HEAD SPRAY NOZZLE MATERIAL: CARBON STEEL SA508,CL1 A. NORMAL & UPSET CONDITION:
Pm
- 1. Normal PRIMARY MEMBRANE Sm = 16,670 @ 575°f 2. Pressure PL + Pb 3.5 Sm 3. OBE PRIMARY MEMBRANE 50,000 28,700 1.5 Sm = 50,000 @ 575°F 4. SRV PLUS BENDING B. EMERGENCY CONDITION:
PL + P b 1.8 Sm 1. Normal 1.8 Sm - 30,000 @ 575°F 2. Pressure
- 3. Chugging PRIMARY MEMBRANE 30,000 28,700
- 4. SRV PLUS BENDING C. FAULTED CONDITION:
PL + Pb 3 Sm 1. Normal 3 Sm = 50,000 @ 575°F 2. Pressure PL + Pb 3 Sm 3. SRV PRIMARY MEMBRANE 50,000 28,700 3 Sm = 50,7000 @ 550°F 4. SSE PLUS BENDING
- 5. LOCA D. MAXIMUM CUMULATIVE USAGE FACTOR: Satisfied per N-415-1
- Not required per NB-3222.
FSAR Rev. 65 Page 3 of 3
SSES-FSAR TABLE 3.9-2(c)
REACTOR WATER CLEANUP HEAT EXCHANGERS Allowable Required Thickness Actual Thickness Part Stresses (in.) (in.)
. (psi) .
.i .
REGENERATIVE cu HX I Shell .858 15,000 1
Shell head .704 171500 I Channel shell .917 15,900
- Tubesheet 2.87 15,900 Tubes .084 14,025 .095 I /
t-------~---+----_..;,
1 Piping .240 15,000 .337 I
. Channel c*over 3.53 17,500 3.75*
lI
- ~
Shell .168 15,000 .375 Shell head .144 17,500 .375 ;
/ Channel shell .917 15,900 1.0* I I Channel cover 3.53 17,500 3.75*
Tubesheet 2.87 15,900 2.875* I 1
! Tubes .056 11,900 *, :
0.65 I I Channel pipi~g_ .240 15,000 .337 I I Shell piping .073 15,000 .322
- Values within 10%
Rev. 53, 04/99 Page 1 of 1
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SSES-FSAR Table Rev, 54 Table 3.9-2d.1 ASME CODE CLASS 1 MAIN STEAM PIPING AND PIPE MOUNTED EQUIPMENT -
HIGHEST STRESS
SUMMARY
- UNIT 2 IDENTIFICATION OF LOCATIONS OF ACCEPTANCE LIMITING CALCULATED ALLOWABLE RATIO LOADING HIGHEST STRESS CRITERIA STRESS TYPE STRESS OR LIMITS ACTUAL/ POINTS - NODG USAGE FACTOR ALLOWABLE POINT NUMBERS ASME B&PV Code Section III, NB-3600 Design Condition: 1. Pressure Eq. 9 1.5 Sm Primary 26,480 psi 26,550 psi 0.997 2. Weight M.S. Line A
- 3. OBE Sweepolet 990
- 4. SRV Service Levels A&B (Normal & Upset) Secondary 32,295 psi 54,252 psi 0.595 1. Thermal M.S. Line C Condition: Sweepolet 641 Eq. 12 3.0 Sm Service Levels A&B Primary Plus 1. Pressure (Normal & Upset) Secondary 54,046 psi 54,252 psi 0.996 2. Weight M.S. Line A Condition: (Except 3. OBE Sweepolet 790 Eq.13 3.0 Sm Thermal 4. SRV Expansion)
Service Levels A&B (Normal & Upset) N/A 0.683* 1.0 0.683 M.S. Line B Conditions: TTJ @
Cumulative Usage Sweepolet 932 Factor Service Level B 1. Pressure (Upset) Primary 30,010 psi 31,860 psi 0.942 2. Weight M.S. Line D Condition: 3. OBE Sweepolet 90 Lower{Eq.9 1.8 Sm 4. SRV Of {EQ 9 1.5Sy Service Level C 1. Pressure (Emergency) Primary 30,110 psi 39,820 psi 0.756 2. Weight M.S. Line D Condition: 3. SBA Sweepolet 75 Lower{Eq.9<2.25 Sm 4. OBE Of {Eq 9 < 1.8 Sy 5. SRV Service Level D 1. Pressure (Faulted) Primary 58,915psi 60,000 psi 0.982 2. Weight M.S. Line A Condition: 3. SSE Sweepolet 790 Eq. 9 <3.0 Sm 4. LOCA
- 5. IBA
SSES-FSAR Table Rev, 54 Table 3.9-2d.1 (Continued)
ASME CODE CLASS 1 MAIN STEAM PIPING AND PIPE MOUNTED EQUIPMENT - UNIT 2 IDENTIFICATION OF COMPONENT/LO HIGHEST ALLOWABLE RATIO LOADING EQUIPMENT WITH AD TYPE CALCULATED LOAD CALCULATED/AL HIGHEST LOADS LOAD LOWABLE
- 1. Pressure Snubber/ 44,848 50,000 0.897 2. Weight M.S. Line A Service Level 3. OBE 120; H42 B Loads 4. Operating transient
- 1. Pressure Snubber/ 42,043 70,350 0.598 2. Weight M.S. Line A Service Level 3. SBA 120; H42 C Loads 4. Operating transient
- 1. Pressure Snubber/ 74,120 91,000 0.815 2. Weight M.S. Line A Service Level 3. IBA 120; H42 D Loads 4. SSE
- 5. Operating transient
- 4. SSE
- 4. SRV
- 4. SSE
- 1. Pressure MSIV Bonnet/ 1,369,899 1,589,000 0.862 2. Weight Line B Bending Moment in-lbs in-lb 3. LOCA MSIV Bonnet
- 4. SSE
- This CUF accounts for the original 40-year plant design. During the period of extended operation, actual fatigue design margins are periodically evaluated in accordance with the Plant Transient and Fatigue Monitoring Program.
FSAR Rev. 65 Page 2 of 2
SSES-FSAR Table Rev. 54 Table 3.9-2e ASME CODE CLASS 1 RECIRCULATION PIPING AND PIPE MOUNTED EQUIPMENT -
HIGHEST STRESS
SUMMARY
- UNIT 1 IDENTIFICATION LIMITING CALCULATED ALLOWABLE RATIO OF LOCATIONS OF ACCEPTANCE CRITERIA STRESS TYPE STRESS OR LIMITS ACTUAL/ LOADING HIGHEST STRESS USAGE FACTOR ALLOWABLE POINTS - NODG POINT NUMBERS ASME B&PV Code Section III, NB-3600 Design Condition:
- 1. Pressure Eq. 9 1.5 Sm Primary 24,097 psi 25,013 psi 0.963 2. Weigh Loop 'B'
- 3. OBE Elbow-784 Service Levels A & B (Normal & Upset) Conditions:
Eq. 12 3.0 Sm Secondary 36,019 psi 50,025 psi 0.720 1. Thermal Loop 'B' Weldolet 508 Service Levels A & B (Normal & Upset) Condition:
Eq. 13 3.0 Sm Primary Plus 49,544 psi 50,025 psi 0.990 1. Pressure Loop 'B' Secondary 2. Weight TTJ@Valve (Except 3. OBE End 782 Thermal 4. Operating Expansion) Transient
- 5. SRV Service Levels A & B (Normal and Upset)
Conditions:
Cumulative Usage Factor N.A. 0.5555** 1.0 0.555 1. Pressure Loop 'B'
- 2. Thermal Half Coupling 727
- 5. SRV Service Level B (Upset)
Condition:
Eq. 9 1.8 Sm & 1.5 Sy Primary 25,298 psi 27,750 psi 0.912 1. Pressure Loop 'B'
- 2. Weight Elbow 784
- 3. OBE
- 4. Operating Transient
SSES-FSAR Table Rev. 54 Table 3.9-2e ASME CODE CLASS 1 RECIRCULATION PIPING AND PIPE MOUNTED EQUIPMENT -
HIGHEST STRESS
SUMMARY
- UNIT 1 IDENTIFICATION LIMITING CALCULATED ALLOWABLE RATIO OF LOCATIONS OF ACCEPTANCE CRITERIA STRESS TYPE STRESS OR LIMITS ACTUAL/ LOADING HIGHEST STRESS USAGE FACTOR ALLOWABLE POINTS - NODG POINT NUMBERS Service Level C (Emergency)
Condition:
Eq. 9 < 2.25 Sm & 1.8 Sy Primary 28,801 psi 33,300 psi 0.865 1. Pressure Loop 'B'
- 2. Weight Weldolet 508
- 3. OBE
- 4. Operating Transient
- 5. SRV
- 6. SBA Service Level D (Faulted)
Condition:
Eq. 9 < 3.0 Sm Primary 41,019 psi 50,025 psi 0.820 1. Pressure Loop 'B'
- 2. Weight Elbow 784 3 .LOCA
- 4. Operating Transient
- 5. SSE
- 6. IBA
SSES-FSAR Table Rev. 54 Table 3.9-2e (Continued)
ASME CODE CLASS 1 RECIRCULATION AND PIPE MOUNTED EQUIPMENT - UNIT 1 HIGHEST RATIO IDENTIFICATION OF COMPONENT/LOAD TYPE CALCULATED ALLOWABLE CALCULATED/ LOADING EQUIPMENT WITH LOAD LOAD ALLOWABLE HIGHEST LOADS Snubber/ 56,085 120,000 0.467 1. Pressure Loop 'A' Service Level 2. Weight (SAI) H46
'B' Loads 3. OBE
- 4. Operating transient STRUT 68,094 87,840 0775 1. Pressure Loop 'A' Service Level 2. Weight (PO1) H50
'C' Loads 3. SBA
- 4. Operating transient STRUT 94,493 121,896 0.775 1. Pressure Loop 'A' Service Level 2. Weight (PO1) H50
'D' Loads 3. LOCA
- 4. Operating transient Discharge Gate Valve/ 7.296 9.8 0.744 1. Pressure Loop 'B' Horizontal Acceleration 2. Weight Operator
- 3. LOCA
- 4. Operating transient Discharge Gate Vale/ 2.2 2.2 1.0 1. Pressure Loop 'A' Vertical Acceleration 2. Weight Operator
- 3. LOCA
- 4. Operating transient By-Pass Valve/ 4.336 4.336 1.0 1. Pressure Loop 'A' Horizontal (East/West) 2. Weight Operator Acceleration 3. LOCA
- 4. Operating transient By-Pass Valve/ 2.605 2.605 1.0 1. Pressure Loop 'A' Horizontal (North South) 2. Weight Operator Acceleration 3. LOCA
- 4. Operating transient By-Pass Valve/ 2.514 2.62 0.960 1. Pressure Loop 'A' Vertical Acceleration 2. Weight Operator
- 3. LOCA
- 4. Operating transient Suction Gate Valve/ 5.24 9.87 0.531 1. Pressure Loop 'A' Horizontal & Vertical 2. Weight Operator Acceleration* 3. LOCA
- 4. Operating transient Recirc. Pump/ 2.338 2.7 0.866 1. Pressure Loop 'A' Horizontal Acceleration 2. Weight Pump Motor C.G.
- 3. LOCA
- 4. SSE Recirc. Pump/ 1.004 2.1 0.478 1. Pressure Loop 'B' Vertical Acceleration 2. Weight Pump Motor C.G.
- 3. LOCA
- 4. SSE
- 5. SRN
- Allowable acceleration applies to operator only; valve body is qualified based upon allowable moment and allowable axial force.
FSAR Rev. 65 Page 3 of 3
SSES-FSAR Table Rev. 51 Table 3.9-2e.1 ASME CODE CLASS 1 RECIRCULATION PIPING AND PIPE MOUNTED EQUIPMENT -
HIGHEST STRESS
SUMMARY
- UNIT 2 IDENTIFICATION OF LOCATIONS OF ACCEPTANCE LIMITING CALCULATED ALLOWABLE RATIO LOADING HIGHEST STRESS CRITERIA STRESS TYPE STRESS OR LIMITS ACTUAL/ POINTS - NODG USAGE FACTOR ALLOWABLE POINT NUMBERS ASME B&PV Code Section III, NB-3600 Design Condition: 1. Pressure Eq.9 1.5 Sm Primary 24,712 psi 25,013 psi 0.988 2. Weight Loop B
- 3. OBE Elbow 765
- 4. SRV Service Levels A&B Secondary 35,891 psi 50,025 psi 0.717 1. Thermal Loop B (Normal & Upset) Weldolet 305 Condition:
Eq.12 3.0 Sm Service Levels Primary Plus 1. Pressure A&B Secondary 49,585 psi 50,025 psi 0.991 2. Weight Loop B (Normal & Upset) (Except 3. OBE Half Coupling 300 Condition: Thermal 4. SRV Eq.13 3.0 Sm Expansion)
Service Levels A&B N/A 0.7938** 1.0 0.7938 Loop B (Normal & Upset) Half Coupling 380 Conditions:
Cumulative Usage Factor Service Level B 1. Pressure (Upset) Primary 26,160 psi 27,750 psi 0.943 2. Weight Loop A Condition: 3. OBE Elbow 575 Lower{Eq. 9 1.8 4. SRV Sm Of {EQ 9 1.5 Sy Service Level C 1. Pressure (Emergency) Primary 26,650 psi 33,300 psi 0.800 2. Weight Loop B Condition: 3. SRV Weldolet 275 Lower{Eq. 9 < 2.25 4. SBA Sm 5. OBE Of {Eq. 9 < 1.8 Sy Service Level D 1. Pressure (Faulted) Primary 41,120 psi 50,025 psi 0.822 2. Weight Loop A Condition: 3. LOCA 4 inch
SSES-FSAR Table Rev. 51 Table 3.9-2e.1 ASME CODE CLASS 1 RECIRCULATION AND PIPE MOUNTED EQUIPMENT - UNIT 2 IDENTIFICATION COMPONENT/ HIGHEST ALLOWABLE RATIO LOADING OF EQUIPMENT LOAD TYPE CALCULATED LOAD CALCULATED/ WITH HIGHEST LOAD ALLOWABLE LOADS
- 1. Pressure Strut/ 57,612 62,800 0.92 2. Weight Loop A Service Level 3. OBE PO1; H50 B Loads 4. Operating transient
- 1. Pressure Strut/ 79,132 86,250 0.92 2. Weight Loop A Service Level 3. SBA PO1; H50 C Loads 4. Operating transient
- 1. Pressure Strut/ 96,643 105,340 0.92 2. Weight Loop A Service Level 3. LOCA PO1; H50 D Loads 4. Operating transient
- 1. Pressure Discharge Gate 7.678 9.8 0.783 2. Weight Loop A Valve/Horizontal 3. LOCA Operator Acceleration 4. Operating transient
- 1. Pressure Discharge Gate 2.118 2.188 0.968 2. Weight Loop B Valve/ 3. LOCA Operator Vertical 4. Operating Acceleration transient
- 1. Pressure By-Pass Valve/ 5.438 5.468 0.995 2. Weight Loop B Horizontal 3. LOCA Operator Acceleration 4. Operating (East-West) transient
- 1. Pressure By-Pass Valve/ 2.539 2.539 1.000 2. Weight Loop A Horizontal 3. LOCA (North-South) 4. Operating transient
- 5. Pressure Vertical 2.376 2.376 1.000 6. Weight Loop A Acceleration 7. LOCA Body
- 8. Operating transient Suction Valve/ 1. Pressure Horizontal & 7.714 9.87 0.782 2. Weight Loop A Vertical 3. LOCA Operator Acceleration* 4. Operating transient Recirc. 1. Pressure Pump/Horizontal 2.386 2.7 0.884 2. Weight Loop A Acceleration 3. LOCA Pump Motor C.G.
- 4. SSE Vertical 1. Pressure Acceleration 1.388 2.1 0.661 2. Weight Loop B
- 3. LOCA Pump Motor C.G.
- 4. SSE
- 5. SRN
- Allowable acceleration applies to operator only; valve body is qualified based upon allowable moment and allowable axial force.
FSAR Rev. 65 Page 2 of 2
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS Unless otherwise specified, all references are to ASME B and PVC Section III (July 1971).
- 1. Body Inlet and Outlet Based on ANSI B31.7-1969, PARA. 1-704.5.1 & ASME BPVC Criteria Flange Stresses fM o + Pb B < 1.5 S m SH =
2 Lg B 4g1 (4/3te + 1)Mo < 1.5 Sm SR =
2 IT J YM o Z S R < 1.5 S m ST =
2 t J Where SH = Longitudinal Hub Wall Stress, PSI SR = Residual Flange (Body Base, Inlet) Stress, PSI ST = Tangential Flange Stress, PSI For Inlet As SH < 27,300 PSI 27,300 PSI See Note 1 (1.5 Sm @ 575°F for A-105GII)
SR<27,300 PSI 27,300 PSI See Note 1 (1.5 Sm)
ST< 27,300 PSI 27,300 PSI See Note 1 (1.5 Sm) Criteria Satisfied For Outlet FSAR Rev. 65 Page 1 of 7
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS As SH < 27,300 PSI 29,100 PSI See Note 1 (1.5 Sm)
SR <27,300 PSI 29,100 PSI See Note 1 (1.5 Sm)
ST < 27,300 PSI 29,100 PSI See Note 1 (1.5 Sm)
- 2. Inlet and Outlet Stud Based on USAS B31.7-1969, PARA. 1-704.5.1 and ASME BPVC Total Cross-Area Requirements Sectional Area of Bolt Shall Equal or Exceed the Greater of Wm 1 WM 2 AM1 = I Sb or AM 2 =I Sa Where:
AM1 = The total required bolt (stud) area for operating condition, in2.
AM2 = Total required bolt (stud) area for gasket seating, in2.
