ML20210B555
| ML20210B555 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/20/1987 |
| From: | Atalay M, Leong D, Wang T CYGNA ENERGY SERVICES |
| To: | |
| Shared Package | |
| ML20210B513 | List: |
| References | |
| DC-1, NUDOCS 8705050355 | |
| Download: ML20210B555 (126) | |
Text
{{#Wiki_filter:I I Job No. : 80023/81060/81061/86064 Doc. No.: DC-1 Revision: 4 April, 1987 SEISMIC REEVALUATION AIO RETROFIT OtITERIA ; For l Yankee Nuclear Power Station Rowe, Massachusetts Prepared for Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Prepared By Cygna Energy Services 2121 N. California B1vd. Suite 390 Walnut Creek, California 94596 Prepared by: 8 D T. V. Wang Date Prepared by: ' I7
, M. . StaYay U '
Dbte Prepared by: N D. K. Leong ( V Uate I l Reviewed by: h d. N % i 2B/77 W.R. Horstman Date Approved by: 4bJL M k/A 53 . P. V4)lenas Date Approved by: Yl)b YJ6f $7 8705050355 870430 PDR ADOCK 05000029
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1 1 TABLE OF CONTENTS Pa ge i
- 1. 0 I N TR O D UC T I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
- 2. 0 SC0PE............................................................... 2 l 3.0 C O D ES A ND S T A ND AR DS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
- 4. 0 R E F ER E NC E DOC UME NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 5.0 L I NE AR PDIF OR MANCE CR ITER I A F OR S TR UCTUR ES A ND EQU I PME NT. . . . . . . . . . . . 11 5.1 Ma t e r i a l Pro p e r t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 5.1.1 Concrete............................................... 11 5.1. 2 Structural Steel....................................... 11 5.1. 3 Masonry................................................ 11 5.1.4 Soils.................................................. 11 5.1.5 Major Mechanical Equipment............................. 12
- 5. 2 Loads Description............................................. 12 5.2.1 Dead Loads............................................. 12 5.2.2 L i ve Lo a d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 5.2.3 Earth Pressure and Groundwater TM)le................... 13 5.2.4 Fluid Loads............................................ 13
- 5. 2. 5 Seismic Loads.......................................... 13
- 5. 2. 6 Thermal L0 ads.......................................... 14
- 5. 2. 7 Pressure loads......................................... 14 5.2.8 Wind Loads............................................. 14 5.3 An a l y s i s Me tho do l o gy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.3.1 An a ly s i s Pro ced u re. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.3.2 Lo c a l S t r e s s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.3.3 Generation of Amplified Response Spectra............... 17 5.4 Acceptance Criteria........................................... 18 5.4.1 Load Cod 3inations...................................... 18 5.4.2 Allowd)le Stresses..................................... 20
- 5. 4. 2.1 Al lowd)l e Stresses for Structures. . . . . . . . . . . . . . 20 5.4.2.2 Allowd)1e Stresses for Major Mechanical Equipment, VC Shell Penetrations, and Tanks........................ 22 5.4.2.3 Acceptance Criteria for Reactor Internals.............................. 25 5.4.3 Al l owda l e De f o rma t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 5.4.4 Da mp i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 Yankee Nuclear Power Station i Seismic Reevaluation Criteria t *b . -iJ6 A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11::"""llll::llll:::::ll11
i l l l TABLE OF CONTENTS (cont'd. ) Page 5.4.5 Alternate Criteria..................................... 27
- 5. 5 Modification of Structures.................................... 27
- 6. 0 S TR UCTUR AL NONL I NE AR PER F ORMANCE CR ITER I A. . . . . . . . . . . . . . . . . . . . . . . . . . . 28 6.1 No n l i n e a r B e h a vi o r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 6.1.1 Lumped Plasticity...................................... 28 6.1. 2 Di st rib uted Pl a sti ci ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 6.1. 3 Stiffness Degradation.................................. 29 6.1. 4 Damping................................................ 29
- 6. 2 Analysis Procedure............................................ 30
- 7. 0 MAS O NR Y WALL PER F OR MA NCE CR ITER I A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
7.1 Purpose and Scope
............................................. 31
- 7. 2 Requi red Fi el d Data for R e vi ew. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
- 7. 3 Analysis and Desi gn Criteri a. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
- 7. 4 Frequency Calculations........................................ 33
7. 5 Wall Inertial Load and Stress Cal culations. . . . . . . . . . . . . . . . . . . . 33
- 7. 6 E q u i pme n t I n e r t i a l Lo a d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 7.7 I n t e r s t o ry D i s p l a c eme n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
- 7. 8 Load Cod)inations............................................. 35 7.9 A l l o w da l e S t r e s s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 7.9.1 YCS S e i s mi c Lo a d i n g. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 7.9.2 N1C S pectrum S ei smi c Loadi n g. . . . . . . . . . . . . . . . . . . . . . . . . . . 37 7.10 Dampi ng Va l ues for Dynami c Analys i s. . . . . . . . . . . . . . . . . . . . . . . . . . . 38 7.11 Materials..................................................... 38 7.12 Stati c and Dynami c Analysi s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 l
1 Yankee Nuclear Power Station ii Seismic Reevaluation Criteria L*I L J L A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11:: lllllll"""""""::::
TABLE OF CONTENTS (cont'd. ) Pa ge 8.0 P I P I NG A N AL YS IS CR I TER I A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 8.1 Load Description.............................................. 51
- 8. 1.1 T h e r ma l Lo a d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 8.1. 2 We i gh t Lo a d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 8.1.3 Pressure load........................................... 51 8.1.4 Seismic Load............................................ 51 8.1. 5 O c c a s i o n a l Lo a d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52
- 8. 2 An a l y s i s Me t h o do l o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52
- 8. 2.1 Geomet ry and Computer Model i n g. . . . . . . . . . . . . . . . . . . . . . . . . . 52 8.2.2 Weight Analysis......................................... 56
- 8. 2. 3 T h e r ma l A n a l y s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 8.2.4 Seismic Analysis........................................ 57
- 8. 2. 5 S eismic Anchor Mo vement Analysis (S AM). . . . . . . . . . . . . . . . . . 59 8.2.6 Pressure Effect......................................... 59 1
8.2.7 Effect Due to Relief Valve Blow-off and Other Occasional Loads.................................. 59
- 8. 3 Acceptance Criteria........................................... 60 8.3.1 Stress Equations........................................ 60 8.3.2 Allowd)le Stresses...................................... 65
- 8. 3. 3 Al l owda l e De f o r ma ti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65
- 8. 4 Small Bore Pipe S tress Analysi s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 8.4.1 Detailed Stress Analysis................................ 65
- 8. 4. 2 Simpl i fi ed S tress Analys i s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66
- 8. 4. 3 F r e qu e n cy Te s t i n g. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 8.5 Buri ed Pipe Stress Analysi s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69
- 8. 6 Main Steam /Feedwater Piping Outside the Vapor Container....... 69 I
- 8. 6.1 Geometry a nd Compu ter Model i n g. . . . . . . . . . . . . . . . . . . . . . . . . . 69 i 8.6.2 Loading Conditions...................................... 71 8.6.3 Acceptance Criteria..................................... 73
- 9. 0 PI PE S U PPOR T DES IGN CR ITER I A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 l
1 9.1 Introduction.................................................. 76 l
- 9. 2 Codes , Standards , and Ref erences. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76
.s 1
Yankee Nuclear Power Station iii Seismic Reevaluation Criteria L'I.LJk A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lill!!!nnnnnnn!!iiritii
, .- , - - , - - - - , - - n .-
TABLE OF CONTENTS (cont'd.) Page
- 9. 3 Loading Description........................................... 76
- 9. 3.1 No r ma l Op e ra ti n g Loa d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77 9.3.2 Emergency / Faulted Loads................................. 77
- 9. 4 Loading Cod >1 nations.......................................... 78 9.5 Frequency..................................................... 78 9.6 Al l ow ab l e S t r e s s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 APPENDIX A Not used.
APPENDIX B Not used. APPENDIX C Fi g. C-1: Structural Analysis Flowchart Fi g. C-2: Large Bore Piping Stress Analysis Flowchart APPENDIX D Table D-1: Material Properties for Structures TM)le D-2: Allowable Loads for Hilti Rolts Te>1e D-3: Recommended Damping Values Te>1e D-4: Material Specifications for Major Mechanical Equipment APPENDIX E Buckling Criteria for Vapor Container Columns APPENDIX F Design Criteria for SSS Building and Buried Piping APPENDIX G Not used. APPENDIX H Not used. APPENDIX 1 Not used. APPENDIX J Not used. APPENDIX K Computer Programs Yankee Nuclear Power Station iv Seismic Reevaluation Criteria L*k k J L A- 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
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1.0 INTRODUCTION
Yankee Nuclear Power Station (YNPS) was designed before the current technology and codes for seismic analysis and design had fully evolved. In the last decade, the state-of-the-art of earthquake engineering has progressed considerably. During this same period, new codes and regulations governing the design of nuclear power plants have been developed and have undergone significant changes. This evolution, while not changing the basic design concepts, has yielded more detailed information concerning the seismic
, behavior of structures, systems and equipment.
Yankee Atomic Electric Company (YAEC) requested Cygna Energy Services to develop this criteria document to be used in completing the seismic reevaluation and retrofit of the plant's critical structures and piping systems in accordance with the PRC Systematic Evaluation Program (SEP). This document estab lishes the evaluation approaches to be used in demonstrating the capacity of the plant to maintain safe shutdown following a postulated seismic event. l l l 1 i i' Yankee Nuclear Pcwer Station 1 Seismic Reevaluation Criteria k J };;m...L 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11.. .. un 0439
l l 2.0 SCOPE The purpose Cf this document is to establish the methodology and criteria for the seismic evaluation of piping, components and structures for the Yankee Nuclear Power Station which are part of the Safe Shutdown System or are otherwise included in the seismic scope. The scope of this program consists of-the following: (a) Perform analyses for thermal, dead weight, anchor movement, pressure, seismic, hydrodynamic and other appropriate loads. These analyses will be based on the modified states of the piping systems and structures where modifications have been proposed and/or irrplemented. (b ) Evaluate the critical piping, components and structures to demonstrate the ability of the plant to achieve hot shutdown. To achieve this goal one of the following procedures is implemented:
- 1) Demonstrate that the response uader Yankee Composite Spectra (YCS, see Section 5.2.5) loads is within code allowables.
- 2) Demonstrate that the response under EC Spectra (EC, see Section 5.2.5) loads is within SEP allowables. Sample evaluations will be used to confirm that seismic response under RC spectrum loading is within SEP allowables or equivalent criteria.
or, on a case-by-case basis:
- 3) Demonstrate that the plant can achieve hot shutdown following a
- seismic event consistent with MC spectra levels.
(c) Design retrofits for critical piping systems and structures to meet the loading conditions specified herein. Yankee Nuclear Power Station 2 Seismic Reevaluation Criteria MJ6 A 11::"""........;;;lll111ll1 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 0439
3.0 CODES AM) STANDARDS The following codes and standards apply to the appropriate sections of this document (except where noted otherwise). (a) American National Standard Code for Pressure Piping, ANSI B31.1,1977. (b ) Nuclear Regulatory Guides: 1. 60 R e v. 1, 1.61 Rev. O, 1.92 Rev. I and 1.122 Re v. 1. (c) American Institute of Steel Construction (AISC), " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," 8th Edition, (d) Americin Concrete Institute (ACI), " Building Code Requirements for Rein-forced Co crete" (ACI 318-77), including 1977 commentary. (e) Not used. l (f) Not used. l (g) American Society of Mechanical Engineers (ASME), " Boiler and Pressure Vessel Code," 1977 Edition. (h) U. S. Nuclear Regulatory Commission (RC), " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," NUREG/CR-0098, May 1978. (1) Not used l Yankee Nuclear Power Station 3 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 n',.....b....d...'.....,t.m 0439
(j) U.S. NJClear Regulatory Commission (NRC), " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, Section 3.7, Washington, DC, Office of Nuclear Reactor Regulation, July, 1981. (k) American Concrete Institute (ACI), " Code Requirements for Nuclear Safety Related Concrete Structures" (ACI 349-76), including supplements. (1) Not used. (m) American Society of Mechanical Engineers (ASME), " Boiler and Pressure Vessel Code," 1983 Edition.
- (n) Not used.
(o) American Concrete Institute (ACI), " Building Code Requirements for Concrete Masonry Structures," (ACI Committee 531-79). (p) Not used. (q) Not used. (r) Not used. (s) Not used. (t) Not used. (u) Not used. l (v) Not used. Yankee Nuclear Power Station 4 Seismic Reevaluation Criteria L*k l J k & 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11llll1811lll1111111llllllllll 0439
(w) " Reevaluation Guidelines for SEP Group 11 Plants (Excluding Structures) (Revision)," United States Nuclear Regulatory Commission Docket No. 50-29, Septenb er, 1982. (x) Not used. (y) Not used. (z) U.S. Nuclear Regulatory Commission (RC), Standard Review Plan, Appendix A to SRP Section 3.8.4, " Interim Criteria for Safety Related Masonry Wall E valuation". (aa) Not used. (a5 ) Not used. (ac) ANSI A58.1-1982, " Minimum Design Loads for Buildings and Other Structures". (ad) " Guidelines for SEP Soil-Structure Interaction Reviews", USNRC Letter LS 05-12-035, Decenb er 15, 1980. Yankee Nuclear Pcwer Station 5 Seismic Reevaluation Criteria 8 023/8 60/81 61/86 64 D c. No. C-1; Ro. 4 tii2nkinintinnitu 0439
4.0 REFERENCE DOCUMENTS The following reference documents are used in carrying out the piping stress and structural analysis efforts: (a) Yankee Atomic Electric Company, " Final Safety Analysis Report, Yankee Nuclear Power Station, Rowe, Massachusetts." (b ) Specification for Piping. YS-497 (S & W, J0-9699) July 15,1959.' Yankee Atomic Electric Company, Yankee Nuclear Power Station, Rowe, Massachusetts. d (c) Hot Service Thermal Insulation for Yankee Atomic Electric Plant. YS - 2304 (S & W, J0-9699) June 1, 1959. Yankee Nuclear Power Station, Rowe, Massachusetts. (d) " Integrated Plant Safety Assessment - SEP, Yankee Nuclear Power Station", Final Report, U.S. Regulatory Commission, NUR EG-0825, June 1983. (e) Drawings in the Plant Drawing System. (f) fRC Safety Evaluation Report for SEP Topic III-6 (NYR 83-21) Fe ruary 1,1983. 4 (g) " Fire Water Tank Seismic Analysis for Yankee Nuclear Power Station", YAEC Report No.1492, Septenber,1985. (h) " Yankee Rowe Valve Acceleration Allowables", Project Memorandum, Job No. 80023, Cygna Energy Services, Decenber 16, 1981. l l 1 I Yankee Nuclear Power Station 6 l Seismic Reevaluation Criteria j kN b ik A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 ! lilillilillllllHilllllmilli ! 0439 )
i (1) Weston Geophysical Corporation, " Geology and Seismology, Yankee Rowe Nuclear Power Plant," January 29, 1979. (j) Wiegel, R. L., Earthquake Engineering, Prentice-Hall, Inc. (Englewood Cliffs,N.J.),1970. (k) Housner, G. W., " Dynamic Pressures on Accelerated Fluid Containers," Bulletin, Seismic Society of America, 47(1), January,1967. (1) U.S. Atomic Energy Comission, Nuclear Reactors and Earthquakes, TID-7024, Washington, DC, Office of Technical Services,1963. (m) Kennedy, R . P. et al., "Recomended Revision to Nuclear Regulatory Commission Seismic Design Criteria", NUREG/CR-1161. (n) Letter from R.E. Vaughn (Soil & Material Testing, Inc. ) to D. Grimes (YAEC), SMT #1030, Septenber 14, 1984 1 (o) Letter from R.E. Vaughn (Soil & Material Testing, Inc. ) to D. Grimes 1 (YAEC), SMT #1036, August 20, 1986. (p) Not used. (q) Newmark, N. H. and E. Roserb lueth, Fundamentals of Earthquake Engineering, Prentice-Hall, Inc., 1971. (r) Not used. (s) Not used. (t) Not used. Tankee Nuclear Power Station 7 Seismic Reevaluation Criteria i kinndinisinntin 8 23/8 6 '82 62'86 64 c"- c-""" 4 1 0439 I i l
(u) Takeda, T., "S tudy of Load-Deflection Characteristics of Reinforced Concrete Beams Stbjected to Alternating Loads," Transactions, Archi-tectural Institute of Japan, Vol. 76, Septenber 1962. , (v) Not used. (w) Roessett, J.M., "The Use of Simple Models in Soil Structure
- Interaction", Civil Engineering and Nuclear Power, Vol. II, Septenber 15-17, 1980, Knoxville, Tennessee.
(x) Veletsos, A.S. and Wei, T. Y. , " Lateral and Rocking Vib rations of Footings", Journal of Soil Mech. and Found. Di v. , ASCE, Vol. 97, Septenber 1971.
- (y) Weston Geophysical Corporation, " Site Dependent Response Spectra, Yankee Rowe," February 1980.
(z) Timoshenko, S. P., and J. M. Gere, Theory of Elastic Sta5111ty, 2nd Ed., McGraw-Hill Book Company, 1961. (aa) Not used. (ab) Not used. (ac) Not used. (ad) Hilti, Architects and Engineers Anchor and Fastener Design Manual, File No. H2189-51, Report No. 873ER. (ae) Bijlaard, P.P., "(1) Stresses in a Spherical Vessel from Radial Loads i Acting on a Pipe; (2) Stresses in a Spherical Vessel from External Moments Acting on a Pipe; and (3) Influence of a Reinforcing Pad on the Yankee Nuclear Power Station 8 Seismic Reevaluation Criteria dijk A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 1111111111lll18ll181111111llll 0439
Stresses in a Spherical Vessel under Local Loading", Welding Research Council Bulletin 49, April 1959. (af) Wichman, K. R., et al, " Local Stresses in Spherical Shells Due to External Loads", Welding Research Council Bulletin 107, March,1979. (ag) Mershon, J.L., et al, " Local Stresses in Cylindrical Shells due to External Loading on Nozzles - Supplement to WRC Bulletin No. 107", Welding Research Council Bulletin No. 297, August,1984 (ah) YAEC Letter to USMC, FYR 86-049, " Ultimate Soil Bearing Capacity at Yankee Nuclear Power Station", May 5,1986. (ai) " Seismic Evaluation of the Yankee Nuclear Power Station Reactor Internals", Impell Corporation, Report No. 02-0570-1204, August, 1984 j (aj) " Yankee Rowe Fuel Bundle Assenbly (Proof of Fabrication) Structural and Fretting Corrosion Tests", Exxon Nuclear Company Report No. XN-76-28, ! Decenber 1976. (ak) YAEC letter to USmC, FYR 81-58, dated April 10, 1981. (al) YAEC letter to USMC, FYR 81-161 dated Decenber 14, 1981. (am) Letter from D.G. Eisenhut (mC) to J. A. Kay (YAEC), dated August 4, 1980. (an) Letter from D.M. Crutchfield (RC) to all SEP Owners (except San Onofre 1), dated June 17, 1981. i l 1 Yankee Nuclear Power Station 9 Seismic Reevaluation Criteria L L k A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 161146111111111111111111111111 0439 _ - - - - _ - - - - - _ - - - _ - - _ - _ _ - - - - ----------- -- _ _ ---- -- ---- --------- - - -_--- _ ----- _ - _l
(ao) Not used. l (ap) ITT Grinnell Catalog PH81, i
~
(aq) YAEC Letter MAG 6/83, D.R . LeFrancois to Sam Swan of EQE, dated January 6, 1983. ] . I 1 I Yankee Nuclear Power Station 10 Seismic Reevaluation Criteria LJ6 A L 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l lilillitilliliittelitilllillli l 0439 l l 1
5.0 LIEAR PERFWMANCE CRITERIA FM STRUCTLRES Am EQUIPMENT This section describes the criteria to be used in the analysis, evaluation and design of modifications of the structures and equipment which are part of the Safe Shutdown System or are otherwise involved in the seimic scope. 5.1 Material Properties The following material specifications govern. Field test data may be used stbject to case-by-case review. 5.1.1 Concrete See Table 0-1 of Appendix 0. These values are obtained from Reference 4(e). 5.1.2 Structural Steel See Table 0-1 of Appendix D. 5.1.3 Masonry See Section 7.11. 5.1.4 Soils The design bearing capacity for all footings is 8 ksf. If seismic or wind loads are included in the evaluation, this bearing capacity can be increased to 10.6 ksf. These capacities were used by Stone & Wester in the original design of YNPS structures (see Stone & Webster Drawing Nos. 1 9699-FC-1A, 9699-RC-40A, 9699-FC-59A, and 9699-FC-62A) and they are still I applicable to all structures except the Reactor Support Structure (RSS) and Vapor Container (VC). Yankee Nuclear Power Station 11 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 (n.m. ..? ' ,,j., 0439
For the RSS and VC footings which are founded on undisturbed den:ie glacial till, the allowable bearing capacity when stbjected to dynamic loads is 20 ksf [Ref. 4(ah)]. 5.1.5 Major Mechanical Equipment See Table 04 of Appendix 0, 5.2 Loads Description 1 5.2.1 Dead Loads Dead loads and their related internal moments and forces, including fixed equipment loads, will be included in the analysis. Equipment weights up to 500 lbs. will be considered as distributed loads; equipment weights more than 500 lbs. will be applied as concentrated loads. 5.2.2 Live Loads Live loads and their related internal moments and forces, including snow loads and any moveable equipment loads, will be included in the analysis. I Yankee Nuclear Power Station 12 Seismic Reevaluation Criteria 8*23'8 "'8 5 2'8"" oc "o oc-t i "" 4 tiiiiiiiihiiiiiiiiiinim 0439
- 5. 2. 3 Earth Pressure and Groundwater Table Loads due to earth pressure will be included in the analysis. Lateral soil load factors to be used in the analysis shall reflect the physical properties of the soil as described in Tab le 4 of Reference 4(y).
