ML20148D155
| ML20148D155 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/27/1978 |
| From: | Reed C COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20148D161 | List: |
| References | |
| NUDOCS 7811020258 | |
| Download: ML20148D155 (1) | |
Text
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~LSCS-FSAR AMENDMENT 39 OCTOBER 1978 i
( LA SALLE COUNTY STATION INSTRUCTIONS FOR UPDATING'YOUR=FSAR Changes: to the 'FSAR'are identified by- a vertical line and revision number-in the right margin of the page. . To update your copy of the LSCS-FSAR, remove and destroy the following pages'and figures and insert pages and figures as indicated.
REMOVE INSERT s
VOLUME 1 Chapter 1.0 Page 1.7-1 Page 1.7-1 Page 1.7-8 Pa les 1.7-8 through 1.7-10 VOLUME 3 Chapter 3.0 .
Pages 3.0-01 and 3.0-02 Pages 3.0-01 and 3.0-02 Pages 3.0-xiii through Pages 3.0-xiii'through 3.0-xxviii 3.0-xxv k<~S
/ Page 3.6-8 Pages,3.6-8 and 3.6-8a .
Page 3.6-31 Pages 3.6-31 through 3.6-31b Figures 3.6-1.and 3.6-2 Figures 3.6-1, 3.6-la, 3.6-2, ,
and 3.6-2a Pages 3.9-1 through 3.9-186-
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Pages 3.9-1 through 3.9-140
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VOLUME.4 ,
i Chapter 3.0 (Cont'd) .
J Page 3.11-4 Page 3.11-4 )
l
-l VOLUME 5
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Chapter 6.0
/v
) Pages 6.0-00 and 6.0-01 Pages 6.0-00 and 6.0-01 1 gh' Page 6.0-vi Page 6.0-vi .
Oc Attachment 6.A,.from title Attachment 6.A,. including title
~ sheet through Figure 6.A-5 sheet, pages 6.A-i' through
(\ 6.A-lii, 6.A-1 through 6.A-103, and Figures 6.A-1.through.6.A-17
'() Chapter 7.0 l
-Pages.7.0-00 and 7.0-01 Pages 7.0-00 and 7.0-01 1
- - . . . . . . - - . - . - - .._...--.....-,,---..a,. =. - . .
e-3 LSCS-FSAR AMENDMENT 39 OCTOBER 1978 if( ) REMOVE INSERT VOLUME 6 Chapter 7.0 (Cont'd)
Pages 7.3-114 through' Pages 7.3-114 through 7.3-116 7.3-116 Figure 7.4-2, Sheets 1 to 5 Figure 7.4-2, Sheets 1 to'5 Pages 7.6-21 through Pages 7.6-21 through.7.6-25 7.6-25 Page 7.6-56 Page 7.6-56 VOLUME 7 Chapter 9.0 Page 9.0-00 Page 9.0-00 Pages 9.4-5, 9.4-6, Pages 9.4-5, 9.4-6, 9.4-6a, and 9.4-7 9.4-6a, and 9.4-7 Pages 9.4-11 and 9.4-12 Pages 9.4-11, 9.4-12, and 9.4-12a VOLUME 9 Chapter 13.0 Page 13.1-10 Page 13.1-10 l Figure 13.1-3 Figure 13.1-3 I Page 13.3-1 Page 13.3-1 1
CHAPTER 14.0 Page 14.0-11 Page 14.0-11 Page 14.0-iv Page-14.0-iv Page 14.2-42 Page 14.2-42 Page 14.2-47 Page 14.2-47 Pages 14.2-54 and 14.2-55 Sheet for pages 14.2-54 and i 14.2-55 l Pages 14.2-58 and 14.2-59 Pages 14.2-58 and 14.2-59 Page 14.2-79 Page 14.2-79 Fage 14.2-88 Page 14.2-88 Page 14.2-98 Page 14.2-98 "
Page 14.2-102 Page 14.2-102 Pages'14.2-119 and 14.2-119a Pages 14.2-119 and 14.2-119a
.Page 14.2-123 Page 14.2-123 Page 14.2-125 Page 14.2-125 Pages 14.2-131 through Pages 14.2-131 through 14.2-133 14.2-133
('Y N#
Pages 14.2-137 and Pages 14.2-137 and 14.2-138 14.2-138 Page_14.2-150 Page 14.2-150 Page-14.2-158 Page 14.2-158 2
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 ,
(} REMOVE INSERT Page:14.2-162- Page 14.2-162' Page'14.2-169 Page 14.2-169 Pages.-14.2-185 and 'Pages 14.2-185 and 14.2-186 ,
14.2-186- -
Page 14.2-191' Page 14.2-191 Page 14.2-193 Page 14.2-193
.Page 14.2-195 Page'14.2-195 Pages'14.2-197. through Pages 14.2-197 through~14.2-200 -
14.2-200 Pages 14.2-206 and 14.2-207 Pages 14.2-206 and 14.2-207 Page 14.2-217 Page 14.2-217' After page. 14.2-223 Pages 14.2-223a and 14,2-223b Chapter 15.0 Page 15.0-9 Page 15.0-9.
Chapter 16.0 s
Page 16.0-1 Page 16.0-1 and the LSCS technical.
specifications beginning with title page and ending with page 6-27
( )' Chapter 17.0 -
Page 17.0-i- Page 17.0-1 Page 17.0-1 .
.Page-17.0-1 .
Pages 17.0-2 through Topical Report CE-1A beginning 17.0-15 with title page and ending with page A-5 )
Appendix B Page B.1-10 Page B.1-10 Pages B.1-35 and B.1-36 Pages B.1-35, B.1-36, and B.1-36a Page B.1-40 Page B.1-40 l Page B.1-72 Pages B.1-72'and B.1-72a I RESPONSES TO NRC QUESTIONS 1 l
Q10 Series Page 010.23-1 Page Q10.23-11 Page Q10-29-1 Pages Q10.29-1 through Q10.29-7 iQ20' Series After Page Q20.0-3 'Page! Q20.0-4 Pages Q21.1'-2 through Q21.1-3a-
,' N Pages Q21.1-2 and Q21.1-3 After Page 021.1-7 Pages Q21.1-7a through Q21.1-7c t After Page 021.71-2. Pages Q21.72-1 through 021.72-6 and page-Q21.73-l'
, 3-
, _ _ - _ _ . . . _. _. . _ . . _ . . . . _ .. , - . _ - _. . . _ .i
LSCS-FSAR AMENDMENT 39 OCTOBER 1978
(')
v REMOVE INSERT Q30 Series Page Q30.0-10 Pages Q30.0-10 and 030.0-11 Page Q31.89-2 Pages Q31.89-2 and 031.89-3 Page 031.134-1 Page Q31.134-1 Page Q31.143-2 Page 031.143-2 Page Q31.156-1 Page Q31.156-1 Page Q31.158-2 Page Q31.158-2 Page Q31.185-1 Page Q31.185-1 After Page Q31.214-1. Pages Q31.215-1 through 031.215-3, 031.216-1 and Q31.216-2, Q31.217-1 and 031.217-2, Q31.218-1 through 031.218-3, and Q31.219-1 Page Q31.220-1 Page Q31.220-1 Page Q31.221-1 Pages 031.221-1, Q31.222-1, Q31.223-1, 031.224-1, 031.225-1, 031.226-1, 031.227-1, Q31.228-1 through 031.228-3, 031.229-1, 031.230-1, Q31 231-1, Q31.232-1, Q31.233-1, Q31.234-1, 031.235-1 through Q231.235-3, 031.236-1, 7s Q31.237-1, 031.238-1, and is_)
Q31.239-1 040 Series Page Q40.0-5 Page Q40.0-5 After Page Q40.109-1 Pages Q40.110-1 through 040.110-7 RESPONSES TO NRC QUESTIONS PART 2 0110 Series i Page Q111.3-1 Page Q111.3-1 Page Q111.6-1 Page Q111.6-1 ;
Page Q111.18-1 Page Q111.18-1 l Page Q111.21-1 Page Q111.21-1 1 Page Q111.25-1 Page Q111.25-1 l Page Q111.29-1 Page Q111.29-1 Page 0111.30-1 Page Q111.30-1 Page 0111.32-1 Page Q111.32-1 Pages 0111.36-1 through Pages Q111.36-1 through 0111.36-7 Q111.36-11 Page 0111.50-1 Page Q111.50-1 Page 0111.52-1 Page Q111.52-1 1 Page Q111.54-1 Page Q111.54-1 l
(~N Page 0111.56-1 Page Q111.56-1 i
(-) Pages Q111.57-1 and Pages Q111.57-1 and 0111.57-2 l l
Q111.57-2 l
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4 i
LSCS-FSAR' AMENDMENT 39 ,
OCTOBER 1978 i
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_a REMOVE INSERT -
'Page Q111.60-1 Page 111.60-1 Page Q111.61-1 _;
Page 111.61-1 ;
Page Qlll.63-'l- Page Q111.63-1 0220' Series Pages Q221.14-1 and Pages Q221.14-1 and'Q221.14-2 i 0221.14 Page Q221.19-1 Page Q221.19-1 0230 Series Page 0231.20-1 Page Q231.20-1 4 Page Q231.22-1 Page Q231.22-1 RESPONSES TO NRC QUESTIONS, PART 3 0420 Series Pages 0421.1-1 and Q421.1-2 .Pages.Q421.1-1 and Q421.1-2 Page Q421.2-1 Pages Q421.2-1 and 0421.2-2 Pages Q421.3-1.and 0421.3-2 Pages Q421.3-1 and Q421.3-2 Page Q421.4-1 'Pages Q421.4-1 through 0421.4-4 Page Q421.5-1 Page Q421.5-1 O Page Q422.13-1 Pages Q422.13-1 and Q422.13-2 0430 Series Page Q432.3-1 Page Q432.3-1 l
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LSCS-FSAR AMENDMENT 39 OCTOBER 1978 J 1.7 ELECTRICAL, INSTRUMENTATION, AND CONTROL (EC&I) DRAWINGS (FSAR)'
The following list indicates the EC&I drawings seven copies of which i 1
are provided under separate cover:
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A. General Electric Drawings REV.
DRAWING NUMBER TITLE NO. DATE 807E161TD Standby Liquid Control System (SLC) 9 4-26-77 807E160TD Feedwater Control System 12 7-05-78 l 807E159TD Control Rod Drive - Hydraulic System 8 5-28-75 807E158TD F? actor Manual Control System 8 5-11-78 807E175TD r eactor Wate.. Cleanup System (RWCS) 13 6-14-77 l 807E156TD Jet Pump Instrumentation System 13 4-25-77 ;
761E792TD Reactor Recirculaf*on System 11 807E153TD Nuclear Boiler Pr ss Instrumenta- 12 9-11-78 tion System 807E170TD Residual Heat Re al System (RHR) 13 6-03-78 807E172TD High-Pressure Cr Spray System 6 11-21-77 (HPCS) 731E302AA HPCS - One Line Diagram 2 3-27-74 807E183TD HPCS - Power Supply 4 3-01-78
,,, 807E171TD Low-Pressure Core Spray System (LPCS) 9 2-27-78
(,) 807E173TD Reactor Core Isolation Cooling System (RCIC) 12 10-14-77 807E155TD Automatic Depressurization System 11 3-07-78 -
(ADS) i 807E152TD Nuclear Steam Supply Shutoff System 20 9-11-78 I 807E154TD Leak Detection System 11 2-27-78 807Elb8TD Process Radiation Monitoring System 11 2-27-78 l 807E169TD Area Radiation Monitoring System - 7 11-12-74 I Unit 1 l 807E164TD Neutron Monitoring System (NMS) -
7 11-21-77 l Startup Range Detector Drive Control 807E162TD NMS - Startup Range 13 10-25-77 ;
807E165TD NMS - Traversing Incore Probe 9 11-04-74 l 807E163TD NMS - Power Range 11 10-25-77 l 807E166TD Reactor Protection System (RPS) 14 6-03-78 807E167TD RPS Interconnection Scheme 5 6-02-75 l 115D6268TD RPS MG Set Control 8 10-14-74 807E151TD Remote Shutdown System 3 12-24-77 828E155TD Off-Gas System 9 3-16-78 1 851E708TD Main Steamline Isolation valve 6 5-18-78 i Leakage Control System (MSIV-LCS) 828E230 Reactor Manual Control System 8 5-11-78 j
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1.7-1 1
L
-LSCS-FSAR' AMENDMENT "39 OCTOBER 1978 i
,m
! REV.
h . DRAWING NUMBER' TITLE NO. DATE lE-1-3653 Cable Pan Routing Reactor-Bldg. E 9-15-77 Plan El..740'-0" 9-26-77
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lE-1-3654 . Cable Pan Routing Reactor Bldg. .D- ,
l Plan El. 761'-0" 1E-1-3655 Cable. Pan Routing Reactor Bldg. F 9-26-77 j PlanuEl. 786l-6" .
1 lE-1-3656 Cable' Pan' Routing' Reactor-Bldg. C 7-05-77 !
Plan El. 807*-0" 1E-1-3657 Cable' Pan Routing Reactor Bldg. B 7-05-77
- Plan'El.
- 820'-6" ' !
lE-1-3661 Cable Pan Routing. Aux. Bldg. Plan .B- 3-02-77 El. 674'-0" and El. 687'-0" 1E-l'-3662 Cable Pan Routing Auxiliary Bldg. E '6-28-77 !