For Inlet AM1 = 9,060 in2 > AM2 = 0.807 in2 9.060 in2 See Note 1 As AM1 = 9,060 in2 (Am, actual stud area Criterion Satisfied)
FSAR Rev. 65 Page 2 of 7
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS For Outlet Am1 = 4.76 in2 > Am2 = 1.55 in2 As Am1 = 4.76 in2 4.76 in2 See Note 1 (Am, actual stud area Criterion Satisfied)
- 3. Nozzle Wall Based on PARA. NB-3530, NB-3540 and NB-3550 for non-standard primary Thickness pressure rating.
- 1. Valve Wall Thickness Criterion tmin. < tA Where tmin. = Minimum Calculated Required Thickness With Corrosion Allowable, in.
tA = Actual Nozzle Wall Thickness, in.
i) For thinnest section near valve seat As tmin. = 0.467 in. 0.467 in. See Note 1.
(ta, actual) ii) For section about mid-section of nozzle As tmin. = 0.468 in.. 0.468 in. See Note 1.
(ta, actual)
FSAR Rev. 65 Page 3 of 7
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS
- 2. Cyclic Rating (NB-3550) i) The thermal cyclic index criterion Nri It = I Ni As It = 0.032 < 1 Criterion Satisifed ii) The fatigue requirement Criterion Na 2000 cycles 2 Peb S P1 = I QP + I + Q t 2 + 1.3Q t1 3 2 K
S P 2 = 0 .4 Q P +I (Peb + Q t 2 )
2 Where:
SP1 = The fatigue stress intensity at the inside surface of the crotch, PSI SP2 = The fatigue stress intensity at the outside surface of the crotch, PSI SP1 = 9,130 PSI > SP2 = 545 PSI The permissible number of startup/shutdown cycles.
NA > 106 > 2,000 cycles Criterion Satisfied FSAR Rev. 65 Page 4 of 7
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS
- 1. Body to Bonnet Stud Stress Criterion Total Cross-Sectional Area of Bolt Shall Equal or Exceed the Greater of Wm 1 Wm 2 Am1 = I Sb or Am 2 = I Sa Am1 = 5.19 in.2 > AM2 = 0.849 in.2 As AM1 = 5.19 in.2 5.19 in.2 See Note 1 (Am, actual stud area criterion satisfied)
- 2. Bonnet Flange Strength (B31.7-1969, PARA. 1-704.5.1) Criterion SH < 1.5 Sm SR < 1.5 Sm ST < 1.5 Sm As SH < 29,100 psi 29,100 PSI See Note 1 (15 Sm @ 500°F A105 GI SR < 29,100 psi 29,100 PSI See Note 1 ST < 29,100 psi 29,100 PSI See Note 1 (1.5 Sm)
FSAR Rev. 65 Page 5 of 7
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS
- 5. DISC Insert Bending Stress of Disc Insert per Roarks formulas for stress and strain, 4th Edition pg. 222-223 superposition of case no. 21 and case no. 22 for flat plates.
Alternatively, finite element analysis may be used to calculate stress.
3W 4 1 4a (m + 1)1n(a / b ) a 4 (m + 3 ) + b 4 (m 1) + 4 S =S +S =
t r r 2 2 2 1 2 4 t avg a (m + 1) + b (m 1) 2 2 2 3W 2 2a (m + 1)1n( a / b ) + a (m 1) b (m 1)
+I 2 2 2 2 t a (m + 1) + b (m 1)
Where: W1 = Pressure, psig W2 = Seat Load, lbs.
a = 1/2 Disc. Insert Outside Dia., 2.782 in.
b = Hub Radius 0.592 in.
t = Average Thickness, 1.022 in.
Sr1 = Stress Due to Pressure, psi Sr2 = Stress Due to Seat Load, psi m = Reciprocal of poisson ratio, 3.333
- 1. The normal operating condition of 80% of design pressure.
St < 45,225 psi (Sm @ 575°F for ASME SA-637 TYPE 718) 45,225 PSI See Note 1 Criterion Satisfied Where W1 = 1,000 psi W2 = 5,422 lbs.
FSAR Rev. 65 Page 6 of 7
SSES-FSAR Table Rev. 38 TABLE 3.9-2g SAFETY/RELIEF VALVES (MAIN STEAM)
ALLOWABLE STRESS CALCULATED STRESS CRITERIA METHOD OF ANALYSIS OR MINIMUM OR THICKNESS REQUIRED ACTUAL THICKNESS
- 2. The maximum possible full flow pressure at 10% above design pressure (the stress is entirely due to pressure).
St < 45,225 psi (Sm) Criterion Satisfied 45,225 PSI See Note 1 Where W1 = 1,375 psi W2 = 0
- 3. The Set Spring Load, Zero Inlet Pressure, Load Temperature (The Stress is Entirely Due to Seat Load, Zero Inlet Pressure).
St < 50,000 psi (Sm @ Room Temperature of ASME SA-637 TYPE 718) 50,000 PSI See Note 1 Where W1 = 0 W2 = 27,110 psi Note 1: The calculated stress was determined by calculation to be less than the allowable stress. The actual thickness is greater than the minimum thickness required as determined by calculation. The actual thickness satisfies minimum thickness requirements for the period of extended operation.
FSAR Rev. 65 Page 7 of 7
TABLE 3.9-2fil (pa~e 1 of 9)
MAI~ STEA..~ lS0LAT!ON VALVE Allowable Stress calculated Stress or Minimum or Actual Criteria Method of Analisis Thickness .i!!!.:J. Thickness ~
Design of Pressure All references are made t~ ASME Boiler & Pressure Vessel Code,Section III.
Retaining parts ~uclear Power Plant Components, 1971 Ed. plus addendum July 1971. Reference the same code for explanation of the sy,ubols used.
t 8ody Minimum Reference paragraph NB 3543, ~onstandard Pressure-Rated Valve, Table NB 3542-1.
Wall !hic.knege for design condition of 1250 psig and 575°F. The primary service rating* 495 1.58 2.12 based on a core diameter of 23.00 in. tm
- 1.539 in. (including a corrosion allowance of 0.12 in).
Bod~e Rule Reference paragraph ~B 3544, Body Shape Rules Radius of Crotch Reference paragraph ~B 3544.l(a), Radius of Crotch criterion r 2 ~ 0,3 tm as r - 1.00 in., tm
- 1,539 + 1.00_~ 0.3 x 1.937
- 0.581 criterion satisfied, 2
corner Radii on Reference paragraph NB 3544.l(b), Corner Radii on Internal Surfaces Internal Surfaces
~riterion r < r as r
- 1.00 in., r
- 1.125 in.+ 1.00 e 1.125 criterion 4 2 4 2 satisfied.
Out of Roundness Reference paragraph 3544.5. Out of Roundness, Figure NB 3545.1-2 The body is not out of round in excess of 51., the requirements of this article are satisfied.
Longitudinal Curvature Reference paragraph ~B 3544,6 Longitudinal Curvature criterion l + l 4 as TL
-r-- -r-- -> ~ ong.
- 27 1 n **
Long. Lat. m rLat,
- 11.5 in ** dn
- 23.00 in ** + .124 ~ .058 criterion satisfied.
Flat Wall Referenc~ paragraph ~8 3344.7, Flat ~all Limitation t.imitat 1.c..,
Since ~o !lat aecti~n~ ~e~e ~uilt into the vdlv~ ~od~ ~esl~n. ch~ re~uirementa .;
~= ~~~g d~:ic!e are ~~tisf~~d.
Re,... .3 ':> , G 7 / 6 ~
T.o\BLE ).9-2D!l (~age 2 of 9)
MAIN ST.EA.'1 ISOLATION VALVE Allowable Stress C&lculated Stress or Minimum 01" Actual Criteria Method of Analrsis Thickness 1.!!!_J thickness ~
Minimum i.. all at 1
Reference para~raph )544.8. Minimum Wall at Weld End
\.leld' End Actual thickness at> 1.937 in. (i.e., 1.937 in,) measured along. The run direction is 1,492 in. Actual run distance is~ 9.5 in.-
Primary Crotch Reference paragraph NJ 3545.l Stress Due to criterion p a D. + O.S) P < S Internal Pressure = A s m m
2 where A* 503 in. 2 , A
- 57 in. , P
- 1375 psig, P
- 12,821 ffl S Ill psi, Sm D 19,400 psi, since Sm> Pm criterion satisfied Valve Body Secondary Reference paragraph ~B 3545.2 Stress Primary Plus Secondary Reference paragraphs SB 3545.2(a)(l), NB 3545.2(a)(2)
Stress Due to Internal Pressure Q
- C p p
_Qt
+ 0.5)P s
Ca e
where CP - 3, r 1
- 11.625 ln., P 8
- 1375 psi, te
- 2.75 fo~ wye-type valve Ca* 1,33 + QP
- 25.965 psi Secondary Stress Due Reference paragraph ~B J54S.2(b), Figures ~"B 3S4S.2-3. NB 33~3.2-3, NB J345.2-6 to Pipe Reaction F S 2 2 Direct o-r Axial Ped*~ where S - 30,750. Fd
- 30 in. , Gd* 183 in.
Load Effect G .
Ped* s&1 psi F S J 5,041 Bending 1.(iad Effect Peb
- Cb -2_ vhere S
- 30,750, Fb
- J40 in. , 29,100 Gb
~.d.
- l). 25 1":'I. * :- - 11.625 tt! * ~- 73 T" - ~ 3.91 fn.
1
-"1.!, -~
- l\, l"_.17 *}. ~ ~ ... ,.
-;)
r C * ;,,fit:i-e i .. !5.)2~ 1:1." r
- 1.l.~~; b .*
t' 1 r +t
.I. "-!
3 te
- 2.75 1n.
- Gb
- 1048.2 in.
... p *
- 9.974 pd 29.100 9,'H4 eb Rev. 35, 07 /84
TABLE 3. 9-2(1"!) (page 1 of 9)
M.t\!N STEA:-! lSOLATION VALVE Allovable Stregs Cal~ulated Stress or Minimum or Actual Criteria ~ethod of Analvsis
- Thickness {in.) Thicluless (_i_n_.J Torsion Lvad Effect Reference paragraphs 3545.~(b)(l), 3545.2(b)(6)(c)
F S )
Pet* 2 _*b~ where Fb
- 340 in .
- S
- 30,750 psi G
t 1
Gt* 2161 in.
Pet* 9,676 psi 29.100 9.676 For special requirement of S
- 2 S
- 41,000 psi
- n Ped* 6,722 ~ 29,100 Peb
- 13,299 _!. 29,100 29, 1_00 13,299 Pet - 12,896 ~ 29,100 sh - 33,733 ~ 58,200 psi 58.-200 33,733 Thermal Secondary Reference paragraph ~B 3545.2(c), Figures ~"B J545.2(c)-2, Stress at Crotch ~B 3545.2(c)(2), ~B 3545.2(C)-3, :1s 3*545,2(c)-3, mJ 3545.2(c)-4 Region QT
- QT + QT 1 2 where re 1
- 3 in ** QT
- 1;100 1
QT
- c6 c 2AT 2 vhere c2
- 0.53 c6
- 210 and 6T 2
- 4.7 2
Qr
- 523 psi, Q7
- 1,623 psi 2
criterion S~
- QP +Ped+ 20T .::, 3 Sm 2
where QP
- 25,965 Ped* 5,Ot.l QT
- 523 2
as 32,052 ~ 58,200 criterion satisfied 58,200 32,052
- -:ormal Duty Valve Reference raragraphs 35t.S. 3, NB 3545. 3(a), :is 3545. 3a, F3t1gue Require~nts H.~urc :-~-1 i*r i Co! l*i<'~ :'- ! * 'j~=" ~- *.*c 1-e :1 o'.i.
Rev. 35, 07/84
J'AB_~~9..:_2.ill f!').'.l~e t. of 9) iAli* STEM! 1S01.AtlON \'AL\'E Allowable Stress Calculat~d Stress or Minimum or Actual Criteria nethod of Analvsis Thickness (i_n_._) Thiclmess .{_i_n_.J Normal Duty Valve S
- 2 Q' p+peb+QT +l.J*~T Sp .. 0.4Q' p+
Fatigue Requirements pl 3 2 3 l 2 (cont'd.)
~ (P b + 20T )
2 e 3 where Q~ P
- 25.965 P - 9,974 K*2, QT
- 1.100 Q~
- 622 psi eO 1 *3
- Sp
- 24,349 Sp
- 21 1 604 S equal to the larger of Sp and Sp + Sa
- 24,_349 psi I 2 a 1 2
~ N
- 50,000.:. 2.000 criterion satisfied 8
Cyclic Loading Re- Reference paragraph 3550 quirements at Valve For the largest temperature change range criterion QP +Ped+ c6 c2c4 ~Tf max~ 3 Sm crotch Where Q' P;:: 25,965 psi, Ped* 5.041 c6 "" 210 at ,Hf max of 171 F, c2
- 0.53 C4
- 0.03, Sm, 19,400
- 31,577 ~ 58,200 criterion satisfied 58,200 31,577 thermal Transients ~ot Excluded bv Code criterion ENri < l
. E!\
Calculate the fatigue usage factor (It) as follows:
Sn Max* Q~ p + P~b + c 6c 3c4 ~Tf max~ Sn max* 36,18 psi 58,200 36,618 Since Sn max~ 3 Sm(* 58.200) the follo~ing equation is used:
51 * .! Q' p + peb + c6 (C3C4+C5) 6Tf1 3
for ~tfi + 45°F. Nri
- 120 s1 ~ 54.719 psi, ~i
- 25,000, Nri,..'Jt *
- 0048
~T:!
- 12l~F ~Ti* 8, s 1
- 83,069 psi, ~l
- 6.000
~:-!.;'\ * .0013
~tf~ * ~J*F. ~r!
- 10, Si* ~8,319 psi, ~l * ~0,000
~= r i ." ::! .. . ()*~05 Rev. 35, 07/84
TABLE ).9-2(h) (pa~e 5 of 9)
~IN STEA.~ ISOLATION VALVE
-1.ll<>vable Stress C3lculaced Stress or Hininium or Actual Criteria ~lethod of Analvsis Thickness (in.) Thickness ! i_n .J X
as I t -~
- -
- ri
- 0. 0066< l criterion satisfied
~1 LOO .0066 Disk Destgn,Calculatlon Reference para~raph NB 3546.J. Table I-1.1, Roark, 3rd Ed., Pages 198, 200, 201 Disk design condi.tions, PR~ 1250 psi at 500°F, Sm* 17,800 psi at 500°F 4 4 Case No. 13 S
- 3\J (a (3m + 1) + b (m - 1) -
2 2 2 4mt (a -b )
2 2 2 2 4 3 b -4(m + 1) a b (ln(a/b))
m where~* 1250 psi, m - 10, t
- 5,875 in., a* 10.75 in., b
- 1.750 inc .*
3 s t
- 9,488 psi 2 .i!l.
Case No. 14 S * ~ (2a (m + 1) ln (b) + (m - 1})
2 2 2irmt a2-b where~* 58,123 lbf' t
- 5.875 in., m - 10, 3
a
- 10.75 in., b
- 1.75 in., s t
- 4,460 psi s - s + st
- 13,949 !. 17,800 psi 17.800 lJ. 949 t tease no. 13 Case No. 14 4 _!& 4 b.
Case ~o. 21 S t'
- 3W {4a (m + 1) lr (b) a (m + 3) + b (m - l)_+_ _4a_:_t:>2_i
") 2 2 4t~ a (m + l) + b (m - l) whe~e W
- 1250 m
- 10/3, t
- 3.125 in., a* 10.75 in ** b
- 7.25 in.
- Sr* 5.761 psi J&
-- [2a 2 1 '
Case No. 22 S
- JW (m + 1, ln (b) + a (m - t) - b- (m - l))
\" ") ., ")
'.!".t- 3. (~ ~ !) + ~- (m - L)
~here w* ~52.?SS ~
- 10/3. t
- J.:!s ~ * :0.75. ~ * :.25 Sr
- tO, 738 pst Rev. 35, 01/84
TABLE J.9-2Qtl (page 6 of 9)
MAIN ST£AM ISOLATION VALVE Allowabl~ Stress Calculated Stress or Minimum or Actual Criteria Method of Analis1s Thickness ~ Thie kness ~
Disk Design Calculation Total stress a S:az + Sr 22
- 16,499 p~i, allowable stress 17,800 psi 17,800 16.~99 (codt'd.) sshear at inner ge dis~
I 2 Sh
- F vhere F
- 264,536 lb, A* 71,18 in. Sh
- J,716 psi sear A sear Allowable shear stress* 0.6 x allowable stress* 0,6 x 17,800
- 10,680 psi 10,680 3,716 F *2 2 Tensile Stress at SA* A where F* 46,342 lb, At* 2.624 in. , S
- 17,661 lb/in thread Relief t Sn* 17,661 psi, Sm* J0,600 psi 30,600 17.661 Bonnet Design Calcula- Reference paragraph NB 3546,1 tions including Seismic accelerations for SSE Minimum Thickness *P
- P + P P
- 16M + 4F fd eg eg -- --
2 1rG3 11G where M
- 335,253 ln.-lb, F
- 46,342 lb, G
- 24,75 in.
- Peg* 209 psi, Pfd
wher~ C
- 0.3, P
- 1,459 psi, S. 17,800 pai, hg
- 2.625 in., w
- 910,144 lb, d
- 24.75 in.
- t
- 4.97 in., t - 4.97 + 0.120
- 5.09 in. (Corrosion allovance is 0.120 in.) 5.09 3.344 Rev. 3S. 07/84
TABLE 3.9-2(h) (pa~e of 9)
MAIN STL\..~ lSOLATION VALVE Allovable Stress Calculated Stress or Minimum or Actual Criteria Method of Anal~sts ThickneH (_i_n-1 ThickAess (_1._n.)
Reinforcement Reference para8raph SB 3643.3 To account for the opening for stem in the bonnet
~
Required Reinforcement A
- 7,207 {n- 8.6844 7.207 r
2 Available Reinforcement A* 8,6844 in Bonnet Studs Design Reference paragraph 3232,1 and Article E-1000 Calculation Bolt used 20 pieces of 2 - 8 UNC Bolts 2
Total bolt area* 53.04 in.