Hydrostatic loads due to groundwater table will be included in the analysis.
- 5. 2. 4 Fluid Loads Fluid loads will be treated as hydrostatic except under seismic condi-tions. For seismic conditions, fluid loads may be computed using the Housner method [ Refs. 4(j), 4(k), 4(1) and 4(q)] or as recommended by NUREG/CR-1161 [Ref. 4(m)]. Alternatively, fluids may be modeled by the fluid elements of ANSYS computer code [ Appendices K(8) and K(9)].
5.2.5 Seismic loads There are two site specific spectra which have been developed for the Yankee Nuclear Power Station. The WC staff and its consultant, the Lawrence Livermore National Laboratory, recommended the EC spectra as documented in Refs. 4(am) and 4(an). The licensee has proposed the Yankee Composite Spectra as documented in Refs. 4(ak) and 4(al). All critical structures and equipment shall be evaluated to demonstrate the ability of the plant to achieve hot shutdown (see Ref. [4d]). This will be achieved using one of the following procedures:
- 1) Demonstrate that the response under YCS loads is within code allowab les. Sample evaluations will be used to confirm that seismic response under EC spectrum loading is within SEP allowables or equivalent criteria.
Yankee Nuclear Power Station 13 Seismic Reevaluation Criteria MJk A 80023/81060/81061/86064 Doc. No. OC-1; Rev. 4 lillittillillimilitilllllill 0439
- 2) Demonstrate that the response under EC loads is within SEP allowables as defined in this document, or, on a case-by-case basis:
- 3) Demonstrate that the plant can achieve hot shutdown following a seismic event consistent with EC spectra levels. This option is elaborated upon in Section 6.
5.2.6 Thermal loads Thermal loads will be considered for extreme environmental and operating conditions. The stress-free temperature is assumed to be 50' F. The coolest outside temperature considered is -20*F and the warmest outside temperature is +120'F, which will both produce a differential temperature of 70' F for the extreme environmental condition. Under the design accident condition, the temperature inside the Vapor Container (VC) is assumed to be 249' F. In this case, the differential temperature is 199' F.
- 5. 2. 7 pressure loads The maximum design pressure will be considered for all pressure vessels, pumps and val ves. Under the design accident condition, the pressure inside the VC is assumed to be 31.5 psi.
5.2.8 Wind Loads See SEP Topic 111-2. Yankee Nuclear Power Station 14 Seismic Reevaluation Criteria
- k. bk A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11111611111lll111lll11111llll1 0439
- 5. 3 Analysis Methodology 5.3.1 Analysis Procedure Figure C-1 shows the general structural analysis steps.
The basic linear elastic analysis technique will be the response spectrum method of dynamic analysis. Soil-structure interaction will be considered on a case-by-case basis. Based on information ob tained from Refs. 4(y) and 4(ah) the plant structures are founded on very dense glacial till. The h situ seismic velocity values range from 6,700 to 7,000 ft./sec. and 1,700 to 2,200 ft./sec. for compressional and shear waves, respectively. The average values of the Young's Modulus of Elasticity and Poisson's ratio are 3 x 5 10 psi and 0.46, respectively. The spring constants of soil used in studying soil-structure interaction shall be calculated using Eq. (3.58) recommended by Newmark & Roserblueth [Ref. 4(q)]. The dashpot constant shall be calculated using Eqs. (3.59) or (3.60) of Ref. 4(q) depending on whether the virtual soil mass is included or not. Alternatively, the spring constant and dashpot constant shall be calculated using Eqs. (21) and (22) recommended by Roessett [Ref. 4(w)] and Veletsos and Wei [Ref. 4(x)]. Adjustment to the above values due to other parameters (such as depth of bedrock) will be considered on a case by-case basis. The inter-connected buildings will be studied case by-case. If buildings are coupled, they will be considered appropriately connected and analyzed as one unit. With a few possible exceptions (where the floor slabs have stbstantial openings) the floor diaphragms will be treated as rigid in plane. Three-dimensional beam elements will be used to describe columns and other beam-type components. The models will describe the stiffness Yankee Nuclear Power Station 15 Seismic Reevaluation Criteria LN b J L & 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 1111111:: .:l 0439
and mass relationship in three-dimensional space. Torsional effects due to asymmetric characteristics are automatically considered in this pro-cedure. In buildings where the potential for further accidental torsion is considered likely, accidental torsion considerations as per NUREG/CR-0098 [Ref. 3(h)] will be included. l The vapor container will be modeled using shell elements to represent the l sphere, and beam elements to represent the beams and columns. In devel- l oping the spherical model, care will be taken to generate a finer mesh at the locations of high stress. The thin shell elements will be modeled to account for the actual thickness of the plate used in the structure which ranges from 7/8 to 3 inches. Additional masses will be applied to selected node points in the models to account for concentrated masses such as hatches and platforms. If the dynamic response of a structure is determined using the response spectrum method, the resulting stresses and displacements will be ob tained from the modal responses using the SRSS method, except for closely spaced modes where Regulatory Guide 1.92 [R ef. 3(b)] will be used. The piping b etween structures will be reviewed for these displacements. The stresses and displacements will be evaluated for their compliance with Section 5.4 5.3.2 Local Stresses The local stresses developed in the VC pipe penetrations and equipment nozzles will be calculated according to WRC Bulletin 107 [Ref. 4(af)]. l Yankee Nuclear Power Station 16 Seismic Reevaluation Criteria Li ( J k A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11lllllllll11llllll11!!1lll111 0439
1 The method proposed in WRC Bulletin 107 has some limitations on its -l applicab ility. The limitations are summarized in Appendix A of the l Bulletin and are based on the nozzle and vessel dimensions, location of l 1 the nozzle on the vessel, etc. These limitations will be considered in I any application of the method given in the Bulletin. If any limitations are violated a finite element analysis or another acceptable method may I be used to perform local stress analysis. i
- 5. 3. 3 Generation of Amplified Response Spectra The amplified response spectra (ARS) will be generated for the analyses of piping systems and equipment. The locations of the ARS will b e selected to cover all the piping systems and the equipment that need to be seismically qualified. The ARS will be generated along the vertical and two orthogonal horizontal directions.
For each of the three orthogonal directions, a statistically independent synthetic ground acceleration time history will be generated using the computer program RCQUAKE [ Appendix K(6)]. These time histories shall be verified to match the YCS or RC spectra according to the criteria specified in USMC Standard Review Plan Section 3.7.1 [Ref. 3(j)]. For the vertical acceleration time history, a scale factor of two-thirds shall be applied. Acceleration response time histories at chosen locations will be calculated using a linear structural time history analysis. These acceleration time histories will then be input to the computer program INSPEC or SPECTRA [ Appendices K(3) and K(18)] to generate ARS. The peaks of the ARS will be broadened by 15% in accordance with USMC Regulatory Guide 1.122 [Ref. 3(b )]. If a nonlinear analysis is performed to calculate structural response in accordance with Section 6.0, acceleration response time histories will be available at locations of interest. These response time histories will then be input to the computer program INSPEC or SPECTRA to generate ARS. Yankee Nuclear Power Station 17 Seismic Reevaluation Criteria N !L A- 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11llltll1lllllll111111111llll1 0439
- 5. 4 Acceptance Criteria 5.4.1 Load Cod)inations The analyses will be performed assuming that the seismic ' event is initi-ated with the plant at normal full power condition. The following load cod >inations will be considered in evaluating the structures and performing structural design of modifications.
U=D+Ro+Po + (E or W or M) (5.1) or U=D+L+To+Ro + Pg (5.2) where: U = Total load to be resisted. D = Dead loads or their related internal moments and forces, including any permanent equipment loads, hydrostatic loads, and lateral soil pressures. It also includes operating static and dynamic heads and fluid flow effects. L = Live loads or their related internal moments and forces, including snow loads, any mo ved>1e equipment loads and crane loads. For equipment supports, it also includes loads due to Wbration and any support movement effects. Note that crane loads and wind loads are not simultaneous. T = Anb ient thermal loads or thermal loads during startup, o normal operating or shutdown conditions, based on the most critical transient or steady-state condition. i ! Yankee Nuclear Power Station 18 l , Seismic Reevaluation Criteria L Jk'A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11lll11lllllll1111!!ll11llll11 0439
)
l R n
= Pipe reactions during the startup, normal operating or shutdown conditions, based on the most critical transient or steady-state condition. The pipe reactions include all piping loads (including seismic and thermal anchor movement effects) transmitted to the component during the Site Specific Earthquake.
P o
= PrMsure equivalent static load within or across a com-partment generated by normal operating or shutdown condi-tions, based on the most critical transient or steady-state condition.
E = Loads generated by the Site Specific Earthquakes defined in Section 5.2.5 including sloshing effects, if appli-cab l e. Three earthquake directions will be considered as per NUR EG/CR-0098 [R ef. 3(h)] except for axisymmetric structures where only one horizontal and the vertical direction earthquakes need be considered. W = Wind load, seep SEP Topic III-2. (Use greater value of W, E or M only.) M = Loads associated with missiles other than those tornado-generated. For major mechanical equipment the load corrb ination is gi ven in Eq. (5.1). Yankee Nuclear Power Station 19
- Seismic Reevaluation Criteria k ID' A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 llllllllllll11i1i11ll11lllll!I 0439
1 1 5.4.2 Allowable Stresses 5.4.2.1 Allowable Stresses for Structures This section applies specifically to linear elastic analyses of structures stbjected to seismic loads. For evaluation of the VC shell away from penetrations, see Ref. 4(f). When demonstrating the capacity of the plant to maintain a hot shutdown condition using the YCS to code option (Section 5.2.5), the following allowable menber stresses shall apply: a) The strength of reinforced concrete menb ers, including strength reduction factors, will be per ACI Code [R ef. 3(d)]. In lieu of the code load factors, the factors shown
. in Section 5.4.1 will be used, b) The menber stresses for steel structures will be evaluated against Part 1 the of AISC specifications [Ref. 3(c)].
c) Brace connections will be evaluated per Part 1 of the AISC specifications [R ef. 3(c)]. In analyzing fastener groups with eccentric loads, either the elastic or ultimate strength method (considering that fasteners rotate ab out an instantaneous center) as provided in Part 4 of AISC Manual [Ref. 3(c)] will be used. d) The allowable loads for Hilti Kwik-bolts will be per Table D-2 of Appendix D. For other types of concrete expansion anchor bolts manufacturer catalogs will be used. l Yankee Nuclear Power Station 20 Seismic Reevaluation Criteria MJL A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11lllllll1ll1llll1111111h1111 0439
,, r
When demonstrating the capacity of the plant to maintain a hot shutdown condition using the IRC to SEP allowables option (Section 5.2. 5), the following allowable menber stresses shall apply: a) The strength of reinforced concrete menbers, including the strength reduction factors, will be per ACI code [R ef. 3(d)]. In lieu of the code factors, the factors shown in Section 5.4.1 will be used. b) The allowable tensile and b ending stresses for steel structures will be increased to 0.95 Fy. Axial compressive stresses of up to 0.95 of critical buckling will be allowed for columns and braces. The critical buckling stress is calculated as follows (Appendix E.1): F cr
=
for (K4/r) > Cc [or F ep < 0. 5 Fy ] (Kt/r)2 (K1/r)2
. F cr = (1- )F yfor K1/r < C c 2C 2 C
2n 2E where: Cc" ( )/ p y c) Brace connections will be evaluated per Part 1 of the AISC Specifications [Ref. 3(c)]. d) The allowable loads for Hilti Kwik bolts will be per Table D-2 of Appendix D. For other types of concrete expansion anchor bolts manufacturer catalogs will be used. 1 Yankee Nuclear Power Station 21 Li ( k A Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lilillilllillifilililll!Illll! 0439
5.4.2.2 Allowable Stresses for Major Mechanical Equipment, VC l Shell Penetrations, and Tanks i i The allowable stresses for the equipment and VC shell penetrations will be per ASME Code [Ref. 3(g)] and RC Guidelines [Ref. 3(w)]. For the VC shell penetrations, Class 1 pressure vessels, heat exchangers, inacti ve valves, and inactive pumps, the acceptance criteria per EC Guideline [Ref. 3(w)] are as follows: Pm < Lesser of 2.4 S ,or 0.7 S u and (P,or P g) + Pb < Lesser of 3.6 S, or 1.05 S u where: P, = General Primary Stress Intensity. This stress intensity is deri ved from the avera ge value across the thickness of a section of the general primary stresses produced by design internal pressure and other specified design mechanical loads, but excluding all secondary and peak stresses. Averaging is to be applied to the stress components prior to determination of the stress intensity values. Pg = Local Menbrane Stress Intensity. This stress intensity is the same as P m except that it includes the effects of discontinuities.
% = Primary Bending Stress Intensity. This stress intensity is deri ved from the linear varying portion of stresses across the solid section l
l Yankee Nuclear Power Station 22 Seismic Reevaluation Criteria ) NI J ' A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 Illittillt!IlllihlIllitlhil l 0439 '
under consideration produced by design pressure and other specified design mechanical loads. Secondary and peak stresses are not included. S m
= Allowable stress intensity at temperature listed in ASME Code, f
S u
= Ultimate tensile strength at temperature listed in ASME Code, Note that for the VC shell penetrations, MC Guideline Equations [Ref. 3(w)] are equivalent to ASME Code Section III, Appendix F, Paragraph F-1331.1 requirements when one considers that the load conbination used is the one given in Eq. (5.3).
For Class 2 inacti ve val ves, the acceptance criteria per EC guidelines are as follows: e, < 2.0 S and ( o, o r og ) + g < 2. 4 S where om = General menbrane stress. This stress is equal to the average stress across the solid section under consideration, excludes discontinuities and concentrations, and is produced only by mechanical loads. g = Bending stress. This stress is equal to the linear varying portion of the stress across the solid section under consideration, excludes Yankee Nuclear Power Station 23 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
, ,,, ,,,,{,y,g,,j,,,
0439 l l
discontinuities and concentrations, and is produced only by ' mechanical loads. og = Local menbrane stress. This stress is the same as og except that it includes the effect of discontinuities. S = ASME Code Class 2 allowable stress value. The allowable stress shall correspond to the metal temperature at the section under consideration. For Class 1 acti ve val ves the acceptance criteria per RC guidelines are as follows: Pm < Greater of 1.25, or Sy (Pm or P g) + % < Greater of 1.8Smor 1.5S y where S y = Yield strength at temperature as listed in the ASME i Code. For Class 2 acti ve val ves, the acceptance criteria per EC guidelines are as follows: j o, < 1. 5S ( o, o r og ) + g < 1. 85. When seismic stresses are below 20% of allowables, ASME Code paragraph NC-3521 can be used to qualify body sections of both Class 1 and Class 2 valves. I l l Yankee Nuclear Power Station 24 l Seismic Reevaluation Criteria l 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 114.....I ...NJ...111 0439 I l
For Class 1 and Class 2 mechanical equipment (including active and inactive valves), bolt stresses shall be limited to: Tension < Lesser of 1.0Sy or 0.70 S u Shear < Lesser of 0.6Sy or 0.42 Su' ~ For the fire water tank, the acceptance criteria will be as described in Ref. 4(g) which uses the AISC Specifications [Ref. (3c)] for the tank shell, roof, floor, and anchor bolt stresses, and ACI-349 [Ref. (3k)] for the foundation and anchor bolt pull-out capacity. Buckling of the fire water tank shell shall be evaluated per ASME Code Section III, Division 1, Stbsection NC. For tanks supported on building floors, the appropriate amplified response spectra (ARS, see Section 5.3.3) generated for the floor will be used. The load conb ination will be per Eq. (5.1). Allowable stresses for the tank shell and support will be per the AISC specifications. Tank shell buckling will be evaluated apply-ing ASME Code Section III, including Code Case N-284 For tank nozzles, WRC Bulletin 297 may be used on a case by-case basis. 5.4.2.3 Acceptance Criteria for Reactor Internals The acceptance criteria for reactor internals are presented in Ref. 4(ai). The criteria are intended to insure that the structural integrity of the fuel, control rods, and internals is maintained and that control rod insertab ility is not impaired after the Site Specific Earthquake. The acceptance criteria for the core support and other stainless steel structures is based on the ASME Boiler and Pressure Vessel Code, Section III, Appendix F. For faulted conditions, the code limits the maximum menb rane stresses to 2.4Sm. For normal Yankee Nuclear Power Station 25 Seismic Reevaluation Criteria 1 fk lillbil... .
....n 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l 0439 1
_-l
operating conditions, menb rane stresses are limited to 5,. Conservatively assuming that normal stresses are at their limit, seismic stresses may not exceed the difference between the faulted and normal allowables. Fuel and control rods are not covered by the ASME Code. The acceptance criteria for fuel assenblies is based on test data (Ref. 4(aj)) which shows that a fuel assenbly can withstand a lateral deflection of 0.6 inches with no permanent deformation and an axial rod force of 9550 pounds without buckling. The control rod acceptance criteria is based on the yield strength of the material. Normal operating stresses are judged to be insignificant. Control rod insertability after a seismic event is assured if there are no permanent deformations of control rods or components which may interfere with control rod insertion (fuel, guide ttbes,
. shroud ttbes).
- 5. 4. 3 Allowable Deformation Deformations will be limited to existing clearances to prevent impact of adjacent structures.