' Plan El. 710'-6" Col. 6-12 lE-1-3663 Cable' Pan Routing Auxiliary Bldg. J 9-08-77 "
Plan El. 710'-6" Col. 12-15' lE-1-3664 Cable Pan RoutingLAuxiliary Bldg. E 6-15-77 Plan El. 731'-0" Col. 6-12 lE-1-3665 Cable Pan Routing Auxiliary Bldg. H 9-08-77 '
Plan El. 731'-0" Col. 12 -! lE-1-3666 Cable Pon' Routing Auxiliary Bldg. C 8-08-77 i Plan El. 749'-0" Col. 6-12 !
lE-1-3667 Cable Pan Routino Auxiliary Bldg. F 7-05-77 l Plan El. 749'-0" Col. 12-15 lE-1-3668 Cable Pan' Routing Auxiliary Bldg. B 6-08-77 Plan El. 768'-0" Col. 6-12 ;
1E-1-3669 Cable Pan Routing Auxiliary Bldg. F 6-08-77 Plan El. 768'-0" Col. 12-15 lE-1-3672' Cable Pan Routing Auxiliary Bldg. B 3-30-77 Sections lE-1-3685 Cable Routing - Outdoor Area H 9-26-77 lE-1-4220AA Schematic Diagram - Residual Heat D 5-12-78 Removal System RH (E12A) Pt. 1.
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1E-1-4220AB Schematic Diagram - Residual Heat' D- 6-27-78
. Removal Pump 1A Sys. RH (E12A) Pt. 2.
lE-1-4220AC Schematic Diagram - Residual Heat E 6-27-78 Removal Pump 1B.Sys. RH (E12A) Pt. 3.
lE-1-4220AD Schematic Diagram - Residual Heat D 6-27-78 1 Removal Pump 1C Sys. RH (E12A) Pt. 4 1E-1-4220AE Schematic Diagram - Residual' Heat C 12-22-77
, Removal. System RH (E12A) Pt. 5 lE-1-4220AF ' Schematic' Diagram'- Residual Heat E- 9-21-78 Removal System RH (E12A) Pt. 6.
lE-1-4220AH Schematic. Diagram - Residual Heat D 6-29-78 Removal System RH (E12A) Pt. 8 J([~
'lE-1-4220AJ-Schematic' Diagram - Residual Heat C 5-12-78 !
Removal System RH (E12A) Pt. 9 lE-1-4220AK Schematic Diagram - Residual Heat F 6-29-78 Removal System RH (E12A) Pt. 10 1.7-8
- a - _ . _ . , . - - __ ~ . ~ . _ _ - _ . . - . . -
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 (3
V REV.
DRAWING NUMBER TITLE NO. DATE lE-1-4220AL Schematic Diagram-Residual Heat E 6-29-78 Removal System RH (E12A) Pt. 11 lE-1-4220AM Schematic Diagram-Residual Heat C 10-14-77 Removal System RH (E12A) Pt. 12 lE-1-4220AN Schematic Diagram-Residual Heat D 5-12-78 Removal System RH (E12A) Pt. 13 lE-1-4220AP Schematic Diagram-Residual Heat C 12-22-77 Removal System RH (E12A) Pt. 14 lE-1-4220A0 Schematic Diagram-Residual Heat D 5-12-78 Renoval System RH (E12A) Pt. 15 1E-1-4220AV Schematic Diagram-Residual Heat C 6-29-78 Removal System RH (E12A) Pt. 20 lE-1-4220AW Schematic Diagram-Residual Heat D 5-12-78 Removal System RH (E12A).Pt. 21 1E-1-4220AX Schematic Diagram-Residual Heat D 5-12-78 Removal System RH (E12A) Pt. 22 lE-1-4220AY Schematic Diagram-Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 23 1E-1-4220AZ Schematic Diagram-Residual Heat E 5-12-78 Removal System RH (E12A) Pt. 24 lE-1-4220BA Schematic Diagram-Residual Heat C 5-12-78
- ]
' Removal System RH (E12A) Pt. 25 lE-1-4220BB Schematic Diagram-Residual Heat D 5-12-78 Removal System RH (E12A) Pt. 26 lE-1-4220BC Schematic Diagram-Residual Heat D 6-29-78 Removal System RH (E12A) Pt. 27 lE-1-4220BD Schematic Diagram-Residual Heat E 6-29-78 Removal System RH (E12A) Pt. 28 1E-1-4220BE Schematic Diagram Residual Heat E 5-12-78 Removal System RH (E12A) Pt. 29 lE-1-4220BF Schematic Diagram Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 30 lE-1-4220BG Schematic Diagram Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 31 lE-1-4220BH Schematic Diagram Residual Heat E 5-12-78 Removal System RH (E12A) Pt. 32 lE-1-4220BJ Schematic Diagram Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 33 lE-1-4220BK Schematic Diagram Residual Heat B 10-14-78 Removal System RH (E12A) Pt. 34 lE-1-4220BL Schematic Diagram Residual. Heat D 5-12-78 Removal System RH (E12A) Pt. 35 lE-1-4220BM Schematic Diagram Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 36 lE-1-4220BN Schematic Diagram Residual Heat C 5-12-78 r^s Removal System RH (E12A) Pt. 37
(-) lE-1-4220BP Schematic Diagram Residual Heat E 6-29-78 Removal System RH (E12A) Pt. 38 lE-1-4220BQ Schematic Diagram Residual Heat E 5-12-78 Removal System RH (E12A) Pt. 39 1.7-9
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 REV.
7_ DRAWING NUMBER TITLE NO. DATE i
/
1E-1-4220BR Schematic Diagram Residual Heat C 5-12-78' Removal System RH (E12A) Pt. 40 lE-1-4220BS Schematic Diagram Residual Heat B 10-14-78 Removal System RH (E12A) P t '. 41 lE-1-4220BT. Schematic Diagram Residual Heat. C 5-12-78 Removal System RH (E12A) Pt. 42 lE-1-4220BU Schematic Diagram Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 43 lE-1-4220BV Schematic Diagram Residual Heat C 6-29-78 Removal System RH (E12A) Pt. 44 lE-1-4220BW Schematic Diagram Residual Heat E 6-29-78 Removal System RH (E12A) Pt. 45 lE-1-4220BX Schematic Diagram Residual Heat E 6-29-78 Removal System RH (E12A) Pt. 46 lE-1-4220BY Schematic Diagram Residual Heat C 5-12-78 Removal' System RH (E12A) Pt. 47 lE-1-4220BZ Schematic Diagram Residual Heat D 29-78 Removal System RH (E12A) Pt. 48 lE-1-4220CA Schematic Diagram-Residual Heat D 5-12-78 Removal System RH (E12A) Pt. 49 lE-1-4220CB Schematic Diagram-Residual Heat B 10-14-78 Removal System RH (E12A) Pt. 50 lE-1-4220CC Schematic Diagram-Residual Heat C 5-12-78 Removal System RH (E12A) Pt. 51 lE-1-4220rD Schematic Diagram-Residual Heat B 10-14-78
(}-
% Removal System RH (E12A) Pt. 52 lE-1-4 21 % ' Schematic Diagram-Residual Heat C 12-22-78 Removal System RH (E12A) Pt. 53 lE-1-4220CF Schematic Diagram-Residual Heat C 10-14-78 I Removal System RH (E12A) Pt. 54 lE-1-4220CG Schematic Diagram-Residual Heat D 5-12-78 Removal System RH (E12 A) Pt. 55 lE-1-4220CH Schematic Diagram-Residual Heat C 12-22-78 Removal System RH (E12A) Pt. 56 lE-1-4220CJ Schematic Diagram-Residual Heat D 5-12-78 ,
Removal System RH (E12A) Pt. 57 '
lE-1-4220CK Schematic Diagram-Residual Heat A 3-31-78 l Removal System RH (E12A) Pt. 58 lE-1-4220CL Schematic Diagram-Residual Heat A 6-27-78 Removal System RH (E12A) Pt. 59 l
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'LSCS-FSAR- AMENDMENT 39 OCTOBER 1978 CHAPTER ' 3 ; 0' (Cont 'd) -
FSAR SECTION QUESTION NUMBER 3.7.1- 130.1' 131.1 131.2 131.3 130.7 3.7.2- 130.3 130;5 '130.6 131.4 131.5
'131.6 131.7 131.8 131.9 131.10 362.24- 130.9- 130.10 130.11 130.12-130.13 130.14 130.15 130.16 L130.17 3.7.3 131.11 3.8.1 131.12 3.8.3 131.13 3.8.4 '131.13 3.8.5 362.8 362.25 3.9.1 111.16 111.17 3.9.2 111.18 111.19 111.20 111.21- 111.22
/~% 111.23 111.2'4 111.25 111.26 111.27.
V. 111.28 111.47 031.204 '111.54 111.55-3.9.3 111.29 111.30 111.31 111.32 111.'33 111.34 111.35' 111.36 111.37 111.38~
111.39 111.48 111.62 .111.59 111.56 111.57 111.58 111.60 111.61 111.63 3.9.4 111.35 3.9.6 111.49 #
3.10 031.52 031.83 040.65 040.67 031.133 031.148 040.96 031.236-3.10.1 031.84 3.10.3 111.40 3.10.5 111.41- 111.42 111.43 l[) .
3.0-01 i
LSCS-FSAR AMENDMENT 39 OCTOBER 1978
.,g CHAPTER 3.0 (Cont'd)
A._.)
FSAR SECTION OUESTION NUMBER 3.11 031.6 031.65 031.83 031.105 040.66 !
040.67 312.20 031.170 031.205 031.154 l 031.169 _031.164 040.94 040.95 040.96 l 040.97 031.210 031.216 031.217' 031.218 1 031.219 031.225 040.110 I 3.11.2 031.4 031.25 031.156 3.11.3 031.5 l
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3.0-02
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l LSCS-FSAR AMENDMENT 39 (
OCTOBER 1978 l
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TABLE OF CONTENTS (Cont'd)
PAGE f l
3.8.4.1.6.1 Station Vent Stack 3.8-41 3.0.4.1.6.2 Main Steam Chase 3.8-41 3.8.4.1.6.3 Equipment Access Building (Reactor Building Airlock for Railroad Car) 3.8-42 3.8.4.1.7 Other Non-Seismic Category I Safety-Related Structures 3.8-42 3.8.4.1.7.1 Solid Radwaste Building - Waste Storage 3.8-42 3.E.4.1.7.2 Lake Screen House - Intake Structure 3.8-42 3.8.4.2 Applicable Codes, Standards, and Specifica-tions _
3.8-43 3.6.4.3 . Loads and Loading Combinations 3.8-43 i 3.8.4.4 ' Design and Analysis Procedures 3.8-44 3.8.4.5 Structural Acceptance Criteria 3.8-44 3.8.4.5.1 Reinforced Concrete 3.8-44 3.8.4.5.2 Structural Steel 3.8-45 l l
3.8,4.6 Materials, Quality Control, and Special Construction Techniques 3.8-45 i
3.8.4.7 Testing and Inservice Surveillance
/~ Requirements 3.8-45
(-} 3.8.5 Foundations 3.8-46 3.8.5.1 Descriptions of Foundations 3.8-46 3.8.5.1.1 Main Building Complex 3.8-46 3.8.5.1.2 Lake Screen House 3.8-46 3.8.5.2 Applicable Codes, Standards, and l Specifications 3.8-46 3.8.5.3 Loads and Loading Combinations 3.8-46 3.8.5.4 Design and Analysis Procedures 3.8-47 3.8.5.4.1 General 3.8-47 3.8.5.4.2 Main Building Complex 3.8-47 3.8.5.4.3 Lake Screen House 3.8-47 3.8.5.5 Structural Acceptance Criteria 3.8-47 3.8.5.5.1 Structural Members 3.8-47 3.8.5.5.2 Stability 3.8-48 l 3.8.5.5.2.1 Main Building Complex 3.8-48 l 3.8.5.5.2.2 Lake Screen House 3.8-49 ;
3.8.5.6 Materials, Quality Control, and Special Construction Techniques 3.8-48 3.8.5.7 Testing and Inservice Surveillance Techniques 3.8-48 3.8.6 References 3.8-49
\
3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 gS 3.9.1 Special Topics for Mechanical Components 3.9-1 3.9.1.1 Design Transients 3.9-1
(_/
3.9.1.1.1 Thermal Transients 3.9-1 3.9.1.1.2 Hydrodynamic. Transients 3.9-1 3.0-xiii l
l LSCS-FSAR AMENDMENT 39 1 OCTOBER 1978 l
)~ TABLE OF CONTENTS (Cont'd) l l
PAGE I 3.9.1.1.2.1. Safety / Relief . Valve ' (SRV) Actuation ,
' Loads (Structural. Excitations) 3.9-2 3.9.1.1.2.2 Loss-of-Coolant Accident (LOCA) Loads 3.9-2 ~ ,
3.9.1.1.3 Annulus Pressurization 3.9-3 3.9.1.1.3.1 Acoustic Loads 3.9-4 3.9.1.1.3.2 Pressure Loads 3.9-4 3.9.1.1.3.3 Jet Loads 3.9-5 -
3.9.1.2 ' Computer Programs Used in Analysis 3.9-5 3.9.1.2.1 Reactor. Vessel 3.9-5 3.9.1.2.1.1 CB&I-Program 711 " GEN 022". 3.9-5 ;
3.9.1.2.1.2 CBSI Program 943 " NAPALM" 3.9-6 3.9.1.2.1.3 -CB&I Program 1027 3.9-6 3.9.1.2.1.4 CB&I Program 846 3.9-6 3.9.1.2.1.5 CBSI Program 781 -'"KALNINS" t3.9-6 3.9.1.2.1.6- CBSI Program 979 "ASFAST" 3.9-7 3.9.1.2.1.7 CB&I Program 766 "TEMAPR" :3.9-7 3.9.1.2.1.8 CBSI Program 767 " PRINCESS" 3.9-7 3.9.1.2.1.9 CBSI Program 928 "TGRV" 3.9-7 3.9.1.2.1.10 CB&I Program 962 "E0962A" 3.9-8' 3.9.1.2.1.11 CBSI Program 984 3.9-8
() 3.9.1.2.1.12 CB&I Program 922 - GASP 3.9.1.2.1.13 CB6I Program 1037 "DUNHAM'S" 3.9-8 3.9-9 3.9.1.2.1.14 CBSI Program 1335 3.9-9 3.9.1.2.1.15 CBSI Programs 1606 and 1657 " HAP" 3.9-9 3.9.1.2.1.16 CFSI Program 1635 3.9-10 3.9.1.2.1.17 CBSI Program 953 3.9-10 3.9.1.2.1.18 GE Program - DYSEA 3.9-10 3.9.1.2.2 Piping 3.9-11 3.9.1.2.2.1 Structural Analysis Program - SAP 4 3.9-11 3.9.1.2.2.1.1 Application 3.9-11 3.9.1.2.2.1.2 Program Organization 3.9-11 3.9.1.2.2.2 Component Analysis / ANSI 7 3.9-12 3.9.1.2.2.3 Pressure Design Minimum Wall Thickness Calculation / THICK 3.9-13 3.9.1.2.2.3,1 Area Reinforcement /NOZARP 3.9-13 3.9.1.2.2.4 Dynamic Forcing Functions 3.9-13 3.9.1.2-.2.4.1 Relief Valve Discharge Pipe Forces Computer Program /RVFORCE 3.9-13 3.9.1.2.2.4.2 Turbine Stop Valve closure Input to LAP /TSMOOD 3.9-13 3.9.1 2.2.5 Integral Attachment /LUGSTR 3.9-14 3.9.1.2.2.6 Piping Dynamic Analysis Program /FDA 3.9-14 3.9.1.2.2.7 Miscellaneous Analyses 3.9-15 3.9.1.2 2.7.1 WTNOZ Computer Program 3.9-15 3.9.1.2.2.7.2 'NOZLOD Computer Program 3.9-15 3.9.1.2.3 ECCS Pumps and Motors 3.9-15 Oc 3.9.1.1.4 RHR Heat Exchangers 3.9-15 i
3.0-xiv
+, - . . . , .. . . . - , - - , .