Normal Operation 1. Pressure stress at Opera~ing Condition W 2 s1 * ~
- 17,159 lb/in. where Wml
- 910,144 lb.
3/4 3/4* SJ.04 in.
2 Aml -
W
~
- 910 0 144
- 32,85 in 2
53.04 32.85 SBo 27,700
- 2. Gasket load at ambient condition with no internal pressure s2 * "'
~
- 2,019 lb/tn. 2 where wm2
- 107,065 lbf
'\
2 3/4* 53.0l. in.
2 Am 2 * ~
W
- 107,065
- 3,06 in 53.04 3.06 s88 35.000 2
Haxi~um tensile stress* 17,159 lb/in.
Thermal 9tress is assumed negligible because the coefficients of thermal expansion of bonnet place and stud a~e the same.
Body Flange Design RefeTence paragraph NB 3647.l Calculations Total flange moment undeT ~peratir.g c~nditions
'.Ii * "71 + ~~ + ~
0
"
- H~ ~* H0
- 0. ;g5 B'P. "n *?.
- 0.591 wh~r~'
- 11.75 in .* ?* l.'39 ?Bi - ~D
- 3~~.QSO l~~* hn
- 2.,13 ln.,
3/4* l.5~4.871 in.-lb "c
- Kc h** He - w-H. he* c;c where W h the high.et' of Wau *nd w 112 Re-v. 35, f7/84 w~,
- o.~a5 c-p + (2b ~ J.l~ cm P)
TABLE J.9-2.Q!}_ (page 8 of 9)
MAIN STEAM ISOLATION VALVE Allowable Stress calculated Stress or Minimum or Actual Criteria ~ethod of Analysis Thickness ~ ThickneH ~
Bod~ Flance Design where G
- 24.75 in., b
- 0.306 in ** m
- 3, y
- 4,300 Calculations (cont'd.)
~ Wml
- 910.144 lb, Wml
- 107,065 lb
~HG* 208,210 lb, hG
- 2.625 in.~ MG O 546,551 in.-lb
~*HT hT, HT* H-H , hT * ~gl*hG 0 2 wh@re H
- 701,934 HD 3 542,080 R
- 1.5 in., g
- 2.625 in., hG
- 2.625 in.
1
- HT* 159,854 lb, hT
- 3.375 in., MT* 539,507 in.-lbf M
- 2,610,929 in.-lb. where~ z 1.524,871 in.-lb, 0
MG* 546,551 in.-lb, ~
- 539,507 in.~lb Total flance lDOOlent under gasket seating condition A+
M
- wcc-c,. Y - <~> s o 2 2 a vhere C
- 30 tn ** 3/4* 2 53.04 in.
- G
- 24.75 in.,
2 A
- 32.86 in. , s
- 35,000 psi at l00°F m a
- W
- l,SOJ,198 lb* H
- 3,014,460 lb-in.
0 Longitudinal Hub Stress Reference Paragtaph NB 3647.l(c) fMo + PB 2 2 Sh *L-- 26,700 21,794 2 4 ~ 21,794 lb/in. < l.S Sfo
- 26,700 lb/in.
Sl 8 go i\adial Stress Reference tA-51(1), Equation (7) of Section VIII of ASME B&PV Code. :971 Edition SP,. (l.33 te +.:.) M,~. l2.J03 "Jsi . 1.5 ::\*.,,.. 26,nfl 26, 7 ()0 12,303 Lt-8 Tan~er.tial Stress 3 T * ,Y!i zsR
- 7.1:6 psi< 1.~ si~
- 26,;oo 26,iOO 7,126 t2e.
- Oi vhere Y
- 4.5, t
- 4.125 tn,, Z
- 2.4, B
- 21.7S tn.
Rev
- 3S
- o7 / 8 4
TABLE 3.9-2fil (page 9 of 9)
MAIS STEAM ISOLATION VALVE Allowable Stress Calculated Stress or Minimum or Actual Criteria Xethod of Analrsis Thickness ~ Thicknese i!!_\_:.l Flaage Stress Cri~eria SH+ SR< S. SH+ SR< S
-l- m 2 m SH+ SR~ 17,049 psi
- ! 17 .soo 17,049 SH+ 5t
- 14,460 psi Stem Calculation Back Seated Stress s - IA vhere F
- 30,697 lb net upward force A= 2.535 in. 2 , the smallest cross sectional area on the stem S = 12,110 psi < 26,880 psi Valve Close Ste-en. Stress s .. !
A vhere F
- 46,959 lb net down force 2
A* ).141 in. , the smallesc cross sectional area on the stem S
- 14,950 psi< 26,880 psi Stem Thread Strength Refe~ence Federal Thread Standard Stem Thread Mating Vith Disk Thread 2.00 in. - 8 L,t - 2 Thread 2
A .. 3.49 in. /inch enp;agement.
51
\. 1
- t,,54 in.
~ ~
- - 'F
- _Sl 2
w~ere F * ~6.959 lbf' ASl
- 6.54 ln. ~ tSd
- 7,177 psi Rev. 35, 07/84
Table Rev. 53 SSES-FSAR TABLE 3.9-2i RECIRCULATION PUMP ALLOW. STRESS CRITERIA ANALYTICAL RESULTS OR ACTUAL THICKNESS
- 1. Casing Minimum Wall Thickness t = 2.69 inches Sall. = 15,075 psi PR tact. = 3.00 inches t C SE 0.6 P Loads:
Normal and Upset Condition where:
Design pressure & temperature t = min. reqd thickness, in.
P = design pressure, psig Primary membrane stress limit: R = max. internal radius, in.
S = allowable working stress, psi Allowable working stress per E = joint efficiency ASME Sect. III, Class C C = corrosion allowance, in.
- 2. Casing Cover Minimum Thickness Ss = 3,380 psi Sall. = 8,750 psi F tact. = 3.5 inches Ss A
Loads:
Normal and Upset Condition F = force A = area at shear point Design pressure & temperature 2
K qa Primary bending & shear stress limit: Sb 2 h
Sb = 5,950 psi Sall. = 1.5 x 15,075 1.5 Sm per ASME Code for Pumps tact. = 7 inches and Valves for Nuclear Power Class I FSAR Rev. 66 Page 1 of 5
Table Rev. 53 SSES-FSAR TABLE 3.9-2i RECIRCULATION PUMP ALLOW. STRESS CRITERIA ANALYTICAL RESULTS OR ACTUAL THICKNESS q = pressure load a = radius of O.D.
Allowable working stress per b = radius of I.D.
h = plate thickness
- 3. Cover and Seal Flange Bolt Areas Cover Flange Bolts Bolting loads, areas and stresses shall be calculated in accordance Sact. = 19,050 psi Sall. = 20,000 psi Loads: with Rules for Bolted Flange Connections - ASME Sect. VIII, Am = 90.2 sq. in. Aall. = 101 sq. in.
Para. UA-49 Am = 79 sq. in. (For Upgraded Gasket) Aall. = 95 sq. in. (For Normal and Upset Condition Upgraded Gasket)
Seal Flange Bolts Design pressure & temperature Design gasket load Sact. = 18,000 psi Sall. = 20,000 psi Bolting Stress Limit: Am = 9.85 sq. in. Aall. = 11.1 sq. in.
Allowable working stress per ASME Sect. III, Class C
- 4. Cover Clamp Flange Thickness Flange Thickness Stress tact. = 9.25 inches Flange thickness and stress shall be calculated in t = 8.9 inches Sall. = 26,250 psi Loads: accordance with Rules for Bolted Flange Connections - ASME Sect.
VIII, Para. UA-51 Normal and upset Condition Design pressure & temperature Design gasket load Design bolting load FSAR Rev. 66 Page 2 of 5
Table Rev. 53 SSES-FSAR TABLE 3.9-2i RECIRCULATION PUMP ALLOW. STRESS CRITERIA ANALYTICAL RESULTS OR ACTUAL THICKNESS Tangential Flange Stress Limit Allowable working stress per ASME Sect. III, Class C
- 5. Seal Cover Sb = 2,870 psi Sall. = 15,075 psi KP Loads: Sb 2 t
I t = 1.10 tact = 2 9/16 Normal and Upset Condition Design pressure & temperature P = Bolt load due to pressure t = Thickness, in.
K = Constant of shape factor
- 6. Seal Chamber Minimum Wall Thickens t= 0.741 inches Sall. = 1.5 x 17,075 psi PR tact. = 1.375 inches t C SE 0.6 P Loads:
Normal and Upset Condition where:
Design pressure & temperature Piping reactions during normal t = min. reqd thickness, in.
operation P = design pressure, psig R = max. internal radius, in.
Combined Stress Limit: S = allowable working stress, psi E = joint efficiency 1.5 Sm per ASME Section III Code C = corrosion allowance, in.
for Pumps and Valves for Nuclear Power Class I FSAR Rev. 66 Page 3 of 5
Table Rev. 53 SSES-FSAR TABLE 3.9-2i RECIRCULATION PUMP ALLOW. STRESS CRITERIA ANALYTICAL RESULTS OR ACTUAL THICKNESS
- 7. Mounting Bracket Combined Stress Lug #1 SC = 2,270 psi 1.5 Sm = 25,013 psi Bracket vertical loads shall I
be determined by summing Lug #2 SC = 24,429 psi Loads: the equipment and fluid weights and vertical seismic Lug #3 SC = 14,178 psi Flooded weight forces. Bracket horizontal DBE horizontal seismic force = 2.50g loads shall be determined by Lug #4 SC = 4,591 psi DBE vertical seismic force = 1.61g applying the specified seismic force at mass center of pump-motor assembly (flooded).
Horizontal and vertical loads Combined Stress Limit: shall be applied simultaneously to determine 1.5 Sm per ASME Code Section III for tensile, shear and bending Pumps and Valves for Nuclear Power stresses in the brackets.
Class I Tensile, shear and bending stresses shall be combined to determine max. combined stresses.
- 8. Stresses Due to Seismic Loads Motor Bolt Tensile Stress:
The flooded pump-motor Loads: assembly shall be analyzed Sact. = 22,471 psi Sall. = 30,800 psi as a free body supported by Operation pressure and temperature constant support hangers Pump Cover Bolt Tensile Stress:
DBE horizontal seismic force = 2.50g from the pump brackets.
DBE vertical seismic force = 1.61g Horizontal and vertical Sact. = 19,417 psi Sall. = 32,000 psi seismic forces shall be applied at mass center of Motor Support Barrel Combined Combined Stress Limit: assembly and equilibrium Stress:
I reactions shall be determined Yield stress for the motor and pump Sact. = 3,307 psi Sall. = 22,400 psi brackets. Loads, shear, and moment FSAR Rev. 66 Page 4 of 5
Table Rev. 53 SSES-FSAR TABLE 3.9-2i RECIRCULATION PUMP ALLOW. STRESS CRITERIA ANALYTICAL RESULTS OR ACTUAL THICKNESS diagrams shall be constructed using live loads, dead loads, and calculated snubber reactions. Combined bending, tension and shear stresses shall be determined for each major component of the assembly including motor support barrel, bolting and pump casing. The maximum combined tensile stress in the cover bolting shall be calculated using tensile stresses determined from loading diagram plus tensile stress from operating pressure.
FSAR Rev. 66 Page 5 of 5
SSES-FSAR TABLE 3.9-2 (j)
REACTOR RECIRCULA~ION SYSTEM GATE VALVES, DISCHARGE I
' STRUCTURAL & MECHANICAL LOADlNG CRITERIA 0.
0.1 Component Loads Design Design Procedure Required Design Value Actual Design Value 0.2
- 1. Bodv and Bonnet ,
loads: Design pressure, 1525 psi
~
1525 psi 1.1 ~endor's Design calculation design temp. 575cF 575°F Used NB-3543, Table NB*3531-4 1.2 Pressure Rating. lb. & NB*3531-5 of ASME Pr= 826 lb. P, = 826 lb.
SECTION 111 Minimum wall thickness, Used ASME SECTION Ill PARA 1.3 inches NB-3543. Table NB-3542-1 tm ~ 2.42 inches trn = 2.4224 inches I
Used ASME SECTION Ill PARA 1.4 Primary membrane stress, psi Prr. $ S"' (500~F) = 19600 psi Pr.,::: 9954 psi NB-3545.1 Secondary stress due to pipe Used ASME SECTION 111 PARA Ped s 1.5 Sm= 29400 psi P~ =4271- psi 1.5 Prb s 1.5 Sm= 29400 psi P~:i = 7581 psi reaction NB*3545.2 {b) (S == 30,000 psi)
Pet ~ 1.5 Sm=: 29400 psi Pai= 7170 psi Primary plus secondary Used ASME SECTION Ill PARA 58,000 psi 1.6 stress due to internal Op= 24218 psi NB-3545.2 {a} {See 1.8 below) pressure Used ASME SECTION 111 PARA 58,000 psi 0-:-, = 5800 psi 1.7 Thermal secondary stress Or2 = 1757 psi NB-3545-2 {c) (See 1.8 below)
QT3 =2044 psi Sum of primary plus Used ASME SECTION Ill PARA Sn = a~ ,.. Pe:: + 20r2 1.8 secondary stress NB--3545.2 =
Sn s 3Sr:, (500°F) 58800 psi Sn= 31674 psi Used ASME SECTION Ill PAR.A 1.9 Fatigue requirements N~ ~ 2000 cyctes N. > 10scycles NB-3545.3 Used ASME SECTION lll PARA 1.10 Cyclic rating NB-3550 1151 11 = .00335 2.0 Bodv to Bonnet Boltina Loads: Design pressure &
Used ASME SECTION 111 PARA temp., gasket loads, stem 2.1 NB-3546.1 NB-3647.1 and operational load, seismic load Sedion VIII *
Used ASME SECTION Ill PARA A~ 2: 42.79 sq. in. Ab::: 55.86 sq. rn .
2.2 Bolt Area NB-3546.1 NB-3647.1 and Section VIII S:i s 27975 psi Sb =21288 psi 2.3 Body Bonnet Flange Stresses Used ASME SECTlON mPARA SH 5 1.5 S,.. (575°F) = 28837 psi Sr1 = 24206 psi 2.3.1 Operating condition NB-3546.1 NB-3647 .1 and SR :S 1.5 S:n (575°F) = 28837 psi SR= 7307 psi Section VIII Sr s 1.5 S111 (575°F) = 28837 psi Sr= 7812 psi Used ASME SECTION Ill PARA =
S11 ~ 1.5 S.,., (100"f) 30000 psi SH =29837 psi 2.3.2 Gasket seating condition. NB-3546.1 NB-3647.1 and SR $1 .5 s'11 (100"F) = 30000 psi SR= 11050 psi Section VIII Sr s; 1.5 Sm (1 oo~F} = 30000 psi Sr= 11815 psi NOTES (1} SECTION Ill =ASME BOILER AND PRESSURE VESSEL CODE. SECTION llt. 1971, *NUCLEAR POWER PLANT COMPONENTS.
(2) VALVE OIFFERENTIAl PRESSURE 200 PSI Rev. 53, 04/99
- Page 1 of 4
SSES-FSAR TABLE 3.9*2 (j)
REACTOR RECIRCULATION SYSTEM GATE VALVES, DISCHARGE STRUCTURAL & MECHAN,CAL LOADING CRITERIA I
0.
0.1 Component Loads Design Design Procedure Required Design Value Actual Design Value 0.2 3.0 Stress in Stem Load Operator thrust and 3.1 torque Calculate slenderness ratio. If Slenderness ratio = 61.1 greater than 30, calculate Actual load on Stem =
3.2 Buckling*on Stem allowable load from Rankine's Maximum allowable load = 82.635 lbs. 34.342 lbs. Therefore no
. . formula using safety factor of 9. buckling .
Calculate stress due to operator 3.3 Stem Thrust Stress S--:- ~Sm= 44100 psi S1;:: 6968 psi th,ust in critical cross-section Calculate shear stress due to 13.4 Stem Torque Stress operator torque In critical cross.section Se :S: .6Sm = 26460 psi Sc= 4285 psi
- 4. Dlsc Analvsis l*U Loads: Maximum differential pressure Calculate maximum stress according to chapter of R. J.
4.2 Maximum Stress Smax ~ 1.55m {575°F) = 28500 psi Sll'.a* = 22191 psi Roark KFormulas for Stress and Strain*
Connections Calculate stresses in the yoke and yoke connections to
5.1 Loads
Stem operational load acceptable structural analysis methods Tensile stress in yoke legs 5.2 Sm1x ~ S"' == 23.300 psi Max. stress = 8869 psi bolts I 5.3 Bending stress of yoke legs Sr,:,; 1.5 Sm"' 34,950 psi So= 6359 psi Rev. 53, 04/99 Page 2 of 4
SSES-FSAR TABLE 3.9-2 (j)
REACTOR RECIRCULATION SYSTEM GATE VALVES, SUCTION STRUCTURAL & MECHANICAL LOADING CRITERIA 0.
0.1 Component Loads Design Design Procedure Required Design Value Actual Oesign Value 0.2'
- 1. Bodv and Bonnet Loads: Design pressure, 1275 psi 1275 psi 1.1 Vendor's design calculation design temp. 575"F 575"F Used NB-3543, Table NB-3531-4 1.2 Pressure Rating . lb. & NB-3531-5 of ASME P, = 695 lb. P,::: 695 lb.