5.4.4 Damping l The damping in structures is a function of the type of material, type of construction, and the state of stress produced by the excitation. Table D-3 in Appendix D lists recomended damping values. Damping values will be chosen from the Table 0-3 according to state of stress in the structure. ( Yankee Nuclear Power Station 26 Seismic Reevaluation Criteria 80023/81060/81061/86064 ooc. No. oC-n Rev. 4 ', iiiiiiidiniiJnn$n 0439 L______________-__--____-
1 l l When soil-structure interaction (SSI) is taken into consideration, ; damping ratios will be calculated by including the damping of soil ! materials and radiation damping. Soil material damping shall be taken as 5%. Use of a higher soil material damping shall be justified on a case-by-case basis. In accordance with Ref. 3(ad), 75% of the theoretical radiation damping along the horizontal and vertical directions shall be used. J. 4. 5 Alternate Criteria In cases where stresses exceed the allowaoles given in Section 5.4.2, the
- Structural Nonlinear Performance Criteria given in Section 6 may be used.
- 5. 5 Modification of Structures Design of modifications shall be based on linear structural analysis and code requirements.
i The load conbinations and allowable stresses in Section 5.4.1 and 5.4.2 shall apply. Structural steel connections will also be designed in accordance with Part 1 of the AISC Specifications, Reference 3(c). Selection of concrete anchors shall be in accordance with Tsale 0-2 of Appendix D. 1 For material properties to be used in the design of modifications, refer to Table D-1 of Appendix D. For the criteria to be used in the design in the new SSS building, see Appendix F. Yankee Nuclear Power Station 27 4g Seismic Reevaluation Criteria 80023/81060/8106i/8m84 ooc. no. oC-1; R o. 4 iiimuninmiJimen 0439
6.0 STRUCTtRAL NoNLIEMt P0tFGtMANCE OLITERIA i This section introduces guidelines to be used in demonstrating that the plant can achieve hot shutdown following a seismic event consistent with .C spectra levels as per Section 5.2.5. Under these circumstances, it is reasonable to allow the structural system to undergo stresses and deformations beyond the linear elastic limit as long as safety is not compromised and the plant is capable of maintaining hot shutdown. Acceptable values for limiting behavior indicators consistent with maintaining function by maintaining hot shutdown are determined on a case-by-case basis. Sections 6.1 and 6.2 that follow pertain to the nonlinear behavior and analysis of the Reactor Support Structure (RSS). Similar analysis procedures may be developed to demonstrate functionality of other structures / systems in ) cases where nonlinear response is anticipated under seismic loads consistent i with the MC spectra levels. The use of nonlinear analysis methods developed in such cases is sthject to a case 4y-case review. 6.1 Nonlinear Behavior The nonlinear behavior of the RSS is generally complicated. For this reason, nonlinear material characteristics are modeled into simple idealized force-deformation curves. 6.1.1 Lumped Plasticity For a ductile model of a structural menber stbject to bending moments, a plastic hinge will form when the maximum bending moment reaches its yield value. For RSS column menbers, the interaction of bending and axial forces shall be considered. Yankee Nuclear Power Station 28 Seismic Reevaluation Criteria it..!niunutn 80023/81080/81061/88064 ooc. no. oC-n R o. 4 0439
4 For a structure constructed of ductile materials, plastic deformation may spread under increasing loads. As a result, plastic hinges will form in the structural model and stresses will be redistributed. 6.1.2 Distributed Plasticity For strain-hardening materials such as reinforcing steel, the plastic deformation will be distributed over a finite length of the menber. The distribution is, in general, not uniform over the menber and it plays an important role in stress distribution through the structure. 6.1.3 Stiffness Degradation For the reinforced concrete RSS, reduction of stiffness will appear due to concrete cracking. Under repeated loads resulting from seismic exci-tations, stiffness degradation of reinforced concrete elements can be sih stantial . A beam element with degrading stiffness is available in the PRA computer program (Appendix K(15)]. Yielding may take place only in concentrated plastic hinges at the element ends. Plastic deformation, strain-hardening, and flexural stiffness degradation are modeled by nonlinear 4 rotational springs at each end. The moment-rotation relationship for each spring is an extended version of Takeda's model [Ref.4(u)]. i 6.1.4 Damping l 4 Table D-3 of Appendix 0 recomends damping values for various types of l materials and differing excitation levels. 4 i , Yankee Nuclear Power Station 29 Seismic Reevaluation Criteria 80023/81060/81061/86064 ooc. so. oc-ti Rev. 4 i ti .1. 'iiiiiitu 0439
- 6. 2 Analysis Procedure t A nonlinear time history analysis is required for the RSS sWjected to seismic excitations.
l A nunber of computer programs suitsle for determining the seismic response of nonlinear structures utilizing step by-step numerical integration are a vailab le. The finite element program to be used in the nonlinear analysis of the RSS is PRA [ Appendix K (15)]. At each step, changes in material or geometry or connectivity properties of the structure can be monitored. In addition, important response parameters such as bending and axial force interaction and stiffness degradation can be modeled using PRA. Important general characteristics of inelastic response of structures can often be understood through study of the response of a single degree-of-freedom medel. Such a model is particularly adequate for a structure such as the RSS where the response indicates that most of the mass is excited in one
- fundamental response mode.
l t i l l l' : l l f Yankee Nuclear Power Station 30 g Seismic Reevaluation Criteria 80023/81060/81081/86064 ooc. no. oc-1; Rev. 4 ti . ihiidoiiiitin 0439
l 7.0 M450f0tY WALL PDtFGtMANCE OtITERIA
7.1 Purpose and Scope
I The purpose of this section is to summarize masonry wall design criteria and to establish a consistent and systematic design approach for the e/aluation of and modifications to existing concrete masonry block walls within the seismic scope. The purpose of the evaluation program is to ensure that designated walls maintain their integrity and function during and after a seismic event, thus p ecluding safety-related equipment in the station safe shutdown system from being jeopardized by failure of these walls. Specifically, this procedure requires that walls be evaluated for the following as applicable: I l
, (1) Wall inertial loads (2) Equipment inertial loads (3) Interstory displacements i
(4) Dead loads (5) Wind loads j Wind and tornado loads and analysis procedures are defined in the response to SEP Topic III-2. f Figures 7-1 and 7-2 establish a systematic approach for performance of the wall evaluation and modification program, e
,t Yankee Nuclear Power Station 31 Seismic Reevaluation Criteria tg1.insnen 8oo23/8 o6o'8 o62/86o64 ooc "o oc- "" 4 0439 i
1
- 7. 2 Required Field Data for Review Information required for the evaluation of existing concrete masonry block walls shall be obtained by field sur vey. Applicable YNPS drawings and specifications shall be stained. Using input from the field survey and j existing drawings, "as builts" of surveyed walls shall be prepared for analysis of existing conditions and for the determination, analysis, and design of wall modifications.
- 7. 3 Analysis and Design Criteria i
! a. In general, the analysis of unreinforced and reinforced masonry i shall be based on uncracked section. Cracking of masonry may be l permitted where the application of di verse wall strengthening materials requires the allowab le masonry stresses be exceeded. Nonlinear analysis may be used if the allcwable stresses in Table 7- ,
- 5 are exceeded under the EC spectra loads.
I
! b. Seismic dynamic analysis shall be performed using the appropriate
) Amplified Response Spectra developed for YNPS. , c. Block wall properties shall be calculated on net section, based on ACI 531-79 and specific projcct data (See Table 7-1). i d. Boundary conditions selected shall be representative of the actual - conditions, and analysis results shall be checked to assure validity l of boundary assumptions. The following criteria shall be adopted: l (1) Seismic loads are reversible and supports must provide restraint in both directions. 4 i (2) Consideration shall be given to the relative stiffness of the , i wall and its supports. l . l I l t l Yankee Nuclear Power Station 32 Seismic Reevaluation Criteria j l g 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 0439
(3) Shear transfer mechanisms shall be used at block to block or block to concrete interfaces. No shear transfer shall be assumed at block to steel hints, expansion joints, or compressible material interfaces. Block to block butt joints shall not be considered to transfer moment. (4) Consideration shall be given to the strength of the existing structure (or portion thereof) to carry the wall reactions. (5) Continuity may be assumed at a corner lap joint between block I walls of different orientations providing the appropriate wall stiffnesses and resulting reactions are accounted for, i
- e. Ducts, conduits, and pipes penetrating the block wall shall be reviewed for possible load transmittal to the wall.
- f. Deflection of miscellaneous steel shapes required to support masonry block walls shall be limited to 1/720 of the span under YCS seismic loads. This deflection limit is not applicable under f9C loads.
- 7. 4 Frequency Calculations jf
- a. Frequency calculations shall be based on elastic plate / beam theory.
j
- b. The mass of equipment and/or wall attachements shall be included in the frequency analysis used in the final design solution.
3
- 7. 5 Wall Inertial Load and Stress Calculations
- a. Wall intertial load and stress calculations for final design shall l be determined by use of a peak broadened amplified response spectrum (ARS) and shall account for significant modal participation.
- t I
Yankee Nuclear Power Station 33 l Seismic Reevaluation Criteria
, ,Q,g 80023/81060/81061/86064 Doc. No. OC-1; Rev. 4 0439
- b. Cantilever walls supported at a floor elevation shall use the applicable ARS for that area and floor elevation. For all other walls, the more conservative floor ARS values above the specific wall (s) shall be used.
- c. Where deemed appropriate, inertial loads on a wall panel may be I assumed as the product of the appropriate acceleration (based on the I
wall natural frequency) and the wall mass (w/g).
- d. Moments, shears, and stresses shall be computed using standard structural analysis and strength of material techniques.
(
- 7. 6 Equipment inertial Load l a. Equipment and attachment loads may be calculated using the mass (w/g) multiplied by the appropriate acceleration (b ased on wall natural fregency).
1 l b. The effects resulting from any heavy equipment and/or attachments j shall be superimposed on a representative uniformly loaded area of i the wall. , 7. 7 Interstory Displacements
- a. Applicable block walls shall consider the effects of the appropriate interstory displacements developed for YNPS.
- b. When computing the effects of interstory displacements, only relati ve displacements between elevations shall be used. The displacement profile used shall be compatib le with the wall
; orientation.
i Yankee Nuclear Power Station 34 Seismic Reevaluation Criteria ti ihiiiiiiiiiiiiiiiiiiSoo23/8 o6o/82o62/86o64 ooc "o oc-ti a'v 4 i 0439
- c. Linear interpolation may be used to determine relative displacements at elevations other than those already developed.
- d. In-plane strain analysis applies to the following:
(1) Full story height walls restrained at top and bottom. (2) Full story height walls restrained at all four sides by the building structure. (3) All partial height walls restrained at the bottom and both sides.
- 7. 8 Load Conbinations 4
- a. Masonry Design i
Load conbinations for masonry design shall be: l (i) 0+W i , (2) D + E' i (3) 0+W t (4) 0+Yj + E' i ) where 0 = dead load of block wall plus weight of any equipment attached, ) including pipe support reactions. W = wind load (exterior walls only) per ANSI A58.1-1982. i i E'= loads due to seimic event. 1 i ! Yankee Nuclear Power Station 35 Seismic Reevaluation Criteria
' 80023/81060/8i061/86064 ooc. no. oc-1; Rev. 4 j idiiikiiniiiniiiiin 0439
W t
= loads due to tornado winds, tornado generated AP and tornado missiles per SEP Topic !!!-2.
Yj = loads due to fluid jet impingement.
- b. Steel Design for Masonry Upgrades i Load conbinations for steel design shall be:
(1) D+W (2) D + E' (3) D+W g (4) D+Yj + E' 1 - ! where D. W. E', W g and Yj are defined in Paragraph 7.8.a.
- c. Concrete Design for Masonry Upgrades j
j Load conbinations for concrete design shall be in accordance with ACI 349-76, Ref. 3(k).
! 7. 9 Allowable Stresses l 7.9.1 YCS Seismic Loading 1
l I a. Masonry Design Allowable stresses in unreinforced masonry and reinforced masonry are summarized in Tables 7-2 and 7-3 respectively. it Yankee Nuclear Power Station 36 ; Li L J 6 A Seismic Reevaluation Criteria i 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 181114811111111881188811681611 I t l l 0439 !
i
- b. Steel Design for Masonry Upgrades Allowable stresses per AISC Specifications [Ref. 3(c)).
- c. Concrete Design for Masonry Upgrades A11ow21e stresses per ACI 349-76 [Ref. 3(k)].
7.9.2 MC Spectrum Seismic Loading
- a. Masonry Design Allowele stresses for inspected masonry per ACI-531-79, increased by the factors in Tele 7-5. A11ow21e stresses will be based on values of 1210 psi (12 in, block) and 1486 psi (8 in. block) for F,' and 1381 psi for mo .
- b. Steel Design for Masonry Upgrades Allowable stresses per Section 5.4.2.1, using the MC spectra to SEP Guidelines option.
- c. Concrete Design for Masonry Upgrades A11ow2 1e stresses per Section 5.4.2.1, using the E C spectra to SEP Guidelines option.
Yankee MJclear Power Station 37 Seismic Reevaluation Criteria
,l,( 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 '
0439
l 7.10 Damping Values for Dynamic Analysis
% of critical damping during YCS event:
- a. Masonry - 5%
- b. Steel - 3% (welded) and 5% (bolted)
- c. Concrete - 5%
% of critical damping during MC spectrum event:
- a. Masonry - 7%
- b. Steel - Table D-3 (Appendix D)
- c. Concrete - Table D-3 (Appendix D) i
) m 7.11 Materials i
- a. Masonry -
l Concrete Block: ASTM C-90 (exterior), ASTM C-129
, (interior)
F'm = 600 psi (specified) for YCS F'm = 1210 psi (test results, Ref. 4(n)) for 12 in, block under SC spectrum F'm = 1486 psi [ test results Ref. 4(o)) for 8 in. block ) under EC spectrum Mortar: ASTM C-270 (Type N),n e = 750 psi (specified) for YCS l l mo = 1381 psi [ test results, i Refs. 4(n) & 4(o)] for SC i, spectrum l i b. Concrete - Concrete Strength: For existing concrete, see Table D-1 of Appendix D. For modifications, j f'c = 4000 psi. i i 4 Yankee PWelear Power Station 38 Seismic Reevaluation Criteria j f f , 80023/81060/81061/86064 Doc. No. OC-1; Rev. 4 l 0439 L___._________. _ . _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _
I Reinforcing Steel: ASTM A15 & A305 (fy = 40 ksi) - existing ASTM A615 (Fy = 60 ksi) - modifications Expansion Bolts: Hilti Kwik-Bolts
- c. Structural Steel -
Steel: ASTM A7 (Fy = 33 ksi) - existing ASTM A36 (Fy = 36 ksi) - modifications l Welding: E70XX Electrodes - modifications I Bolting: ASTM A325F - existing and modifications 7.12 Static and Dynamic Analysis The computer program package MCAUTO STRUDL [ Appendix K(16)] or equivalent acceptable program may be used to analyze designated walls for both static and dynamic loadings. The block walls shall be modeled as grids on hybrid (plane , stress - plate bending) finite elements in conbination with beam elements. The initial wall models shall include " preliminary fixes / modifications" as appropriate. Final wall models shall include design fixes / modifications as approved for construction. Static analyses will be performed to account for the following loadings: (a) dead load (b) wind load (c) tornado AP (d) earthquake (vertical direction) (e) earthquake (horizontal direction: N-S & E-W) (f) interstory displacements (N-S & E-W) i t Dynamic analyses will be performed using STRUDL DYNAt. or an equivalent program to account for the following horizontal seismic loadings: (a) amplified floor response spectra (N-S) 1 Yankee Nuclear Power Station 39 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 3g , ! ItemlettilllMillleil 0439 l l
. l l l i (b ) amplified floor response spectra (E-W) a
- A modal extraction will be performed and the N-S and E-W amplified floor response spectra will be applied individually to the finite element / beam
- model. Equipment and/or wall attachment masses, as appropriate, shall be
) considered in the dynamic analysis. Damping values shall be in accordance with Section 7.10 herein. Stresses resulting from horizontal seismic loadings shall be conbined by the square root sum of the squares (SRSS). Stresses resulting from the vertical seismic loading will be absolutely added to the
- SRSS of the horizontal seismic loadings.
4 i Stresses resulting from the static and dynamic loadings shall be conbined in accordance with the design load combinations of Section 7.8. The resulting i conb ined stresses shall be equal to or less than the allowable stresses permitted in Section 7.9. The conbined stresses may be calculated manually l where appropriate. i , e i l l J 1 i i
; e i
i i 1 I Yankee Nuclear Power Station 40 i Seismic Reevaluation Criteria
- ' L h A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 4
INIHilillllleillNHililllll 0439 i
r FIGtstE 7-1 ELL EVALETIO11 & IWOIFICAT10ll pit 0stAM 1 ' I (1) i Wall Identifteetion Design Critetta Cetain Exist &ng Design j
!aput (1) Drawinas & Specifications 4
Oetaan Field "As-Built-j Dets
- Perform Preliminary __
Evaluation o t
, Prepare "As-Built
- Dwns.
n , (1) Deteraane Wall hedificettees Equissent Data T (2) - Setesic ARS taput l 4 j Prepare Structurel - l j Destan Criteria Docueent
! C l
f e l l j TAtc Approvat n l Well Analysis - (Figure 7 2)
~
l 1 1 I
- ala" '
- l- - - - - - - - - J L - - -
- w**-*r.iute.47 a
i l , D.. tan modifs.stione 4 l i i Prepare Declas Drewinas !, l o l~~~~~~~~~*FieldConfirmattenet *~~ "Aa-Built' Dwge. Destao & "As*Dullt' Drewleae Chetbed I
- t a
- Desten Calculettene ---* Modifisetten Draw &nge checked & Filed Cheeked i
- o previnne tee wd (t h vase (2) By CvGNA 4
- Yankee Nuclear Power Station 41
! Seismic Reevaluation Criteria ! ' b 6 ' 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l lismillMMulmill 0439
1 FIGtstE 7-2 WALL ANALYSIS l l
'from figure 7 1
/ Calculate Well Properties s f %
/
/ ,
l / \ y
,* Define boundary Condittena *% \
g l / ' 7% % N
/ m %
$#s#p/p i
1
/ ,
% s *% k l *~~~~~~~~~~~~~~
I l~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
- Prepare Computer leedel l
l
""I I
L l~* Calculate Stette 1Aade Calculate Masses t l l I
' i l~ ~ Perfers Stette Analyste Perfors Dyneate Analyste lL _ I
- teed Lead
- 16erth-$euth AAS g e wind Lead
- Laet-West AAS l
- Tornade AP l
- Interstory Diopt. ft w 6 s-si !
. v.riac.1 s.recen.
!
- sorisontet sarthquake it w 6 u 81 !
I I
~~ ~~~
! Evaluate output , Svaluate output I Cateulate Combined Stressee Evaluate Forces & Stresses i l l l l l
....__l I...
See Figure 7 1 Yankee Nuclear Power Station 42 Seismic Reevaluation Criteria dO6 4 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 m60mmtemtemittil 0439 1
FIGLRE 7-3 ILLUSRATION OF AlloidABLE SMESSES _ h .
.;q
'* in at a 6 6 Fa..te6f'(e*(k!')
4. n'
' r.
-.. ./
****** \
I ] i .
/*
y, ..
/ .
l Q. ,- L f a e e.= ..a n u . ii. .se mae9 e .Ta. 4.s .a.k... avan w Ma mast,s m.. .e. ..a ., e . F, . 6.eK. was . 6e S t. F, . S S K. .. f 6 ps.i
$ e .e . .se ae e s...e ee .se ..e9 8 Fg . p.4 4. no . es p 4. It* h 9 8. e e. 4p6L m :, = .u.
. .i hh ra to i .
m e..: r.. . . ,..i .. Ais .<.-,s....
+. -- + . . . <. i<
- 6
. ......, . .s . . ..
. .. . . . .e . .