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 rh _
TABLE OF CONTENTS (Cont 'd)
PAGE 3.9.1.2.5 Computer Programs Used in Analyses of BOP-Piping and~ Equipment '3.9-16 3.9.1.3 Experimental Stress Analysis 3.9-16' 3.9.1.4 Consideration for the Evaluation of the Faulted Condition. 3.9-16:
- 3. 9. 2 ' Dynamic, System' Analysis and Testing .
3.9-16 3.9.2.1 Preoperational Vibration and Dynamic 'ffects Testing on Piping 3.9-16 3.9.2.1.1 Steam Flow 3.9-17 3.9.2.1.2 Turbine Stop Valve Closure 3.9-18 3.9.2.1.3 Relief Valve Operation 3.9-18 3.9.2.2 Seismic Qualification'of Safety -
.Related Mechanical Equip.nent 3.9-18 -
3.9.2.2.1 Tests and' Analysis Criteria and Methods 3.9-19 3.9.2.2.1.1 Random Vibration Input. 3.9-19 3.9.2.2.1.2 Application of. Input Motion 3.9-20 3.9.2.2.1.3 Fixture Design 3.9-20 3.9.2.2.1.4 Prototype Testing' .
3.9-20 3.9.2.2.2- Seismic ~ Qualification of. Specific NSSS Mechanical Components 3.9-20
( 3.9.2.2.2.'1 Jet Pumps 3.9-20 3.9.2.2.2.2 CRD and'CRD Housing 3.9-20 3.9.2.2.2.3 Core Support, Fuel-Support, and CRD Guide Tube 3.9-21 3.9.2.2.2.4 Hydraulic Control Unit (HCU) 3.9-21 3.9.2.2.2.5 Fuel Channels 3.9-21 3.9.2.2,2.6 Recirculation Pump and Motor Assembly. 3.9-21 3.9.2.2.2.7 ECCS Pump and Motor. Assembly 3.9-22 3.9.2.2.2.8 RCIC Pump Assembly. '3.9-22 3.9.2.2.2.9 RCIC Turbine Assembly .
3.9-22 3.9.2.2.2.10 Standby Liquid Control Pump and Motor Assembly 3.9-24 3.9.2.2.2.11 RHR Heat Exchangers- .
3.9-24 3.9.2.2.2.12 Standby Liquid Control Tank 3.9-25 3.9.2.2.2.13 Main Steam Isolation. Valves 3.9-25 3.9.2.2.2.14 Main Steam' Safety / Relief Valves 3.9-25 3.9.2.3 Dynamic: Response of Reactor Internals Under. Operational Flow Transients and Steady-State Conditions 3.9-25
.3.9=.2.4 Preoperational Flow Induced Vibration Testing of Reactor Internals 3.9-27 3 9.2.4.1
. Preoperational Tests 3.9-28 13.9.2.5 Dynamic System' Analysis of Reactor c
Internals Under Paulted Conditions 3.9-28 3.9.2.5.1 Safety Evaluation 3.9 3.9.2.5.2 Evaluation Methods 3.9-29 h()- '
c3.9.2.5.2.1 Input for Safety ~ Evaluation 3.9-29 "3.9.2.5.2.2 Events to be Evaluated 3.9-29 l
3.0-xv u - . - - - - - -. .= , s._ _ . . - , . -_ -- - - . . ~ . . - .
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 g
() TABLE OF CONTENTS (Cont'd)
PAGE 3.9.2.5.2.3 Pressure Differential During Rapid Depressurization 3.9-30 3.9.2.5.3 Recirculation Line and Steamline Break 3.9-30 3.9.2.5.3.1 Accident Definition 3.9-30 3.9.2.5.3.2 Effects of Initial Reactor Power and Core Flow 3.9-30 3.9.2.5.3.3 Conclusions 3.9-31 3.9.2.6 Correlation of Reactor Internals Vibration Tests with the Analytical Results 3.9-31 3.9.2.6.1 Analysis Methods Under LOCA Loadings 3.9-32 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3.9-32 3.9.3.1 Case 1 3.9-32 3.9.3.1.1 Loading Combinations and Stress Limits 3.9-33 3.9.3.1.1.1 Design Loading Combinations 3.9-33 3.9.3.1.1.2 Plant Conditions 3.9-33 3.9.3.1.1.2.1 Normal Condition 3.9-34 3.9.3.1.1.2.2 Upset Condition 3.9-34 f- 3.9.3.1.1.2.3 Emergency Condition 3.9-34
' )3
( 3.9.3.1.1.2.4 Faulted Condition 3.9-34 3.9.2.1.1.3 Correlation of Plant Condition with Event Probability 3.9-35 3.9.3.1.1.4 Safety Class Functional Criteria 3.9-35 3.9.3.1.2 Design Stress Limits 3.9-35 3.9.3.1.2.1 Stress Level for Seismic Category I Components 3.9-35 3.9.3.1.2.2 Field Run Piping 3.9-36 3.9.3.1.2.3 Stress Levels for ASME Code Class 2 and 3 3.9-36 3.9.3.2 Pump and Valve Operability Assurance 3.9-36 3.9.3.2.1 ECCS Pumps 3.9-37 3.9.3.2.1.1 Analysis of Loading, Stress, and Acceleration Conditions 3.9-37 3.9.3.2.1.2 Pump Operation During and Following Vibratory Loading 3.9-37 3.9.3.2.2 SLC Pump and Motor Assembly and RCIC Pump Assembly 3.9-38 3.9.3.2.3 RCIC Turbine Assembly 3.9-38 3.9.3.2.4 ECCS Motors 3.9-40 3.9.3.2.5 NSSS Values 3.9-41 3.9.3.2.5.1 Main Steam Isolation Valve (MSIV) 3.9-41 3.9.3.2.5.2 Main Steam Safety / Relief Valves 3.9-42 3.9.3.2.5.3 Standby Liquid Control Valve (Explosive Valve) 3.9-43
(,_) 3.9.3.3 Design and Installation Details for Mounting of Pressure of Relief Devices 3.9-43 3.9.3.4 Component Supports 3.9-44 3.0-xvi
1 LSCS-FSAR AMENDMENT 39 OCTOBER 1978
) TABLE.OF CONTENTS (Cont'd)
PAGE L3.9.4 Control-Rod Drive Systems (CRDS)
~
3.9 H3.9.4.1 Descriptive'Information of CRDS .3.'9-45 ,
3.9.4.2 . Applicable CRDS' Design' Specifications 3.9-45 l 3.9.4.3 Design. Loads, Stress' Limits,;and Allowable
,j Deformations 3.9 3.9.4.4 CRDS Performance Assurance Program 3.9-45
-3.9.5 Reactor Pressure Vessel Internals 3.9-46.-
3.9.5.1 Design Arrangement 3.9-46' 3.9.5.1.1 Core' Support Structure. 3.9 3.9.5.1.2 Core Shroud . .
3.9-47' 3.9.5.1.3 Shroud Headiand Steam Separator / Assembly 3;9-48 3.9.5.1.4 Core Support l Plate 3.9-48 3.9.5.1.5 Top Guide- 3.9-48 .
3.9.5.1.6 . Fuel Support .
3.9-48 j 3.9.5.1.7 Control Rod Guide Tubes. '3.9-49 3.9.5.1.8 Jet-Pump Assemblies 3.9-49 3.9.5.1.9 Steam' Dryers 3.9-50 3.9.5.1.10 'Feedwater Spargers 3.9-50~
3.9.5.1.11 Core Spray' Lines 3.9-50 3.9.5.1.12 Vessel Head Cooling Spray Nozzle 3.9-51
() 3.9.5.1.13 Differential Pressure and Ligdid Control Line 3.9-51 3.9.5.1.14 Incore Flux Monitor Guide Tubes 3.9-51 3.9.5.1.15 Surveillance' Sample Holders 3.9-52 3.9.5.1.16 Low-Pressure Coolant. Injection Lines 3.9-52 3.9.5.2 Design Loading Conditions 3.9-52 3.9.5.2.1 Pressure Differential'During Rapid Depressurization 3.9-52 3.9.5.2.2 Recirculation Line and Steamline Break 3.9-52 3.9.5.2.2.1 JAccident Definition 3.9-52 3.9.5.2.2.2 Effects of Initial. Reactor Power and Core Flow 3.9-53 ,
3.9.5.2.2.3 Response of Structures Within the Reactor Vessel to Pressure Differences 3.9-54 3.9.5.2.2.4 Conclusions '3.9-55 3.9.5.2.3 . Dynamic. Loads .
3.9-55 3.9.5.2.4 Safety Evaluation 3.9-55 3.9.5.2.4.1 Evaluation! Methods .
3.9-55 3.9.5.2.4.1.1 Inputifor Safety Evaluation 3.9-56 3.9.5.3 Design Loading Categories- 3.9-56
- 3.9.5.3.1 ' Stress, Deformation, and Fatigue Limits- '
for. Reactor Internals (Except Core Support Structure) .
3.9-56 3.9.5.3.2 Stress, Deformation, and Fatigue Limits
'for Core Support.: Structures 3.9-56 j] 3.9.5.4 Design Bases .
3.9-57 3.9.5.4.1. Safety Design Bases- . '3. 9-5 7 -
'3.9.5.4.2 Power Gentration Design Bases 3.9-57 3.0-xvii
- ., , - .... - , . - . . ~ .- - . .-. . -. _ ,_
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 3 TABLE OF CONTENTS ~(Cont'd) . i PAGE-3.9.5.4.3 , Fuel 1 Assembly Restraints 3.9-57 3.9.5.4.4' Material Selection '3.9-57 3.9.5.4.5 Radiation Effects -3.9-58 3.9.5.4.6 . Vibration' Measurements of Reactor Internals:3.9-58
'3.9.5.4.7 ' Accident Conditions 3.9-59^ l 3;9.5.4.8:- Inspection and_ Testing 3.9-59 3.9.6 Inservice Testing of Pumps and Valves 3.9-59; 3.9.~6.1 . Inservice Testing of Pumps 3.9-61 3.9.6.2 Inservice Testing of Valves 3.9-61 3.9.7 ' References 3.9-62 i 3.10 SEISMIC QUALIFICATION.OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.10-1 3.10.1 Seismic Qualification Criteria :3.10-1 3.10.1.1 Seismic Category I Equipment Identification 3.10-1 -
3.10.1.2 Seismic Design' Criteria- 3.10-1 3.10.1.3 -Reactor Protection, Engineered Safety ~ ,
Features, and Emergency Power Circuits 3.10-2
/~T 3.10.1.4 Cable Tray and Bus Duct Supports:
(sl.
Criteria 3.10-2 3.10.2 Methods and Procedures for Qualifying (
Electrical Equipment and Instrumentation 3.10-3 ;
3.10.3 Methods and Procedures of Analysis or Testing 7 of Supports of Electrical Equipment'and Instrumentation 3.10-4 3.10.3.1 NSSS Equipment 3.10-4
{
3.10.3.2 GE-Supplied Non-NSSS. Equipment Analyses and Testing .
3.10-7 3.10.4 Dyn _.aic Analysis by Response Spectrum Methods 3.10-8 -
3.10.5 Sample Seismic Static Analysis 3.10-12 c
3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT 3.11-1 >
3.11.1 Equipment Identification and Environmental Conditions. 3.11-1 3.11.1.1 Engineered Safety Features and Reactor .
Protection System 3.11-1 3.11.1.1.1 Safety-Related Equipment and Components -
'3.11-1
. l' 3.11.1.1.2 _ Inside Primary Containment Safety-Related Equipment and Components - .