SECTION Ill Minimum wall thickness Used ASME SECTION Ill PARA 1.3 tn 2= 2.03 inches lm:: 2.0254 in.
inches NB-3543, Table NB-3542-,
Used ASME SECTION 111 PARA 1.4 Primary membrane stress. psi P,,, s Sm (S00°F) = 19600 psi P,,, = 13419 psi NB-3545.1 11 Peo ~ 1.5 Sm= 29.400 psi Ped = 4391 psi Secondary stress due to pipe Used ASME SECTION 111 i PARA 1.5 Pcb s 1.5 Sm= 29.400 psi P.,b; 10740psi reaction NB-3545.2 (b) (S = 30,000 psi)
Pt1 ~ 1.5 Sm= 29,400 psi P,, = 9220 psi Primary plus secondary Used ASME SECTION Ill PARA 58,800 psi 1.6 stress due to intemaf NS.3545.2 (a) (See 1.8 below}
Q;, = 23527 psi pressure On::: 3400 psi Used ASME SECTION Ill PARA 58.800 psi 1.7 Thermal secondary stress NB-3545.2 (c) (See 1.8 below) 012 = 11 2 7 psi On= 1332 psi Sum of primary plus S" = 011 + P,cs + 2QTI S, =30332 psi Used ASME SECTION IU PARA 1.8 secondary stress NB-3545.2 S., ~ 3Sm (500°F) = 58800 psi N. > 1o~ cycles Used ASME SECTION Ill PARA 1.9 Fatigue requirements N, ~ 2000 cycles NB-3545.3 USED ASME SECTION III PARA 1.10 Cyclic rating 1, $1.0 11 = .002617 NB-3550 2.0 Body to Bonnet BoltJnq loads: Design pressure &
Used ASME SECTCON Ill PARA temp., gasket loads, stem 2.1 NB-3546.1 NB-3647.1 and operational load, seismic load Section VIII (design basis earthquake)
- Used ASME SECTION Ill PARA Ab~ 35.70 sq. in. Ab = 55.86 sq. in .
I 2.2 Bolt Area NB.3546.1 NB-3647.1 and Section VIII Sb 2= 27975 psi Sb = 17881 psi Used ASME SECTION Ill PARA !
2.3 Body Bonnet Flange Stresses NB-3546.1 NB-3647.1 and Section VIII Used ASME SECTION Ill PARA SH 1.5 s.,, (575°F) = 28837 psi S-.i = 21000 psi 2.3.1 Operating condition N B-3546. 1 N B-364 7 .1 and SFl 1.5 Sm (575°F) = 28837 psi SR= 5677 psi Section VIII Sr 1.5 Srn (575:.F} = 28837 psi Sr= 8049 psi Used ASME SECTION Ill PARA SH~ =
1.5 Sm (100°F) 30000 psi SH = 27342 psi 2.3.2 Gasket seating condition NB-3546.1 N~3647.1 and SR~ =
1.5 Sm {100°F) 30000 psi SR= 9351 psi Section Viii Sr~ 1.5 Sen (100;:,F} = 30000 psi S-:- =13257psi NOTES (1) SECTION Ill::: ASME BOILER ANO PRESSURE VESSEL CODE. SECTION Ill. 197,. "NUCLEAR POVv'ER PLANT COMPONENT" (2) VAlVE DIFFERENTIAL PRESSURE 50 PSI Rev. 53, 04/99 Page 3 of 4
SSES-FSAR TABLE 3.9-2 (j)
I REACTOR RECIRCULATION SYSTEM GATE VALVES, SUCTION i' STRUCTURAL & MECHANICAL LOADING CRITERIA I
I o.
0.1 I Component Loads Design Design Procedure Required Design Value : Actui!I Design Value 0.2 I I 3.0 Stress in Stem Load Operator thrust and 3.1 torque Calculate slenderness ratio. If Slenderness ratio = 61.1 greater than 30, calculate 3.2 Buckling on Stem Maximum allowable load = 82.635 lbs. Actual load on Stem= 17047 allowable load from Rankine's lbs. Therefore no bu~ling.
formula using safety factor of 9.
Calculate stress due to operator 3.3 Stem Thrust Stress S-:- s S11 = 44100 psi Sr = 3473 psi thrust in critical cross-section Calculate shear stress due to 3.4 Stem Torque Stress operator torque in critical Sc ~ .6 Sm = 26460 psi S: = 2140 psi cross-section
- 4. Oise Analvsis Loads: Maximum differential 4.1 pressure Calculate maximum stress according to chapter of R. J.
4.2 Maximum Stress Roark -Formulas for Stress and SIT1,1lll $ 1.5 Sm (575°F) = 28500 psi s,.,... = 18393 psi Strain:
I Rev. 53, 04/99 Page 4 of 4
SSES-FSAR Table Rev. 55 TABLE 3.9-2L STANDBY LIQUID CONTROL PUMP LIMITING ALLOWABLE CALCULATED STRESS STRESS STRESS CRITERIA/LOADING COMPONENT TYPE (PSI) (PSI)
The standby liquid control pump has been designed and fabricated to the 1968 P&V Code for Class 2 component.
Pressure boundarv parts:
- 1) Fluid cylinder - SA 182-F304 Sv = 30,000 psi
- 2) Discharge valve stop stuffing box Sv = 30,000 psi and cylinder head extension SA 479-304
- 3) Discharge valve cover, cylinder Sv = 30,000 psi head and stuffing box flange plate ,
SA 285 GR. C
- 4) Stuffing box gland , ASTM A461 , Sv = 90,000 psi GR. 630
- 5) Studs, SA 193B7 Sv = 105,000 psi
- 7) Studs, cylinder tie, SA 193-B7 SA= 105,000 psi
- 8) Pump holddown bolts, SAE GR. 1 TA= 15,000 psi QA= 12,000 PSi
- 9) Power frame , foot area , cast iron SA= 15,000 psi
- 10) Motor holddown bolts, SAE GR. 1 TA= 15,000 psi QA= 12,000 psi
SSES-FSAR Table Rev. 55 TABLE 3.9-2L (Continued)
STANDBY LIQUID CONTROL PUMP LIMITING ALLOWABLE CALCULATED STRESS STRESS STRESS CRITERIA/LOADING COMPONENT TYPE (PSI) (PSI)
Normal & Uoset Condition Loads:
- 1. Desiqn pressure 1. Fluid Cylinder General Membrane 17,800 See Note (3)
- 2. Desiqn temperature 2. Discharne valve stop General Membrane 17,800
- 3. Operatinq basis earthquake 3. Cylinder head extension General Membrane 17,800
- 4. Nozzle loads' ' 1 4. Discharge valve cover General Membrane 17,800
- 5. Dead weight 5. Cylinder head General Membrane 17,800
- 6. Thermal expansion 6. Stuffing box flange plate General Membrane 17,800
- 7. SRV discharqe 7. Stuffinq box qland General Membrane 35,000
- 8. Cylinder head studs Tensile 25,000
- 9. Stuffinq box studs Tensile 25,000 Emeraencv or Faulted Condition:
- 1. Desiqn pressure 1. Fluid cylinder General Membrane 21 ,360 4,450
- 2. Design temperature 2. Discharge valve stop General Membrane 21 ,360 13,600
- 3. Weight of structure 3. Cylinder head extension General Membrane 21 ,360 13,600
- 4. Thermal expansion 4. Discharge valve cover General Membrane 21 ,360 8,150
- 5. Safe shutdown earthquake 5. Cylinder head General Membrane 21 ,360 8,150
- 6. SRV 6. Stuffinq box flanqe plate General Membrane 21 ,360 10,390
- 7. LOCA 7. Stuffinq box qland General Membrane 42,000 11,420
- 8. Nozzle LoadsP > 8. Cylinder head studs Tensile 25,000 18,820
- 9. Dowel pins('> Shear onl/'> 23,400 19,400
- 10. Studs, cylinder tie Tensile1 25,000 8,685 11 . Pump holddown bolts Shear 12,000 9,415
- 12. Pump holddown bolts Tensile 15,000 11,675
- 13. Power frame-foot area Shear 15,000 1,850
- 14. Power frame-foot area Tensile 15,000 11,390
- 15. Motor holddown bolts Shear 12,000 3,020
- 16. Motor holddown bolts Tensile 15,000 5,290
- 17. Motor frame-foot area Shear 15,000 2,070
- 18. Motor frame-foot area Tensile 15,000 4,125 FSAR Rev. 60 Page 2 of 5
SSES-FSAR Table Rev. 55 TABLE 3.9-2L (Continued)
STANDBY LIQUID CONTROL PUMP LIMITING ALLOWABLE CALCULATED STRESS ACCELERATION ACCELERATION CRITERIA/LOADING COMPONENT TYPE (G) (G)
Faulted Condition Dynamic Loads SLC Pump Assembly Equivalent static 1.75g 0.45g acceleration (vertical) 1.75g 0.73g (horizontal) l 1 2.
SSES-FSAR Table Rev. 55 TABLE 3.9-2L (Continued)
STANDBY LIQUID CONTROL PUMP LIMITING ALLOWABLE CALCULATED CRITERIA/LOADING COMPONENT STRESS LOADS LOADS Nozzle Load Definition:
Units: Forces - lbs.
Moments - ft. - lbs.
Allowable combination of forces and moments are as follows:
Where:
F1 = The largest absolute value of the three actual external orthogonal forces (Fx, Fy, F,) that may be imposed by the interface pipe ,
and ,
M = The largest absolute value of the three actual internal orthogonal moments (Mx, My, M,) permitted from the pipe when they are combined simultaneously for a specific condition .
SUCTION :
Normal and Upset Condition Loads: Fa = Allowable value of F1 when all Fa = 770 lb. 207 moments are zero .
- 1. Desiqn pressure Mo = Allowable value of M1 when Mo = 490 ft .-lb. 388
- 2. Desiqn temperature all forces are zero .
- 3. Dead weiqht DISCHARGE :
- 4. Thermal expansion
- 5. Operatinq Basis Earthquake F0 = 370 lb. 173 M0 = 110 ft .-lb. 95 FSAR Rev. 60 Page 4 of 5
SSES-FSAR Table Rev. 55 TABLE 3.9-2L (Continued)
STANDBY LIQUID CONTROL PUMP LIMITING ALLOWABLE CALCULATED CRITERIA/LOADING COMPONENT STRESS LOADS LOADS Erner~ of Faulted Condition Loads:
- 1. Design pressure I
- 2. Design temperature SUCTION:
- 3. Dead weight
- 4. Thermal expansion F0 =920Ib 207
- 5. Safe Shutdown Earthquake M0 = 590 ft.-lb. 388
- 6. SRV
- 7. LOCA DISCHARGE:
- 8. Nozzle Loads F0 =440Ib 173 MQ= 130 ft.-lb . 95 NOTES :
(1) Nozzle loads produce shear loads only.
(2) Dowel pins take all shear.
(3) Calculated stresses for emergency or faulted condition are less than the allowable stresses for the normal and upset condition loads , therefore, the normal and upset condition is not evaluated .
(4) Operability: The sum of the plunges and rod assembly, pounds mass times 1.75, acceleration is much less than the thrust loads encountered during normal operating conditions. Therefore , the loads during the faulted condition have no significant effect on pump operability.
FSAR Rev. 60 Page 5 of 5
SSES-FSAR Table 3.9-2m Standby Liquid Control Tank Criteria Method of Analysis Allowables Actuals
- 1. Shell Thickness Loads: Normal and Upset Brownell and Young Design Pressure and Temperature "Process Equipment Design" 0.01542 in. 0.1875 in.
t = ___E!L.
SE- 0.6 P 1
Stress Limit ~SME Section Ill 30,000 psi 1,602 psi
- 2. Nozzle Loads Loads: Normal and Upset The maximum moments due to pipe Design Pressure and reaction and maximum forces shall Temperature not exceed the allowable limits. F0 (lb.) Mo (ft.-lb.)
I Fo(lb.) M:1 (ft. . lb.)
Overflow Nozzle 440 330 298[ 1) 167(2} 21s< 11 203 121 Discharge Nozzle 440 330 298111 15]12> 218(1) 203<2>
I Loads: Faulted Dead The maximum moments due to pipe Weight, Thermal Expansion reaction and maximum forces shall and SSE Earthquake not exceed the allowable limits. FQ{lb.) Mo (ft.*lb.} Fo (lb.) M" (ft. . lb.)
Overflow Nozzle 528 360 31311> 1761<1 231 111 219C2>
I l .
Discharge Nozzle 528 360 313()) i75!Z) 231(!) 219 12 )
- 3. Anchor Bolts :ASME Section Ill 18,750 psi 9.617 psi
- 4. Dynamic Loads Equivalent Static (Horizonta) 1.5g 0.41g (Vertical) 1.5g 0.53g
- 1. SSE
- 2. SRV
- 3. LOCA (1)
Unit 1 (2)
Unit 2 I
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR TABLE 3.9-2n (i} RHR Pumps and (ii) Core Spray Pumps
! Location Loading Condition . Crtteria Calculated Stress (psi} Allowable Stress (psi} I I
(i) RHRPUMPS : I FAUL TED CONDITION ASME Boiler & Pressure Suction Barrel Snell Design Pressure Vessel Code. Section Ill Static Loads 17,672 21,000 Dynamic Loads Design Pressure ASME Boiler & Pressure Stuffing Box P.ipe Static Loads Vessel Code, Section Ill 8,451 18,750 Dynamic Loads i Nozzle Shell Intersection FAULTED CONDITION Design Pressure Static Loads ASME Boiler & Pressure Vessel Code; Section Ill 15,193 (Suction)
- 20. 392 (Discharge) 34.650 34.650 Dynamic Loads FAULTED CONDITION ASME Boiler & Pressure Discharge Elbow or Design Pressure Vessel Code, Section Ill 10,926 i 21 ,000 Suction Pipe (Maximum) Static Loads (Disch. Elbow} j FAULTED CONDITION Bolting Loads & Stresses Motor Stand Dynamic Loads per ASME B&PV, Section 5,717 21,000 Static Loads Ill, Subsection NF FAUL TED CONDITION Botting Loads & Stresses Motor Bolting Dynamic Loads per ASME B&PV. Section 1 5.977 30,000 Static Loads 111, Subsection NF (ii) C~RE SPRAY_PUMPS ..
FAULTED CONDITION ASME Boiler & Pressure Suction Barrel Shell Design Pressure Vessel Code, Section Ill Static Loads 13.499 21,000 Dynamic Loads
! Design Pressure ASME Boiler & Presst1re Stuffing Box Pipe Static Loads Vessel Code, Section Ill 6,766 18,750 Dynamic Loads FAULTED CONOITrON ASME Boller & Pressure Nozzle Shell Intersection Design Pressure Vasser Code, Section Ill 33,417 (Suction) 34,650 Static Loads 31,562 (Discharge) 34.650 Dynamic Loads FAULTED CONDITION ASME Boiler & Pressure Discharge Elbow or Design Pressure Vessel Code, Section Ill 11,061 21,000 Suction Pipe (Maximum) Static Loads {Disch. Elbow)
FAUL TED CONDITION Bolting Loads & Stresses Motor Stand Static loads per ASME B&PV, Section 4.505 21.000 Dynamic Loads Ill. Subsection NF FAULTED CONDITION Bolting Loads & Stresses Motor Bolting Static Loads per ASME B&PV. Section 1,960 30.000 Dynamic Loads Ill. Subsection NF Rev. 53, 04199 Page 1 of 1
SSES-FSAR TABLE 3.9-20 RHR HEAT EXCHANGER Allowable Min. Actual Thickness Minimum Criteria Method of Analysis Required Thickness
- 1. Closure Bolting Bolting loads and stresses calculated per "Rules for Bolted Loads: Normal and Upset Flange Connections" ASME Section 1111971, Winter 1972 Addenda Design pressure and temperature Design gasket load Boltina Stress Limit Allowable working stress a. Shell-to-tube sheet bolts 1-3/8" dia. 1-3/8" dia.
per ASME Section Ill
- b. Channel cover bolts 1-3/8" dia. 1-3/8" dia .
- 2. Wall Thickness Shell side ASME Section Ill Class 2 and TEMA Class C Loads: Normal and Upset Tube side ASME Section 111 Class 3 Design pressure and and TEMA Class C J temperature I
Stress Limit I
I ASME Section Ill a. Shell 0.905 in. 1.0 in:
- b. Shell cover 0.895 in. 0.895 in. min .
- c. Channel 0.932 in. 1.0 in.*
- d. Tubes 16 BWG 0.049 in. 0.065 in.*
- e. Channel cover 8.98 in . 9.00 in.*
- f. Tube sheet 7.42 in. 7.50 in.*
- Stresses within 10% of allowable.
ii Rev. 53, 04/99 Page 1 of 4
SSES-FSAR TABLE 3.9-20 RHR HEAT EXCHANGER Allowable Nozzle Forces and Moments Calculated Nozzle Criteria Met~od of Analysis Force in lbs., Forces and Moments
-Moment in in.-
lbs.
- 3. Nozzle Loads The maximum moments due See below. El+ Mi= 0.86 (Inlet)
Design pressure and to pipe reaction a*nd the maximum forces shall not Fo* Mo I
tempera tu re exceed the allowable limits. Fi + Mi = 0.53 (Exhaust)
Fo Mo Dead weight, thermal Primary stress less than 1.5 Maximum allowable piping loads for expansion design ASME Section 111 allowable emergency conditions (including DBE) shall basis earthquake not exceed the following relationship for each nozzle.