= ~
- t,,. . w. ~ **:;* %
. _ or it_1':-- %
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> . / .._ g,,.%,.j ,,::*a,, 5 ; O -
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- _ ;ii. .e,. p. . se,me ,. go,.4, se y, p, . nose ,.:
l Grade M s8 eel): I. em Se,00s pel l6,ede Is lew col.fece e) Yankee Nuclear Power Station 43 Seismic Reevaluation Criteria Ob h ' 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 , IlliteltlMitullitilli 0439
TABLE 7-1 CONCRETE BLOCK MALL PROPERTIES TYPE NOMINAL SIZEIII WEIGHT (2) AREA (NET) Ixx = Iyy(3) Sxx = Syy(3) MASS DENSITY (4' 4 3 (T x H x L) (psf) (in2/in.) (iq /in.) (in /in.) (1b-sec2 /in.4) i Normal 8x8x16 55 2. 5 25.76 6.76 1.462 x 10~4 Wei ght (Hollow) Normal Weight 12x8x16 74 3. 0 77.45 13.32 1.364 x 10-4
! (Hollow)
I j NOTES
. (1) Assume 3 cell hollow concrete masonry unit, and mortar bedding on face shell thickness only. ! - (2) Includes 5 psf for nominal existing masonry reinforcing.
(3) Based on actual thickness. l (4) Based on equivalent solid thickness. l n s j FST = l'/[($"cMu) ; l g l El *(ll"CMd) l N W Y N Yb W Y & /h n
- h.
W c
. l -
7~,N= x l i R Ph//7)/g7/yypp h ' l" (S'C M") ! . , l'/e'Oz cMu) i I i _ L uoru I j ( L.) l 1 Yankee Nuclear Power Station 44 i Seismic Reevaluation Criteria i L b h a 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lilitimelillisteellillitel
- 0439 i
TABLE 7-2 ALLOWABLE STRESSES IN UIREllFGtCED MASOIRY (YCS SPECTRA) New Masonry (3) Existing Masonry (1,7) Description S allow. S max. S allow. (2) U allow. (3) Compressi ve Axial (4) 0.22f' , 1000 132 330 Flexural 0.33f', 1200 198 495 Bearing On ful1 area 900 150 375 On one-third area or less 0.25f'T 0.375f 1200 225 562 m Shear Flexural mes ers (5,6) 1.1 { 50 27 35 Shear walls (5) 0. 9 { 34 22 29 Tension Normal to bed joints Hol1ow units 0,5 /m g 25 14 14 Solid or grouted 1. 0 /m g 40 27 27 Parallel to bed joints (6) Hollow units 1. 0 /mg 50 27 41 Solid or grouted 1. 5 /m g 80 41 62
! Modulus of Elasticity (Em) (1) 1000 f', 2.5 x 10 6 .810 x 10 6 .810 x 10 6 Modulus of Rigidity (G)(8) E,/2(1+ v) 1.0 x 10 6 .350 x 10 6 .350 x 10 6
)
NOTES (Cont'd Next Page) l I l 4
- Yankee Nuclear Power Station 45
- Seismic Reevaluation Criteria
' ' b L A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 1611111lll1111111lll11111111ll l
0439
NOTES (1) Based on f', = 600 psi and m o
= 750 psi (inspected masonry) for interior walls (C-129). E m is based on actual block and mortor strength (test results). Units are lbs/in 2, (2) For normal and severe environmental loads, allowables are not factored.
(3) For abnormal and extreme environmental loads, allowables are factored in accordance with Table 7-4 (4) Multiply values by ( 1- ( 40t
)) or signWcant Mcal load.
(5) Use net bedded area with these stresses. (6) For stack bond construction, use 2/3 of specified values. (7) Refer to Figure 7-3 for illustration of stresses. (8) A value of 0.15 is used for Poisson's ratio (v). I ti Yankee Nuclear Power Station 46 Seismic Reevaluation Criteria 4 80023,81osof81081,880s4 ooc. no. oc-1; Rev. 4 kninildinen 0439 l
l TABLE 7-3 ALLOWABLE STRESSES IN REIEGtCED MASOfstY (YCS SPECTRA) New Masonry (3) rv4 e
- 4 n n u2ennry (!)
Description S allow. S max. S allow. (2) U allow. (3- l Compressi ve l Axial (4) 1000 132 330 0.22f'm F1exural 0.33f', 1200 198 495 Bearing On full area 900 150 375 On one-third area or less 0.25f'T 0.375f , 1200 225 562 Shear No Shear Reinforcement Flexural Menbers 1.1 { 50 27 35 Shearwalls M/Vd>l 0. 9 { 34 22 29 M/Vdc1 Reinforcement taking entire shear
- 2. 0 { (5) (5) (6)
Flexural Menbers Shearwalls
- 3. 0 { 150 73 110 M/Vdal 1. 5 { 75 37 48 Reinforcement M/Vd<1
- 2. 0 { (5) (5) (6)
Bond (7) Plain Bars 60 60 S0 Deformed Bars 140 140 186 Tension Grade 40 20,000 .9 Fy Grade 60 24,000 .9 Fy Joint Wire .5 Fy or .9 Fy 30,000 Compression .4 Fy .9 Fy Modulus of Elasticity (Em)(1) 1000 f', 2.5 x 10 6 .810 x 10 6
.810 x 10 6 Modulus of Rigidity (G)(8) E,/ 2(1+v) 1.0 x 10 6 .350 x 106 .350 x 10 6 Notes (See Next Page)
Yankee Nuclear Power Station 47 A'{ ; , , Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 llllllll1111lllllllllI11111111 0439
NOTES (1) Based on f',= 600 psi and mg = 750 psi (inspected masonry). E, is based on actual block and mortor strength (test results). Units are lbs/in 2, (2) For normal and severe environmental loads. Allowables are not factored. l (3) For abnormal and extreme environmental loads, allowables are factored in accordance with Table 7-4 h (4) Multiply values by ( 1-( 4 ) 3) for significant load. (5) Refer to ACI 531, Table 10.1 for allowable stresses. (6) Factor allowables in accordance with Table 7.4 (7) From Ref. 3'i). (8) A value of 0.15 is used for v Yankee Nuclear Power Station 48 Seismic Reevaluation Criteria liimnipr4 n iintnnntil 8 23f81080f81081/8s084 00c. No. 0C-1 Rev. 4 j 0439 l
TABLE 7-4 MASOfRY ALLOWABLE STRESS FACTORS (l*2) TYPE OF STRESS FACTOR Axial or Flexural Compression (1) 2. 5 Bearing 2.5 Reinforcenent stress except shear 2.0 but not to exceed 0.9 fy Shear reinforcement and/or bolts 1.5 Masonry tension parallel to bed joint 1. 5 Shear carried by masonry 1.3 Masonry tension perpendicular to bed joint (No Increase) NOTES (1) Applice)le for reinforced and unreinforced masonry. (2) Factors shall apply to 8) normal and extreme environmental load conditions only (tornado or YCS seismic event). t l
.i Yankee Nuclear Power Station 49
- Seismic Reevaluation Criteria L*b b J k A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 1 lll1181!!!::' ':!!!Illi 1
0439 ! I
4 TABLE 7-5 ALLOWABLE STRESSES IM UREIEGtCED MASORY (RC SPECTRA) ACI 531 Increase EC Spectra (Normal) Factor Allow. (per SRP 3.8.4, Appendix A) Compression Axial .22 F ', 2. 5 . 55 F ', Flexual . 33 F ', 2.5 .83 F 'm Bearing On full area . 25 F ', 2. 5 .62 F ', On one-third .375 F', 2. 5 .95 F'm area or less
, Shear Flexural 1.1 % 1. 3 1.43 ,K menb ers Tension Normal to bed joints, hollow units 0. 5 /mg 1. 3 0.65fmo Parallel to bed joints, hollow units 1. 0 (m g 1. 5 1.505
. \
I Hodulus of Elasticity (E ,) .810 x 106 psi Modulus of Rigidity (G) .350 x 106 psi Material Properties F'm (12" block) = 1210 psi, (8" block) = 1486 psi, mn = 1381 psi l Yankee Nuclear Power Station 50 Seismic Reevaluation Criteria LT[ j g i. 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lillllllif.!!!!!!!!!!!!!!!!!!! 0439 l
8.0 PIPING ANALYSIS OtITERIA This section describes the criteria to be used in the stress analysis of the piping systems which are part of the Safe Shutdown System or are otherwise included in the seismic scope. 8.1 Load Description The following load cases shall be considered for the piping stress analysis. In addition, local stress concentration due to integral supports shall be evaluated. 8.1.1 Thermal Load Loads due to steady state temperature effects, including thermal anchor mo vements. 8.1.2 Weight Load Loads due to the weight of the pipe, its contents, and its insulation. 8.1.3 Pressure Load Loads due to the steady state internal design pressure. 8.1.4 Seismic Load i Loads due to earthquake excitations, including both seismic inertia effects and seismic anchor mo vements. Pipe stresses due to these loadings are limited to the values in Section 8.3. l Yankee Nuclear Power Station 51
;gg,, Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lI1llllllllllll1111111llll1111
- 0439B l
8.1.5 Occasional Load l Loads due to relief valve blow-off, snow, wind, and other occasional loads.
- 8. 2 Analysis Methodology This section contains the general methodology used in the computer analysis of piping systems. This methodology is to be applied to the analysis of all safe shutdown piping with the following exceptions:
o Small bore piping may be analyzed by either the methods outlined here or by hand calculation methods. Section 8.4 discusses the evaluation of small bore piping, o Buried piping will be evaluated using hand calculation methods and is discussed in Section 8.5. O The main steam /feedwater piping outside the vapor container is supported on a series of light structural frames which may further amplify the seismic motions and promote interaction between pipes on the support structure. Section 8. 6 discusses the evaluation of the main steam /feedwater piping. 8.2.1 Geometry and Computer Modeling For the purpose of computer analysis, the piping systems will be idealized by three-dimensional linear elastic models. All supports and anchors are assumed rigid if stiffness and/or frequency requirements of Section 9.5 are met. Reduced stiffness shall be calculated and justifiied on a case by-case basis. Yankee Nuclear Power Station 52 !
- g{ ,
Seismic Reevaluation Criteria i 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 IIlllilllilillilll11111lll1ll 0439B
(a) Each problem shall be considered from anchor to anchor. If an anchor to anchor problem exceeds program limitations, it may be broken up into smaller problems with adequate overlap areas. The overlap areas will be chosen to properly represent the truncated portion of piping. The results of the multiple computer runs in the overlap areas will be enveloped to assess boundary or loading conditions. All usage of overlap criteria will be justified on a case-by-case basis. (b) The geometry and restraint conditions shall be modeled in accordance with as-built isometrics. (c) The piping analysis shall be performed using Yankee Piping Specifications (YS-497 and YS-4652), YAEC flow diagrams, Yankee Insulation Specifications (YS-2304), and vendor catalog data. Pipe and pipe support material properties for the specified analysis conditions are stained from Appendices A, B, and -C of Reference 3(a). (d) Branch lines may be decoupled from run piping if the following three criteria are met: o The ratio of moments of inertia (run/b ranch) is 25:1 or greater. o There are no anchors or supports on the branch line close to the branch point which serve to restrain the run pipe. o There are no nozzles on the branch line close to the branch point which restrain the run pipe. When analyzing the branch line, the point of decoupling from the run line shall be considered an anchor. The deflections of the run piping at the decoupling point shall be input as anchor motions for the branch line analysis. Yankee Nuclear Power Station 53 ge , 33 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l IIlllllillililllilllilillil!!! 0439B
(e) Equipment nozzles and penetrations shall generally be considered as anchor points in the analysis. On a case by-case basis, nozzle flexibility will be considered using )RC Bulletin 297 [ Reference 4(ag)]. In these cases, adequate documentation of justification for this usage shall be provided. All equipment is assumed to be properly supported. Loading shall be summarized and compared to allowables when availab le. Thermal anchor movements shall be calculated by conventional methods based on system operating or design temperature. (f) Valves shall be modeled as follows: o Thickness of the valve body shall be considered as twice the connecting pipe wall thickness. o Manually operated valves and check valves shall be modeled with the mass of the valve concentrated at the centerline of the pipe at the valve node points. o Valves with eccentric mass, such as motor- and air-operated val ves shall be modeled as eccentric mass points. Where a vailab le, the actual center of gravity and operator weight shall be used from the manufacturer drawings or data. Otherwise, the valve / operator shall be modeled using one of the following methods: (1) If the valve and operator weights are known, the valye weight shall be lumped at the pipe centerline, and the operator wei ght shall be lumped at the center of gravity of the operator. (2) The total weight of the valve shall be concentrated at a poMt one-third (1/3) the distance from the valve assenbly to the c ,terline of the operator (one-third of the " stem length" measurements as noted on the valve l i data form). The second method shall not be used for small-bore motor-operated valves.
- Yankee Nuclear Power Station 54 Seismic Reevaluation Criteria ;
hm..y , ,::..f ... 3ngg 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 I i < 0439B ! \ l
Seismic accelerations of the val ves shall be summarized. The allowable valve accelerations shall be [ Reference 4(h)]: Resultant horizontal acceleration < 4.25 g Vertical accelerations < 3.0 g (g) Flanges shall be considered as additional lumped weights. Flange thicknesses shall be assumed to be the same as that of the pipe for purposes of modeling stiffness. Additional flange information may be obtained from ANSI B16.51977. (h) Stress intensification factors for tees, reducers, flanges, elbows and couplings (half and full) shall be considered as per Appendix 0 of Reference 3(a). (i) Penetrations shall be analyzed as follows: . Grouted penetrations: A bilateral restraint condition shall be considered to exist on either side of the penetration for all load cases. Generally, axial restraint of the pipe shall not be considered unless the pipe has a welded collar which is enbedded in the penetration. However, for piping exerting relatively low axial loads on the penetration, the grout may still afford axial restraint. In these cases, axial restraint may be taken if the bond between the pipe and the grout remains intact. This check will be made if credit is taken for the axial restraint. Ungrouted penetrations: At ungrouted penetrations, deflection of the pipe less than 1/4" shall be considered acceptable. Where deflections exceed 1/4", further review of actual penetration clearances shall be initiated. Deflections shall be based on the conbined thermal, deadweight, and seismic conditions. Yankee Nuclear Power Station 55 g Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l l11lll111ll1ll ll ll1111111 0439B
(j) The moduli of elasticity at various temperatures for ferrous and non-ferrous materials shall be taken from Appendix C, Tables C-1 and C-2 of the Reference 3(a). (k) The Poisson's ratio shall be taken as 0.3 for all metals at all temperatures. (1) The hot modulus of elasticity (Eh) values shall be used for seismic and SAM (seismic anchor movement) analyses, and the cold modulus of elasticity (E )c values shall be used for deadweight, thermal and TAM (thermal anchor movement) analyses. (m) The flexibility of attached equipment shall be considered in the pipe stress analysis where necessary. (n) The computed stresses of various load cases shall be summarized. 8.2.2 Weight Analysis Weight analysis shall be performed considering the weight of the pipe, its contents, insulation, and concentrated masses (such as pipes supported off the pipe, flanges, and valves). 8.2.3 Thermal Analysis Thermal analysis of the piping system shall be performed based either on the maximum design temperatures designated in the Yankee Piping Specifications, on YAEC flow diagrams or stress isometric drawings, or en pipe operating temperatures as supplied by YAEC. Effects of thermal movements from equipment nozzles, anchors, penetrations, and connecting I ) l l ' l Yankee Nuclear Power Station 56 g{ Seismic Reevaluation Criteria I 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 I himimimQmmm I 0439B
piping shall be analyzed. The thermal anchor movement (TAM) stress due to normal operating temperature shall be added to the thermal expansion stress to obtain the total thermal stress. Zero deflection criteria may be used based on the recommendations of WRC-300. 8.2.4 Seismic Analysis (a) The basic dynamic analysis technique will be the response spectrum modal superposition method using lumped mass models. Sufficient mass points shall be used in the computer model to adequately represent the mass distribution. When available, the ' Automatic Mass Point Spacing' option shall be used. When not available, the maximum span length between mass points shall not exceed: L < f {g} h} (Eq. 8.2.4) Where: f = Cutoff frequency, Hz E = Young's modulus, psi I = Pipe moment of inertia, in 4 w = Weight / unit length of pipe + contents, lb/in L = Maximum allowable span between mass points, in For rod hanger supports, when the uplift due to seismic load (include thermal load if it is upward) is larger than the weight load, the effect of the rod hanger support on the system shall be considered as follows: Two seismic analyses shall be considered. In the first analysis, the rod hanger shall be considered ef fecti ve. In the second analysis, the particular rod hanger support will not be included in the model. The results of the two analyses will be enveloped. 1 Both seismic inertia analysis and seismic anchor movement analysis shall be performed. Yankee Nuclear Power Station 57
'g'{yJ.l. , Seismic Reevaluation Criteria
,,,.,HHm 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l
)
0439B )
1 l l (b ) Application of Spectra: Three directions of earthquake will be considered (two horizontal components and one vertical component). The total response due to each of the three (3) components of earthquake shall be calculated first. These individual responses shall then be conbined by the SRSS method (square root of the sum of the squares). The procedures to be used in conb ining the modal responses and responses due to spatial components of earthquake shall be as follows:
- 1. The modal responses for each component of earthquake shall be conbined by taking into consideration the modes with closely spaced frequencies in accordance with Reg. Guide 1.92
[Ref. 3(b )], Stb sections 1.2.1, 1.2.2, or 1.2.3.
- 2. The system's total responses to the three (3) spatial com-ponents of earthquake are then conbined by the SRSS method.
For piping systems spanning several floors or with pipe supports connected to support structures attached to different floors, the response spectra for the analysis of the piping system shtll be the envelope of the floor response spectra of all the floors in vol ved. Howe ver, on a limited basis, the use of multi-level response spectra may be applied. When this is done, a review of the cumulative effects of other techniques used will be performed to ensure a conservative analysis approach. ! ASME Code Case N-397 [ Reference 3(m)] may be used as an alternative to the spectral broadening as described in USPRC Regulatory Guide 1.122. Yankee Nuclear Power Station 58 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 IIllilllilli!!!!!!!"!!!!!! 0439B
l l (c) Cut-off frequency and minimitm nunber of modes: A cut-off frequency of 33 cps, with no less than 10 modes, shall be. considered in the analysis. An equivalent static seismic analysis, using a constant spectral acceleration at the 33 cps cut-off frequency, shall be performed when the contributions of hi gher modes (>33 cps) are significant. The results of the static analysis shall be conbined by SRSS with the dynamic results. (d) Damping values: For either the YCS or RC seismic event, damping as specified in USMC Regulatory Guide 1.61 or ASME Code Case N-411 [ Reference 3(m)] shall be used. The two different dampings (Regulatory Guide 1.61 or N-411) may not be used concurrently on the same piping prcb lem. 8.2.5 Seismic Anchor Movement Analysis (SAM) The seismic anchor movement load condition shall be considered for both stress and support load evaluations. SAM will generally be conbined with seismic inertia loads by sunrning absolutely. However, the SRSS method may also be used with RC spectra loadings. 8.2.6 Pressure Effect The effect of internal pressure shall be considered in computing longi-tudinal stresses. 8.2.7 Effects Due to Relief Valve Blow-Off and Other Occasional Loads The effects due to relief valve blow-off and other occasional loads shall be calculated and applied as external forces to the piping. Yankee Nuclear Power Station 59 Seismic Reevaluation Criteria [g....j ' .1,, 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 ( 0439B i
- 8. 3 Acceptance Criteria 8.3.1 Stress Equations Stresses in the piping systein must not exceed the allowtle stress limits of Refereno 3(a) for YCS loading. The acceptance criteria shall be considered satisfied when the requirements of the following equations are met.