Insf.de Secondary Containment 3.11-1 3.11.1.1.3 Safety-Related Equipment and Components
=
Outside the Primary Containment 3.11-1 ,
'( );
.3.11.1.1.4 Auxiliary Equipment 3.11-1 E3.11.1.2 Environmental-Bases 3.11-1 '
j3.11.1.2.1 Plant Operational Environmental Conditions 3.11-1 3.0-xviii
,.-,,,u--e ,+n- .,n-- , . , - - - , . , , , . - . ..+~-nn,,- ,-n - - - - - , -,,vv-e-
-LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE OF CONTENTS (Cont'd) -
PAGE 3.11.1.2.2 Accident Basis Environment 3.11-2 )
3.11.1.2.2.1 Accident Basis Environment - Inside i Primary Containment 3.11-2 l 3.11.1.2.2.2 Accident Basis Environment - Outside '
)
3.11-2 3.11.'1.2.2.2.1 Bulk. Atmosphere Conditions 3.11-2 3.11.1.2.2.3 Accident Basis Environment - Outside Secondary Containment 3.11-2 3.11.2 Qualification Tests and Analyses 3.11-3 3.11.2.1 Environmental Qualification of Electrical Equipment .
3.11-3 3.11.2.1.1- Equipment Identification 3.11-3 3.11.2.1.2 Tests 3.11-3 3.11.2.1.3 Analysis 3.11-3 -
3.11.2.1.4 Environmental Qualification Procedures for Electrical Equipment 3.11-4 3.11.2.2 Environmental Qualification Procedures for Electrical. Equipment 3.11-4 !
3.11.2.2.1 Worst Case Environment Analysis 3.11-4 3.11.2.3 Environmental Qualification Procedures for O- Mechanical Equipment' 3.11-4
- 3.11.2.3.1 Worst Case Environment Analysis 3.11-4 3.11.2.3.2 Qualification of Mechanical Equipment 3.11-4 3.11.2.3.3 Auxiliary Environmental Equipment 3.11-5 ,
3.11.2.4 Safety-Related Equipment and Components 3.11-5 3.11.3 ' Qualification Test Results 3.11-5 3.11.4 Loss of Ventilation 3.11-6 3.11.4.1 Control Room Air-Conditioning and Ventilation System 3.11-6 3.11.4.2 Emergency Switchgear Room Ventilation System 3.11-6 3.11.5 Estimated Chemical and Radiation Environment 3.11-6 ATTACHMENT 3.A PROCEDURES FOR CALCULATING MISSILE STRIKE AND PENETRATION PROBABILITIES- 3.A-1 ;
O-3.0-xix
- ,, _ . . - . _ . _ _ _ _ _ - _ _ . _ ~ . . __ - . - . .
LSCS-FSAR AMENDMENT 22 MAY 1977
.(3 CHAPTERJ3.0 - DESIGN OF STRUCTURES, COMPONENTS,
\. / EQUIPMENT, AND SYSTEMS LIST OF TABLES NUMBER '
TITLE PAGE 3.2-1 Structures, Equipment, and Component Classifications 3.2-3 3.2-2 . Code Requirements for Components and Systems Ordered Prior to July 1, 1971 3.2-25 3.2-3 Code Requirements for Components and Systems Ordered After July.1, 1971 .3.2-26 3.2-4 Code Requirements-for Components and Systems Ordered After July 1, 1974 3.2-27 4 3.3-1 Dynamic Wind Pressure', q,,for. Seismic Category I Structures 3.3-6.
3.3-2 Equivalent Static Force, p, for Seismic Category I Structures 3.3-7 3.5-1 18-Inch Last_ Stage Bucket, _1800-rpm '
Low-Pressure Turbine - Hypothetical Missile Data 3.5-12 3.5-2 Probabilities of Low Trajectory Turbine Missile DamageLUnder Destructive Overspeed Failure 3,5-14
() 3.5-3 Summary of'High and Low Trajectory Missile Damage Under Destructive Overspeed Failure 3.5-15 3.6-1 Postulated Rupture Orientation-and Number '
of Design-Basis Breaks Used in Pipe Rupture.
Analysis Inside Containment 3.6-30 3.6-2 Recirculation Piping Systems Operating Stresses at Break Locations 3.6-31 3.6-3 Main Steam System - Results of Dynamic Analysis for Postulated Pipe Rupture 3.6-32 3.6-4 Feedwater System - Results of Dynamic Analysis for Postulated Pipe Rupture 3.6-34 3.6-5 Results of Dynamic Analysis for Postulated Pipe Rupture 3.6-36 3.6-6 Material Properties of Pipe Whip Restraints 3.6-38 3.6-7 Summary of Restraint Data for Recirculation Piping 3.6-39 3.7-1 Critical Damping Ratios for Different Structure or Component .
3.7-40 3.7-2 Periods and Participation Factors 3.7-41 3.7-3 Probable Maximum Displacements 3.7-43 3.7-4 Periods and Participation Factors 3.7-44 3.7-5 Comparison of Response 3.7-45
("1 3.7-6 Comparison of Calculated Seismic Loads to Design Seismic Loads of Seismic Category I Equipment, SSE Condition 3.7-46 3.0-xx
l LSCS-FSAR AMENDMENT 39
, OCTOBER 1978 iO x/ '
LIST OF TABLES (Cont'd)
NUMBER TITLE PAGE 3.7-7 Number of Dynamic Response Cycles Expected During a Seismic Event 3.7-48 3.7-8 Fatigue Evaluation Due to Seismic Load 3.7-49 3.7-9 Comparison of the Max 3 mum Seismic Loads of Reactor Pressure Vessel and Internals 3.7-50 3.7-10 Solid Torsional Modal Properties 3.7-51 3.8-1 Containment' Penetrations 3.8-50 3.8-2 List of Specifications, Codes, and Standards 3.8-56 3.8-3 Load Definitions and Combinations for Primary Containment and Drywell Floor 3.8-61 3.8-4 Allowable Stresses and Strains -
Containment Liner Plate and Anchorages 3.8-63 3.8-3 Predicted Response Readings for Containment Under 52 psig Test Pressure 3.8-65 3.8-6 Load Definitions and Combinations for Class MC Containment Components (Other Than Piping Penetrations) 3.8-67 73 3.8-7 Physical Properties for Materials to be
(_) Used for Pressure Parts or Attachments to Pressure Parts - MC Components 3.8-68 3.8-8 Load Definitions and Combinations for Reactor Pedestal 3.8-69 3.8-9 Seismic Category I Structures Load Combination - Structural Steel Elastic Design 3.8-70 3.8-10 Seismic Category I Structure Load Combination - Reinforced Concrete Structures Other than Containment 3.8-72 3.8-11 Definitions of Structural Terminology 3.8-73 3.8-12 Loading combinations for Penetration Sleeves and Head F.4.ttings 3.8-77 3.8-13 Allowable Stresses for Penetration Sleeves and Head Fittings 3.8-78 3.9-1 Pressure Differentials Across Reactor Vessel Internals 3.9-64 3.9-2 Reactor F. essure Vessel Support Components 3.9-65 3.9-3 Reactor Vessel and Associated Equipment _
3.9-68 3.9-4 Fuel Assembly (Including Channel) 3.9-76 3.9-5 Main Steam Piping Highest Stress Intensities or Stress Intensity Range - Class 1 Pipe 3.9-77
()
s.
3.0-xxi
LSCS-PSAR AMENDMENT 39 OCTOBER 1978 i rs L'IST OF TABLES (Cont'd)
NUMBER TITLE PAGE 3.9-6 Recirculation Piping System Highest Stress Intensities or Stress '
Intensity Range - Class 1 Pipe 3.9-81 3.9-7 Reactor Vessel Support Equipment and_CRD Housing Support 3.9-83 3.9-8 Main Steam Safety / Relief Valves 3.9-87 i 3.9-9 Main Steam Isolation Valve 3.9-94 3.9-10 Recirculation Pumps 3.9-110 !
3.9-11 Structural and Mechanical Loading I criteria 3.9-116 I 3.9-12 Hydraulic Control Unit Piping 3.9-122 3.9-13 RHR Pump 3.9-123 3.9-14 RHR Heat Exchanger 3.9-125 3.9-15 Low-Pressure Core Spray Pump 3.9-127 l 3.9-16 RCIC Turbine 3.9-129 )
3.9-17 RCIC Pump 3.9-132 l 3.9-18 Recirculating Pipe and Pump Restraints 3.9-134 3.9-20 Standby Liquid Control Pump 3.9-137
(~) 3.9-21 Standby Liquid Control Tank 3.9-139 x' 3.9-22 Jet Pump 3.9-141 3.9-23 Active ASME Class 1, 2, and 3 Pumps 3.9-142 3.9-24 List of Safety-Related Active Valves -3.9-143 3.9-25 Load Combinations and Acceptance Criteria 3.9-147 3.9-26 Deformation List 3.9-150 3.9-27 Primary Stress Limit 3.9-151 3.9-28 Buckling Stability Limit 3.9-154 3.9-29 Fatigue Limit 3.9-155 3.9-30 Core support Structures Stress j Categories and Limits of Stress j Intensity for Normal and Upset ,
Conditions 3.9-156 !
3.9-31 Core Support Structures Stress ;
Categories and Limits of Stress Intensity for Emergency Conditions 3.9-160 !
3.9-32 Core Support Structures Stress Categories and Limits of Stress Intensity for Fault Conditions 3.9-163 3.9-33 Applicable Thermal Transients 3.9-166' ;
3.10-1 Class 1E Equipment Requiring j Qualification 3.10-20 !
3.10-2 Seismic Qualification Test Summary for !
La Salle Class 1E Control Panels and rm Local Panels 3.10-38 ;
i%f 3.10-3 Standard Enclosures 3.10-42 !'
3.10-4 Seismic Design Verification Data Sheet 3.10-43 3.0-xxii i f
L------ 4 ,,3,, b & -O '4++
. . 3 * ++ 4 34s.-E 4 s44-. J n w- k. 40 .J ~ ><a4 )
- LSCS-FSAR AMENDMENT 39 L
- OCTOBER 1978
' LIST OF TABLES '(Cont'd)
- NUMBER TITLE PAGE-3.11-1 ' Pressure, Temperature, Relative Humidity, .. i
.and Environmental' Conditions Inside '
Primary Containment - Abnormal Conditions .3.11-7 3.~11-2 -Pressure, Temperature,' Relative Humidity, and Environmental Conditions ~Inside Secondary Containment ~ Abnormal Conditions Design-Basis ~ Loss-of-Coolant ~
Accident' 3.~11-11:
3.11-3 Auxiliary' Equipment' Environmental Conditions. ..
. 3.11-16. '
3'.11-4 Environmental Conditions-(Plant)-
Operational _ 3.11-18 3.11-5 . Radiation Tolerance' of Materials : .
3.11-23 3.11-6 Pressure,-Temperature Relative Humidity, and Environmental' Conditions Outside the Secondary Containment ESF Non-GE-Furnished Electrical Equipment 3.11 .i (I.
i
?)
3.0-xxiii
. c. . .._._ -_ _. ..u..._.. , , . .. , _ . . _ . . u.c__ - _ _ . . _ . , . - - , . , -,_;....;.-.._.,-,_.,_
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 o
O CHAPTER 3.0 - DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS LIST OF FIGURES NUMBER TITLE 3.2-1 Group Classification Diagram 3.2-2 Group Classifications for Instrument Piping.