Allowable limits Fi + Mi ~ 1 (design basis) Fo Mo
_fil_ -1:!L ~ ~ where Ft (jbs.} is the maximum ot the three (3) orthogonal forces Fx = 15,500 15,500 21,500 21 ,500 Fy::: 15,500 15,500 21,500 21,500 (Fx Fy Fz)
Fz = 15,500 15,500 21,500 21,500 Mx = 60,000 60,000 100,000 100,000 and Mi (ft.-lbs.) is the maximum of the My::: 60,000 60,000 100,000 100,000 three (3) orthogonal moments Mz = 60,000 60,000 100,000 100,000 (Mx My Mz)
Rev. 53, 04/99 Page 2 of 4
SSES-FSAR TABLE 3.9-20 RHR HEAT EXCHANGER Allowable Min. Thickness Actual Criteria Method of Anal~sis Reguired (Qsi) _JQfill
- 4. SuQ~Ort Brackets & SAP-4G Finite Element Attachment Welds Computer Code Stress Analysis Loads: Faulted Design pressure and temperature, dead weight; nozzle loads, safe shutdown earthquake Stress Limits Allowables as per ASME a. Lower Bracket Welds Section Ill, Subsection NF
{Upset Condition} - Bending Stress 14,438 5,990
- Shear Stress 21,000 3,304
- b. Upper Bracket Welds
- Bending Stress 14.438 9,887
- Shear Stress 21,000 3,197
- 5. Anchor Bolts Bolt loads calculated using SAP-4G finite Element Computer Code Loads: Faulted Design pressure and temperature, dead weight, nozzle loads, safe shutdown earthquake Bolting Stress Limits Allowable stresses as per Lower Support Bolting ASME 111, Appendix XVII Tension - 2461.1 Tension 29,000 13,115 Shear -2461.2 Shear 11,990 8,647 Rev. 53, 04/99 Page 3 of 4
SSES-FSAR TABLE 3.9*-20 RHR HEAT EXCHANGER Allowable I Min. Actual Criteria Method of Analtsis Thickness .JQ.§!1 Reguired U2fill
- 6. She!I Adjacent to Su~~ort Finite Element and Localized Shell Brackets Stress Analysis Loads: Faulted i
Desig*n pressure and temperature, dead weight, nozzle loads, safe shutdown earthquake I
Shell Stress Limit Allowables as per ASME a. Maximum Principal Stress 28,875 22.432 Section 111, Subsection NC adjacent to upper support
- b. Maximum Principal Stress 28,875 23.463 adjacent to lower support l
- 7. Shell Finite Element Dynamic Stress Analysis Loads: Faulted Design pressure and temperature, dead weight, nozzle loads, safe shutdown earthquake I
Stress limits Allowable as per ASME - Principal Stress 19,250 18,560 Section 111, Subsection NC .I (Upset Condition)
=
Rev. 53, 04/99 Page 4 of 4
SSES-FSAR TABLE 3.9-2(p)
RWCU PUMP The following is a summary of design calculations by Hayward Tyler Fluid Dynamic*s Ltd. on the ,RWCU Pump:
Page 1 of 2 Required Allow11blfl Actual ASM; ~2d~ Qa)gu.JD:thm Thi~kDlHlA Unl Strg§§ ,gsi} Thicknei1 {inl Pump Part Body 0.8833 17,500 2.081 Drain Nozzle 0.0421 17,500 0.181 Suction Nozzle 0.2513 17,500 0.422 Suction End 3.688 17,500 3.713 Suction Bend 0.2518 15,000 0.43 Discharge Branch 0.1693 11 tsoo 0.8615 Pump/motor Flange Studs 1.3393 (NDtD 1) 25,000 1.376 {Note 11 Motor Part Motor Case 3,09 17,500 3.21 End Plate 1.5943 17,500 2.07 Gasket Seating 2.21 17,500 3.00 Thermal Neck 0.1871 17,500 1.3457 Motor Case Shelr Body Pump End 0.54 17,500 0.995 Cover End 0.6058 17,500 2.8025 Motor Cover (Fill, Drain & Purge Nozzle) 0.0421 17,500 0.1986 Flange 2.585 17,500 2.826 Motor Case/Motor Cover Studs 1 .3393 (Note l J 25,000 l .375 (Note 1)
Note 1: Engagement Length Rev. 52, 11/97
SSES-FSAR TABLE 3.9-2{p) (Continued)
Page 2 of 2 Reguir~d Allow1blg Actual ASM~ ~iMh2 ~1ha1.lati20 Thi~kn111 Unl §tr1H l2sil ThickDHi Unl ..
Heat Exchanger Part Heat Exchanger Shell Body 0.2404 17,500 0.331 Nozzles 0.4775 17,500 0.884 Vent Nozzle 0.0401 17,600 0.199 HP Nozzles 0.0884 17,500 0.47725 Heat Exchanger End Cover Body 0.0216 17,500 0.5575 Heat Exchanger End Cover End Plate 0.2076 17,500 1.12 Heat Exchanger LP Tubes 3,461 {Not* 2) 1,450 (Note 21 END ~Q~EB - ELAMQE A~ME ~2df! ~DJ~ulatism ~Dl~ulatcd Slt@H 8112w1blo Strn~~
(psi) mfill O~er!!tiog ~2odilion SH 2,354 (1.5S} 26,250 SR 1,307 {1.05} 17,500 ST 962 (1.05) 17,500 s~ *;;:-;-:--------:- 1,830 (1.0S) 17,500 2
SH+ ST -. 1,658 (1.0S) 17,500 2
Gaskgl Stating Condition . ... . ..
SH 11,916 {1.55) 26,250 SR 6,617 {1.05) 17,500 Sr 4,865 (1.0S) 17,500 SH+ SR . - ... 8,267 (1.0S) 17,500 2
SH+ ST .. 8,391 (1 .OS) 17,500 2
Note 2: Pressure, {psi}
Rev. 52, 11/97
SSES-FSAR Tabla Rev. 36 TABLE 3.9-2q RCIC TURBINE Criteria The highest stressed components*of the RCIC Turbine assembly are identified.
Allowable stresses for pressure retaining components are based on ASME B&PV Code, Section Ill.
Normal Condition: Faulted Condition:
Pressure Boundary Castings: SA216-WCB@ 500°F Pressure Boundai;, Castings: SA216-WCB@ 500°F S = 17,500 psi S = 17,500 psi SA (General Membrane) = 0.8 x S SA = (General Membrane) = 1.2 x 0.8-x S SA (Bending) = 1.5 x 0.8 x S SA= (Bending) = 1.8 x 0.8 x S Pressure Boundary Bolting: SA193-B7@ 500°F Pressure Boundary Bolting: SA 193-B7 @ 500°F SA= 1.0 x S S = 25,000 psi SA= 1.0 x S S = 25,000 psi Alignment Taper Pins: AISI 4037, Re 28-35 Alignment Taper Pins: AISI 4037, Re 28-35 T8 == 61, 100psi Sy = 106t000 psi Ta= 61, 100psi Sy= 106,000 psi LOADING COMPONENT LIMITING STRESS ALLOWABLE CALCULATED TYPE STRESS(PSI) STRESS(PSI)
Normal Condition Loads: Castings: 1) Stop valve General Membrane 14,000
- 1. Design Pressure 2) Governor valve General Membrane 14,000
- 2. Design Temperature 3) Turbine inlet Local Bending 21,000 { 1)
- 3. Deadweight 4) Turbine case Local Bending 21,000
- 4. Inlet Nozzle Loads Pressure boundary bolts Tensile 25,000
- 5. Exhaust Nozzle Loads Alionment taper pins Shear 61.100 Faulted Condition: Castings: 1) Stop valve General Membrane 16,800 9,800
- 1. Design Pressure 2) Governor valve General Membrane 16,800 13,200
- 2. Design Temperature 3} Turbine inlet Local Bending 25,200 15,300
- 3. Deadweight 4) Turbine case Local Bending 25,200 18,000
- 4. Inlet Nozzle Loads Pressure boundary bolts Tensile 25,000 20,100
- 5. Exhaust Nozzle Loads Alignment taper pins Shear 61,100 46,880
SSES-FSAR.
Table Rev. 36 TABLE 3.9-2q RCIC TURBINE Nozzle Load Definition: Allowable Nozzle Load Criteria:
Turbine vendor ~as defined allowable nozzle loads for the turbine assembly. The above calculated stresses assume these allowable nozzle loads have been satisfied.
Normal Loads:
- 1 . Design Pressure Inlet: F = (2620-M)
- 2. Design Temperature/Thermal Expansion 3
- 3. Deadweight (6000- M)
Exhaust: F = - -
3 Where: F = Resultant force (lbs.)
M = Resultant moment (ft.-lbs.)
Faulted Loads:(3)
- 1. Design Pressure Inlet: F = (7000 - M)
- 2. Design Temperature/Thermal Expansion 4.7
- 3. Deadweight .
0.34 Where: F::: Resultant force (lbs.)
M = Resultant moment (ft. -lbs.)
NOTES:
(1) Calculated stresses for the faulted condition are lower than the allowable stresses for the normal condition. Therefore the normal, upset, and emergency conditions are not evaluated.
(2) Operability: Analysis indicates that shaft deflection with faulted loads is 0.008 inch; which is fully acceptable. And, maximum bearing loads under faulted conditions are acceptable. Furthermore, the turbine assembly has been seismically qualified via dynamic testing. This qualification included demonstration of start-up and shutdown capabilities, as well as no load operability during seismic loading conditions.
FSAR Rev. 56 Page 2 of 3
SSES-FSAR Table Rev. 36 TABLE 3.9-2q RCIC TURBINE (3) The nozzle toads for the Upset loading condition, as determined by the piping analysis, shall satisfy the allowable criteria for the Faulted loading I condition.
FSAR Rev. 56 Page 3 of 3
SSES-FSAR Table Rev. 36 TABLE 3.9*2r RCIC PUMP Criteria The critical components of the RCIC pump assembly are identified.
Allowable and calculated values are based on ASME Section Ill or VIII, as applicable.
COMPONENT ALLOWABLE CALCULATED (1) ft2 fv2 Pump Hold Down Bolts: - -2 + - -
Ftb Fvb 2 1 0.502 Anchor Bolt: Shear Stress 10,000 psi 7,422 psi Tensile Stress 21,600 psi 17.090 psi Pump Outer Case 17,500 psi 7,052 psi Pump Outer Case at Discharge Nozzle 26,250 psi 7,855 psi Discharge Nozzle 17,500 psi 3,600 psi Suction Elbow 18,000 psi 7,900 psi Pump Shaft (2) 32,000 psi 5,975 psi Impeller Key 9,000 psi 4,810 psi Mounting Feet 17,500 psi 6,208 psi Mounting Feet Weld Stress 9,625 psi 6,208 psi Pump Pedestal Weld Stress 9,625 psi 1,868 psi Base Plate & Plate Stiffener 21,600 psi 13,818 psi Outboard BearinQ - Axial (2) 17,200 lb. 1,323 lb.
Inboard BearinQ - Axial (2) 7,670 lb. . 376 lb .
Seal Circulation Piping - 1/2" 15,000 psi 10,000 psi
-3/4" 15,000 psi 4,630 psi Bypass Piping 15,000 psi 8,592 psi Shaft: Relative Radial Displacement Between Shaft & Sleeve (2) .0055 in. .00383 in.
Relative Radial Displacement Between Shaft & Mech. Seal (2) .0055 in. .0008 in.
Angular Misalignment at Coupling (2) .017 Rad. .0022 Rad.
Impeller: Relative Radial Displacement Between Impeller & Casinq (2) .0075 in. .00084 in.
Suction End and Discharge End Bolting 25,000 psi 20,740 psi FSAR Rev. 56 Page 1 of 3
SSES-FSAR Table Rev. 36 TABLE 3.9-2r RCIC PUMP Nozzte Load Definition:
Allowable nozzle loads for the pump have been defined. The above calculated stresses assume these allowable nozzle loads have been satisfied.
Units: Forces = lbs.
Moments = ft~lbs.
The allowable combinations of forces and moments are as follows:
Fi
- +Mj- < I Fa Mo -
Where:
Fj = Largest absolute value of the three actual orthogonal forces (Fx. Fy. Fz) imposed by the interface pipe.
M = Largest absolute value of the three actual orthogonal moments (Mx1 My, Mz) imposed by the interface pipe.
FO = Allowable value of Fi when all moments are zero.
M0 = Allowable value of M, when all forces are zero.
Nozzle Allowable Load (3)
Suction F0 = 2906 M0 = 3688 Discharge Fo = 4450 Mo== 5200 NOTES:
(1) Calculated values are due to the highest faulted condition loads and are.less than 1.2 time~ the upset allowables. Therefore, the normal plus upset and emergency conditions are not evaluated.
FSAR Rev. 56 Page 2 of 3
SSES-FSAR Table Rev. 36 TABLE 3.9-2r RCIC PUMP (2) Operability is addressed by the evaluation of the relative displacements between rotating and stationary components. shaft stress, and bearing loads under faulted condition loads. All criteria are met. Therefore, operability is ensured.
(3) Nozzle load allowables are applicable to all loading conditions.
FSAR Rev. 56 Page 3of 3
~
SSES-FSAR TABLE 3.9-2s NIMS Rev. 36 REACTOR REFUELING & SERVICING EQUIPMENT (i) NEW FUEL STORAGE RACKS ALLOWABLE CRITERIA I LOADING I LOCATION I STRESS
.7 ULT I CALCULATED STRESS
- 1. NEW FUEL STORAGE RACKS FAULTED CONDITION A 11 11 Stress due to normal upset or 1. Dead Loads 1. Beam (Axial) 1. 26,000#/in 2 1. 18,905#/in2 emergency loading shall not cause a 2. Full Fuel Load in rack 2. Beam (Trans.) 2. 26,000#/in 2
- 2. 7 ,005#/in 2 failure so as to result in a critical 3. -S.S.E. 3. Combined 3. 26,000#/in 2 3. 25,91 0#/in 2 arra . A- Thermal (not a~rnlicable}
- 2. SOURCE OF ALLOWABLE STRESS (.7 ULT)
- a. ASTM B308 Alloy 6061-T6
- b. ASME Code - Boilers and Pressure Vessels, Sect. 111, NA
- c. Product Safety Standards for BWR-6-Mark 111, Sect. VI, A. (3)
- d. ASME - Pressure Vessels and Piping : Design and Analysis, Volume One, Page 69.
- e. ASTM code for Boilers and Pressure Vessels was selected on the premise that data used from this source .would necessarily be on the conversative sice as applied to the fuel storage rack calculations.
- 3. S. S. E. loads derived by dynamic analysis . Total stress refers to combined earthquake and thermal load at highest expected pool temperature. Earthquake stresses obtained by square root of the sum of the squares method for a response due to tri-axial excitation. Stress given is the highest in the total structural array. The calculated stresses are conservative since they are based on the original fuel assembly type which has the greatest mass of all fuel assemblies in use or in storage at SSES.
Rev. 55 Page 1 of 3
SSES-FSAR TABLE 3.9-2s NIMS Rev. 36 REACTOR REFUELING & SERVICING EQUIPMENT (i) NEW FUEL STORAGE RACKS ALLOWABLE CRITERIA LOADING LOCATION. STRESS I CALCULATED
(.7 ULT) STRESS
- 4. NEW FUEL STORAGE RACKS FAUL TED CONDITION '*B" (Location - See Not Applicable l Not Applicable Par. 6, Below)
Stresses due to normal upset or I (See Below, Par. 5) emergency loading shall not cause a failure so as to result in a critical array.
- 5. FAULTED CONDITIQN_'13_: I Condition "B is an emergency condition in which the stress limit is equal to the yield strength at 0.2% offset. The racks were tested to determine their capability to safely withstand the accidental, uncontrolled, drop of a fuel bundle from its fully retracted position into the weakest portion of the rack.
- 6. METHOD OF TESTING: I Four (4) rack castings were subjected to impact loads ranging from 1908 ft. lbs. To 4070 ft. lbs .. which were generated by dropping simulated fuel bundles weighing 660 lbs. from heights varying from 3.0' to 6.17'. Racks were aligned in pairs and simulated bundles were dropped on both racks at the flange area. Both center impact and end impact tests were conducted. (Two (2) of the racks were X-ray examined prior to testing. Strain gauges were mounted on racks to ascertain max. strain and accelerometers were mounted on bundles to determine G" loads.}
- 7 TEST RESULTS: I A total of nineteen (19) tests were performed with drop height increased at each test.
First failure occurred due to a central impact on rack No. 3 from a max. height of 6.17',
(Test #13). Racks #1 and #2 both failed from a center impact caused by a load dropped from a height of 5.33', (Test #19). Accelerometer readings are not available due to the inability to adequately affix the accelerometer to the simulated fuel bundle.
Rev. 55 Page 2 of 3
SSES-FSAR TABLE 3.9-2s NIMS Rev. 36 REACTOR REFUELING & SERVICING EQUIPMENT (ii) FUEL PREP MACHINE OPERABILITY ACCEPTABLE CRITERIA LIMITING LOAD PRIMARY ALLOWABLE CALCULATED ASSURANCE COMBINATION LOADING STRESS (psi) STRESS (psi) DEMONSTRATED BY AISC Code; Fu= 75 000 1
ASME, Sect. Ill Fy == 30,000 Sm200 = 17,800 Chain 302 Stainless Side Plates 17-4 PH or 17-7 PH Rollers Normal Condition Static Axial Load 17,800 12,300 Analysis Upset Condition N+OBE+SRV Axial Load 24,000 18,900 Analysis Emergency Condition Analyzed As Upset Cond .- Axial Load -- N/A --
Faulted Condition N+SSE+SRV+LOCA Axial Load 52,500 32,200 Analysis Note: The calculated stresses are conservative since they are based on the original fuel assembly type which has the greatest mass of all fuel assemblies in use or in storage at SSES.
Rev. 55 Page 3 of 3
~-~ -- ---~~- --~ - . ---
SSES-FSAR TABLE 3.9-2t HIGH PRESSURE COOLANT INJECTION PUMP Calculated Allowable I
! Location Loading Condition Criteria Stress Stress Material (psi) (psi}
Pressure Boundary Parts Closure Bolting (Main) Erne rgency/F aulted Allowables Based on 15,498 25,000 A-193 GR.B7
- 1. Pressure NormaVUpset Tensile Stress Closure Bolting {Booster) Condition Per ASME 15,878 25.000
- 2. Des_ign Temperature B&PV Code Section Cesing Wall (Main) 3. Seismic for Pressure 12.050 14,000 A-216 GR. WCB
- 6. Nozzle Loads I
Non Pressure Boundary Parts Pump Bolts (Booster) EmergencyIF aulted Allowables Based on 15,870 21,000 A=307 GR.B7 (Tensile) 1. Pressure ASME B&PV Code Su - 60,000 psi Booster Pump Mounting Foot S~dion IV @ 140"F
- 2. Design Temperature Pump Bolts (Main) 19,918 21,000
- 3. Seismic 0.5 Su for Bolts (Tensile)
Main Pump Mounting Foot 4. Nozzle Loads 0.4 Sy for Pins Dowel Pins (Booster) 5. SRV 20,978 33.600 A*193 GR.87 (Shear) Sy =105,000 psi
' (Shear)
Main Pump Taper Pin NOTE: Eight (8) anchor bolts, each carries the stresses for both units mounted on a common base plate.