(a) The effects of pressure, weight, and other sustained loads must meet the following requirements:
+ 0.751 (Eq. 8.3.1-A) 4t n Z MA'b h where:
P = Internal design pressure, psi Do = Outside diameter of pipe, in. t = Nominal wall thickness of components, in. n i = Stress intensification factor. The product of 0.751 shall never be taken as less than 1.0. Z = Section modulus of the pipe, in3 , MA = Resultant moment loading on cross section of the pipe due to weight and other sustained loads, in-lb. , S h
= Basic material allowable stress, psi, taken from l Appendix A of Reference 3(a) at the design temperature ,
defined by the Yankee Piping Specifications YS-497 and i YS-4652 or at the pipe operating temperature. Stress intensification factors "1" shall be per Appendix D of . Reference 3(a). Stress intensification factors for non-standard ! components may be stained by finite element analyses as required. { t !' l 60 Yankee Nuclear Power Station Seismic Reevaluation Criteria 4 , 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 i I[ IllililllllilllllIllililllil l 0439B l--. ___, __ ._ , . . . . . _ _ _ _ _ , _ _ _ _ _ _ , _ _ _
Note: The pressure stress term in Equations 8.3.1-A, -B, and -D may be replaced by the following: Pd2 Dg 2- d2 where d = inside diameter of the pipe. (b ) The effects of pressure, weight, other sustained loads, and earthquake must meet the following requirements: PD o 0.75i M A , 0.751 g8 < 1. 8Sh (Eq. 8.3.1-B) 4t n Z Z for the Yankee Composite Spectra, where M B
= Resultant moment loading on the cross section due to earthquake loads, in.-lbs. All other terms are the same as in Equation 8.3.1-A.
For the FRC spectra, the piping systems must be shown to remain functional under the effects of pressure, weight, other sustained I loads, and earthquake. As a first check, the SEP allowables will be used in Equation 8.3.1-B as follows: 1.8S h for Class 1 and 2.4S h for Class 2 and 3 piping. In cases where these allowables are not satisfied, the equivalent strain at any point in the system except equipment and val ve nozzles, threaded connections, the specially fab ricated tee and fork components in the Main i Steam /Feedwater piping outside the VC, shall be limited to 1%. Stress at those nozzles and threaded connections shall meet the SEP allowab les. The strain calculations will be used and justified on l a case-by-case basis. Yankee Nuclear Power Station 61 Seismic Reevaluation Criteria gd ygggg 1 ;;;j,;; 3 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 0439B
l Strain shall be calculated as follows: So For carbon steel - c= < 0.01 ) i 6.67a l For stainless steel - c= < 0.02 E where c = equivalent strain, in/in E
= ANSI B31.1 Equation 12 stress, psi Note: If SAM stresses are not added to Equations 13 and/or 14, they will be included here.
E = Young's modulus, psi In addition, the following checks shall be made: For stainless steel large bore Sch 40 piping - c< 00.4t" (wrinkling limit) o-tn where: Do = outside diameter of pipe, in, tn = nominal pipe wall thickness, in. For carbon and stainless steel " seismic < 44ksi (fatigue limit) Note: If SAM Stresses are not added to Equations 13 and/or 14, they will be included here. Yankee Nuclear Power Station 62 g , Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lilililllililliifilllitilli 0439B
1 l Elastically calculated displacements in the area of inelastic b eha vior shall be multiplied by 3.33 to account for added deformation. The effect of the added displacements shall be reviewed. For strinless steel piping with strains greater than 0.01, elastically calculated flanged joint, nozzle and support loads shall be multiplied by the following factor, X: X = 1. 0 + 170. 0 ( c - 0. 01) (c) Thermal Expansion Stress (SE ) 1M SE" (Eq. 8.3.1-C) z where: MC = The range of resultant moments due to thermal expansion, i n. -lb s. Also include moment effects of anchor displacement due to earthquake if anchor displacement effects were omitted from Eq. 8.3.1-B. For the EC load case, since Equation 8.3.1-C is a Level A/B stress comparison, and since the RC Spectra represent much greater load cases than an OBE, 60% of the SAM effects will be used in M .c This 60% is intended to limit the SAM results to the same levels as an OBE code analysis would. However,- if the 60% SAM conb ination is used at any point, the strain criteria shall not be used at the same point. f s Yankee Nuclear Power Station 63
- g Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
- killill!!""" "!!!!I!!!
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SA
= Allowable stress range for expansion stress.
=
f (1.25 Sc + 0.25 Sh) where: S = All wable stress of the specific material at c 70'F, psi S h
= Allowable stress of the specific material at maximum temperature as defined by the Yankee Piping Specifications YS-497 and YS-4652 or at . the pipe operating temperature, psi f = Stress range reduction factor See Equations 8.3.1-A and -B for terms not listed.
(d) Sustained Plus Thermal Expansion Stresses: The effects of pressure, weight, other sustained loads and thermal expansion must meet the requirements of the Equation 8.3.1-D: PD
+ 0.751 g (Eq. 8.3.1-D) 4t n Z A+Z "C ' ( h + bA )
4 Terms as previously defined. (e) The requirements of either Equation 8.3.1-C or Equation 8.3.1-0 I must be met. Yankee Nuclear Power Station 64 Seismic Reevaluation Criteria g J ' ,,,j,,, 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 0439B I _.
(f) Although only the response spectrum analysis method is considered in this criteria, it does not preclude the possibility of using the time history analysis method, if the situation warrants its application. Specific criteria for time history analysis will be provided when the need arises. 8.3.2 Allowable Stresses Allowable stress values to be used for safe shutdown piping systems are given in Appendix A of Reference 3(a). Those values shall be used for piping stress analyses. For material allowable stress values not availab le in Appendix A of Reference 3(a) or for material yield stress values at design temperatures, Reference 3(g) shall be used. The appropriate allowable stress values shall be taken from tables contained in Appendix I of the reference. 8.3.3 Allowable Deformations Deformations will be limited to existing clearances to prevent impact of adjacent components. Also see Section 8.2.1(i).
- 8. 4 Small Bore Pipe Stress Analysis This section applies to piping with a nominal outside diameter of 2" or smaller. The stress qualification for small bore piping shall be performed using one of the methods outlined in Sections 8.4.1 and 8.4.2.
8.4.1 Detailed Stress Analysis For detailed stress analysis, the same procedures and methods as those for large pipe stress analysis shall be followed (Sections 8.1 through Yankee Nuclear Power Station 65 Seismic Reevaluation Criteria 4 "* * ; * *
- IIlllilill lilllilililllIlli 0439B
1 1 8.3). In addition, connections at elbows, tees, reducers, couplings, and nozzles shall be considered as socket welded unless otherwise noted on the design isometrics. A threaded connection shall be considered where noted on the drawings. 8.4.2 Simplified Stress Analysis This is an alternative approach to the detailed stress analysis method and is used to evaluate relatively simple piping geometries, such as cantilevers or continuous piping spans with minimal branches. Each span of a piping system (spans are generally separated by guides) is evaluated by simplified thermal, seismic, and weight stress analyses. Span lengths and support locations are investigated to ensure that the requirements of piping flexibility and hi gh natural frequency are met. Stress intensification factors are included as required. (a) Weight Stress - Weight stress is calculated by performing static span analysis. Concentrated loads, such as valves or risers, are included as required. 4 (b) Thermal Stress - Thermal stress shall be kept to an acceptable level by providing a minimum offset to absorb thermal expansion and thermal anchor movement. Offset is defined as the length of piping in a plane perpendicular to the direction of movement. The offset piping shall be unrestrained in the direction of movement. A large nunber of the small bore piping problems are analyzed because they are required to maintain primary and secondary pressure boundaries. The safety related portions of the piping are limited to those portions up to isolation valves. In these cases, it is unrealistic to assume that the maximum design temperature will be maintained in the entire line. More realistic thermal Yankee Nuclear Power Station 66 Seismic Reevaluation Criteria 4' g$.imi. ....h mii 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 0439B 1
distributions will be postulated where necessary, to d>tain less severe thermal expansion stress distributions. (c) Seismic Stress - Seismic pipe stresses shall be evaluated by the equi valent static method. Span stresses are calculated assuming simply-supported spans between supports. For the initial evaluation, the peak of the response spectrum is used to accelerate the simply-supported span. Since 100% mass participation is assumed for this static "first mode" which maximizes span stress, a dynamic amplification is not used. If the piping is overstressed using the peak acceleration from the response spectra, an evaluation is performed to determine if frequency testing or detailed analysis should be the next course of action. Zero period' acceleration (the acceleration at 33 cps, in this case) is used to accelerate the pipe span. If the span is qualified by this method and appears to be relatively rigid (high fundamental frequency), frequency ' testing is warranted. If the span is not qualified by this method, detailed analysis or addition of supports is necessary. For some piping systems, frequency testing will be performed for three orthogonal directions to ob tain less conser vati ve accelerations for the equivalent static analysis. The testing methods are described in Section 8. 4. 3. The acceleration corresponding to the tested fundamental frequency will be used in the piping analysis, if the fundamental frequency is greater than the frequency at the peak acceleration. Agai n, the dynamic amplification factor for these cases is taken as 1. 0. If the fundainental frequency of the piping is below the frequency at the peak acceleration, the peak acceleration is retained in the analysis. Yankee Nuclear Power Station 67 g Seismic Reevaluation Criteria "o sundusanntn8oo23/8 o6o/82o62'86o64 ooc 1 i oc-u "" 4 0439B
Seismic anchor movement (SAM) analyses are performed by applying the required displacements at the anchor points and calculating the induced stresses. (d) Pressure Stress - Longitudinal pressure stress shall be computed per the ANSI B31.1 Code [ Reference 3(a)] requirements using the design pressures documented in the Yankee Piping Specifications, flow diagrams, and isometrics. (e) Acceptance Criteria - the calculated stresses for each load case are conbined as specified in ANSI B31.1, 1977 [ Reference 3(a)). Equations 11 through 14 are evaluated for all prob lems. See Section 8.3 for details. Exceptions taken to the acceptance criteria presented in Section 8.3 are as follows: (1) For pressurizer spray piping problems 2 and 3, the smaller value of
- 1. 8S y and 2.25S h shall be used as the stress allowab le in conjunction with the stress indices in Section NC-3654 of the 1983 version of the ASME Code (Equation 9), Reference 3(m).
8.4.3 Frequency Testing Frequency testing may be performed to determine the fundamental frequency of pipe spans to obtain more realistic acceleration levels for the stress analyses. The pipe spans shall be instrumented with accelerome.ters in each of the three orthogonal directions used in the stress analysis. Indi vidual frequency response functions will be generated by recording I the pipe span response of impact excitation or step relaxation excitation with a spectral analyzer. Impact excitation will be used for relatively ri gid lines by striking the piping with a hammer. Step relaxation lE l Yankee Nuclear Power Station 68 l Seismic Reevaluation Criteria ' 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 b..d(m{ 04398 l
excitation will be used for relatively flexible lines by displacing the pipe and releasing it to vibrate freely. At least four individual tests for each response direction and location will be made. To obtain the final frequency response functions to be used in the stress analyses, the individual frequency response functions will be averaged and plotted. The frequency of the first significant response peak will be used as the fundamental frequency of the span for the tested direction.
- 8. 5 Buried pipe Stress Analysis Buried piping shall be analyzed per Section 6.2 of Appendix F.
- 8. 6 Main Steam /Feedwater Piping Outside the Vapor Container 8.6.1 Geometry and Computer Modeling 9
The main steam /feedwater piping outside the VC is supported on a series of li ght structural frames. The mass and stiffness of the support structure and piping runs are of similar magnitude; therefore, the rigid support assumption generally used in the majority of the piping analyses is not valid here. Additionally, a nunber of pipes may be supported by the same frame, promoting interaction between the pipes. For the above reasons, the main steam /feedwater piping and support structure shall be analyzed together in one problem. Non-seismic piping shall be modeled and analyzed if it affects the response of the seismic piping and the support structure. The conbined main steam /feedwater piping and support structure system is large and complex, making it prohibitively costly and time-consuming to analyze a detailed model of the entire system. To minimize the size of the actual seismic analysis model, stbstructuring methods may be used to Yankee Nuclear Power Station 69 Seismic Reevaluation Criteria 4 80023/81060/81061/86064 ooc. No. oC-1; Rev. 4 hinndniintunntn 0439B
l l model the structures connected to the MS/FW piping systems such as the i VC, PAB, Turb ine Building (TB), and Crane Support, as well as some portions of the MS/FW Support Frame. These structures may be modeled by one or more stbstructures. Each frame will first be modeled in detail, considering all structural menbers (including all vertical and horizontal menbers and cross braces), connections, and other attached piping and equipment. Consideration will be given to the flexibility of frame menbers on which the piping systems are directly supported and to the supporting scheme on each frame to determine whether the main steam and feedwater pipes will interact on that particular frame. If the supporting scheme promotes interaction, the attachment points of the piping will be retained, and the detailed frame model will then be condensed to retain only the important menbers and the mass and stiffness matrices of the remaining menbers represented by the stbstructures. This simplified model will then be used for the i I system analysis. To ensure the proper application of the stbstructure technique, the portion of the structure modeled by stbstructure will be compared to the full model. For example, as shown in Fig. 8.1, a frame is modeled with a stb structure. Depending on the dynamic characteristics of the frame and the nunb er of connections between the pipe supports and frame, the amplified response spectra (ARS) at one or more pipe support attachment points will be generated using the original model plus a portion of the piping system. Sthsequently, ARS at the same locations will be generated from the model containing the stbstructure plus the same portion of the , piping system. Tne comparison of the ARS generated using these two j models will provide an assessment of the stbstructure accuracy. The support frames and the main steam and feedwater lines are connected I to several structures. These structures may be modeled as stbstructures ! 4 Yankee Nuclear Power Station 70 g Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 ll1lll1111111111111111$111 0439B
8 or treated as anchor points as discussed below and in Section 8.6.2. For the PAB and Turbine Building, the walls and structural penetrations are stiff compared to the support frames and piping. Piping and/or support stucture anchors will be taken at those attachment points as applicab le. The piping penetrations for the VC, however, are relatively flexible by comparison. For those penetrations, equivalent stiffnesses will be calculated. In all cases, the size and thickness of the reinforcing pad on the VC shell will be considered when calculating the equivalent stiffnesses. The relative stiffness of the piping section exiting the VC and the reinforcing pad will be considered in determining the local stiffness of the VC shell. The connection of the main steam and feedwater piping to the VC resenbles a nozzle penetrating a spherical vessel. The local stiffness of the nozzle attachment will be determined either by finite element method or using the formulas and figures provided in WRC Bulletin 49 [Ref. 4(ae)]. 8.6.2 Loading Conditions i ( The main steam /feedwater piping and support structure will be analyzed for deadweight, anbient wind, thermal, and site specific spectral loads (EC Spectra loads), including the translational anchor motions associated with each of the four load cases. The intent of this evaluation is to ensure the adequacy of the system to function during and af ter a seismic event to attain a safe shutdown. The thermal conditions for normal operation will be used in conbination with the applicable deadweight, anbient wind, and seismic loads for evaluation of the piping and support structures. Thermal anchor motions of the anchor points will be included as appropriate. l Yankee Nuclear Power Station 71 Seismic Reevaluation Criteria Li d. 6 A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 2r "' "'''''1 04398
As discussed in Section 8.6.1, the VC, PAB, and TB may be modeled using stbstructures or treated as anchors. If these buildings are modeled using stbstructures, the ground spectra will be applied at the base of the structure. If the building structures are treated as anchors, amplified response spectra will be generated for each of the anchor points in the system model in accordance to Section 5.3.3. PVRC damping (ASME Code Case N-411) will be used in analyzing the piping systems. Under PRC loads, the appropriate damping ratio for the VC, PAB, and TB is 7%. If these buildings are modeled using stbstructures, they will be stbjected to the ground spectra with PVRC damping. Since all these-structures have the fundamental frequency of less than 10 Hz, their damping ratio will be 5% for the first few modes and even lower for the higher modes. In addition, the effect of the relative displacements between buildings is accurately calculated. Therefore, modeling these buildings with the stbstructuring technique is very conservative. Seismic response spectrum analyses will be performed for the system. The { seismic load will be input as independent support motions (ISM) at the j appropriate anchor points (multi-level response spectrum method). The l ISM input may be used in conjunction with stbstructuring technique. More specifically, the ground spectra will be input at the base of the buildings modeled using stbstructures, and appropriate ARS will be input at the anchor points. As suggested in NUREG-1061, the response between the support levels will be conb ined by ab solute summation, and conb ination within a level will be performed by square-root-sum-of-the-squares. The Reg. Guide 1.9210% grouping method will be used to conbine modal effects. ll l Yankee Nuclear Power Station 72 l Seismic Reevaluation Criteria l d I ;i i. 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 llll11!!!!!!!!::""'"lllllll i 0439B
Translational seismic anchor motions will be considered for each building as appropriate and as specified in the Section 8.2 if the buildings are i I modeled as anchors for the piping and support structure. In addition to the building motions, relative seismic anchor motions of adjacent support frames will be considered due to seismic wave propagation. To determine i the proper differential motion to be applied to adjacent supports, the seismic wave will be propagated along the surface using a shear wave velocity of 1800 fps. Two real earthquake records (Taft and El Centro) i scaled to a ZPA of 0.199 will be used to calculate a maximum relative displacement between any two anchor points, as well as the distribution of the anchor motions between a series of supports. A series of maximum displacements between supports will be used in the analysis unless a reduced set of displacements can be justified. Response from the seismic , wave load case will be conbined with the standard seismic anchor motion ! i response using the SRSS method. 8.6.3 Acceptance Criteria Acceptance criteria for the main steam /feedwater piping will be as specified in Section 8.3. The standard ASME/ ANSI B31.1 equations will be used first in the qualification. In the initial , evaluation, the piping will be compared to the SEP criteria as I follows: l DW + Anbient Wind < Sh Eqn 11 i
< DW + PRC Spectra + Anbient Wind ( 2.4Sh Eqn 12 I
Thermal + TAM + SAM 4 Sa Eqn 13
- DW + Thermal + TAM + SAM + Anbient Wind ( Sa + Sh Eqn 14 t
Yankee Nuclear Power Station 73 Seismic Reevaluation Criteria A.d. k 1- 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 ll111llll1llllll11llI111111111 0439B l l
I 60% SAM may be used in Equations 13 and 14, provided that strain criteria were not used for Equation 12 at that point or at the points imediately adjacent to that point. In this case and at all ' l other times, 100% SAM will be used. The strain criteria and detailed code equation application criteria are specified in Section 8.3. These strain criteria and requirements apply to the seismic sections of the main steam and feedwater piping and to the non-seismic sections which have an effect on the response of the seismic piping or the frames which support the seismic piping, t I Yankee Nuclear Power Station 74 gg . Seismic Reevaluation Criteria liinnmuniiinhnen 8 3/8 6 '8 6 '86 64 c"- c- i Rav 4 !' 0439B
+
PAB m l MS/FW SUPPORT FRAME
.. N. .
i 4 . I it I e e a PIPING SYSTEMS WITH ORIGINAL FRAME PAB' DOTTED LINES REPRESENT g MASS AND STIFFNESS MATRICES CONNECTING PIPE ATTACHMENTS
\ N
\ \
\ \
I x '
\'/ _
's l
l [N - Or PIPING S STplS WITH ORIGINAL FRAME MODELED BY SUBSTRUCTURE I 1 Figure 8.1 Sibstructure Modeling Technique Yankee Nuclear Power Station 75 Seismic Reevaluation Criteria kNI J k A 80023/81060/81061/86064 Doc. No. OC-1; Rev. 4 111111111111111111111111111111 0439B l
9.0 PIPE SUPP(RT DESIGN CRITERIA 9.1 Introduction The following criteria shall be used to evaluate or redesign pipe supports at the Yankee Nuclear Power Station,
- 9. 2 Codes, Standards and References The following codes shall be used for the design of pipe supports:
9.2.1 ASME, Reference 3(g), Section III, Subsection NF, 1977 edition. 9.2.2 ANSI, Reference 3(a). 9.2.3 AISC, Reference 3(c). 9.2.4 ITT, Reference 4(ap). 9.2.5 Hilti, Reference 4(ad). 9.2.6 Other manufacturer's pd>l1shed catalogs.