3.3-1 Annual Extreme Wind Velocities 100-Year Recurrence Interval 3.3-2 Differential Pressure and Velocity Distribution for the Design-Basis Tornado 3.3-3 Dynamic Wind Pressure Distribution for the Design-Basis Tornado 3.3-4 Resultant Surface Pressure for the Design-Basis Tornado for Rectangular Flat-Topped Structure 3.4-1 Flood Control Basement Floor Plan 3.4-2 . Flood Control Upper Basement Floor Plan 3.4-3 Lake Screen House Windwave Action 3.4.4 Doors Affected by the PMP - Ground Floor Plan 3.5-1 Turbine Placement and Orientation 3.5-2 Section Showing Low Trajectory Missile Strike Area
()
3.5-3 3.5-4 (DELETED)
Reactor Building Missile Barriers 3.5-5 Auxiliary Building Missile Barriers 3.5-6 Diesel-Generator Building Missile Barriers 3.6-1 Recirculation Loop-A with Postulated Breaks 3.6-1a Piping Element Diagram - Recirculation Loop A l 3.6-2 Recirculation Loop-B with Postulated Breaks 3.6-2a Piping Element Diagram - Recirculation Loop B l 3.6-3 Main Steamline-A, Postulated Breaks and Restraint Locations 3.6-4 Main Steamline-B, Postulated Breaks and Restraint Locations 3.6-5 Main Steamline-C, Postulated Breaks and Restraint Locations 3.6-6 Main Steamline-D, Postulated Breaks and Restraint Locations 3.6-7 Feedwater Subsystem-1, Postulated Breaks and Restraint Locations 3.6-8 Feedwater Subsystem-2, Postulated Breaks and Restraint Locations 3.6-9 High-Pressure Core Spray - Postulated Breaks and Restraint Locations 3.6-10 Low-Pressure Core Spray - Postulated Breaks and Restraint Locations-3.6-11 Residual Heat Removal System, Line 53B - Postulated
("N Breaks and Restraint Locations k- 3.6-12 Residual Heat Removal System, Line 40BA - Postulated Breaks and Restraint Locations 3.0-xxiv
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 O
LIST OF FIGURES (Cont'd)
NUMBER TITLE 3.6-13 Residual' Heat Removal System, Line 40BB - Postulated Breaks.and Restraint Locations 3.6-14 Reactor Core Isolation Coolant - Postulated Breaks and Restraint Locations 3.6-15 Residual Heat Removal System I 3.6-16 Residual Heat Removal System II - Postulated Breaks and Restraint Locations 3.6-16a Reactor Water Cleanup - Postulated Breaks and Restraint Locations 3.6-17 Forcing Functions Model Associated with Pipe Whip Dynamic Analysis of the Reactor Recirculation System 3.6-18 Forcing Function Model Associated with Pipe Whip Dynamic Analysis of High Energy Systems Inside Containment 3.6-19 Generic Representation of Pipe 3.6-20 Model Representation of Piping 3.6-21 Model Representation of Piping with Built-in Ends 3.6-22 Linear Deflection of Pipe 3.6-23 Models for Dynamic Analysis of Pipe Whip r~)s
\_ 3.6-24 Second and Third Stages of Motion 3.6-25 Time Dependent Loadings on Pipe Whip Restraints 3.6-26 Restraint Configuration 3.6-26a Diagram of Restrain Frame i 3.6-27 Typical Pipe Whip Restraints Used on All High Energy Lines Inside Containment Except Reactor Recirculation 3.6-28 Stress-Strain Relations of Pipe Restraint Material 3.6-28a Stress-Strain Curve for Recirculation Pipe Material 304 Stainless Steel 3.7-1 N-S Horizontal Synthetic Time History 3.7-2 E-W Horizontal Synthetic Time History 3.7-3 Synthetic Time History and Design Spectrum Comparison for N-S Components (2% Damping) 3.7-4 Synthetic Time History and Design Spectrum Comparison for E-W Components (2% Damping) 3.7-5 Synthetic Time History and Design Spectrum Comparison for N-S Components (5% Damping) 3.7-6 Synthetic Time History and Design Spectrum Comparison for E-W Components (5% Damping) 3.7-7 Typical Shear Structure 3.7-8 Slab at El. 694 ft. 6 in.
3.7-9 Horizontal Building Model 3.7-10 Vertical Dynamic Model 3.7-11 Modified Vertical Model
() 3.7-12 Reactor Containment Shear and Moment Diagram (x-Excitation OBE) 3.0-xxv
LSCS-FSAR' AMENDMENT.39 OCTOBER 1978-LIST OF FIGURES '(Cont 'd)' l NUMBER TITLE 3.7-13 Reactor Containment Shear and Momen: Diagram (y-Excitation OBE) 3.7-14 Horizontal Floor Response Spectra North-South Component (OBE) El'. 694 ft 6 in. (Slab No.~1)
React., Aux., Turb.~ Bldg., Htr. Bay 3.7-15 Horizontal Floor Response Spectra East-West Component (OBE) El. 694 ft 6 in. (Slab No.'1)
React., Eaux., Turb. Bldg., Htr. Bay y 3.7-16 Horizontal Floor Response Spectra. North-South- l Component (SSE) El. 694 ft 6 in. (Slab.No.11)
React., Aux.,.Turb. Bldg., Htr. Bay . ,
3.7-17 Horizontal Floor Response Spectra East-West ?
Component (SSE).El.,694~ft.6 in. .(Slab No. 1).
React., Aux., Turb. Bldg., Htr. Bay- . ,
3.7-18. Horizontal Floor Response Spectra North-South l Component -(OBE) . El.184 3 f t 6 in. ' (Slab' No. 8) H Reactor Building 3.7-19 Horizontal Floor Response. Spectra. East-West Component (OBE) REl. - 843 f t 6 in. (Slab j r,g No. 8) Reactor Building
(_/ 3.7-20 Horizontal Floor Response Spectra. North-South Component (SSE) eel. 843 ft 6 in. (Slab.No. 8)
Reactor Building Horizontal Floor Response Spectra East-West .
3.7-21 Component.(SSE) El. 843 ft 6 in. (Slab No. 8) Reactor Building 3.7-22. Horizontal Floor Response Spectra North-South Component (OBE) El. 786 ft 6 in. Reactor Containment 3.7-23 Horizontal. Floor Response Spectra East-West Component (OBE)'El. 786 ft 6.in. Re ac tor'.'
Containment 3.7-24 Horizontal Floor Response Spectra North-South Component (SSE) El. 786 ft'6 in; Reactor Containment !
3.7-25 Horizontal Floor Response Spectra East-West Component-(SSE) El. 786 ft 6 in. Reactor Containment.
3.7-26 Vertical Response Spectra (OBE) Reactor Building Wall E1.:710 ft 6 in. to.El. 694 ft 6 in.
3.7-27 Vertical Response Spectra.(OBE) Reactor ~ Building 1 Slab El. 710-ft 6 in, and El. 694 ft 6 in.
3.7-28 Vertical Response Spectra (SSE) Reactor Building Walls El. 710 ft 6 in, and El. 794 ft 6 in.
3.7-29 Vertical Response Spectra (SSE) Reactor Building
-( )- Slab El. 710 ft 6 in. and El. 694'ft 6.in.
3.7-30 Vertical Response Spectra -(OBE) Reactor Building.
, Walls El. 843 ft 6 in. to El. 807 ft 0 in.
3.0-xxvi
r o i
, LSCS-FSAR AMENDMENT 39 l OCTOBER 1978 i LIST OF FIGURES (Cont'd)
~
~
l NUMBER TITLE j 3.7-31 ' Vertical Response Spectra (OBE) Reactor Building !
Slab El. 843 ft 6 in. to El. 807 ft.0 in. 4 3.7-32 Vertical Response Spe'ctra (SSE) Reactor Building !
Walls El, 843 ft 6 in..to El. 807 ft 0 in.
3.7-33 Vertical; Response Spectra (SSE) Reactor: Building Slab El. 843 ft 6 in. to El. 807 ft'0 in. l 3.7-34 Vertical, Response Spectra (OBE)LReactor Containment-Shield El.1843 ft 6 in, to El. 786 ft 6 in. 1 i
3.7-35 Vertical Response Spectra (SSE) Reactor Containment Shield El.'843 ft 6 in, to.El. 786 ft 6 in. ,
3.7-36 Reactor' Pressure-Vessel and Internals Seismic Model dR 3.7 Density of Stress Reversals 3.7-38 Typical Mass Model: for Cable Pans ,
'3.7-39 Finite ElementLSoil'Model' l 3.7-40 Effective Soil Column for Calculating Soil 1 Torsional Mode.
3.7-41 ' Foundation Level OBE, 2% Damping 3.7-42 Foundation Level SSE, 5% Damping 3.7-43 SSE N-S Spectrum Curves (5% Damping) s 3.7-44 Slab #2, OBE N-S, 2% Damping 3.7-45 Slab #8, OBE N-S, :2% Damping l 3.7-46 Slab #2, OBE E-W, 2% Damping l 3.7-47 Slab #8, OBE E-W, 2% Damping i 3.7-48 Slab #2, SSE N-S, 5% Damping ;
3.7-49 Slab #8, SSE N-S, 5% Damping
.3.7-50 Slab #2, SSE.E-W, 5% Damping j 3.7-51 Slab #8, SSE E-W, 5% Damping 3.7-52 Foundation Level OBE Vertical, 2% Damping 1
.3.7-53 Foundation Level SSE Vertical, 5% Damping 3.8-1 Reactor Containment 3.8-2 Primary. Containment 3.8-3 Downcomer' Pipe-Vent System 3.8-4 Drywell Floor Containment Wall Junction j 3.8-5 Drywell Head and Ring Girder Connection 3.8-6 Reactor-Building Foundation Bottom Reinforcing Plan ~
3.8-7' Top Reinforcement Base Slab' ,
3.8 LSuppression Chamber Basement Liner 3.8-9 Baseline Anchorage Detail .
3.8-10 Typical Reinforcing Details'of the. Containment j 3.8-11 Typical Layout of Hoop and Meridional Tendons 1 3.'8-12 Tendon Termination at Buttress Reinforcing Details'at Containment Wall Base Slab.
3.8-13 Junction q '
. 3.8-14 Primary Containment Liner. Anchor Detail-U(]); 3.8-15 Location of. Containment ~ Penetrations !
~3.8-16 Reinforcement Around Personnel Lock and Equipment Hatch 3.0-xxvii ,
.- . . . - - - . . . . - -- -. . .~ , . ,
4' f
LSCS-FSAR LAMENDMENT 39 OCTOBER 1978 f LIST.OF FIGURES (Cont'd).
NUMBER TITLE 3.8-17 Tendon Deflection Around Large Penetration Sleeve-3.8-18_ LType I Flued Head-3.8-19 - Type'II Flued Head ,
-3.8-20 iType III Flued. Head 3.8-21 Electrical Penetration ,
3.8-22 ContainmentLWall Embedments 3.8-23 Reactor Stabilizer Structure 3.8-24 Containment Wall Embedments
~
'3.8-25 Beam or Slab-Support Corbel-3.8-26 . Containment Thin.Shell' Analytical Model -i 3.8-27 ~ Containment' Shear Moments and Forces'- Dead Load 3.8-28 Containment' Shear Moments and Forces'- Post tensioning 1 3.8-29 Containment Shear Moments and Forces - Pressure 3.8-30 Containment Interaction' Diagrams 3.8-31 Instrumentation' Scheme for Structural: Integrity Test' 3.8 Containment Proof Test' Crack Mapping Locations Personnel ~ Access Lock' 3.8-33' 3.8-34 Equipment Hatch-3.6-35 DELETED 3.8 DELETED
.(). 3.8-37 3.8-38 DELETED DELETED 3.8-39 Typical Reinforcing Details of Drywell Floor 3.8 Reactor Stabilizer-Bracket 3.8-41 Reactor Pedestal <
3.8-42 Reactor Pedestal Stretchout i 3.8-43 Reactor Shield Stretchout 3.8-44 Foundation Mat Plan ;
3.0-45 ' Foundation Mat Sections 3.8-46 Foundation Details 3.9-1 RPV and Internals Vertical Dynamic Model
,3.9-2 Reactor VessellInternals t 3.9-3 Reactor Internals Flow Paths 3.9-4_ Fuel Support Pieces ;
-3.9-5 Jet 1 Pump i 3.9-6 Steam Dryer 3.9-7 Pressure Nodes Used for Depressurization Analysis 3.9-8 Transient Pressure Differentials Following a Steamline Break 3.10-1 Typical Vertical Board- ,
-3.10-2 Instrument Rack
,3.10-3 Typical' Local ~ Rack-3.10-4 NEMA Type 12' Enclosure 1 3.10-5 Static Seismic. Analysis of Standard Enclosures 3.10-6 Illustrations for Sample Panel Frequency-
'(])E Analysis -
3.0-xxviii 4
-,4s- - , - - , , . , - ,#e4r,y.--. ...y_. ,,,, .ew ra4.-7,.g,u , - ,rm.,, .c ,w, .as+ r w w w w - r wm m " a--pMvr* *=re---*a fr- w +
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 Piping Run - a main or branch run.
(])
Terminal End - piping originating . at structure or components (such as vessel and equipment nozzles and structural piping anchors) that act as rigid constraint to the thermal expansion.
Typically, the anchors assumed for the piping code stress analysis are considered terminal ends. The branch connection to the main run is one of the terminal ends of a branch
.run, except where the branch run was classified as part of a main run as defined above.
3.6.2.1.1 Break Location in ASME Section III Class 1 Piping Runs Postulated pipe break locations are selected in accordance with the intent of Regulatory Guide 1.46, USNRC Branch Technical Position APCSB 3.1, Appendix B, and as expanded in NRC Branch'
, Technical Position MEB 3-1. For ASME Section III, Class 1 piping systems, the postulated break locations are as <
follows:
- a. The terminal ends of the pressurized portions of the run. (Terminal ends are extremities of piping runs that connect to structures, equip-ment, or pipe anchors that are assumed to act as rigid constraints to free thermal expansion O of piping. A branch connection to a main piping run is a terminal end for a branch run, except when the branch and the main run is modeled as a common piping system during the piping stress analysis.)
- b. At intermediate locations between the terminal ends where the maximum stress range'between any two load sets (including zero load set),
according to Subarticle NB-3600 ASME Code Sec-tion III for upset plant conditions and an in-dependent OBE event transient, exceeds the following:
- 1. If the stress range calculated using Equation (10) but is not of the Code exceeds 2.4.S,ll
, no brerks wi be postulated greater than unless the cumul 3 S@tive usage factor exceeds 0.1.
- 2. The stress ranges, as calculated by Equations I (12) or (13) of the Code, exceed 2.4 S or if the cumulative usage factor exce$ds 0.1 when Equation (10) exceeds 3 S,.
.m c. In the event that two or more intermediate loca- m U tions cannot be determined by stress or usage 3.6-8
i r
LSCS-FSAR: AMENDMENT:39-OCTOBER 1978-
.' factor-limits, a total of-two intermediate loca-
'tions shall be identifiedcon at.. reasonable basis, hs 7: ' '
for each piping run:or branch run. .
(Reasonable basis.shall be one.or more of the following: ,
l.- Fitting locations.