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR iI TABLE 3.9-2u CONTROL ROD DRIVE (Index Tube)
Allowable Calculated Primary Stress Criteria Loading Type Stress Stress (psi) (psi) (ABS)*
Allowable Primary Membrane Stress plus Bending Stress is based on ASME Boiller and Pressure Vessel Code, Section Ill for Type 316 Stainless Steel@
250CF Sm::: 20000 psi For Normal and Upset Condition: For Normal & Upset Condition: General Membrane 28.500 18,700 11
- 1. Normal loads >
Sa*1owo =Sy for General Membrane 2. Scram General Membrane 42,500 32,700 and Bending a 1.49 for General Membrane 3. OBE and Bending
- 4. SRV (2) (2)
For Emergency Condition: For Emergency Condition : General Membrane
- 1. Normal Loads P> and Bending (2)
- 2. Chugging
- 3. SRV
- 4. Scram For Faulted Condition: For Faulted Condition : General Membrane 56,500 <29,400 1
- 1. Normal Loads < >
S,11cw =0.80 Su for General 2. SSE General Membrane 61,560 <32.700 (3)
Membrane and Bending
- 3. Chugging 4
-= 2.16 S,. for General Membrane 29.400 ( )
and Bending 4 . SRV NOTES:
{1)
Normal loads include pressure. temperature. weight and mechanical loads.
(2)
Less severe than the upset condition.
(3)
Unit 1 (4)
Unit 2 The points of highest stress using the absolute sum (ABS) methodology are given. The ABS approach results in higher values than the square root sum of the squares (SRSS) methodology for corresponding load combination.
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 3.9-2v CONTROL ROD DRIVE HOUSING Allowable Calculated Criteria Loading Primary Stress Type Stress Stress (psi) (psi)
Primary Stress Limit - The Normal and Upset Condition Maximum membrane 16,600 15,710 allowable primary membrane Loads: stress intensity stress is based on the ASME occurs at the tube to
- 1. Design Pressure Boiler and Pressure Vessel tube weld near the Code,Section III, for Class I 2. Stuck Rod Scram Loads center of the housing Vessels, for Type 304 3. Operational Basis for the normal, upset Stainless Steel. Earthquake, with Housing and emergency Lateral Support Installed. conditions.
For Normal and Upset Condition:
Sm = 16,660 psi @ 575°F For Faulted Conditions:
- Faulted Condition Loads: 39,800 16,700 Slimit = 2.4 Sm 1. Design Pressure
= 2.4 x 16,600 2. Stuck Rod Scram Loads
= 39,984 psi
- 3. Design Basis Earthquake, with Housing Lateral Note: Analyzed to Support Installed Emergency Condition Limits.
- 4. Jet Reaction
- 5. Annulus Pressurization.
- Analyzed to Normal upset Condition Limits.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 55 Table 3.9-2w JET PUMPS CRITERIA LOADING STRESS TYPE ALLOWABLE CALCULATED CONDITIONS STRESS (psi) STRESS (psi)
Primary Membrane Plus Bending Stress Based on ASME B&PV Code Section III MATERIAL: TYPE 304SS A. For Service Levels A & B - NORMAL & UPSET CONDITION:
Sm = 16,900 psi @ 1. Deadweight PRIMARY MEMBRANE 25,350 12,700 550oF 2. Pressure PLUS BENDING Pm + Pb 3. SRV Slimit = 1.5 Sm 4. OBE B. For Service Level C - EMERGENCY CONDITION:
Pm + Pb 1. Deadweight PRIMARY MEMBRANE 38,025 16,010 Slimit = 2.25 Sm 2. Pressure PLUS BENDING
- 3. SRV
- 4. SBA C. For Service Level D - FAULTED CONDITION:
- 1. Deadweight PRIMARY MEMBRANE 60,840(1) 38,416 Pm + Pb
- 2. Pressure PLUS BENDING 50,800(2)
Slimit = 3.0 Sm
- 3. Chugging
- 4. SRV
- 5. SSE D. MAXIMUM CUMULATIVE FATIGUE USAGE FACTOR ACCEPTANCE CRITERIA: <1.0:
UNIT 1 PROJECTED CUMULATIVE USAGE FACTOR: = 0.94 UNIT 2 PROJECTED CUMULATIVE USAGE FACTOR: = 0.67 These CUFs account for the original 40-year plant design. These components are presently inspected on a regular basis. These inspections continue during the period of extended operation to ensure components continue to perform their intended functions.
(1) Unit 1 (2) Unit 2 FSAR Rev. 66 Page 1 of 1
SSES-FSAR Table Rev. 37 TABLE 3.9-3 NSSS SEISMIC CATEGORY I ACTIVE PUMPS AND VALVES IDENTIFICATION AS SHOWN ON COMPONENT NAME APPLICABLE FIGURES B 21 F 022 Main Steam Isolation Valves B 21 F 028 Safety Relief Valves B 21 F 013 C 12 F 010/180 (Unit 1)
C 12 F 011/181 (Unit 1)
Control Rod Drive Globe Valves C 12 F 010A/B (Unit 2)
C 12 F 011A/B (Unit 2)
Recirculation System B 31 F 031 (Discharge and Bypass) Gate Valves B 31 F 032 Standby Liquid Control Pump C 41 C 001 Standby Liquid Control Valve C 41 F 004 RHR Pump E 11 C 002 Core Spray Pump and Motor E 21 C 001 RCIC Pump E 51 C 001 RCIC Turbine E 51 C 002 HPCI Pump E 41 C 001 HPCI Turbine E 41 C 002 FSAR Rev. 60 Page 1 of 1
SSES-FSAR TABLE 3.9-4 {Page 1 of 21)
APPLICABLE THERMAL TRANSIENTS (PRE STARTUP LEAK TEST) 130 CYCLES CONDITION - TEST Initial Final Temp. Rate Temp.
Pipeline Temp. °F Temp. °F Time °F/Hour Of Main Steam Line 70 100 30 min 60 30 Head Spray (RHR) 70 100 30 min 60 30 Recirculation Suction 70 100 30 min 60 30 Recirculation Discharge 70 100 30 min 60 30 Bottom Drain 70 100 30 min 60 30 Standby Liquid Control 70 100 30 min 60 30 100 50 Step 10 min 50 Duration 50 100 Step 50 Core Spray 70 100 30 min 60 30 Feedwater 70 100 30 min 60 30 CRDHS Return 70 100 30 min 60 30 Remarks: After temperature is raised to 100°F, reactor pressure is increased to 1250 psig and then decreased to O psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 2 of 21)
APPLICABLE THERMAL TRANSIENTS (STARTUP) 120 CYCLES CONDITION - NORMAL Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 100 551 - 100 451 Head Spray (RHR) 100 551 - 100 451 Recirculation 100 551 - 100 451 Suction 551 543 Step - 8 543 527 Step 16 Recirculation 543 527 Step - 16 Discharge Bottom Drain 543 527 Step - 16 Standby Liquid 543 527 Step - 16 Control Core Spray* 100 406 - 100 306 only occurs 10 times 406 50 Step - 356 50 406 Step - 356 406 551 - 100 145 Feedwater 100 551 - 100 451 551 90 Step - 461 90 420 30 min 660 330 CRDHS Return 100 50 Step - 50 Remarks: Reactor pressure increases from O to 1038 psig at rate of temperature increase.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 3 of 21)
APPLICABLE THERMAL TRANSIENTS (DAILY POWER REDUCTION AND ROD PATTERN CHANGE) 10,400 CYCLES CONDITION - NORMAL Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 551 Head Spray (RHR) 551 551 Recirculation Suction 527 527 Recirculation Discharge 527 527 Bottom Drain 527 527 Standby Liquid Control 527 527 Core Spray 527 527 Feedwater 420 354 15 min 264 66 354 420 15 min 264 66 GROHS Return 50 50 Clean Return 435 435 Remarks: Reactor pressure remains at 1038 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 {Page 4 of 21)
APPLICABLE THERMAL TRANSIENTS
{WEEKLY POWER REDUCTION) 2000 CYCLES CONDITION - NORMAL Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 551 Head Spray {RHR) 551 551 Recirculation Suction 527 527 Recirculation 527 527 Discharge Bottom Drain 527 527 Standby Liquid 527 527 Control Core Spray 527 527 Feedwater 420 324 30 min 192 96 324 420 30 min 192 96 CRDHS Return 50 50 Cleanup Return 435 435 Remarks: Reactor pressure remains at 1038 psig .
Rev. 491 04/96
SSES-FSAR TABLE 3.9-4 (Page 5 of 21)
APPLICABLE THERMAL TRANSIENTS (TURBINE TRIP 100 PERCENT BYPASS) 10 CYCLES CONDITION - UPSET Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 551 Head Spray (RHR) 551 551 Recirculation Suction 527 495 1.5 min 1280 32 495 527 4 min 480 32 Recirculation 495 527 4 min 480 32 Discharge Bottom Drain 495 527 4 min 480 32 Standby Liquid 495 527 4 min 480 32 Control Core Spray 495 527 4 min 480 32 Feedwater 420 100 1.5 min 12,800 320 100 420 4 min 4800 320 CRDHS Return 50 50 Remarks: Reactor pressure remains at 1038 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 6 of 21)
APPLICABLE THERMAL TRANSIENTS (PARTIAL FEEDWATER HEATER BYPASS}
70 CYCLES CONDITION - UPSET Initial Frnal Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 551 Head Spray (RHR) 551 551 Recirculation Suction 527 517 2 min 300 10 517 527 4 min 150 10 Recirculation Discharge 517 527 4 min 150 10 Bottom Drain 517 527 4 min 150 10 Standby Liquid Control 517 527 4min 150 10 Core Spray 517 527 4 min 150 10 Feedwater 420 265 1.5 min 6200 155 265 420 3min 3100 155 CRDHS Return 50 50 Remarks: Reactor pressure remains at 1038 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 1 ot 21)
APPLICABLE THERMAL TRANSIENTS (SCRAM - T/G TRIP FEEDWATER ON - MSIV OPEN) 40 CYCLES CONDITION - UPSET Initial Flnal Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 565 10 sec 5040 14 565 543 15 sec 5280 22 543 400 - 100 143 400 551 - 100 151 Head Spray (RHR) 400 551 - 100 151 Recirculation Suction 527 400 . 100 127 400 551 - 100 151 551 543 Step - 8 543 527 Step - 16 Recirculation Discharge 543 527 Step - 16 Bottom Drain 543 527 Step - 16 Standby Liquid Control 543 527 Step - 16 Core Spray 543 527 Step - 16 Feedwater 420 275 1 min 8700 145 275 100 15 min 700 175 100 250 Step - 150 250 420 30 min 340 170 CRDHS Return 50 50 Remarks: Reactor pressure increases to 1163 psig all relief valves open. Pressure decreases to 240 psig and then increases to 1038 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page a of 21)
APPLICABLE THERMAL TRANSIENTS (ALL OTHER SCRAMS) 140 CYCLES CONDITION - UPSET Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 543 15 sec 1920 8 543 400 - 100 143 400 551 - 100 151 Head Spray {RHR) 400 551 - 100 151 Recirculation Suction 527 400 - 100 127 400 551 . 100 151 551 543 Step - 8 543 527 Step - 16 Recirculation Discharge 543 527 Step - 16 Bottom Drain 543 527 Step - 16 Standby Liquid Control 543 527 Step - 16 Core Spray 543 527 Step - 16 Feedwater 420 275 1 min 8700 145 275 100 15 min 700 175 100 250 Step - 150 250 420 30 min 340 170 CRDHS Return 50 50 Remarks: Reactor pressure decreases to 240 psig and then increases to 1038 psig.
J Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 9 of 21)
APPLICABLE THERMAL TRANSIENTS (RATED POWER)
SEE BELOW CONDmON - NORMAL Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F1Hour Temp.
OF Main Steam Line 551 551 Head Spray (RHR) 551 551 Recirculation Suction 527 527 Recirculation Discharge 527 527 Bottom Drain 527 150 1 hr 377 377
- 240 Times 150 527 Step . 377 Standby Liquid Control 527 60 Step . 467
- 10 Times 60 527 60 min 467 467 Core Spray 527 527 Feedwater 420 420 CRDHS Return 50 50 Cleanup Return 435 435 Remarks: Reactor pressure remains at 1038 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 10 of 21)
APPLICABLE THERMAL TRANSIENTS (SHUT DOWN) 111 CYCLES CONDITION - NORMAL Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 375 - 100 176 375 330 10 min 270 45 330 100 - 100 230 Head Spray (RHR) and 551 375 - 100 176 RHR Return 375 50 Step 15 sec 325 Duration 50 300 Step 250 300 100 - 100 200 Recirculation Suction 527 551 Step - 24 551 375 - 100 176 375 330 10 min 270 45 330 100 - 100 230 Bottom Drain 330 100 - 100 230 Standby Liquid Control 330 100 -
- 100 230 Core Spray 330 100 . 100 230 Recirculation Discharge 527 551 Step - 24 551 375 - 100 176 375 300 Step - 75 300 260 10 min 240 40 260 100 - 100 160 Feedwater 420 265 30 min 310 155
- 5 step changes to 100°F and back 265 420 Step - 155 during cooldown .
420 551 100 131 551 100 . 100 451 CRDHS Return 50 50 Remarks: Reactor pressure decreases from 1038 psig to 0 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 11 of 21)
APPLICABLE THERMAL TRANSIENTS (LOSS OF FEEDWATER PUMPS - MSIV CLOSED) 10 CYCLES CONDITION - EMERGENCY Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 571 3 sec 24,000 20 571 565 10 sec 2,160 6 565 536 3 min 580 29 536 565 73 min 24 29 565 505 7 min 514 60 505 400 - 100 105 400 551 - 100 151 Head Spray {RHR) 400 551 - 100 151 Recirculation Suction 527 300 30 min 454 227 300 551 - 100 251 551 543 Step - 8 543 527 Step - 16 Recirculation Discharge 543 527 Step - 16 Bottom Drain 527 300 3.7 min 3,681 227 300 505 23 min 535 205 505 300 3 min 4,100 205 300 505 73 min 169 205 505 300 7 min 1,757 205 300 551 - 100 251 551 543 Step - 8 543 527 Step - 16 Standby Liquid Control 543 527 Step - 16 Core Spray 543 527 Step 16 Rev. 491 04/96
SSES-FSAR TABLE 3.9-4 (Page 12 of 21)
APPLICABLE THERMAL TRANSIENTS (LOSS OF FEEDWATER PUMPS - MSIV CLOSED) (continued)
Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Feedwater 420 551 Step - 131 551 40 Step - 511 40 551 23 min 1,333 511 551 40 Step - 511 40 551 51 min 601 511 551 40 Step - 511 40 300 5 min 3120 260 300 551 - 100 251 551 100 Step - 451 100 250 Step - 150 250 420 30 min 340 170 CRDHS Return 50 545 10 min 2,970 495 Remarks: Reactor pressure increases to 1218 psig. All relief valves open. Pressure decreases to 1163 psig and relief valves close. RCIC initiates and pressure decreases to 913 psig. RCIC trips off on high level and pressure increases to 1163 psig and one relief valve opens and then closes as pressure decreases at rate of 100°F/hr.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 13 of 21)
APPLICABLE THERMAL TRANSIENTS
{REACTOR OVERPRESSURE DELAYEO SCRAM) 1 CYCLE CONDITION - EMERGENCY Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 583 2 sec 57,600 32 583 543 30 sec 4,800 40 543 400 - 100 143 Head Spray (RHR) 543 400 - 100 143 Recirculation Suction 527 562 11 sec 11,455 35 562 400 - 100 162 Recirculation Discharge 562 400 - 100 162 Bottom Drain 562 400 - 100 162 Standby Liquid Control 562 400 - 100 162 Core Spray 562 400 - 100 162 Feedwater 420 276 1 min 8640 144 276 100 15 min 704 176 100 250 Step - 150 250 420 30 min 340 170 CRDHS Return 100 50 Step - 50 Remarks: Reactor pressure increases to 1350 psig. All relief valves and safety valves open.
Pressure decreases to 240 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 14 of 21)
APPLICABLE THERMAL TRANSIENTS
{SINGLE SAFETY/RELIEF VALVE SLOWDOWN) 8 CYCLES CONDITION - EMERGENCY Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 375 10 min 1,056 176 375 100 ~
100 275 Head Spray (RHR) 375 100 - 100 275 Recirculation Suction 527 375 10 min 912 152 375 100 - 100 275 Recirculation Discharge 375 100 - 100 275 Bottom Drain 275 100 - 100 275 Standby Liquid Control 375 100 - 100 275 Core Spray 375 100 - 100 275 Feedwater 420 276 1 min 8640 144 276 100 15 min 704 176 CRDHS Return 50 50 Remarks: Reactor pressure decreases to 0 psig with one relief valve or safety valve open.