- 9. 3 Loading Description All loadings obtained from piping stress analysis shall be used for support desi gn.
l Yankee Nuclear Power Station 76 Seismic Reevaluation Criteria L 'I J L A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 11llllll111111lll1111lll111111 04398
9.3.1 Normal Operating Loads 9.3.1.1 Dead Weight, D - includes all gravity loads, such as the weight of the pipe, its contents and insulation, the weight of supporting menb ers, and the loads due to steady internal pressure. Generally, the support self-weight is negligible when compared to the applied loads and is therefore not included. 9.3.1.2 Thermal, TH - Loads generated by restrained thermal expansion.
- 9. 3.1. 3 Thermal Anchor Movement, TAM - loads applied at supports due to thermal anchor movement.
- 9. 3.1. 4 Friction, FL - Friction loads are to be applied in the direction of thermal mo vement. Its magnitude shall be the friction coefficient times the sumation of the pipe dead load, thermal load and loads due to thermal anchor movement, 0+TH+ TAM, applied orthogonally to the direction of movement. The friction-coefficient for steel-on-steel shall be at least 0.3 and for Teflon 0.07. If friction loads cause support component.
o verstress, comparison will be made between the pipe thermal movement (Apipe) and the support deflection (asupt) due to friction loading. If o pipe<asupt, then friction loading shall be reduced by the ratio of o ipe/ p Asupt* 9.3.2 Emergency / Faulted Loads 9.3.2.1 Seismic Loads - The Yankee Composite Spectra will be used as the design event for supports. For piping systems evaluated to the IRC Spectra, pipe supports will be evaluated against a functionality criteria, as described in this Section. Values stained from SRSS (square root of sum of squares) of the 1 Yankee Nuclear Power Station 77 Seismic Reevaluation Criteria L ( Ji A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 111111lll11111111llll1llllll11 0439B
l l l individual responses of each of the three components of earthquake load shall be used to review the support. Seismic loads shall be considered to act in both the positi ve and negative restraining directions. 9.3.2.1.1 Seismic loads considered are (a) the Yankee Composite Spectra (YCS) or (b) the EC Spectra (RC). 9.3.2.1.2 Seismic Anchor Movement, SAM - Loads cbtained from seismic anchor movement of either YCS or EC (SAMYCS and SAMEC) shall be considered. For supports, the full SAM loads must be used, not 60% as allowed in the pipe stress criteria (Section 8.3.1-c)
- 9. 4 Loading Conbinations Load Case load 1 D + TH + TAM + FL 2 D + TH + TAM + YCS + S AMYCS 3
D + TH + TAM + EC + SAMRC Load Case 3 will be used for the confirmatory analyses ' and selected example analyses. Thermal and TAM loads will only be added to the load conb inations if they increase the magnitude of the design load. The larger of lDj and lD + TH + TAM l will be used as the deadweight + thermal contribution to the design load.
- 9. 5 Frequency For supports which are assumed to be rigid in the large bore piping analyses, the natural frequency of a seismic restraint with its tributary pipe mass must be greater than 33 Hertz in the pipe's restrained direction. The mass used to calculate the natural frequency shall include the weight of the restraint, Yankee Nuclear Power Station 78 Seismic Reevaluation Criteria LiI J i 1- 80023/81060/81061/86064 Doc. No. DC-1 ; Rev. 4 lllllllll111111lll1111lllll!!!
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restrained pipe, pipe insulation, fluid, pipe attachments, and valves. Any rational analysis may be used to calculate the natural frequency. The natural l frequency calculations of pipe restraints do not have to include the flexi-bility of the building structure. I For the purpose of determining the natural frequency of sntbbers and their frames, consider the sntbber to exhibit stiffness qualities which would make them a rigid link b etween the pipe and the supporting structure. The supporting structure, from the building's frame to the sntbber, shall be designed such that the natural frequency is at least 33 Hertz. For supports which are assumed to be rigid in the small bore analyses, either I stiffness or deflection criteria must be met. For supports checked to the stiffness criteria, stiffnesses must exceed 1500 lbs/in and 5000 lbs/in for piping less than or equal to 1-inch in diameter and piping greater than 1-inch in diameter respectively. For supports checked to the deflection criteria, support structure deflections due to the total design load at the point of restraint are acceptable if less than 1/16". Deflections greater than 1/16" are acceptable only if the total displacement is less than 1/8" for the design loading but less than 1/16" for primary loading. Deflection of nodes other than loading points can be larger than 1/8" provided that they are reasonable and that all stress limitations of the structure have been satisfied.
- 9. 6 Allowable Stress All pipe supports shall be evaluated using the methods, requirements, and allowables of the AISC Specifications for Steel Construction, including increases in the AISC allowables for faulted load conbinations, as prescribed by Appendix F of the ASME Boiler & Pressure Vessel Code, Section III. The normal load case 1 (see Section 9.4) shall be evaluated against the AISC 3 allowab les. The emergency / faulted load cases shall be evaluated using a 50%
increase in the AISC allowables up to a maximum stress of 0.9Fy. These emergency / faulted allowab les are acceptab le under the functionality { Yankee Nuclear Power Station 79 l Seismic Reevaluation Criteria i L iJa A- 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lilll!!!!!!!!"""""illlill 0439B o
requirements of Appendix A of USRC Standard Review Plan 3.9.3. Limiting the emergency / faulted allowables to 0.9F y ensures support functionality. For non-symmetric structural steel sections such as angles, principal axis properties shall be used for all evaluations. For anchor bolts and other catalog items, vendor-specified allowables shall be used with appropriate factors of safety. Table 9.6-1 summarizes -the component allowables used for the qualification of pipe supports at YNPS. For some catalog items, the allowables are given at elevated temperatures; however, the components may be used in supports stbjected to much lower design temperatures. In these cases, the catalog allowables may be increased by a ratio of material allowables at the design temperature versus the catalog temperature. When evaluations to the RC spectra are performed, they will be first done to the allowables discussed herein. However, when these are not satisfied, additional refinements will be considered and reevaluations performed. These refinements will include: a) Increased anchor bolt allowables. A safety factor of less than 4.0 but greater than 2.0 shall be permitted for interim operation. b) Increased anchor bolt allowables. A safety factor of less than 4.0 but greater than 2.0 shall be permitted for long term operation provided that
- 1) the base plate in question has at least four anchor bolts, and not more than half are simultaneously stb jected to tensile loads, l
- 2) loads greater than that associated with a factor of safety of 4.0 can be redistributed to adjacent supports, Yankee Nuclear Power Station 80 ,
Seismic Reevaluation Criteria l Li I J 6 A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 )
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- 3) Adjacent supports remain elastic and, if appropriate, meet the factor of safety of 4.0 under the new load, and
- 4) the anchor bolts meet visual and ultrasonic (for enbedment) requirements.
c) Conbinations of inertia and SAM loads by SRSS. Application of refinements a) and b) will be justified on a case-by-case basis. 1 l
- I l Yankee Nuclear Power Station 81 Seismic Reevaluation Criteria 8 '8 6 '8 685 64 ' " - -"""4
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1 T&le 9-1 i Allow &le Stresses for Pipe Supports i SERVICE LEVEL
. STRESS LOAD CASE 1 LOAD CASES 2 AND 3 Value KSI(l) Value KS1(l)
Tension 0. 6 Fy 19.8 0. 9 Fy 29.7 i j Shear 0. 4 F y 13.2 0.6 Fy 19.8 We Crippling 0.75 F y 24.8 0.9 F y 29.7 Compression F a Smaller of 1.5 Fa or 2/3 F cr 1 Bending 0.6 F y 19.8 0.9 F y 29.7 Bearing 0. 9 F y 29.7 N/A NA Bolts Tension Allowele Tension 1.5 X (Allowable Tension per AISC per AISC) Shear Allow 2 1e Shear 1.5 X (Allowele Shear per AISC per AISC)
- Anchor Bolts Hilti (2) (2)
Star Slug-in (3) (3) Through Bolts (4) (4) i Toggle Bolts (5) (5) l l Yankee Nuclear Power Station 82 l , Seismic Reevaluation Criteria l L 1 L A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 '
- 11lll1lll111m1111111111lll11 1 0439B l . - . . _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , . , _ _. _.__ . _ . . . - _ . . . _ - . - _ . _ , . .
Table 9-1 (continued) SERVICE LEVEL STRESS LOAD CASE 1 LOAD CASES 2 AND 3 Value KS I Value KSI Welds Shear 0.3 F y 21.0 0.45 F y 31.5 (Fillet) (Weld Metal) (Weld Metal) 0.4 Fy 13.2 0. 6 Fy 19.8 (Base Metal) (Base Metal) (Full or Partial Tension 0.6 Fy 19.8 0.9 F y 29.7 Penetration (Base Metal) (Base Metal) Conbined Stress Per AISC Per AISC Catalog Items Catalog Values Catalog Values (6) NOTES: (1) The numerical allowable values shown above are for A33 steel only. For other types of steel, the allowables should be obtained per the AISC manual. (2) The allowable Hilti expansion anchor loads shall be obtained using 1/4 of the a verage ultimate tensile and shear loads shown in Tab le D-2 of Appendix D. The allowable loads for shell-type anchors shall be obtained using 1/5 of the average ultimate tensile and shear loads as specified in the vendor catalog. An interaction ratio will be computed considering the computed tensile and shear loads (T and V, respectively) and the allowable i tensile and shear loads (TA and V g , respectively) as follows
)
Yankee Nuclear Power Station 83 Seismic Reevaluation Criteria L IJ6 A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 Eiiililllil 04398 l
(TT )5/3 +(VV )5/3 < 1. 0 A A Additionally, the applied shear, V, shall not exceed 40% of the tensile allowab le. If the applied shear exceeds the 40%, the exponent shall be reduced from 5/3 to 1. Bolt spacing and edge distance shall be considered per vendor requirements. (3) Allowable loads for Star Slug-in anchors shall be -as given in Table 9-2. These values are based on information provided in Ref.4(aq). (4) Pipe supports shall be attached to masonry walls by either through bolts, toggle bolts or other types of connectors developed specifically for use with hollow masonry. Through bolts are preferrable and will be used where practicab le. Allowables for through bolts shall be per the AISC Code, 8th Edition. (5) Pipe supports shall be attached to masonry walls using toggle bolts if through bolts are not practical. Allowables for toggle bolts shall be limited to manufacturer values. (6) If catalog allowables are given for emergency / faulted loading conditions, they will be used instead of the allowables calculated here. i
- I I
f Yankee Nuclear Power Station 84 Seismic Reevaluation Criteria A ( )i A. 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
;;;:"""'!!!!!!!llilillllll!
0439B
Te>le 9-2 Allowable loads for Star Slug-in Anchors Anchor 00,in Tension Shear Allowdale (lb ) Allowdale (lb ) 1/4 400 200 3/8 1,000 500 1/2 2,000 2,000 5/8 3,000 1,200 3/4 5,500 (4) 4,800 7/8 6,200 (4) 3,800 1 7,000 (4) 4,200 Notes: (1) These values are based on information provided in Ref 4(aq) (2) Allowdyle loads assume that the recommended nud)er of units were used. (3) Allowdales are based on a Safety Factor of 4 (4) Tension allowables for 3/ 4" 4, 7/ 8" 4, and 1" 4 anchors are based on 4,000 psi concrete. All other values based on 3,000 psi. i I Yankee Nuclear Power Station 85 ) , Seismic Reevaluation Criteria j L LJa A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 ' 1111lllllllll1111llllll1111111 l 0439B
- APPENDIX C
- Figur -1 STRUCTUML ANALYSIS FLOWCHART s, =
o = === es.o.l 9.. 4m.m.ni 80erts.ne.1 i .' u. . .T."..' i S.wir.
. l
,as
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4 ( mgy I ArtWe... l
- s. 7 N i r La... sen ... -
'T 7.'"'
i v miwr
; A .,.= i I aa.'r.= \ Lu.
I 1 es.as , !
' s. _ e .e.
i
- s. o.s ser... l l seww. l mL hs.s '
i
**ema j
l i s.. i i ev.n.e. . i i 1 i t i ll o.cer.= po.e enssw on j I w I . I s e.ai i- 1
! Yankee Nuclear Power Station l l
i L -b Seismic Reevaluation Criteria a lullteillfilliililllilitilill 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 C-1 l t 4
. .,. - , . < - . - - , - - --w--------'*"*"-*"'" ' - ~' *
APPENDIX C FIGURE C-2 LARGE BORE PIPING STRESS ANALYSIS FLOWCHART Criteria Development and Procedure Preparation v Digitiration and Enveloping of Spectra e f Prepare Computer - p_____1______ Determination of ,i Model and Stress T REVISE 'l Pipe Support Isometrics
~
MODEL [, Stff,,f,ngs,s_,,,,,jl e Perform Stress Analysis
- Weight
- Thermal
- Internal Pressure
- Seismic
- Anchor Movement SUPPORT OVERLOADED
- Static-Seismic OR UPLIFT (if required) e CVER STRESSED r _______..________ ,
Perform Stress Check - ADD SUPPORT i Analysis and Review 8 and Summarize loads of Pipe Supports 8 a L.(Include Local Stresses) J - RESULTS 0.K. '
- ~ '
SUPPORTS 0.K, ,
~
'f (FIX UPLIFT /0VERLOADED -
SUPPORT) Complete Calc. documentation and Q.A. Requirements l l-MA 11111111lllllllll111lll1111lll Yankee Nuclear Power Station Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 C-2 i l ,
APPENDIX Dr TABLE D-1 MATERIAL PROPERTIES ) FOR STRUCTURES i MATERIAL 5 tu1LDl4G 1.0. STRUC TUe 4L ST((L COntatit RE!W0ecimG sit (L SOIL ==
- 1. Diesel Gen. ASTM All & A 305 8 tst tidg. 8 Accum. ASTM A7 (Fy = 33 kst) f' = 3.000 pst. Int. Gr. 10.6 tsf (utth W or t)
Tank Enclosure (Fy = 40 ksi)
- 2. Turbine building A$TM A7 (Fy = 33 tsi) A) Footings & Grade 8 tsf a Pedestal Seems, f' = 2500 pst. 10.6 ksf (with W or ()
B) Precast EMS & Wall
$nteld, f' = 2500 pst. ASTM A15 & A 305 C) All Other' Cast In. Int. Gr.
- place, f' = 3000 pst. (Fy = 40 ksi)
D) Turbine bpport Met
& Pedestal, 3000 pst.
f"' =
- 3. Spent Fuel Poc1/ . A$TM A7 (Fy = 33 tst) f' = 3000 psi ASTM All & A 305 8 tsf Spent Fuel Chute Int. Gr. 10.6 ksf (with W or ()
(Fy = 40 kst)
- 4. Steel vapor A) Plate Material. ASTM 20 ksf Container A.300. Class A.201 Grade B. Fy = 32 tst.
B) Steel Colves A$TM A.283. Grade C Fy = 30 tst C) fle Rod Asse=ely f ' = 3000 ps t ASTM All & A 305 All! C.1020 8 4320 (hedestal and Int. Gr. Fy = 30 kst Footings) (Fy = 40 tsi) p-( Assuae ta'e as colues) /y
- 0) Base Plate .s ASTM A284. Grade 8. FT=27 151 ,-
() Anchor Bolts e y' Fg = 20 att (Fye32:st) f F, = 10 kst .' F g 20-tsi'{$.S.) p 25 ksi (0.5.) '- Reactor Support Structure ASTM A.300. Class 4 201 Grade B. Fy = 32 tst A) Footing and Grade SMS. f' = 3000 pst 20 asf 3)Pedestafs.Cols. ASTM All & A 305 Walls 8 All Others Int. Gr. f' = 4000 pst (Fy = 40 ksi)
- 6. Primary Aun. ASTM A7 (Fy = 33 tst) f,' = 3000 pst ASTM All & A 305 8 tsf Sutiding sad Int. Gr. 10.6 tsf (with W or ()
Radioactive (Fy = 40 kst) Tunnel
- 7. MS/FW Support ASTM A7 (Fy = 33 ksi) f' = 3000 pst A$7M All & A 305 8 tsf e Structure Int. Gr. 10.6 tsf (with W or ()
(Fy = 40 4:1) Continued next sheet Yankee Nuclear Power Station Seismic Reevaluation Criteria Nb k A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 D Mlll111111118lll18lll111111 (
) TABLE D-1 (Continued)
- 8. Modifttattent* ASTM AM (Fy = 36 tst); Mtitt hatt Solts A$fM A615 8 tsf E7011 Electrode; Mud Met f', a 2000 pst Grade 40, 60 10.6 tsf (with W or E)
Solts: ASTM A3N. or as Duct tant f'g *=3000 pst otherwise spectfted All Others f g m pst ]
- 9. Fire fast A) Plate Material. A5fM f"' = 4000 pst ASTM A615 8 tsf A.283. Grade C Grade 60 10.6 tsf (with W or E)
I (Fy = 60 kst)
- 8) Stacing System
)' A 36 , C) Anchor Bolts Al$1 1141 1
- Appitcable to all structures.
** See discussions in Section 5.1.4 l
l r j l i 1 Yankee W clear Power Station . 4 eg[ g , Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 D-2 Illillilililllilillillimmi
APPENDIX D: TABLE D-2 ALLOMABLE LOADS FOR HILTI 90LTS ** KWIK-80LT AVERAGE ULTIMATE TENS 1LE & SHEAR LOADS
- Concrete Strenoth 2000 PS! 4000 PSI 6000 PSI Diameter Embedment Tension. Shear Tension Shear Tension Shear 2 3/4" 5410 11198 6600 11562 7700 13500 5/8" 6250 11198 9100 11562 9560 13500 3 1/2" 11198 12000 11562 14500 13500 4 1/2" 7000 13378 14300 15437 20300 15437 5 1/2" 7550 61/2" 8025 13378 16000 15437 2iO00 . 15437 13378 17000 15437 21000 15437 7 1/2" 9000 3 1/4" 13257 10150 17133 10860 18102 3/4" 8155 4" 9700 13257 13400 17133 13700 18102 5' 11700 13257 16500 17133 17600 18102 6' 13800 15195 18000 18466 22500 21009 7' 15800 15195 21000 18466 23600 21009 8' 16000 15195 23000 18466 23600 21009 9" 16000 15195 23500 18466 23600 21009 l' 4 1/2" 14000 27355 16000 26879 20500 32112 5' 15500 27355 18900 26879 23441 32112 6' 17600 27355 23441 26870 23441 32112 7" 18200 27355 23441 26879 23441 32112 8' 18200 27355 23441 34491 23441 363e4 9" 18200 27355 23441 34491 23441 36304 10" 18200 27355 23441 34491 23441 36394 1 1/4" 5 1/2" 19000 36750 23000 35680 31200 45195 6 1/2" 21600 36750 27100 35680 36500 45195 7 1/2" 23600 36750 31100 35680 42000 45195 8 1/2' 25100 39843 34600 35680 44400 47098 91/2' 26200 39843 37800 35680 44400 47098 10 1/2" 26800 39843 40900 35680 44400 49596 Notes For 1/2" Biltis consult NRC IE Notice 86-94.
- Tension values obtained from best fit curve through mean values of test data. Curves and test data contained in A. A. Hanks Report No. 8784 (HILTI No. TR.111 A).
Shear values are minimum mean values at each embedment based on fatture across threaded section of the anchor.
~~ ** The allowable loads should be 1/4 of the ultimate loads IIsted above.
Yankee Nuclear Power Station ( sigillilillililliffillllill Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 D-3
APPENDIX D TABLE D-3 DAMPING VALUES
- Stress Level Type. and Condition Percentage of Structure Critical Damping Working stress, a. Vital piping See Section 8.2.4 no more than about i b. Welded steel, prestressed 2 to 3 yield point concrete, well reinforced concrete (only slight cracking)
- c. Reinforced concrete with 3 to 5 considerable cracking
- d. Bolted and/or riveted 5 to 7 steel, wood structures
- with nailed or bolted joints.