- 2. Highest stress.or usage factor locations.. )
Where more than two such' intermediate locations are possible using the. application of the above>
reasonable basis, those two locations possessing the greatest damage potential.will be used. A break- '
at :each end of a ' fitting may be classified as two discrete break locations where the stress'. analysis t
- is sufficiently detailed to differentiate-stresses ~
at. each postulated break.) ,
Conformance to the.above pipe break = criteria is demonstrated '
'in Figures 3.6-1, 3.6-la, 3.6-2, and 3.6-2a.
() 4 l
l l
1 1
i I
3.6-8a 1
i
-,--,.-,.,,,,~3.,- -+-,.--...#..,.,me.-- c.-.,. --., %.w,-.,..# .w-,, ,,,m. ,..% -
..-~w.., +. w - r,, ,5e....-.7,#, ..--.w,,w--, v .W ww,... w-
LO.; 0 D '
TABLE 3.6-2
' RECIRCULATION PIPING SYSTEM OPERATING STRESSES AT BREAK LOCATIONS
- STRESS RATIO PER;
, EO . -(10 ) EQ. (12) EQ. (13 )
~
BREAK ' JOINT. Sn Sc S USAGE BREAK BREAK BASIS
'ID** NO.*** 3Sm- 3Sm 3Sm FACTOR TYPE PARA. NUMBER
-A. Recirculation Loop A
-S1 lI 1.01 0.13 0.59 0.O Cire. : Terminal End
- 3. 6. 2.1.1. (a)
I w
- 36-c, 36 L 9I 1.36 0.42 0.70 0.39 Circ., 3. 6. 2.1.1. (b) (2)
.m e.
s long. m 6
i
- b. D c,D6L 36J- 1.52 0.17 0.89 0.80 circ. 3.~ 6. 2.1.1. (b) (2 ) O
" _long., i F2 60I 1.04 0.22 0.68 0.02- Circ. 3. 6. 2.1.1. (c) 7 i F8 52I 1.36 0.15 0.73 0.63 ~ Circ. 3. 6. 2.1.1. (b) (2) $-
~ D9 - 41I~ 'O.42 .0.02 0.31 0.0- Circ. 3. 6. 2.1.1. (c) 1 F19 68I 1.57' O.45 0.73 0.91 Circ. 3. 6. 2.1.1. (b) (2)
F25 76I 1.13 0.30 0.68 0.04 Circ'. 3. 6. 2.1.1. (c) op.
03:
< 8 trj .
O 2:
} *The-value of 2.4Sm for Recirculation Piping Material is'39,960'. psi ~. -E E
- For Break'ID'see Figure 3.6-1. Ny i
- For' joint.no., see. Figure 3.6-la. ga 3
-a w a C3 D 4
4
O O O 1
TABLE 3.6-2 (Cont ' d)
- STRESS RATIO PER:
BREAK- JOINT' Sn Sc S USAGE ' BREAK BREAK BASIS ID** NO.*** '3Sm 3Sm 3Sm FACTOR . TYPE PARA. NUMBER F6 64J 0.78 0.13 0.54 0.0 Circ. Terminal End
- 3. 6. 2.1.1. (a)
F12 56J 0.90 0.23 0.54 0.0 circ. 3.6.2.1.1 (a)
F17 48J 1.00 0.16 0.59 0.01 Cire. 3. 6. 2.1.1 (a )
4 F23 72J 0.84 0.16 0.56- 0.0 Circ. 3. 6. 2.1. l ' (a) $-
tn m
F29 80J 0.81 -0.18 0.55 0.0 Circ. 3.6.2.1.1 (a) h
- p b B. Recirculation Loop B M S1 lI 1.01 0.16 0.54 0.0 Circ. Terminal End 3.6.2.1.1 (a)
D6 c, D6L 36J- 1.55 0.17 0.88 0.95 Cire . , . 3.6.2.1.1 (b) (2)
- long.
F2 60I 1.13 0.33 0.69 0.03 Circ. 3. 6 . 2.1.1 (c )
F8 52I 1.49 0.33 0.75 0.72 Circ. 3 . 6 . 2 .1. l (b ) (2) o ; ,
o 8
O2
- The Value of 2.4Sm for Recirculation Piping Material is 39,960 psi. E@
- For Break.ID,_see Figure 3.6-2. Wg-
- For joint ID,-see Figure 3.6-2a. He-w-
MW CO W i
_ _ _ _ _ . _.__.-___._.__._-_-_____.___.___.__r_-____ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . - _ . _ - - _ _ -
_ _ _ _ - -c~ ~?-~ '
-w r --u --
~ w --
,-w - 4 '"
' TABLE 3.6-2 (Cont'd)-
EQ. (10) EQ. (12)' EQ . (13 ) ..
. BREAK JOINT Sn Sc -S USAGE BREAK BREAK BASIS-
--I D ' 'NO. 3Sm 3Sm -3Sm FACTOR TYPE PARA." NUMBER D9 41I. 0.42. 0.02 0.31' O.0. :Cire.. 3 . 6 . 2.1. l (c )
F19 68I .1.46 0.42 0.71 0.39'. -:Cire. 3. 6.2 1.1 (b) (2)
'F25 76I- 1.17 0.30 0.67 0.05 -Circ. 3 6. 2.1.1 (c ) '[
~
F6 64J- 0.82- 0.20 0.54 0.0 Cire. Terminal'End 1 3 . 6. 2.1. l (a ) 6tn w .o ,
S6J 0.20 0.54 1 Circ. 3. 6. 2.1.1 (a ) - (n
- . . . F12 0.89- 0.0) m 4 ..
w F17- :48J. 1.09- 0.23 0.62 0.01 ' Circ.- 3 . 6 . 2 .1.1. (a ) . $ -- ..
o- m
'F23- '72J J0.84 0.10 0.55- 0.0- Cire. 3. 6. 2.1.1 (a ) - ' ,
i' .. .
[ 3. 6 . 2.1.1 (a )
~
F29: .80J 0.81 0.18 .0.55 ~ 0. 0 Circ.
4 4
.hh]
eM
+ -OZ-WO i M3 i MM
- Z. .
- He-
! w.
QW
' co w
?
k
+
0 h.
3 'D
...,.,:. ~-.,,--;- :-...,..-_., ..-_..-..-,c , . _ _ , . .
t AMENDMENT 39 OCTOBER 1978-r 0
- [} - 770 m
\J . <
si .
- .so
,,o. .;. .
,2,
- s. . .
F6 : .
e D. %)
i s % :
l g- C kj l RHH ILOOP ' A' ON LY) () INDICATES POSTULATED
\ CIRCUMFERENTIAL BRE AK "o'c^$ 'o87u'^7'c
( RWR 0 ' LONGITUDINAL BREAK AT ANY ANGLE PAR ALLEL TO PIPC Cg Q BREAKID O '
p V v .
LA SALLE COUNTY STATION FIN AL S AFETY AN ALYSIS REPORT s' -
FIGURE 3.6 '
RECIRCULATION LOOP-A WITH POSTULATED BREAKS
.- , , y:n
- ,, , , ,. .-- , . i . 1, .
, , . i l
\1 .. i i
/ e ,
~
Y . gg 70 h 33 270* K 1 38 1 208 h - on 99 g .
_ p/ 90" 4 -
'l 180* 2 u 3 21B 6 . t 68 i -
3 62- $
L COC9 DIN ATE o g
l 25B 78
$9 ,
~' '
g 60 58 3 IS SSB
, 258 '582 IC
~ %,
568}
%499 '
a
( lli
[l55 33B 154 AtlXILI ARY COC3 DIN ATE I12 SYSTEM 91
, fik 358 8
I44 115 10 o < ,
?
- n- 298 28B 4.30 428 948 332-1861 93B 1 153 n 417] 14 38
'O F "
40B l39B 78 12 40 %
119 _ ) g is 778 123 78] - 150 l .- 308 139 o.
898 i36 149 35 l
l J (22 22E 16 IN 3
C l -
^
B '
798 ~8tB .gg ,93 l -
124 22 s 3 368 90B S #
\ gg , gg \" qf b863 3;g 23 k26 . f j 43B i
. 835 ) 128
% 2g
_13: %
,277 28 [ [30
- 448 + 460 8B
- I3I t 878 I * >3 132 23
];, } _ ~ ,9
"! L,;
~ ~ _1: . J ' _ .1 n / t mi . 2 n an. e 4 - 1
~-
t I e ! = I '"
I AMENDMENT 39 0CTOBER 1978 i
S 4 + No. l
'v^
g,g 2B mT :ei s sNuaBER [
178 gh80J RES iR AIN I EOUIPM EN T NitMBER SPRING HANGER ll AA, ;
138
+7 7i 70 y #
!46 76 00 HGR V ~
SIB
- g5 y f-
. A" YY-B T V I
'" b 7 S28 43' BOUNDARY J
93 ELEM NO '
30 P,JPE AND SUSPENSION SYMBOLS
.A 508 .
39 o
6B --: .
~
'O' 75B 37 ice
~
107 7,
2 .
0 . 83 36J 73g 1 7 ,3 6081 68 '0 f.
e 36 "SS 's Xes 72 e6 104 1 _
34
/ 87 103 102
% 2B 63B 101
. 68 708 6 ** 78 93,f s 66B ~
LA SALLE COUNTY STATION FIN AL S AFETY ANALYSl'* REPORT FIGURE 3.6-la PIPING ELEf1ENT DIAGRAM -
RECIRCULATION LOOP-A l
(T i' ~ ~t- :e_- L - A.:~~' ~
AMENDMENT 39 OCTOBER 1978 oc
,~~1 .
e vo* -
I w 1 s, g ov
) o re 6
U Y a s,
)
~ l
)
b
{} HHR (*) INDICATES POSTULATED
( CIRCUMF E R ENTI AL BRE AK 8 h INDICATES POSTULATED
( ) RWR LONGITUDINAL BRE AK AT ANY ANGLE PAR ALLE L TO PIPE C g C BREAKID D
9
\
v LA SALLE COUNTY STATION FIN AL S AFETY AN ALYSIS REPORT s
FIGURE 3.6-2 J
RECIRCULATION LOOP-B WITH POSTULATED BREAKS
~ s
g , .g , 1iil i1- a 'l "* I ' 'l'
( ..
o Y 88
, - A 3 98 4
I o JP 238 ' -
m<
P RY COORDIN CE '
3
. Fg <
a -258a
, y' 578 f-
_I O B .
Fc . 4 - gg0I 53 C
0258 5
. S F-A A 5 9
-1 AUXILI AR Y C00RSINATE S/ STEM 268 7BX 7 348 IM r~ 338 f Is4 =
0 Nh-
. RECRC *
- 28Bk 298 h N I h 418 #
152 . 2B '
/
11 15 3 1 151 12 408 ,
~
14 7 15 0 108 898 ,,
19 8 l
' 16 h t-14 15
/ h
% i35 94
$368 - 9th 32B <#
i /; 20
- 708 Q 21 %, .
i 458 , 1
- 438 26 G3lBX Y ' 4d24WK 2B , J '
. Q%
l L ,
l 4
w
'l 9
l
.n i l a i - 8 ;
- i 5 l
- 2 '
I
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AMENDMENT 39 0CTOBER 19?8 4 + e. _ I ,
A " i1NE o Sj28 HB E
M EOU!PMEN- l SCRING I Ry,g i4B \'80J RESTR AIN~ N .; V B E 4 # h*MGER J bfG t, l
? 8
-} ,
7 5 71 4 Y 46 h ,
77 5
CONSTAN-SPRT HGR bNUOOCR I MO GATE VALVE j
g '
i u :
i a,e 548) A p p j.
f33 p%hTb,f4- $28'a t
HAND WHEE'_
GATE VALVE CHECK
'!/,L V E f
- PIPE AND SUSPENSIC'i SYF90LS k NOTES:
L ELEMENT f"JMBERS tilth '.ETTER SUFFIX 'X' ARE NOT EFFECTIVE IN TrE ANALYSIS.
30 4 78 ^y\
488 N
37 86 60B o -I 85 7 82 "
799 o 98
- 688 ' 7:02 8 58 g , g4 i ' 9 4 36) 8 '8 101 p IC g 89 95 p658 g3,g% g 628 33' y 9, 90 4 1
~ l}928 g / ..fg g 5
g32 x 30 29 LA SALLE COUNTY STATION FIN AL S AFETY ANALYSIS REPORT FIGURE 3.6-2a PIPING ELEMENT DIAGRAM -
RECIRCULATION LOOP-B .
k I
i . t io i : ,, i
r L SC S-FSAR - AMENDMENT 39 '
OCTOBER 1978 MECHANICAL SYSTEMS AND COMPONENTS
-Q 3.9 3.9.' special Topics for Mechanical Components 3.9.1.1 Desion Transients This subsection describes the ' transients which were used to demonstrate the design of the ASME Code Class 1 core supports, reactor components, piping . systems and mechanical. equipment. .The-transients and combinations of transients are classified with respect to the component operating condition categories identified as " normal," " upset,"." emergency," " faulted," or
" test" in the ASME Boiler and Pressure Vessel Code, as applicable.
3.9.1.1.1 Thermal Transients The thermal transients used in the design and f atigue analysis of ASME Code Class .1, core supports, reactor components, piping systems and mechanical equipment are listed in Table 3.9-33.
These thermal transients were derived from those established for the RPV nozzles and are tabulated for the major piping systems connected to the RPV. Other reactor coolant pressure boundary lines whicn connect to these major ' lines were assumed to have identical thermal transients..