Rev. 49. 04/96
SSES-FSAR TABLE 3.9-4 (Page 1s of 21)
APPLICABLE THERMAL TRANSIENTS (AUTOMATIC DEPRESSURIZATION) 1 CYCLE CONDITION - EMERGENCY Initial Final Temp. Rate Plpellne Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 375 3.3 min 3,200 176 375 281 - 300 94 Head Spray (RHR) 375 281 - 300 94 Recirculation Suction 527 375 3.3 min 2,764 152 375 281 - 300 94 Recirculation Discharge 375 281 - 300 94 Bottom Drain 375 281 - 300 94 Standby Liquid Control 375 281 - 300 94 Core Spray 375 281 - 300 94 Feedwater 420 276 1 min 8640 144 276 100 15 min 704 176 CRDHS Return 50 50 Remarks: Reactor pressure decreases to 35 psig with auto-blowdown relief valves open.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 16 of 21)
APPLICABLE THERMAL TRANSIENTS (IMPROPER START OF COLD RECIRCULATION LOOP) 1 CYCLE CONDITION - EMERGENCY Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time aF/Hour Temp. °F Main Steam Line 551 551 Head Spray (RHR) 551 551 Recirculation Suction 527 130 Step 26 sec 397 Duration 130 527 Step 397 Recirculation Discharge 527 13D Step 34 sec 397 Duration 130 527 Step 397 Bottom Drain 527 527 Standby Liquid Control 527 527 Core Spray 527 268 Step 34 sec 259 Duration 268 527 Step 259 Feedwater 420 420 CRDHS Return 50 50 Remarks: Reactor pressure remains at 1038 psig.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 11 of 21)
APPLICABLE THERMAL TRANSIENTS (SUDDEN PUMP START IN COLD LOOP) 1 CYCLE CONDITION - EMERGENCY Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 551 Head Spray (RHR) 551 551 Recirc*ulation Suction 527 527 Recirculation Discharge 527 130 Step 397 34 second 130 527 Step Duration 397 Bottom Drain 527 350 Step 177 350 527 Step 177 Standby Liquid Control 350 527 Step 177 Core Spray 527 527 Feedwater 420 420 GROHS Return 50 50 Remarks: Reactor pressure remains at 1038 psig.
Rev. 491 04/96
SSES-FSAR TABLE 3.9-4 (Page 18 of 21)
, APPLICABLE THERMAL TRANSIENTS (IMPROPER START WITH RECIRCULATION PUMPS OFF) 1 CYCLE CONDmON - EMERGENCY Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam line 100 551 - 100 451 Head Spray (RHR) 100 551 - 100 451 Recirculation Suction 100 551 . 100 451 Recirculation Discharge 100 551 - 100 451 551 130 Step 34 sec 421 Duration 130 551 Step 421 Bottom Drain 100 551 5 min 5,412 451 Standby Liquid Control 100 551 5 min 5,412 451 Core Spray 100 551 - 100 451 Feedwater 90 551 - 100 461 551 90 Step - 461 90 420 30 min 660 330 GROHS Returns 50 50 Remarks: Reactor pressure increases to 1038 psig as temperature increases.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 19 of 21)
APPUCABLE THERMAL TRANSIENTS (PIPE RUPTURE AND SLOWDOWN) 1 CYCLE CONDITION - FAULTED Initial Final Temp. Rate Pipeline Temp. °F Temp. °F Time °F/Hour Temp. °F Main Steam Line 551 281 15 sec 64,800 270 Head Spray (RHR) 551 281 15 sec 64,800 270 Recirculation Sucfion 527 281 15 sec 59,040 246 Recirculation Discharge 527 281 15 sec 59,040 246 281 223 35 sec 5,966 58 223 50 Step 90 sec 173 Duration 50 130 Step 80 Bottom Drain 527 281 15 sec 59,040 246 281 273 35 sec 822 8 273 50 Step 90 sec 223 Duration 50 130 Step 80 Standby Liquid Control 50 130 Step 80 Core Spray 527 406 10 sec 43,560 121 406 50 Step 90 sec 356 Duration 50 130 Step 80 Feedwater 420 281 15 sec 33,400 139 CRDHS Return 50 50 Remarks: Reactor pressure decreases from 1038 psig to 35 psig in 15 seconds.
Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 20 ot 21)
APPLICABLE THERMAL TRANSIENTS (BECHTEL CRITERIA) 1/2 SSE Cycles ((?perating Basis Earthquake) Condition-Upset Expected number of equivalent 1/2 SSE in life of pipe 5 system Average duration of strong motion vibration 1/2 SSE 15 sec Average number of maximum seismic load cycles of pipe system for each 1/2 SSE 10 Total lifetime number of maximum seismic load cycles of piping system 50 SSE Cycles (Design Basis Earthquake) Condition-Faulted Expected number of equivalent SSE in life of pipe sy~tem 1 Average duration of strong motion vibration SSE 15 sec Rev. 49, 04/96
SSES-FSAR TABLE 3.9-4 (Page 21 of 21)
APPLICABLE THERMAL TRANSIENTS
( GENERAL ELECTRIC CRITERIA)
FOR NSSS PIPING
- 1. 112 SSE Cycles Condition - Upset Expected number of equivalent 1/2 SSE in life 1 of pipe system Average duration of strong motion vibration 30 sec 1/2 SSE Average number of maximum seismic load cycles of pipe system for each 1/2 SSE 10 Total lifetime number of maximum seismic load cycles of piping system 10
- 2. SSE Cycles Condition-Faulted Expected number of equivalent SSE in life of 1 pipe system Average duration of strong motion vibration 30 sec SSE Average number of maximum seismic load cycles of pipe system for each SSE 1 Total lifetime number of maximum seismic load cycles of piping system 1
- 3. Turbine Stop Valve* Closure (TSVC) Condition-Upset 120 cycles
- 4. Relief Valve Lift* (RVL) Condition-Upset 20,100 cycles
- not applicable to recirculation piping due to negligible effect.
Rev. 49, 04/96
Table 3.9-5 LIST OF COMPUTER PROGRAMS USED IN BOP MECHANICAL SYSTEMS AND COMPONENTS Security-Related Information Table Withheld Under 10 CFR 2.390
SSES-FSAR TABLE 3.9-6 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 1,2, AND 3 COMPONENTS (NON-NSSS)
Condition Design Loading Combinations(1)
Design PD Normal PD + DW Upset (a) PO + DW + (OBE2 + SRV x2)1/2 (b) PO + DW + (RVC2 + OBE2)1/2 (c) PO + DW + FV (d) PO + DW + OBE + RVO Emergency (a) PO + DW + (OBE2 + SRV2ADS + SBA2)1/2 (b) PO + DW + (FV2+ OBE2) 1/2 Faulted (a) PO + DW + (OBE2 + SRV2ADS + IBA2)1/2 (b) PO + DW + (SSE2 + SRV2ADS + IBA2)1/2 (c) PO + DW + (SSE2 + DBA2)1/2 NOTE: (1) As required by the appropriate subsection, i.e. NB,NC or ND, or ASME Section III Division I.
Other loads, such as loads from thermal transient, thermal gradients, and the anchor point displacement portion of the OBE or SRV may require consideration in addition to those primary stress producing loads listed.
Legend: PD - Design Pressure PO - Operating Pressure DW - Dead Weight OBE - Operating basis earthquake (inertia portion)
SSE - Safe shutdown earthquake (inertia portion)
SRVx - Loads due to Safety Relief Valve Blow - Axisymmetric or Asymmetric SRVADS - Loads due to Automatic Depressurization SRV Blow-Axisymmetric SBA - Small Break Accident IBA - Intermediate Break Accident DBA - Design Basis Accident FV - Transient response of the piping system associated with fast valve closure. Transients associated with valve closure time less than 5 seconds are considered.
RVC - Transient response of the piping system associated with relief valve opening in a closed system.
RVO - Sustained load or response of the piping system associated with relief valve opening in an open system or last segment of the closed system with steady state load.
SBA, IBA AND DBA include all event induced loads, as applicable, such as chugging, condensation oscillation, pool swell, drag loads, annulus pressurization, etc.
For the NSSS load combination, see Table 3.9-2.
Rev. 47, 06/94 Page 1 of 1
SSES-FSAR
~'l"ABLE 3.9-7 :
.; DESIGN CRITERIA FOR ASME CODE CLASS 1 YAl.VES
~ ...CNON-NSSS}
Condition Str111 Umltt Design NB-3521 111 Normal and Upset NB-3200 or NB-3600111 (Standard Design Rufes)
Emergencyf21 NB-3526"'
Faulted'2 1 NB-3527"'
(1)
As specified by ASME Code Section 1111 1971 thru Winter 1972 Addenda.
C21 Where valve function must be ensured (active valve) during the emergency or faulted conditions, the specified emergency or fauhed condition for the plant 1ha11 be considered the normal condition for the valve.
ell As specified by ASME Code Section Ill, 1971, Winter 1973 Addenda.
Rev. 47, 06/94
SSES-FSAR
~tT
~;
ABLE 3.9-8
.* DEStON CRITERIA FOR ASME CODE Cl.ASS 2 ANO 3 VUSELI DESIGNED TO NC-3300 AND ND-3300 Condition Streaa Um1t1"1C21 Design and Normal The vesse1 shall conform to the requirements of NC-3300 and ND-3300.
Upset, Emergency, The vessel shell conform to the requirements of ASME Code Case 1607-1.
and Faulted (11 As specified by ASME Code Section 111, 1971 thru 1972 Winter Addenda.
(21 For the Diesel Generator E Facility, the design criteria for class 3 vessef s Is governed by ASME code, Section Ill, 1980 edition through eummer 1981 addenda.
Rev. 47, 06/94
SSES-FSAR t TABLE 3.9-9 I
- . DESIGN CRITERIA FOR ASME CODE CLASS 2 VESSELi
{ DESIGNED TO ALTERNATE RULES OF NC-3200 Page 1 of 2 Condition Stress Limits<1K2)
Design and Normal The vessel shafl conform to the requirements of NC-3200.
Upset<3> P. :!. 3 S.,
Pm :s; 1.1 Sm (Pm or Pl)+ Pb $ 1.65 Sm Emergency Pm $ greater of 1.2 Sm or 1.0 S 1 (Pm or PL)+ Pb $ greater of 1.8 s..,
or 1.5 Sw Faulted <4> Pm :9' 2.0 Sm (P"' or PL) + Pb :s: 2.4 Sm (1) Definition of symbols:
p"' :a:: General primary membrane stress tntenslty. This stress Intensity Is derived from the average value across the soTid section under consideration. Excludes
. discontinuities and concentrations. Produced only by pressure and ~ther mechanical loads.
pl. : Local primary membrane stress Intensity. Same as P111 except that discontinuities are considered.
pl> = Primary bending stress intensity. Component of primary stress Intensity proportjonat to distance from centroid of IORd section. Excludes discontinuities and concentrations. Produced only by pressure and other mechanlca1 loads.
P. C Secondary stress intensfty range. Developed by constraint of adjacent parts or by self-constraint of a structure. Considers discontinuities but not concentrations.
Produced by mechanlcat loads and by thermal expansion.
S"' = Design stress Intensity value, Appendix I. Table 1-1.0.
s,, = Yield strength value. Appendix I, Tabte 1-2.0.
(2) These Umits do not take into account eJther local or generaJ buckling that mJght occur In thin..waH vessels. Such buckJlng shan be considered for upset conditions, but need not be considered for emergency or faulted conditions unless required by the design specification.
Rev. 4?, 06/94
SSES-FSAR TABLE 3.9*9 I DESIGN CRITERIA FOR ASME CODE CLASS 2 VESSELS DESIGNED TO ALTERNATE RULES OF NC-3200 Page 2 or2 Condition I Stress Llmfts(1K2)
(3) Fatigue analysis requtrements of NC-3219 and Appendix XIV Bhan also be considered.
(4) As an alternative to satisfying these timltsl the faulted condition stress Omits of Appendix F may be applied provided that a complete analysls In accordance with NC-3211.1(c) Is performed.
Rev. 47, 06/94
SSES-FSAR
.; . tABLE 1.9-10 T>ESION CRITERIA FOR ASME CODE CLASS 2 AND 3 PIPING Condition Stress Llmttsf1)
Design, Normaf, Upset, and The piping shan conform to the requirements of Section 111, Emergency (2} paragraphs NC-3600 and N0-3600.
Faultedf2) The piping shan conform to the requirements of ASME Code Case 1606.
(1) As specffied by ASME Code Section Ill. 1971 through 1972 Winter Addenda. For diesel generator 'E' facility piping on the auxmary and air start skids the applicable code Is ASME Section Ill, 1980 through Summer 1981 Addenda. See Table 3.2*3 for later code ed"rtions used for snubber enmination or other piping modifications.
(2) Functional capability of essential piping wt11 be assured per Rodabaugh Criteria for emergency and faulted conditions only, Ref. 3.9-8.
Rev. 47, 06/94
SSES-FSAR rABLE S.9*11
,*:; DESIGN CRITERIA FOR ASME COO£ CLASS 2 AND 3 PUMPS Condition Stress Llmtt.f2M3>
Design and Normal The pump shall oonform to the requirements of Section Ill, Paragraphs NC-3400 and N0-3400.
Upset, Emergency<1> and Faultecf'>. The pump shan conform to the requirements of ASME Code Case 1636-1.
(1) VVhere pump function must be ensured (active pumps) during the emergency or faulted condition, the pumps nozzle toads due to the specified emergency or faulted plant conditions shall be considered in satisfying the normal con<frtion stress lmlts for the pump.
(2) As specified by ASME Code Section Ill, 1971 and 'Mnter 1972 Addenda.
(3) For pumps mounted on the diesel generator 'E' auxiliary skld, the design criteria Is governed by ASME Code, Section Ill, 1980 editiOn through Summer 1981 Addenda.
Rev. 47, 06/94
SSES-FSAR iTABLE 3.9-12
~ DESIGN CRITERIA FOR ASME CODE CLASS 2 AND I VALVES Condition Stress Llmttaf2M3)
Des;gn and NormaJ The vaJve ahaH conform to the requirements of Section Ill, Paragraphs NC-3500 and N0-3500.
Upset. Emergency"> and Faulted' >1 The va1ve shall conform to the requirements or ASM E Code case 1635--1.
(1) VVhere valve function must be ensured (active valve) during the emergency or faulted condition, the specified emergency or faulted conditions for the plant shan be cons1dered as the normal condition for the valve.
(2) As specffted by ASME Code Section m, 1971 and Vv1nter 1972 Addenda.
(3) For valves supplied on the diesel generator *E* auxmary and air start skids, the design criteria is govemed by the ASME Code. Section Ill. 1980 edition through Summer 1981 Addenda.
Rev. 47, 06/94
SSES-FSAR D£SIQN LOADING COMBINATIONS FOR SUPPORTS FOR ASME CODE CLASS 1, 2, AND 3 COMPONENTS'"
tNON-NSSS)
Condrtlon Design loading ComblnetJona'1' Allowable Stres,*a 1 Hydrostatic Test (a) HTDW 0.8 Sy*
Normal and Upset (a} OW + TH + (OBE 2 + SRV:1l 1/2 (b) OW + TH + (RVC 2 + OBE 2) 1/2 (c) OW + TH + FV (d) DW + TH + OBE + RVO Emergency (a} DW + TH + (OBE 2 + SRV2.-os + SBA 2)1/2 (b) OW + TH + (OBE 2 + FV2 ) 1/2 Faulted (a) OW + TH + (SSE 2 + SRV 2.-os + IBA 2) 1/2 (b) OW + TH + (OBE 2 + SRV2ADS + IBA') 1/2 0.9 Sy (c) DW + TH + (SSE 2 + OBA 2) 1/2
- Snubbers, compensating struts and struts compJy wJth an the requJrements of ASME Section Ill, Subsection NF. (They are not commercially avetlable to the requirements of ANSI 831.1.)
NOTES: (1) Loads due to OBE, SSE, SRV_., SRVADS, SBA, IBA and OBA lnc1ude both Inertia portion end anchor motion portion when response spectre method ts used. The loads from inertia portion and anchor motion are combined by the method of Square Root of the Sum of Squares (SRSS).
- 12) The allowable stress shall be limited to two-thirds of the critical buckling stress.
(3) Supports on the diesel generator E and air start skids comply with all the requirements of ASME, Section Ill, Subsection JF, 1980 edition through Summer 1981.
Legend: HTOW - Piping dead weight due to hydrostatic test TH
- Reaction et the support due to thermal expansion of the pipe Sy
- Yield stress Sh - Allowable stress per ANSt B31.1 See Table 3. 9-6 for additional nomencJature.
See Table 3.9-2 for NSSS support.
Rev
- 4 7 , 0 6 / 9 4
SSES-FSAR TABLE 3.9-16 VALVE QUALIFICATION TEST RANGE Size of au.n.
QualJfic:edon &tencf1 To 1Mte Valve la**
led -~ 1.6 2 3 4' 6 8 10 12 14 16 18 20 22 24 26 28 30 16 Valve X X X 2 X X X 4 X X X 8 X X X X 12 X X X X X
-t4 . : )(
16 X X X X )( X
- 1f X X )(
20 X X X )( X X X X X X X. X X X X X 24 X X X *x X X )(
X )( X X X X X 28 X X X X X X X so )( X X X X )(
36 X X )( X Rev. 47, 06/94
Table 3.9-16 HISTORICAL INFORMATION LISTING OF DYNAMICALLY QUALIFIED EQUIPMENT Security-Related Information Table Withheld Under 10 CFR 2.390
Table 3.9-17 HISTORICAL INFORMATION DIESEL GENERATOR A-D SEISMIC TEST OR ANALYSIS SUBMITTAL CHART Security-Related Information Table Withheld Under 10 CFR 2.390
SSES-FSAR NIMS Rev. 48 Table 3.9-18
SUMMARY
COMPARISON - PROJECT SPECIFICATOIN -10, "GENERAL PROJECT REQUIREMENTS FOR A SEISMIC DESIGN AND ANALYSIS OF CLASS 1 EQUIPMENT AND SUPPORTS"*
SPEC. G-10 IEEE-344-1975 REMARKS 1 . Analysis Equipment is classified as (1) structurally simple and (2) structurally There are two methods of analysis; one approach is based upon Reg. Guide 1.100 Rev. 1 of August t 977 indicates that the static coefficient of complex. Structurally simple equipment is one which can be equivalent static analysis and the other on dynamic analysis. For IEEE-344-1975 of 1.5 is acceptable for verifying structural integrity of frame type adequately represented as a single degree of freedom system or the static coefficient analysis, no detennlnatlon or natural frequency structures such as columns, beams that can be represented by a simple model.
equipment whose fundamental frequency is greater than 33 cps . is made but the response of equipment is assumed to be the peak F01 eQIJipmenl having coofigurations other lhM framo typo structure iustilication Otherwise the equipment is structurally complex. of the RRS at a conservative value of damping multiplied by a static should bo provided for tho uso ol static coefficient.
coefficient of 1.5 to take into account the effects of both multifrequency excitation and mullifrequency response . A lower value of staUc coefficient can be used. if II can be shown to yield conservative results.