At or jUst below a. Vital' piping See Section 8.2.4 yield point b. Welded steel, prestressed 5 to 7 concrete (without complete loss in prestress)
- c. Prestressed concrete with 7 to 10 no prestress left '
t
- d. Reinforced concrete 7 to 10
- e. Bolted and/or riveted steel, 10 to 15 wood structures, with j bolted joints
!g f. Wood structures with nailed 15 to 20 joints
- Source: NUREG/CR-0098 (except piping) - '
Yankee Nuclear Power Station 7
........t.3 Seismic Ree valuation Criteria->i ~ ~ . ~ <~ ~ D-4
i l i ! TABLE D-4 MATERIAL SPECIFICATIONS F(R MAJ(R ECHANICAL EQUIPMENT EQUIPMENT MATERIAL SPECIFICATION , Reactor Pressure Vessel (RPV) Head, Vessel Shell & Bottom Carbon Steel, SA-302 Grade B Bolting Flange Carbon Steel, SA-105 Grade 2 Closure Studs Carton Steel, SA-193 Grade B16 Vessel Support Carton Steel, SA-212 Grade B Steam Generators Head, Shell & Bottom Carbon Steel, ASTM A212 Grade B Tihes Stainless Steel, ASTM A-212, Type 304 Pressurizers Heads & Shell Course Carton Steel, SA-302 Grade B Vessel Support Carton Steel, SA-212 Grade B Main Coolant Pumps Casing Stainless Steel, Type 304 Main Coolant Loop Iso. Valves Valve Body Stainless Steel, SA-351, Type 304, Grade CF8 Yoke Carbon Steel, SA-216, Grade WCA (Assumed) Bonnet Stainless Steel, Type 304 Main Coolant Pump Disch, Check Valves Valve Body Stainless Steel, SA-351 Grade CF8 Cap Stainless Steel, Type 316 Studs ( 20 per valve) Stainless Steel, Type 316 Main Coolant Bypass Piping Iso. Valves Valve Body Stainless Steel, ASTM A351, Grade CF8M Yoke Carbon Steel, ASTM A216, WCB Yoke Studs Carbon Steel, ASTM A193 Grade B7 Pressurizer Spray Valves Valye Body & Stem Stainless Steel, Type 316 J Main Steam Non-return Valves Valve body & Bonnet Carbon Steel, ASTM A216 Grade WCB Studs Carton Steel, ASTM A193 Grade B7 l Main Steam Code Safety Valves l Valve Body Carton Steel, ASTM A216 Grade WCB i 1 Yankee Nuclear Power Station D-5 Seismic Reevaluation Criteria
.....d.1...1.. 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 111.... . ..n.. ...
TABLE D-4 MATERIAL SPECIFICAT10115 F(R MAJ(R ECHANICAL EQUIPENT (Continued) EQUIPMENT MATERIAL SPECIFICATION Main Feed Check Valves Valve Body & Bonnet Carbon Steel, ASTM A216 Grade WCB Bonnet Studs Carton Steel, ASTM A193 Grade B7 Pressurizer Relief & Block Valves Valve Body & Bonnet Stainless Steel, ASTM A351 Grade CF8M Bonnet Studs & Eye Bolts Carbon Steel, ASTM A193 Grade B7 Loop Isolation Valves Valve Body & Bonnet Stainless Steel, ASTM A351, Grade CF8M Yoke Carton Steel, ASTM A216 Grade WCB Yoke Studs Carbon Steel, ASTM A193 Grade B7 Loop Isolation Check Valves Valve Body and Cover Stainless, ASTM A182 Type 316 Cover Studs Carbon Steel, ASTM A193 Grade B7 Shutdown Cooling Iso. Valves Valve Body & Bonnet Stainless Steel, ASTM A351 Grade CF8M Yoke Carbon Steel, ASTM A216 Grade WCB
, Yoke & Bonnet Bolts Carton Steel, ASTM A193 Grade B7 Feed & Bleed Heat Exchanges Shell Stainless Steel, ASTM A351 Grade CF8 Tth es Stainless Steel, ASTM A213, Type 304 l
l 1 4 1 Yankee Nuclear Power Station D-6 4 Seismic Reevaluation Criteria IIlii;;,,, :::::::.:::lll 111 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
1 l APPENDIX E BUCKLING CRITERIA FGt VAPOR CONTAllER COLUMNS The critical buckling stress of the VC columns will be calculated according to AISC Specification [Ref. 3(c)]: r ] Fy for (kt/r) < C c *} F p = [1 - c nE (E.2) F cr
= for (kt/r) > C c (ki/r)2 where (kt/r) is the effective slenderness ratio of the column and C
c =(
) / . Equation E.2 is also known as Euler's buckling formula.
Fy For slender columns p( kt > C ), the buckling stress is not sensitive to the e strength of material, residual stresses, small eccentric loading and geometric imperfection. Therefore, Euler's formula (Eq. E.2) will be used without any reduction. The columns in the intermediate slenderness range (such as VC columns with kt/r = 44.5) are most sensitive to residual stresses and eccentric loading. Short columns (suen as VC columns with k1/r = 14) are affected most by variation in the strength of material and eccentric loading. The effects on critical buckling stress due to residual stresses, > geometric imperfection and eccentric loading will be discussed as follows: The reduction in the critical buckling stress due to residual stesses can be theoretically corrected by replacing the elastic modulus, E, with tangential modulus, E t
, as plotted by the dashed curve in Fig. E.1. As shown in Fig.
E.1, the solid curve representing Eq. E.1 is considerably lower than the dashed curve. The difference between these two curves represents the additional reduction in critical buckling stress due to georaetric Yankee Nuclear Power Station E-1 Seismic Reevaluation Criteria { 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 IIlllillfilililllllIlllililli
imperfection. The test data shown in Fig. E.1 indicates that Eqs. E.1 and E.2 are conservative. The slenderness ratios of the VC columns are also indicated , in Fi g. E.1. The allowable buckling stress to be used in evaluating the VC columns for load conbinations without EC spectrum follows Part 1 of AISC specification as discussed in Section 5.4.2.1. The allowable buckling stress to be used for the load conbination including MC spectrum is 0.95 times the value calculated by Eqs. E.1 or E.2. As shown in Fig. E.2, this reduction factor can cover the column eccentricity up to 1.0 inch. In addition to these conservative factors, additional justification for using 0.95 is discussed as follows: A. If the stress developed in the diagonal tie rods is below allowable, the VC is adequately braced against sidesway. The value k A/r can be calculated using k=1. A more appropriate value of k=0.8 as recommended i in Table C1.8.1 of Ref. 3(c) can also be used in this case due to the end restraining conditions. l4 B. The Fy specified is typically less than the actual. i C. If the axial compressive stress developed in the columns is small under YCS or EC Spectrum loads, ductile behavior of the columns can be 1 guaranteed. 1 D. 16 columns allow for stress redistribution. f Therefore, it can be concluded that the criteria described in this section will not overestimate the true buckling stresses of the VC columns and the integrity of the VC will not be jeopardized even if the buckling stresses are
- slightly overestimated.
- I Yankee Nuclear Power Station E-2 g Seismic Reevaluation Criteria y ........ , ,, 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
FIGURE E-1 e s ,
- l. g 6 a LEGEND: c I - Beam Fy = 38.5 ksi o
* ,a A H - Shape F = 40 ksi 1.0 e Y
\ '
o Mild Steel Rod . N ' e \ p , y2 Et 9 = 0.282", Fy=39.2 ksi ( Kl/ r) Z Proportional Limit
- 1 -
0.75 l 3 F er
=
[1 IEI/I)2 C Y 2~
\
]F t c d'
(F y = 39 ksi) n
. m .s 0.5 - - - - - - - - - - - - - - - - - - 6 o
C I I l l
$ kl/r of VC Columns .
I 0.25 - F cr =7E (Fy=39ksi) (K1/r)2 , !! I I i l 1 f e i 1 C 200 50 100 c 150 (K1/r) COMPARIS0N OF BUCKLING STRESS 4 i
= = = ==
Yankee Nelear Power Station r - , Seismic Reevaluation Criteria E-3 Li ( ) 6 A5 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 Ill:: '"":::::::
FIGURE E-2 ECCENTRIC LOADING y y F , cr =0.95[1-{c/rg)p,
,._7 s 30 s
'c2 = 0.1 r
(From Timosheni o) 20 4 I i s" 5 ! l l F cr
= 0.95 2E (K /r)'
I 50 100 c 150 200
- c
( _) Fy = 36 ksi and E = 29,000 ksi For the columns of SVC, 0.D. = 42", t = 7/8", I = 23410 in d A = 113.05 in2, r = 14.543", and c = 21" ('C/r 2) = 0.1 + e = 1.01" 1 See Reference 4(z) l 2== Yankee Nelear Power Station g~27"V Seismic Reevaluation Criteria
~
E-4 r8 L-Q
- 6. J a al 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 i lililllllllllilllllililllillli l
APPElWII F DESIGN CRITERIA FOR SAFE SHUTOOWN SYSTEM BUILDING AND BURIED PIPING YANKEE ATOMIC ELECTRIC COMPANY FRAMINGHAM, MASSACHUSETTS i-I ! Yankee helear Power Station Seismic Reevaluation Criteria 4( bliftiHfelillisillHilIlle 80023/81060/81061/86064 Doc. fio. DC-1; Rev. 4 F-i
Table of Contents Section Pa ge 1.0 GE NER A L . . . l. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 2.0 S C O PE OF W0R K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1
- 3. 0 CODES AND S TA ND AR DS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-2 4.0 R EF ER E NCE DOC UME NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-2
- 5. 0 MAT ER I AL PR O PER T I ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-3 6.0 LOADS AND LOAD COMB I NAT IONS . . . . . . . . . . . . . . . . . . . . . . F-4
- 7. 0 ANALYS IS AND DES IGN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-10 i
l 1 1 Yankee Nuclear Power Station seismic Reevaluation Criteria LN k ) k & 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 F-il lueilleillfilllilll#llitelli l
1.0 EIERAL This design criter;ia provides the technical basis for the analysis and design of the safe shutdown system structure, and its associated buried conduit, buried piping, buried storage tank and equipment foundations for the Yankee Nuclear Power Station at Rowe, Massachusetts. The safe shut-down system building will provide an alternate means of safely shutting down the plant in an emergency condition. 2.0 SCOPE OF W(RK This criteria will cover the design of the safe shutdown system building and the following associated components. 2.1 The foundation pads and/or anchor bolts for the: f
- a. Primary make-up pump i
- b. Secondary make-up pump i c. Motor control center
- d. Diesel Generator set
- e. 500 gallon boron addition tank.
- 2. 2 The underground (buried) piping from:
- a. The primary make-up pump discharge to the charging system tie-in. This 2" 4 schedule 160 line to the building penetration at the primary auxiliary building.
- b. The secondary make-up pump discharge to the steam generator blowdown piping. This 2" ( schedule 80 line to the building penetration at the primary auxiliary building,
- c. The secondary make-up pug recirculation to TK-55. This 1" 4 recirculation line to the nozzle at the tank.
- d. The secondary make-up pug suction from TK-55. This 4" (
i schedule 40 suction line to the nozzle at the tank. I
- e. The 4000 gallon storage tank to diesel day tank. The 1" 4 schedule 40 suction and recirculation lines.
All the piping discussed will be designed to the safe shutdown building penetration. The routing and design inside the building is I not covered by this appendix. l Yankee Nuclear Power Station F-1 g[ , , Seismic Reevaluation Criteria j bilillilllllilllillilillllli 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 m ,_ .
(
- 2. 3 The underground (buried and encased in concrete) conduit from the safe shutdown structure to the emergency diesel generator building.
4 From there it will be run above ground to the vapor container and a will not be designed as part of this appendix. i 2.4 The underground 4000 gallon fuel oil tank'.
; 3.0 CODES AIS STAlettDS' j The following codes and standards apply to the appropriate sections of
] this document.
- a. American National Standard Code for Pressure Piping. ANSI B31.1, j 1977.
- b. American Institute of Steel Construction (AISC), " Specification for i the Design, Fabrication and Erection of Structural Steel for
; Buildings", 8th Edition.
- c. American Concrete Institute of Steel Construction (AISC), "Cuilding Code Requirements for Reinforced Concrete" (ACI 318-83).
- d. American Welding Society (AWS), " Structural Welding Code", D1.1-75.
- e. U. S. Nuclear Regulatory Commission (PRC), " Standard Review Plan for l the Review of Safety Analysis Reports for Nuclear Power Plants",
j NUREG-0800, Section 3.7, Washington, D.C., Office of Nuclear Reactor ' Regulation, July,1981.
- f. American Concrete Institute (ACI), " Code Requirements for Nuclear
- Safety Related Concrete Structures" (ACI 349-76), including Supplements.
4.0 REFERENCE DOCLMENTS l 1 The following reference documents shall be used in carrying out the l structural analysis and design efforts: , i
- a. Yankee Atomic Electric Company " Final Hazard Summary Report",
Yankee Nuclear Power Station, Rowe, Massachusetts.
- b. Stone & Webster Contract Drawings for Yankee Nuclear Power Station.
{ c. Weston Geophysical Corporation, " Geology and Seismology, Yankee Rowe j Nuclear Power Plant", January 29, 1979, i i Yankee Nuclear Power Station ' F-2 Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 Hilli ilH H Nililill 1
l 1 l
- d. " Seismic Design of Buried Piping", Iqbal and Good 11ng, 2nd ASCE i Specialty Conference on Structural Design of Nuclear Plant Facilities, December 1975.
- e. " Seismic Analysis of Underground Structural Elements" Shah and Chu, Journal of the. Power Division, July 1974
- f. Corner and Lada, " Supports and Restraints for Piping and Equipment".
Catalog dated 1977
- g. "More on Flexibility Analysis of Buried Pipe", E.C. Goodling, ASME
~
Paper 80-C2/PVP-67 5.0 MATERIAL PROPERTIES The following material specifications shall govern.the design: Concrete f'c = 3000 psi Structural Steel ASTM A-36 Reinforcing Steel ASTM A-615 Grade 60 Conduit 4" 4 rigid steel conduit Piping Primary make-up pump to charging system tie-in: 2" 4 schedule 160 stainless steel A-376-TP-304 Secondary make-up pump to steam generator blowdown piping: 2" 4 schedule 80 carbon steel A106 Grade B Secondary make-up pump recirculating to TK-55: 1" 4 schedule 40 carbon steel A106 Grade B Secondary make-up pump suction to TK-55: 4" 4 schedule 40 carbon steel A106 Grade B The 4000 gallon storage tank to diesel day tank piping: 1"4 schedule 40 carbon steel A106 Grade B 1 l
=-
Yankee Nuclear Power Station F-3 r4y l[ j Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l hi!!!!!!!"'""""::!!!I!!! .
Buried Storage Tank Fiberglas Doors Commercial grade hollow metal Lou vers Commercial grade steel or aluminum
- 6. 0 LOAD AND LOAD COW IMATIONS 6.1 Concrete Structures 6.1.1 Loads D Dead load including all permanent construction such as piping, pipe supports and support structures L Live load and its related internal moments and forces will be included in the design H Loads due to earth pressure will be included in the design. Soil lateral load factors to be used in the analysis shall reflect the physical properties of the soil. Hydrostatic loads due to ground water table will be considered in the design (both static and dynamic).
W Wind loads resulting from the design wind velocity for the site will be included in the design. E Earthquake loads and effects as defined by the Yankee y Composite Spectra inertia forces will be considered in the design. T Tornado loads and missiles which could be generated ; due to tornado will be considered in the designs. 6.1. 2 Load Combinations l
- 1. U = 1.4D + 1.7L
- 2. U = 1.4D + 1.7L + 1.9E
- 3. U = 1.4D + 1.7L + 1.7W lf 4 U = 1. 2D + 1. 9E
, 5. U = 1.2D + 1.7W i
- 6. U 1.00 + 1.0L + 1.0T I .
E Yankee Nuclear Power Station F4 r - Seismic Reevaluation Criteria Pj 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 l h.eg {ga 1m..i.m..imdI.gigiidig lT
Resistance to lateral earth pressure H is included in the design, required strength shall be at least equal to U = 1.4D + 1.7L + 1.7H except where D or L reduce the effect of H, 0.90 shall be substituted for 1.4D and zero value of L shall be used to determine the greatest required strength U. 6.1. 3 Minimum Factors of Safety for Foundations In addition to the load combinations specified above, the following load conbinations shall be checked against sliding and overturning due to earthquake and wind. <
- a. D+H+E
- b. D+H+W Minimum Factor of Safety Load Combination Overturning Sliding
- a. 1. 5 1. 5
- b. 1.5 1.5 6.2 Piping 6.2.1 Loads TH Thermal load due to steady state temperature effect W Load due to pipe weight, contents and insulation P Pressure load due to steady state internal pressure T Tornado loads and missiles which could be generated due to tornado will be considered in the design E Load due to earthquake inertia forces as defined by the Yankee Composite Spectra 6.2.2 Load Combinations
- 1. Sustained Loading Stress Evaluation PD 751 M 4t Z h
- 9' n
Yankee Nuclear Power Station F-5 T4y 5 [] Seismic Reevaluation Criteria - 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 hipumydI qinni, l
where P = internal pressure, psig Do = outside diameter of pipe, inches
'tn= nominal wall thickness of pipe, inches
'M, = result moment loading on cross-section due to-weight and other sustained loads, inch-lbs Z= section modulus, inches 3 i= stress intensification factor. The product of
.751 shall not be taken as less than 1.0.
Sh= basic material allowable stress at maximum (hot) temperature; from ANSI B31.1 allowable stress tables.
- 2. Occasional 1.oading Stress evaluation PD
+
751 M,
+
.7514 4t Z Z ' KSh ( ANS I B31.1 Eq. 12) n Terms are as previously defined, except K = 1.8 for evaluation of E loading Mb= resultant moment loading on cross-section due to earthquake inertia only. Moments shall be calculated statically for above ground piping using the peak ground response factored by 1.5 to account for higher modal contributions.
- 3. Additive Stress (thermal) Evaluation iM Se= <
S, W B31.1 Eq. 13) 7 Terms are as previously defined, except M c
= range of resultant moment on the pipe cross-section due to thermal expansion of the system and thermal anchor movements. Consider moment Yankee Nuclear Power Station F-6 r '
Seismic Reevaluation Criteria (,4 {g-] h,,[,,,],,,, 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
effects of earthquake anchor displacements as full range and combine with thermal moments. Sa= the allowable stress range for thermal expansion stresses, where: S, = f (1.25Sc + 0.25Sh ) S c
= basic material allowable stress at system (cold) temperature from the ANSI B31.1 allowable stress tables.
f= stress range reduction factor for cyclic conditions which will be assumed as 1.0. 4 Sustained Plus Thermal Expansion Stress PD .751M iM
+
7
+
7
< (S +
S,) (ANSI B31.1 Eq.14) terms as previously described. Use only if Eq. 13 fails.
- 5. As piping will be buried the following additional provisions shall be satisfied. The buried piping shall be analyzed for seismic stresses at penetra-tions, elbows, and straight sections using static methods.
Straight Sections Axial and bending stress shall be calculated using the following equations. EV EV EV Axial Stress = or or 2c, c c p era 0.3849 era era Bending stress = 2 or OP ', 2 2 c, c c Yankee Nuclear Power Station F-7 rg,2 i f 3 Seismic Reevaluation Criteria l lillllllillililifilllil$lil80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
where E = pipe modulus of elasticity R = pipe outside radius V, = maximum ground velocity am = maximum SSE acceleration
- cs.cp,cp = Shear, Compression and Rayleigh wave velocities, respectively Axial and bending stresses shall be conbined absolutely to represent the total seismic stress.