3.9.1.1.2 Hydrodynamic Transients The hydrodynamic transients associated with safety / relief valve (SRV) actuations ' and postulated LOCA events are described in the DFFR (Ref erence 7) . These hydrodynamic transients were defined subsequent to design, procurement and delivery of the LSCS mechanical systems and components. The effects of the hydrodynamic transients on plant structures are presented in the LSC3-DAR (Ref erence 1) .
The static and dynamic load. ef f ect on the mechanical systems and components resulting from the hydrodynamic transients were evaluated in a design assessment. This design assessment evaluation considered the dynamic response of the containment and structures located on the containment base mat and the -response on mechanical systems and components resulting from structrual excitations. The load combinations used for the design assessment evaluation are defined in Table 3.9-25. Results reported in Tables 3.9-2 through 3. 9-22 reflect the design-basis loads and the loads from hydrodynamic transients, including.a postulated LOCA of reactor vessel saf e-end weld to pipe rupture (annulus pressurization) .
Descriptions of the hydrodynamic transient loads and the shield wall' annulus pressurization loads are provied in the following paragraphs.
3.9-1
- .-~. _ ~ .. _ _
~. --. . ._ -. .- . _ _- - .-
LSCS-FSAR' AMENDHENT 39 -
OCTOBER 1978 I) 3.9.1.1.2.1 Safety / Relief Valve ( SRV) Actuation Loads (Structural Excitations)
The actuation of SRV's causes pressure disturbances in the sup-pression pool water that produce. oscillatory transient pressure ;
forces on the suppression pool boundary. These pressure oscillations result in structural excitations which impart dynamic responses to the attached piping and equipment.
Safety / reflief valve hydrodynamic transients and associated calculational procedures are f urther described in References 1 ,
a nd 7. Structural response loads 'due to the following SRV actuations were ' evaluated:
- a. actuation of all valves,
- b. actuation of the automatic depressurization system (ADS) valves,
- c. the actuation of two adjacent valves, and 1
- d. actuation of a single valve (subsequent actuation) .
Reference 7 describes the characteristics of the SRV load as a f unction of the piping configuration and the discharge device
(]) (rams head or quencher) located at the exit of the SRV line.
quencher device typically produces lower dynamic loads. The LSCS The design assessment evaluation was based on loads calculated with a j rams head device, thus producing bounding SRV loads for equipment I analysis.
3.9.1.1.2.2 Loss-of-Coolant Accident (LOCA) Loads The postulated LOCA event gives rise to several hydrodynamic phenomena which cause transient pressure loads on the suppression pool boundary. These phenomena include main vent clearing, pool !
swell, condensation oscillation, and chugging. A brief
. description of each follows:
- a. Main Vent Clearing 1
Following a postulated LOCA, the drywell pressure j increases due to blowdown of the reactor system, i Pressurization of the drywell will cause the water iaitially in the vent system to be accelerated out through the vents. During this water expulsion process the resulting water jets cause impingement ;
loads on local containment structures. However, this water clearing process has oeen found to produce ]
insignificant structrual response loads on the drywell piping and equipment.
( 3.9-2
$ .I LSCS-FSAR AMENDMENT 39 OCTOEER 1978
- b. Pool Swell,;
Following main vent clearing, an air / steam bubble forms at the vent exit - This causes a. hydrostatic pressure increase in the pool water.' resulting in a loading' condition on the pool boundaries. :The' steam condenses in the pool. However, the continued-addition and expansion cf the drywell air. causes the pool volume to swell, resulting in the rise of the
, pool su :f ace and associated. drag and impact loads on' L e urrounding structures. References 1 and 7 provide f arther details and' calculational procedures. This
. phenomenon has been found to produce negligible structural response -loads for . piping and equipment. ,
- c. Condensation Oscillation Loads: '
Evaluation of test results for the steam condensation cycle has revealed the occurrence of a dynamic ' load - during high mass-flow of steam into the suppression pool. This low-pressure, symmetric, sinusoidal pressure fluctuation occurs over n low-frequency range which actn on the pool boundary. These ' fluctuating pressures excite the structure proceeding - low-frequency responses on the drywell piping systems f(*) andLequipment. .Ref erence 7 ' (Revision 3). provides
~
v f urther description and calculation procedures. ; Condensation. oscillation has been found ta) produce I negligible structural response loads for piping and . equipment. '
~
- d. ,Chuggin_g The application of-chugging-loads is described in the
" Mark II Phase I - 4T Tests Application Memorandum u submitted to the 'NEC in June 1976. This Application
-Memorandum was used .for' the LSCS Design Assessment )
Evaluation to expedite licensing review. Additional l methods for'the application of chugging loads are l being developed {NEDO-24014, _ June 1977; NEDE-21669-P, 1 February 1978). -These new methods for the application of multivent chugging loads provide a i realistic. load definition, and the results of these new methods, when completed, will , provide the final design-basis load definition. 3 . 9.1. 1. 3 Annulus Pressurization l Annulus pressurization reiers to the loading on the shield wall and reactor vensel caused by. a pcstulated pipe rupture at the A (~ '
. reactor pressure vessel nozzle ' safe-end to pipe weld. The pipe 1
i rupture. assumed is an instantaneous guillotine ruptur e which allows; mass / energy release into the drywell and annular region
- 3. 9- 3
i LSCS- FSAR AMENDMENT 39 OCTOBER 1978 (]) between the biological shield wall and the reactor pressure vessel- (RPV) . The mass and energy released during this postulated pipe rupture
. cause s:
- a. A' rapid asymmetric decompression acoustic . loading of the annular region between the vessel' and shroud from the pipe hreak at or beyond the vessel nozzle safe-end weld.
- b. A transient asymmetric differential pressure within. I the annular region between the biological. shield wall i and the . reactor pressure vessel (annulus press urization) .
- c. A jet stream release of the reactor _ pressure vessel inventory and the impact of the ' ruptured pipe against the pipe whip restraint attached to the biological' shield wall.
The results of the mass and' energy release evaluation are then - ! used to produce a dynamic structural-analysis (f orce-time history) of the RPV and shield. wall. The force-time history . 1 output from the dynamic analy. sis is subsequently used to compute (')
%J loads on the reactor components. l The postulated pipe rupture at the weld between recirculation or feedwater piping and the reactor nozzle safe-end leads to a high flow rate of water and steam mixture into the annulus between the RPV and the shield wall. Calculation of the mass / energy release .
is perf ormed using the generic method for short-term mass releases. This method is described in Attachment 6. A to Chapter 6.0, where a sample calculation is also provided. This mass energy release also results in acoustic pressure and jet loads. 3.9.1.1.3.1 Acoustic Loads Because the boiling water reactor (BWR) is a two phase system that' operates at or close to saturation pressure (1000 psi) , the differential pressure across the reactor shroud is of short duration, and the BWR system is not subjected to a significant shock-type load with respect to structural supports. This short-duration acoustic load is confined to a bending moment and shear force on the reactor pressure vessel and reactor shroud support. 3.9.1.1.3.2 Pressure ' Loa ds The pressure responses of the RPV-shield wall annulus for a recirculation suction line postulated rupture and a f eedwater line postulated rupture were investigated using the RELAP4 O. computer ~ code. An asymmed ~ic model, using several nodes and flow Epaths, was developed for tue analysis of the recirculation and 3.9-4
LSC6-FS AR AMENDMENT 39 OCTOBER 1978 O f eedwater line ruptures. Further description of these analytical models and detailed. discussion of the analyses may be found in section 6.2. 3.9.1.1.3.3 Jet Loads structural loads on the vessel and internals, jet thrust, jet impingement and pipe whip restraint loads were considered in conjunction with the above mentioned pressure loads. Jet thrust refers to vessel reaction force which results as the jet stream-of liquid is released from the rupture. Jet impingement refers to the jet stream . force which leaves the broken pipe and impacts the vessel. The- pipe whip restraint load is the force which - results when the energy absorbing pipe whip restraint restricts the pipe separation to less than one full pipe diameter. These jet loads are calculated as described in Reference 8. 3.9.1.2 Computer Programs Used in Analysis The following sections discuss computer programs used in the analysis of specific components. The GE computer programs are. maintained either by General Electric or by outside computer program developers. In either case, the quality of the programs a nd the computer results are controlled. For each program, one or more individuals are assigned. Their duties are: O' a. to keep abreast of the capability, the software i contents and the theory of the program, i l
- b. to run test -cases and maintain the reliability of the program, and
- c. to advise users on the proper usage of the program and the correct interpretation of computed results.
All necessary modifications are coordinated and verified by the responsible individuals. Thus, users' confusion .over the changes is avoided and the high reliability of these programs is maintained. . 3.9.1.2.1 Reactor Vessel The computer programs used in the preparation of the reactor vessel stress report are identified and their use summarized in the following paragraphs. 3.9.1.2.1.1 CBSI Program 711 "GENOZZ" The GENOZZ computer program is used to proportion barrel 'and double taper type nozzles to comply with the specifications of (~% the ASME Code Section III and contract documents. The program - (_/ will either design such a configuration or analyze the configuration input into it. If the input configuration will not 3.9-5
'L SCS-FS AR ' AMENDMENT 39 l OCTOBER 1978
' {) comply with the specifications, the program will modify the design and redesign _it to yield an acceptable result.
3.9.1.2.1.2- CBSI Program 943 " NAPALM" ! The basis for the program NAPALM, Nozzle Analysis Program--All Loads Mechanical, is to _ analyze ~ nozzles for mechanical loads and find -the maximum stress . intensity and location. Specified
. locations. are analyzed from the point of application 'for 'the mechanical loads.- At each location the maximum stress intensity is calculated. for both the inside and outside surfaces of the nozzle. The program gives the maximum 1 stress' intensity for both the inside and outside surf aces. of the nozzle as well' as its angular location. The principal stresses are also printed. The i stresses resulting from each component of loading (bending, a xial, shear, and L torsion) are printed, as well as the loadings which caused these stresses.
3.9.1.2.1.3 CBSI Program 1027 This program is a computerized version of the analysis _ method contained in the Welding Research Council Bulletin F107, December 1965. Part of this program provides for the determination of the shell stress intensities (S) at each of four cardinal points at both the upper and lower 'shell plate surfaces (ordinarily considered outside and inside surfaces) around the perimeter of.a loadedL , attachment on a cylindrical or spherical vessel. With each-determination _ of S, the components of that are also rdetermined. . (2 normal stresses, ox .and cy, and one shear stress T) . This ) program 'provides the same information as . the manual calculation, ! and the input data is essentially the geometry of the vessel 1. and ; attachment. 3.9.1.2.1.4 CBSI Program 846= This program computes the required thickness of a hemispherical
~
- head with a large number of circular parallel penetrations by means of the area replacement method in accordance with the ASME Code, Section III. i In cases were the_ penetration has a counterbore, the thickness'is determined so that the counterbore does not penetrate the outside -
surface of the head.- 3.9.1.2.1.5 CBSI' Program 781 "KALNINS"
~
This program is a thin elastic shell program for shells of a revolution. This program was developed by Dr. A. Kalnins of-ss . Lehigh University. Extensive revisions and improvements have
.1 'been made by Dr.' J. Endicott to yield the _CBSI version of this
- 3. 9- 6 x
v v *=*rt*r' r"v-- v -t r ia r e-e * ~'vte w"we ev a ' *-r**+ -wr'- **'v* *'ns'*-t*sw+"e- w -* ,-*~=< - w --* -v me-rr*"'=--*'e=-' - - - * * ** -
, , - . - . . . - _ ~ ~. - . .
LSCS-FS AR - AMENDMENT 39 OCTOBER 1978 () program. 9. The ' basic- method of analysis was published in Reference 3.9.1.2.1.6 -CBSI Program'979 "ASFAST" ASEAST Program (Program 979) -performs the stress analysis of axisymmetric, bolted closure flanges between head and cylindrical shell. The KALNINS thin shell program (Program 781) is used to establish the shell influence coefficient and to perform detail stress a nalysis .of the vessel. The stresses and the deformations of the vessel.can be computed-for any combination of.the following axi-symmetric loading:
- a. preload condition, [
- b. internal pressure, and
- c. . thermal load.
3.9.1.2.1.7 CBSI Program 766 "TEMAPR" This program will reduce any arbitrary temperature gradient through the wall thickness. to. an equivalent linear gradient. The ([) resulting equivalent gradient will have the same average temperature and the same temperature-moment as the given temperature distribution. Input consists of plate thickness and actual temperature distribution. The output contains the average temperature and total gradient through the wall thickness.- 3.9.1.2.1.8 CBSI Program 767 HPRINCESS"
- The PRINCESS computer program calculates the maximum alternating stress amplitudes from a series of. stress values by the method in "
3 Section III of the ASME Pressure Vessel Code. 3.9.1.2.1.9 CBSI Program 928 "TGRVn l The TGRV program is used to calculate temperature distributions in structures or vessels. Although it is primarily a program for solving the heat conduction equations, some provisions have been _ made for including radiation and convection effects at the surfaces of the vessel. The TGRV program is a greatly modified version of the TIGER heat transfer program written about 1958 at Knolls - Atomic Power Laboratory by A. P. Bray. There have been many versions of TIGER in' existence, including TIGER II,- TIGER II' B, TIGER IV, and TIGER V, in addition to TGRV. O
- 3. 9- 7
9 i I LSCS-FSAR AMENDMENT 39 OCTOBER 1978 h Thi program utilizes:an electrical network analogy to obtain the temperature; distribution of any_ given system as a f unction of j time. .The finite _ difference representation of the three-' L dimensional equations- of heat _ transfer are repeatedly solved for j small-time increments and continually summed. Line ar - mathematics are' used to solve tthe mesh network f or every time interval. Included.in:the analysis are the three-basic forms-of heat transfer, -conduction, radiation and convection, as well as internal ' heat. generation._ Given ' any odd-shaped structur'e which can be . represented by a three-dimensional; field and its geometry,. physical properties,
' boundary conditions, and internal. heat generation rates, TGRV will calculate and give as output the' steady-state or trans' ient
~
i temperature distributions in the structure as a function of time. 1 3.9.1.2.1.10 cBSI Program 962 - "E0962A" Program E0962A is one.of a group of programs (E0953A, E1606A, E0962A, E0992N, E1037N, 'and E0984N) which are used together to determine the temperature _ distribution and stresses in pressure vessel' components by the finite element method. 1 Program-E0962A is primarily a plotting program. Using the nodal l temperatures calculated by Program E1606A or Program E0920A and i
/~i - tne node . and element cards for the finite- element model, it I
\~'
calculates and plots lines of constant temperature .(isotherms) .