For equipment which is structurally simple due to single degree of For the dynamic analysis the equipment shall be modeled to best freedom system the seismic load consists of a static load represent its mass distribution and stiffness characteristics. and this corresponding to the equipment weight times lhe acceleration model is used to detem1ine if the equipment Is rigid or flexible . II selected from the response spectrum cvrve for the natural there Is no response in the frequency range below the high-frequency or the equipment. If the equipment frequency is not frequency asymptote (ZPA) of the RR$, ii is oonsidered rigid . Then known the acceleration shall correspond to the maximum value of the seismic forces on each component of the equipment are the response spectrum. obtaine<I by concentrating its main at Its center of the gravity and multiplying the values of main and the appropriate maximum floor acceleration. If flexible, the model can be analyzed using response spectrum model analysis technique or time history analysis. The response or interest is determined by combining each model response considering all significant modes by SRSS. The absolute sum of similar effects should be considered for closely spaced modes which are those with frequencies differing by 10% or less.
In the analysis the effects of each of the two major horizontal directions and the vertical direction should be considered.
For equipment which is structurally simple due to fundamental frequency greater than 33 cps the seismic load sh.Ill consist of a static toad corresponding to the acceleration at 33 cps selected from appropriate response spectrum curve with an increase of 50% .
- The siesmic design of the deisel gerierator 'E" facility confirms to project specification C-1041 or Cooper Energy S<:r,iccs Standrud No. OS-140 anti IEEE Standard 344-75 in liou of project specification G-10.
Rev. 55 Page 1 of 3
SSES-FSAR NIMS Rev. 48 Table 3.9-18
SUMMARY
COMPARISON - PROJECT SPECIFICATOIN - 10, "GENERAL PROJECT REQUIREMENTS FOR A SEISMIC DESIGN ANO ANALYSIS OF CLASS 1 EQUIPMENT AND SUPPORTS'"'
SPEC. G-10 IEEE-344-1975 REMARKS For equipment which is structurally complex for the analysis purposes, the equipment shall be idealized by a lumped mass model. Frequencies and mode shapes are determined for the vertical. and two orthogonal horizontal directions. Spectral acceleration per mode shall be obtained from appropriate response spectrum curves, the value chosen lo be lhe largest value on the curve when the frequency is varied by +/-0%. The results of the individual modes shall be combined by the square root of sum of the squares method. For closely spaced modes which have frequencies that do not differ by more than 10% the responses of all these modes are combined by sum of the absolute values before combining with other modes by SRSS method .
- 2. Damping The damping values specified are 1/2 % for OBE and 1% f<X SSE The allowable damping values for the equipment are 2% for the Reg. Gulde 1.61 lor equipment also allows da~ing values of 2% tor OBE and 4%
in the response curves attached with the mc1terial requisition and OBE and 3% for SSE. Damping values higher than these cc1n be lor SSE and higtlcr values am permitted in a dynamic seismic analysis for equipment specifications. If there is evidence (such as test results) used if justified by documented test data. If equipment damping is documented test data are provided.
of different damping values these can be used. not known a value of 5% is recommended.
- 3. Testing Testing for both OBE and SSE loads must be done unless it can be Seismic qualifications tests designed to show adequc1cy or shown for a partirular item that the SSE is a more severe condition performance during and fullowlng a SSE must be preceded by one than the OBE. or more OBE test . The number of tests shall be justified for each site or shall oroduce the equivalent effect of 5 OBE 's.
- 4. Testing Perform frequency sweep at line amplitude acceleration input Exploratory test may be run in the form of low level continuous varying frequency (at sweep rate of 1/6 of forcing frequency per sinusoidal sweep (such as 0.2g) at a rate no greater than 2 octaves minute) and determine all resonant frequencies below 33 cps and per minute over the frequency range equal to or greater than lhal to first resonant frequency above 33 cps if below 65 cps. which the equipment is to be aualified.
Rev. 55 Page 2 of 3
SSES-FSAR NIMS Rev. 48 Table 3.9-18
SUMMARY
COMPARISON - PROJECT SPECIFICATOIN - 10, "GENERAL PROJECT REQUIREMENTS FOR A SEISMIC DESIGN AND ANALYSIS OF CLASS 1 EQUIPMENT AND SUPPORTS"*
SPEC. G-10 IEEE-344-1975 REMARKS
- 5. Testing Input acceleration, a, for each of horizonlal and vertical directions at The maximum acceleration of shake table should be at least equal Reg Guide 1.100 indicates that tho uso of factor 1.5 in IEEE-344. ond the conccp1 each reasonant (frequency is determined by, a= 1.5 x 1-(ff/fe)- Sa to ZPA on RRS. For equipment with more than one predominant that TRS need no1 envelope the RAS as a consequence of using 1.5 should not, in where. SA Is spectral response acceleration for equipment frequency the shake motion should provide TRS acceleration at the the absence ol justi1ication, be considered acceptable.
frequency fe. and ff is the forcing frequency of the shaking device test frequency of 1.5 times that of RRS or less if justified. Ttie (ff 0.8 te) . The factor 1.5 is used to account for the possible choice of the preceding factor (with largest values of 1.5) is excitation of other modes. applii;able to broadband RRS. As a consequence the TRS need not envelope RRS provided proper justification is given.
- 6. Testing Duration of excitation of the input acceleration shall not be less than The duration of each test shall be least equal the strong motion The duration of design earthquake for the Susquehanna project is 20 seconds.
30 seconds. oortion of the ori<Jinal time historv used to obtain RRS for the SSE.
Rev.55 Page 3 of 3
l-*-**---- ~ --- *--- -,-* . . - HIST0RICALJ~E0RMATION SSES-FSAR Table Rev. 46 TABLE 3.9-20 BOP PIPING SYSTEMS POWER ASCENSION TESTING Page 1 of 5 Piping System Code(s) SC/HE/ME Temp.>.200°F Thennal E,cpanslon Dynamic Transient Steady State VlbraUon Remarks (1) Test (2) Test (3} Test (4)
Main Steam ASME 111-2 Yes Yes Yes Yes Main Slop Valve Closure and SRV B31.1 Opening Transients sen HE Extraction Steam B31.1 Yes N/A(5) NIA NIA SC II HE Condensate & Refueling B31.1 No NIA NIA NIA Water Storage SCII ME Feedwater ASME 111-1.2 Yes Yes Yes Yes Power Ascension Tesl for Safety Related 831.1 Piping Portion Only HE. SCI SCII Air Removal and Seal B31.1 Yes N/A NIA N/A Steam SCII HE Service Water ASME Ill- 2,3 No NIA NIA N/A A Portion of the System has B31.1 Ternperalum >200°F Bui is Less than SC I & SC II 300°F ME Raw Water Treatmenl 831.1 No N/A NIA N/A SCII ME Lube Oil & Diesel Oil ASME 111-3 No N/A NIA NIA Storage & Tr;insfer 831.1 SC 16' II ME -
Auxiliary Steam B31.1 Yes N/A N/A NIA sen.HE FSAR Rev. 58
c**-* . -----*---*-.. -----***----.. .*--* *- ***- - ... -* H1sfbRic*AL.°fNFORMAff6N- **--*-- -... *-*- -____ . _.
SSES-FSAR Table Rev. 48 TABLE 3.9*20 l
BOP PIPING SYSTEMS POWER ASCENSION TESTING Page 2 of 5 Piping Systom I Code(s) SC/~E/M~l (1)
Temp. > 200°F I Thermal Expansion Test {2) I Dynamic Transient Test {3)
I Stead. y St*a*te Vlbratl011
__ _ Test (4)
Remarks Fire Procection SCll,ME No NfA N/A NIA __ _]
Process Sampling 831.1. No N/A NIA NIA sen. ME Chlorination 831.1, SC II No NIA N/A NIA ME Compressed Air B31.1, SC II No NIA N/A NIA ME 1nslrument Gas ASME Ill -2.3 No NIA N/A NIA 631.1. SC I SCll,ME f~~ P\.lm? T \Jrt:ir-e 'o'3"\.""\ I SC\\ '(~ NJA NIA N}A.
Steam HE Makeup Water 831.1, SC II No N/A N/A NIA ME Valve Steam Leakoff B31.1 SC II Yes NIA N/A N/A HE Acid Injection 831.1, SC II No N/A NIA NIA ME Hydrogen Storage B31.1, SC ll, ME No N/A NIA NIA Diesel Engine Auxiliaries ASME 111-3 Yes Yes(151 N/A N/A Emergency Diesel Exhaust Has 831.1, SC I & II, ME T >300°F and Thermal Expansion Test Pertormed (See Remarks)
MSIV Leak~ge Control ASME Ill -1,2 Yes N/A N/A N/A B31.1, SC I SCll. HE Reactor Recirc 831.1, SC II No N/A N/A NIA Mo\OJIGenerator ME *~ . -- - _ _ , 1 . - , . . * ~ -~~' ~~
f--------- ~--
FSAR Rev. 58
r**-* ~. _-___ *. -... -. ~-- -_-. ---* _ --_ *-** -----*-* .... _-* .._ -* --* -- - .HISTORiCALiNFORMATION. - ,.. _.. **- ___ _
SSES-FSAR Table Rev. 48 TABLE 3.9-20 SOP PIPING SYSTEMS POWER ASCENSION TESTING Page 3 of 5 Piping System Code(s) SC/HE/ME Temp. > 200°F Thonnal Expansion Dynamic Transient Steady State Vibration Remarks 1 Test {2 Test (3 Test (4 High Pressure Coolant ASME 111-1,2 Yes Yes Yes Yes HPCI Turbine Stop Valve Closure Injection 831.1, SC I Transients tor Steam Supply, Steady SC II. HE, ME State Vibration for Stearn Supply and Turbine Exhaust HPC! Pump Suction and discharge lines under steady state vibration.
Reactor Core Isolation ASME llI-1,2 Yes Yes NIA Yes Steady State Vibration for RCIC Steam Cooling B31.1. SC I Supply and Turbine Exhaust RCIC pump SC II, HE, ME suction and discharge piping under steady state vibration test Reactor Water Cleanup ASME 111-1.2 Yes Yes *NJA Yes( 6) Steady State Vibration for RWCU Line B31.1, SC I Inside Containment SC U, HE, ME Residual Heat Removal ASME 111 ~ 1,2,3 Yes Yes NIA Yes<li) Majorily oJ the System has Normal (lndudes Head Spray) B31.1, SC I & II, Operating Temperature less lhan 300°F.
HE.ME Thermal Expansion Tests arc done for SCI Systems WAA T >300°F. Steady State Vibration for Inside Containment Piping and RHR Pump Oischarqe.
(See Remanr.s)
Cleanup Filler ASME 111-2 No N/A NIA NII'.
Oeminer:alizer B31.1, SC II ME Control Rod Drive ASME Ill -2 No NIA N/A Yes(6l CRO insert/Withdrawal pipe.
831.1. SC I, SC II.
ME Standby Liquid Control ASME 111- 1,2 No N/A NIA NIA Only a small portion of the Line near 831.1. SC I & II. RPV Ms Temperature >200<>F.
HE, ME (See Remarks)
OL-~*-
Core Spray ASME 111-1,2 Yes Yes< 11 N/A Yes 161 Sleady State Vibration For Core Spray 831. 1. SC I & II Pump Discharge.
HE.ME . -- . -- --* -...... ~----- - - - - - -~- --* . *-*-----**--- -*--*-
FSAR Rev. 58
r- ___ _,.. '-..- .. -*- . - - .. -*** .. ****--------. . . . .*.*. - - H*1si"ORICAL.iNFORMAfl6N . -- .. . *-* -
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SSES-FSAR .
Table Rev. 48 TABLE 3.9-20 SOP PIPING SYSTEMS POWER ASCE:NSION TESTING Page 4 of 5 Piping System Code(s) SC/HE/ME Temp.> 200°F Thermal Expansion Dynamic Transient Steady State Vibration Remari(;s m Testm Test (3) Test{4l Fuel Pool Cooling. ASME Ill - 3 No NIA NIA NIA Cleanup & Deminera1izer B31.1, SC I SCll,ME OJ ~
-~*.
Containment ASME 111-2 No N/A N/A NIA Atmospheric Control 831.1, SC I & II ME Solid Radwaste and B31.1 No NIA N/A NIA Rad1Naste Solidification SCI!
ME Off-Gas Recombiner ASME 111-3 Yes NIA NIA N/A 831.1, SC I SCII, HE Ambient Temperature 631.1. SC II No NJA NIA NIA Charcoal Off-Gas ME Treatment Chilled Water ASME 111-3 No NIA N/A N/A B31.1, SC I SC II. ME FS/\R Rev. 58
I i ...... .., _ , _ 1 __ __ . _ . .. . HISTORiCAL-INFORMAffbN --
SSES-FSAR Table Rev. 48 TABLE 3. 9-20 BOP PIPING SYSTEMS rowER ASCENSION TESTING Page 5 or 5 Piping System Code(s} SC/HE/ME Temp. > 20Qc>f Thennal Expansion Dynamic Transient Steady State Vibration Remarks 1 Tcst(2 Test (3 Test (4 NOTES: (1} Code(s): ASME 111 Boiler and Pressure Vessel Code. -1, -2, or -3 Denotes Nuclear aass 1. .2 or 3 Piping.
SC lorll Denote Seismic Calegory I or II HE: Denotes High Energy Piping System i.e. Pressure ~275 PSI or Temperature ~200° During Normal Plant Operation ME: Denotes Moderate Energy Piping System *
(2) Thermal Expansion Tests for the indicated systems corresponds to lest de$cription ST-JS, Chapter 14.
(3) Dynamic Transient Tests for the indicated systems corresponds to test description ST-39, Chapter 14.
(4) Steady State Vibration Tests for the indicated systems corresponds to test description ST-40, Chapter 14.
(5) NIA - Denotes Not Applicable and it means test is not performed for lhe reasons given below:
A) For Thennal Expansion Tests: Either the system is not safety-retaled or the normal operating lem perature is less than 300°F .
B} For Dynamic Transient Test : Either the system is not safety-related or the system docs not oxpcrionco any signif1ec1nt transients.
C) For Steady-State Vibration Tests: Either the system is not safety-related or no significant vibration is expected.
(6) Test may be clone during Peroperational Test Program.
(7) Far the effect of RPV exp~nsion oniy. No flow in the Cote Spray iine.
FSAR Rev . 58
SUSQUEHANNA INSIDE STEAM LINE BREAK 8 HIGH POWER
- 1. SHROUD SUPPORT
- 2. CORE PLATE
- 3. UPPER SHROUD 6
1 2 3 TIME AFTER BREAK._.
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FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TRANSIENT PRESSURE DIFFERENTIALS FOLLOWING A STEAMLINE BREAK AT 105% RATED STEAM FLOW 100% RECIRCULATION FLOW FIGURE 3.9-1, Rev. 48 Auto-Cad Figure Fsar 3_9_1.dwg
1800 1600 1400 1
1200 g
I- 1000
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0 0.05 0.10 0.15 0.20 0.25 030 0.35 0.50 TtME Ind FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TYPICAL RELIEF VALVE TRANSIENT FIGURE 3.9-2, Rev. 48 Auto-Cad Figure Fsar 3_9_2.dwg
STEAM DRYER LIFTING LUG VENT AND HEAD SPRAY STEAM DRYER ASSEMBLY STEAM SEPARATOR ASSEMBLY JET PUMP ASSEMBLY CONTROL BLADE FUEL ASSEMBLIES CORE PLATE JET PUMP RECIRCULATION WATER INLET RECIRCULATION WATER OUTLET VESSEL SUPPORT SKIRT SHIELD WALL NOTE THIS FIGURE SHOWS GENERAL LOCATION AND CONFIGURATION CONTROL ROD DRIVES OF MAJOR COMPONENTS ONLY CONTROL ROD DRIVE
- ; - - - HYDRAULIC LINES IN-CORE FLUX MONITO FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REACTOR VESSEL CUTAWAY FIGURE 3.9-3, Rev. 48 Auto-Cad Figure Fsar 3_9_3.dwg
REACTOR CORE SPRAV COOi.iNG
- ARGERS 0
CORE SHROUD NOZZLE DOWNCOMER FLOW SUCTION CHAMBER THROAT (MIXING SECTlONI FROM RECIRCULATION P\JMP FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REACTOR INTERNALS FLOW PATHS FIGURE 3.9-4, Rev. 48 Auto-Cad Figure Fsar 3_9_4.dwg
FUEL ASSEMBL V KRIPHERAL FUEL SUPPORT CNON-AEMOVABLEI ORIFICE GUIDE CORE PLATE ASSEMBLY ORIFICED FUEL SUPPORT CONE ORIFICE SHOWN)
ORIFICE SPRING
,ERIPHERAL FUEL SUPPORT FUEL ASSEMBLY OAIFICED FUEL SUPPORT GUIDE TUBE AND FUEL SuPPORT ALIGNMENT PIN CORE PLATE ORIFICE 11 of 41 COOLANT FLOW s:,...-------- CONTROL Fl:00 GUIDE Tue E CENTER LINE OF FUEL SUPPORT FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FUEL SUPPORT PIECES FIGURE 3.9-5, Rev. 48 Auto-Cad Figure Fsar 3_9_5.dwg
DRIVE NOZZLE SUCTtON INLET THROA1" INLET DIFFUSER OVTLET _ _ ___.
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT JET PUMP FIGURE 3.9-6, Rev. 48 Auto-Cad Figure Fsar 3_9_6.dwg
ITEAJ.t00ME0 ITEAM OR YEA 0
LOWED 'LEIIIUM FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PRESSURE NODES USED FOR DEPRESSURIZATION ANALYSIS FIGURE 3.9-7, Rev. 48 Auto-Cad Figure Fsar 3_9_7.dwg