This combined stress included in ANSI B31.1 Eq.12 for code stress evaluation. Elbows Axial and bending stresses shall be calculated as follows: Axial force = S=c mAE - fL Axial stress = A S Bending stress = 3AZ where, em = maximum soil strain A= area of cross section, pipe E= pipe modulus of elasticity f= slip resistance L= effective slippage length A= beam parameter Z= pipe section modulus Axial and bending stresses shall be combined absolutely (total seismic stress) and included in ANSI B31.1 Eq.12 for code stress evaluation. (
=
Yankee Nuclear Power Station ' F-8 Seismic Reevaluation Criteria r4 Amnk'g i ysM..am23f 81osoisiosifasou ooc. No. oC-1; Ro. 4
t l l-Penetrations Piping stresses shall be calculated based on penetrations acting as anchors.
. External Pressure loading The requirements of ANSI B31.1, Para.104.1.3 shall be statisfied in addition to those of Section 7.3.
6.3 Conduit As the conduit will be encased in concrete and buried, the loads and load combinations for concrete structures will apply.
- 6. 4 Storage Tank 6.4.1 Loads H Loads due to earth pressure. Hydrostatic loads due to ground water table.
E Earthquake loads and effects as defined by the Yankee Composite Spectra inertia forces. T Tornado missile protection requirements. L Live load forces.
-, F Internal fluid forces.
6.4.2 Load Combinations H+L+F < S I H+L+FiE < 1. 5S H+L+F+T < 1. 5S S is the allowable stress in structural steel
-I sections, plates, bolts and welds permitted by the AISC code, excluding the provisions of section 1.5.6. In no case shall the allowable stress exceed
.9Fy.
Yankee Nuclear Power Station F-9 e Seismic Reevaluation Criteria 8"23/82o5o/82oS2/85o aac "' "c-2 ""
- innnig4 5 y }rgdiMin
- 6. 5 Equipment Foundations 6.5.1 Loads D , Dead load of equipment E Earthquake loads and effects as defined by the Yankee Compose Spectra The vibration characteristics of the equipment will also be considered in the design of the foundations.
6.5.2 Load Conbinations The concrete foundations will use the same combinations as previously described. 6.6 Doors and Louvres The design will account for both wind and tornado generated missiles.
. 7.0 ANALYSIS AND DESIGN PROCEDtRE 7.1 Safe Shutdown Pumphouse Structure This structure, composed of concrete, will be designed for the following loads.
D The weight of a concrete slab, roof, walls and future attachments. Roof L 50 psf snow load 4000# concentrated loads placed over the pumps for servicing. Ground Slab L 300 psf I l
~
Yankee PtJclear Power Station F-10 Seismic Reevaluation Criteria rg' p t [ 3 8 " 23/82060/8 052/8505* =c ,
- oc-2 5 ""
- kunnn{innenen i l
General Structure W 110 mph normal wind H Lateral earth pressure. Ground water maximum level will be assumed at grade. 1 E The site ground response spectra (YCS ground) will be used. The vertical direction will be 2/3 of the horizontal motion spectra (see Attachment A). T 85 nph Tornado wind velocity. The missile design is based on a 2x4 wood plank weighing 20 pounds at 75 ft/sec. The applicable load conbinations from Section 6.0 will be applied against normal allowable values from the codes.
- 7. 2 Buried Conduit The conduit will be encased in concrete and buried. The design of this duct bank will account for:
L Truck load AASHO H-20 or 100k cement truck These conditions and others previously mentioned will be combined and applied against normal allowables.
- 7. 3 Buried Piping
, The piping covered by this criteria will be directly buried.
Additional external loads which will be accounted for are: L Truck load AASH0 H-20 or 100k cement truck
- , These conditions and others previously mentioned will be combined and applied against normal allowables.
- 7. 4 Storage Tank (Deleted)
- 7. 5 Equipment Foundations
- 8 The vibrational characteristics of the pumps and diesel will be accounted for in the design. The diesel foundation will be isolated for the floor slab to limit vibration interference.
i Yankee Nuclear Power Station p.11 T8'NJ L' IA 1
- 6. Seismic Reevaluation Criteria tilliiiiiiiiiiiiiiiiiiniilill 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
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FREQUENCY IHZ) Input: Spectre output: Lusti.a a= t. z. YCS GROUND GROUND i seas s eirutan 2'3'5'7 N/A eirection seas s er.. s HORIZONTAL N/A . .h / A m.n an.ei.,in at a. a. v. a. tinie a.. f pree.ree e, MM Cue a.i Nel 83033 GROUND RESPONSE SPECTRA n ,iens e, g,g gg o.i.q f. ,,,, . A-1 4erem e, g . aste 4 g e... no. 0 _ ATTACHMENT A r Yankee Nuclear Power Station 3 Seismic Reevaluation Criteria dr LU L O 80023/81060/81061/86064 Doc. NO. DC-1; Rev. 4 lillllillilillllllilllllilllll F-12 i
APPEM)IX K COMPUTER PROGRAMS
- 1. "MOST, A Computer Program For Mode Superposition Time-History Analysis",
Version 1.1, Earthquake Engineering Systems, Inc.*, Novenber,1979.
- 2. "MOST, A computer Program For Mode Superposition Time-History Analysis",
Version 2.0, Cygna Energy Services, Novenber,1981.
- 3. "INSPEC, A Computer Program For Calculating Spectra", Version 1.2, Earthquake Engineering Systems, Inc., October,1980.
9
- 4. "INSPEC, A Computer Program For Calculating Spectra", Version 2.0, Cygna Energy Services, Novenber,1981.
- 5. " BATS, A Computer Program For Analysis of Multi-Story Frame and Shear Wall Building Systems", Version 6.0, Earthquake Engineering Systems, Inc. , Feb ruary, 1980.
- 6. "PRCQUAKE, An Automated Command File Procedure for Generating Design Time Histories", Versions 1.0 and 2.0, Cygna Energy Services.
- 7. " SAP IV, A Structural Analysis Program For Static and Dynamic Response of Linear Systems", Version 1.0, Earthquake Engineering Systems, Inc.,
Octob er, 1980.
- 8. "ANSYS, Engineering Analysis System, User's Manual", Version 3, Swanson Analysis Systems, Inc., July,1979.
't
- Earthquake Engineering Systems, Inc. is the old name of Cygna Energy Ser vices. Ownership, philosophy and staffing of the firm remains the same. ,
l l Yankee Nuclear Power Station K-1 g[M i Seismic Reevaluation Criteria lilililllilillllilllilililill! 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
- 9. "ANSYS, Engineering Analysis System, User's Manual", Versions 4.1 and 4.2, Swanson Analysis Systems, Inc., March 1,1983.
i
- 10. "0 RAIN-20, A General Purpose Computer Program For Dynamic Analysis of ]
Inelastic Plane Structures", NISEE/ Computer Applications, August,1985.
- 11. "ANSR-I", NISEE/ Computer Applications, Decenber,1975.
- 12. "ADLPIPE, Static and Dynamic Pipe Design and Stress Analysis", Arthur D.
Little, Inc. , July,1979.
- 13. "PIPESD", Version 6.0, CDC, August, 1979.
14 Mahin, S. A. and Bertero, V.V., " RCCOLA, A Computer Program for Reinforced Concrete Column Analysis," Department of Civil Engineering, University of California at Berkeley, August ,1977.
- 15. "PR A: A Computer Program for Pipe Rupture Analysis", Versions 2.0 and 3.0, Cygna Energy Services.
- 16. " ICES STRUDL, User Manual", McDonnel Douglas Automation Company, Octob er, 1982.
- 17. "EPLATE: A Nonlinear finite Element Program for Analysis of Baseplates",
Cygna Energy Services.
- 18. "S PECTR A: An Interactive Program for Response Spectra Generation Data Management", Versions 1.2 and 2.0, Cygna Energy Services.
( Yankee Nuclear Power Station K-2 g[ , , , Seismic Reevaluation Criteria lillilillililililllillllllilij 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4
a. b ANALYSIS OF THE MAIN STEAM AND FEE 0 WATER PIPING AND SUPPORT STRUCTURE OUTSIDE THE VAP(R CONTAINER '
- 1. 0 INTRODUCTION The main steam /feedwater piping outside the VC is supported on a series of light structural frames. The mass and stiffness of the support structure and piping runs are of similar magnitude; therefore, the rigid support assumption generally used in the majority of the piping analyses for YNPS is not valid here. Additionally, a nunber of pipes are supported by the same frame, promoting interaction between the pipes.
For the above reasons, the main steam /feedwater piping and the support structure will be either analyzed together in one problem or analyzed separately with interaction explicitly considered. This document outlines the methods and criteria which will be used in the analysis and qualification of the main steam /feedwater piping and support structure outside the VC. Both seismic and non-seismic sections of the piping will be analyzed. The extent of non-seismic piping to be included will be determined based on its effect on the response of the seismic piping and the support structure. Certain modifications may be designed to limit the areas of interaction between the seismic and non-seismic sections. Specific analysis details and evaluation criteria are discussed below.
- 2. 0 MAIN STEAM /FEEDWATER PIPING ANALYSIS 2.1 Geometry and Computer Modeling The conbined main steam /feedwater piping and support structure system is large and complex, making it prohibitively costly and time-consuming to analyze a detailed model of the entire system. To minimize the size of the actual seismic analysis model, stbstructuring methods may be used to model the support structure frames and attached plant structures to limit the degrees of freedom to a manageable nunber.
One example of the stbstructure modeling technique is using the superelements provided in the ANSYS computer code. The superelement is a group of previously assenbled elements that is treated as a single element. This element does not have a physical shape but it is made of a mass matrix and a stiffness matrix to interconnect the degrees of freedom retained by the superelement. Guyan reduction is l 1 YAN: 0599
, , = _
l
1 i used to develop the reduced mass and stiffness matrices. Using this technique, the structures connected to the MS/FW piping systems such as the VC, PAB, Turbine Building (TB), and Crane Support, as well as some portions of the MS/FW Support Frame, may be modeled by one or more superelements. Use of the stbstructure technique reduces both the static and dymanic degrees of freedom of the model. Further reduction in the dynamic degrees of freedom can be achieved using a dynamic condensation technique such as the master degree of freedom technique used in the ANSYS computer code. This can be performed after superelements and other elements have been assenbled to form the entire structural model. The ststructuring techniques associated with other approved programs may be used as well. The dynamic condensation technique as described above may not be available in other programs. Each frame will first be modeled in detail, considering all structural menbers (including all vertical and horizontal menbers and cross braces), connections, and other attached piping and equipment. Consideration will be given to the flexibility of frame menbers on which the piping systems are directly supported and to the supporting scheme on each frame to determine whether the main steam and feedwater pipes will interact on that particular frame. If the supporting scheme promotes interaction, the attachment points of the piping will be retained, and the detailed frame model will then be condensed to retain only the important menbers and the mass and stiffness matrices of the remaining menbers represented by the stb structures. This simplified model will then be used for the system analysis. To ensure the proper application of the stbstructure technique, the portion of the structure modeled by substructure will be compared to the full model. For example, as shown in Fig.1, a frame is modeled with a stbstructure. Depending on the dynamic characteristics of the frame and the nunber of connections between the pipe supports and frame, the amplified response spectra (ARS) at one or more pipe support attachment points will be generated using the original model plus a portion of the piping system. SW sequently, ARS at the same locations will be generated from the model containing the stbstructure plus the same portion of the piping system. The comparison of , the ARS generated using these two models will provide an 1 assessment of the stbstructure accuracy. 2 YAN: 0599
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)
l l 1
l i The analytical model will consist of the.eight main steam and feedwater lines appropriately supported and interconnected with real ; elements and s4 structures representing the support frames. After- ' i the system analysis.is completed, the support frame will be analyzed per.Section 3. If modifications to the supporting scheme or to the support frames are required, the frame models will be reevaluated to assess the interaction and to determine the necessary degrees of freedom for the revised configuration. The revised support frame - models will then be regenerated and recondensed. This . iterati ve process will be- repeated until an acceptable supporting scheme -is determined. j The support frames and the main steam and feedwater lines are j connected to several structures. These structures may be modeled as . stbstructures or treated as anchor points as discussed below and in Section 2.2. For the PAB and Turbine Building, the walls and structural penetrations are stiff compared to the support frames and 1 piping. Piping _and/or support stucture anchors will be'taken at those attachment points as applicable. The piping penetrations for the VC, however, are relatively flexible by comparison. For those penetrations, equivalent stiffnesses will be calculated. In all cases, the size and thickness of the reinforcing pad on the VC shell will be considered when calculating the equivalent.stiffnesses. The relative stiffness of the piping section exiting the VC and the i reinforcing pad will be considered in determining the local , stiffness of the VC shell. t
! The. connection of the main steam and feedwater piping to the VC . resenbles a nozzle penetrating a spherical vessel. The local I
stiffness of the nozzle attachment will be determined either by finite element method or using the formulas and figures provided in WRC Bulletin 49 (ptblished in 1959). j The remaining modeling aspects of the main steam /feedwater piping and support structure model will be as specified in the Seismic 4 Retrofit Criteria document, DC-1, Revision -3.
- t j 2. 2 Loading Conditions
! The main steam /feedwater piping and support. structure will be analyzed for deadweight, anbient wind, thermal, and site specific ! spectral loads (IRC Spectra loads), including the translational i anchor motions associated with each of the four load cases. The l intent of this evaluation is to ensure the adequacy of the system to i J 3 , YAN: 0599
. L N JL.A.
. mmm em man l l , _ ..._ ._ _ _ _ _ . .-_ ___ __ ._ _ . _ . . . . _ , _ . _ . ___ _
O function during and after a seismic event to attain a safe shutdown. The thermal conditions for normal operation will be used in cosination with the applicable deadweight, ant >1ent wind, and seismic loads for evaluation of the piping and support structures. Thermal anchor motions of the anchor points will be included as appropriate. As discussed in Section 2.1, the VC, PAB, and TB may be modeled using s@ structures or treated as anchors. If these buildings are I modeled using stbstructures, the ground spectra will be applied at the base of the structure. If the building structures are treated as anchors, amplified response spectra will be generated for each of the anchor points in the system model in accordance to the Seismic Reevaluation and Retrofit Criteria, DC-1, Rev. 4. PVRC damping ; (ASME Code Case N-411) will be used in analyzing the piping systems. Under NRC loads, the appropriate damping ratio for the VC, . PAB, and i TB is 7%. If these buildings are modeled using stbstructures, they will be stbjected to the ground spectra with PVRC damping. Since all these structures have the fundamental frequency of less than 10 Hz, their damping ratio will be 5% for the first few modes and even lower for the higher modes. In addition, the effect of the relative displacements between buildings is accurately calculated. Therefore, modeling these buildings with the stbstructuring l technique is very conservative. Seismic response spectrum analyses will be performed for the system. The seismic load will be input as independent support motions (ISM) at the appropriate anchor points (multi-level response spectrum method). The ISM input may be used in conjunction with stbstructuring technique. More specifically, the ground spectra will be input at the base of the buildings modeled using stbstructures, and appropriate ARS will be input at the anchor points. As suggested in NUREG-1061, the response between the support levels will be co61ned by absolute summation, and co@ination within a level will be performed by square-root-sum-of-the-squares. The Reg. Guide 1.9210% grouping method will be used to cosine modal i effects. Translational seismic anchor motions will be considered for each building as appropriate and as specified in the Seismic Retrofit Criteria Document, DC-1, Revision 4 if the buildings are modeled as l
- anchors for the piping and support structure. In addition to the l
building motions, relative seismic anchor motions of adjacent 4 l YAN: 0599 ddi A 111llllll11lllllllll111111llll
3.0 MAIN STEAM /FEEDWATER PIPING SUPP(RT FRAE ANALYSIS 3.1 Methodology and Load Conbination The detailed model of the MS/FW support frame is developed before it is condensed into stbstructure (or sibstructures) and integrated with the piping model. For each piping analysis (such as deadweight, thermal, etc.), the displacements at the retained degrees of freedom of the stbstructure are available. The full set of displacements and menber forces of the MS/FW support frame can be calculated using the displacements of the retained degrees of freedom. The MS/FW support frame will be evaluated for the following load conb ination: Absolute Sum of (Algebraic sum of DL, Thermal and TAM), (SRSS of Response Spectrum Analysis and SAM), and Anbient Wind Load The absolute value of the menber forces will be used, except that the sign of the axial forces due to seismic and wind loads may be taken as tension or compression for the evaluation of different elements. For example, the seismic axial force will be taken as compression when evaluating the adequacy of columns, but it will be taken as tension when evaluating the footing uplift.
- 3. 2 Acceptance Criteria Acceptance criteria for the MS/FW support frame will be as specified in the Seismic Retrofit Criteria Document, DC-1, Rev. 4, Section i 5.4. The menbers, connections and foundations will be evaluated for the most critical load case as discussed in Section 3.1 of this document.
4.0 NON-RETIRN VALVE ENCLOSLRE ANALYSIS The columns and beams of the non-return valve enclosure structure are an integrated part of the MS/FW support frame and will be evaluated per Section 3. The non-structural part of the enclosure structure, such as the roof and siding will be either modeled with the piping / frame system or analyzed separately using the Seismic Reevaluation and Retrofit Criteria, DC-1, Rev. 4. l 6 YAN: 0599 l k.i i ;L.A l .. . . . . . . . . . . . m. . =m m w y -
m support frames will be considered due to seismic wave propagation. To determine the proper differential motion to be applied to adjacent supports, the seismic wave will be propagated along the surface using a shear wave velocity of 1800 fps. Two real earthquake records (Taft and El Centro) scaled to a ZPA of 0.199 will be used to calculate a maximum relative displacement between any two anchor points, as well as the distribution of the anchor motions between a series of supports. A series of maximum displacements between supports will be used in the analysis unless a reduced _ set of displacements can be justified. Response from the seismic wave load case will be coM)ined with the standard seismic anchor motion response using the SRSS method.
- 2. 3 Acceptance Criteria Acceptance criteria for the main steam /feedwater piping will be as specified in the Seismic Retrofit Criteria Document DC-1, Revision 4. The standard ASME/ ANSI B31.1 equations will be used I first in the qualification. In the initial evaluation, the piping will be compared to the SEP criteria as follows:
DW + AM)ient Wind t Sh Eqn 11 DW + NRC Spectra + Ad)ient Wind < 2.4Sh Eqn 12 Thermal + TAM + SAM < Sa Eqn 13 l DW + Thermal + TAM + SAM + AM)ient Wind < Sa + Sh Eqn 14 60% SAM may be used in Equations 13 and 14, provided that- strain criteria were not used for Equation 12 at that point or at the points immediately adjacent to that point. In this case and at all other times,100% SAM will be used. The strain criteria and detailed code equation application criteria are specified in the Seismic Retrofit Criteria Document, DC-1, Revision 4. These strain I criteria and requirements apply to the seismic sections of the main steam and feedwater piping and to the non-seismic sections which have an effect on the response of the seismic piping or the frames which support the seismic piping. 5 YAN: 0599
- Li l J L A
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i I e If the enclosure structure is modeled separately, loads and load coM)inations will be per Sections 5.2 and 5.4.1 of DC-1, Rev. 4 The response spectra to be used as input loading will be calculated. If the sd) structure technique is used to model all the buildings (PAB, VC, and j TB) connected to the piping systems and support frame, the modal characteristics calculated from the piping analysis can be used to generate the ARS. If any building is taken as an anchor, a separate model which contains the equivalent masses and springs of the building should be used to calculate the modal characteristics used in the ARS generation. The analysis methodology and acceptance criteria will be per Sections 5.3.1 and 5.4 of DC-1, Rev. 4 The non-return valve enclosure structure l is not part of the Safe Shutdown System, and the sole purpose of the analysis is to demonstrate overall integrity of the enclosure structure, such that it does not jeopardize the non-return valves on the main steam piping. i l l l 1 YAN: 0599 q L*k '( - ) k A. l
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I s I h PIPING SYSTIMS WITH ORIGINAL FRAME MODELED BY SUBSTRUCTURE Figure 1 Stbstructure Modeling Technique 8 YAN: 0599 Li I.D 6 A llll11lllllllllllllll1111lll11
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