~
i These isotherm plots are used as part of the stress report to I present the results of- the thermal analysis. They are also.very 1
-useful in determining at which. points in time the thermal' stresses should be determined.
In -addition to its plotting capability, the program can also determine the temperatures of. some 'of the nodal points by interpola tion. This feature <of the program is intended primarily for use with the compatible TGRV and finite element models that are generated:by Program E0953A. 3.9.1.2.1.11 CBSI Program 984 Program 984.is used.to calculate the' stress intensity of the i stress- diff erences, on a component level, between two dif ferent stress conditions. The calculation of che stress intensity of stress component differences (the range of stress intensity) is required by'Section III of the ASME Code. 3.9.1.2.1.12 ' CBSI Program 992 -_G_ ASP The' GASP computer program,. originated by Prof. E. L. Wilson of the University of California at Berkeley, uses the finite-element cr~g7 method to determine the Latresses and. displacements of plane or V axisymmetric structures of' arbitrary geometry. i
- 3. 9- 8
. . - . , . , _ . _ . . - . - . - . _ . _ .- ~ _ . _ . _ , _ _ _ _ -,_._-._-- .1
( LSCS-FSAR- 'AMENbMENT 39. OCTOBER 1978- o t E(])~ -This ' program ' determines the stresses and: displacement of plane or
'axisymmetric structures 'using the ' finite element. method. The structuresLmay have arbitrary geometry' and have linear or
' nonlinear. material properties. .The loadings may1be thermal, mechanical,' accelerational, or a Leombination of these..
The structure lto,beranalyzed is broken up into a finite number of discrete elements .or " finite elements" which are interconnected l at . finite number or : " nodal points" or " nodes. " JThe. actual. loads on the structure :are. simulated by statically equivalent ' loads
? acting at . the ' appropriate nodes. LThe' basic input toLthe program consists of - the geometry ' of . the ' stress ' modeliand the boundary condi tions. - The: program then 'gives the' stress components at the
- center of each. element and the displacements at the nodes, consistent with.the prescribed boundary conditions.
3.9.1.2.1.13 .CB&I Program 10 37 "DUNH AM' S"' DUNHAM'S program is a finite ring ' element stress analysis prog ram. It will! determine the stresses and displacement of axisymmetric_ structures of arbitrary geometry subjected to either. axisymmetric loads or nonaxisymmetric loads represented by a I Fourier 's eries. This program is similar to the' GASP program (CBSI f 992) . The major differences:are.that DU'NEAM'S can handle nonaxisymmetric 1 loads- (which requires that each node have three degrees of i freedom) , and the material properties _ f or DUNHAM'S must be - constant. As in GASP, the loadings may- be thermal, mechanical, and accelerational. ! 3.9.1.2.1.14. CBSI Program 1335 This program models the baffle plate as. a_ continuous circular - l plate for the purpose of computing stresses . in the ' shroud' j support.. The program allows the baffle plate to be . included in l l CBSI Program 781 as. two isotropic parts and an- orthotropic ' portion at the middle 1(where the diffuser holes are -located). 3.9.1.2.1.15 CBSI Programs 1606 and 1657 " HAP" The HAP program is an 'axisymmetric nonlinear heat analysis program. It is a finite-element program and is used ' to determine nodal temperatures in a two-dimensional or axisymmetric body subjected to transient disturbances. , Programs 1606 and = 1657 are I identical except that 1606 has a-larger storage area allocated- l and can.thus be used to solve larger problems. The model for program 1606 is compatible with CBSI stress programs 992 and 1037. ti l
-A/- 1
- 3. 9- 9 l
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 {} 3.9.1.2.1.16 CBSI Program 1635 Program 1635 offers the following three features to aid the stress analyst in preparing a stress report:
- a. Generates punched card input for program 767 (PRINCESS) from the ' stress output of program 781 (KALNINS) .
b. Writes a stress ' table in a f ormat such that it can be incorporated into a final stress report.
- c. Has the option to remove through-wall thermal bending stress and report these results in a stress table similar to the one mentioned above.
3.9.1.2.1.17 CBSI Program 953 The program is a general purpose program, which does the following:
- a. It prepares input cards for the thermal model.
- b. It prepares the node and element cards for the finite l element model. l
() c. It sets up the model in such a way that the nodal points in the TGRV model correspond to points in the finite element model. They have the same number so that there is no possibility of contusion in transferring temperature data from one program to the o ther. i 3.9.1.2.1.18 GE Program _DySEA This program is a General Electric proprietary program developed specifically to compute siesmic and dynamic responses on the reactor pressure vessel structrues, internals, and reactor pedestal and shield wall complex. It calculates the dynamic response of linear structural systems by either temporal model superposition or response spectrum method. Fluid-structure interaction effect in the reactor pressure vessel is taken into account by way of hydrodynamic mass. The DYSEA program was based on the SAP-IV program (see Subsection
- 3. 9.1. 2. 2.1) with added capability to handle the hydrodynamic mass-effect. Structural stiffness.and mass matrices are formulated similar to SAP-IV. Solution is obtained in time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmark's method. A response spectrum solution is also available as an option.
3.9-10
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 /~T 3.9.1.2.2 Piping O . The. computer programs used in the analysis of NSSS piping systems within 13E's scope of supplies are identified and their use summarized in the following subsections. 3.9.1.2.2.~1 Structural Analysis Program - SAP 4 3.9.1.2.2.1.1 Application The structural analysis program SAP 4 for- the static and dynamic analysis of linear structural systems is the result of several years.research and-development experience. The program has proven to be a very flexible and efficient analysis tool. The first version of the SAP Program was published in Reference '10. An improved static analysis program namely SOLID SAP, or SAP 2, was presented in Reference 11. Work was then started on a new static and dynamic analysis program. The program SAP 3 was released toward the end of 1972 (Reference 11). SAP 4 has the additional analysis capability of out-of-core direct integration for the time history analysis. (Reference 12) . 3.9.1.2.2.1.2 Program Organization The structural. systems to be analyzed may be composed of combina-r- tions of a number of diff erenct structural elements. The program (_)s presently contains the- following element types:
- a. three-dimensional truss element,
- b. three-dimensional beam element,
- c. plant stress and plane strain element,
- d. two-dimensional axisymmetric solid,
- e. three-dimensional solid,
- f. thick shell element,
- g. thin plate or thin shell element,
- h. boundary element, and
- 1. pipe element (tangent and bend) .
These structural elements can be used in a static or dynamic analysis. The capacity of the program depends mainly on the total number of nodal points in the system, the number of eigenvalues needed in the dynamic analysis, and the computer used. There is practically no restriction on the number of () elements used, the number of load cases, or the order and band width of the stif fness maxtrix. Each nodal point in the system 3.9-11 e
. . . . 3- _ , ,
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 (~') can have from zero to six displacement degrees of freedom. The \> element stiffness and mass matrices are assembled in condensed form, therefore, the program is equally efficient in the analysis of one , two , or three-dimensional systems. The f ormation of the structure matrices is carried out in the same way as for a static or dynamic analysis. The static analysis is continued by solving the equations of equilibrium followed by the computation of element stresses. In a dynamic analysis the choice is between:
- a. frequency calculations only,
- b. frequency calculations f ollowed by response history analysis,
- c. frequency calculations followed by response spectrum analysis, and
- d. response history analysis by direct integration.
To obtain the frequencies and vibration mode shapes, solution routines are used which calculate the required eigenvalues and eigenvectors directly without a transformation of the structure stiffness matrix and mass matrix. This way the program operation rs and necessary input data for a dynamic analysis is a simple () addition to what is needed for a- static analysis. 3.9.1.2.2.2 Component Analysis / ANSI 7 l The ANSI 7 Computer Program determines stress and accumulative ; usage factors for thermal weight, seismic relief valve lift and i turbine stop valve closure (as applicable) conditions of loadings derived from the Structural Evstem Analysis, in accordance with NB-3600 of ASME Boiler and Preodure Vessel Code Section III. For Class 1 stress analysis, the program generates and prints hoop, bending, thermal discontinuity, linear temperature gradient, and nonlinear temperature gradient components of stress for each equation of suoarticle NB-3600 of Section III. Load combination results from possible load sets for Class 1 stress equa tions. The total stress (sum or component stresses) and the stress ratio (total stress divided by appropriate stress intensity limit) is printed for each Class 1 equa tion. The total stress (sum of the component stresses) and the stress ratio (total stress divided by the appropriate stress intensity limit) is printed for each one of the equations 9, 10, 12, and 13 of NB-3600. The alternating stresses and usage f actor are calculated per NB-3653.6. (VD 3.9-12
LSCS-FSAR AMENDMENT 39 i OCTOBER 1978 ) () 3.9.1.2.2.3 Pressure Design Minimum Wall Thickness Calculation / THICK l The computer program THICK performs minimum wall thickness calculation to the rules of 1974 edition of Section III, NB-3641.1 of the ASME Boiler and Pressure Vessel Code. The wall thicknesses are computed from the OD Calculation Equation (Equation 2.2-1) and ID Calculation Equation (Equation 2. 2-2) .
. The larger of the two wall thicknesses result governs.
3.9.1.2.2.3.1 Area Reinforcement /NOZARP The computer program NOZARP (Nozzle Area Reinforcement Program) performs an analysis of the required reinforcement area for opening s. The calculations performed by NOZARP are in accordance with the rules of the 1974 edition of Section III, NB-3643.3 and the proposed Code Case N/D 73-23 of the ASME Boiler and Pressure vessel Code. The rules of this subdivision are based on the area replacement technique, that is, if a portion of pipe material that is subjected to membrane stress due to pressure is removed, that material must be replaced in close proximity to the area of I removal. N^ZARP takes input data for the run pipe, branch pipe, and nozzle and performs the necessary calculation to indicate that the nozzle is either acceptable or not acceptable. In addition, NOZARP prints the calculation results to confirm the {} acceptable conclusion. 3.9.1.2.2.4 Dynamic Forcing Functions 3.9.1.2.2.4.1 Relief Valve Discharge Pipe Forces Computer Program /RVFORCE The relief valve discharge pipe connects the relief valve to the suppression pool. Under normal circumstances, the discharge end of the pipe is under water and the remainder of the pipe is filled with air. The water may be drawn up into the pipe if the air is less than atmospheric pressure. When the valve is opened, the transient fluid flow causes time dependent forces to develop in the pipe wall. This computer program computes the transient fluid mechanics and the resultant pipe forces. 3.9.1.2.2.4.2 Turbine Stop Valve closure Input to SAP /TSMOOD TSMOOD is written to serve as the guidelines concerning the use of several computer-based tools for calculating fluid functions acting on a main steamline due to turbine stop valve closure. These tools consist of the following:
- a. MOODY 8s Main' Steamline Program - A program which calculates forcing functions due to turbine valve closure.
O 3.9-13
y i
'LSCS-FSAR AMENDMENT 39 OCTOBER 1978
' ([) - b. EHPTSV - A time-share routine 1 which calculates the initial fluid properties along the main _steamline for the. case of steady flow with friction, the condition just r before turbine valve closure.
- c. DIRECTION. COSINES .R) use MOODY's main steamline program,.the-total length of the steamline is.needed, as well as:the distance from the reactor vessel to the.end of each straight section along the main 1 steamline for which force output is desired. To-subsequently use SAP for a dynamic analysis, the. j direction' cosines associated with each forcing f unction must be known. The DIRECTION COSINES calculates all the above information.
- d. CUT - A batch program removes' bias portion of the forcing :f unctions - before the output of MOODY's Main Steamline program can be used in SAP.
3.9.1.2.2.5 Integral Attachment /LUGSTR i The computer program "LUGSTR" was prepared to evaluate the stress in the pipe walls that are produced by loads applied to. the integral attachments.- The program was prepared based on the Welding Research Council Bulletin 198, including the evaluation (]) due to stress range and fatigue analysis. 3.9.1.2.2.6 Piping Dynamic Analysis Program /PDA The pipe whip analysis was performed .using the PDA computer program. PDA is a computer program. used to determine the response of a pipe subjected to the thrust force occurring af ter a pipe break. - The program treats the situation in terms of generic pipe break configuration, which -involves a straight, uniform pipe fixed at one end and cubjected. to a time-dependent thrust force at the other end. ~ A typical restraint used'to reduce the resulting deformation is also-included at a location between the two ends. _ Nonlinear and time-independent stress-strain relations are used for the pipe and the restraint. Similar .to the popular plastic hinge concept, bending of the pipe is assumed to occur only at the fixed end and at the location supported by the restraint. Shear deformation is also neglected. . The pipe bending moment-deflection' (or rotation) relation used for.these locations is obtained from a static nonlinear cantilever beam analysis. Using the moment-rotation relation, nonlinear equations of motion of the pipe are formulated usin; an energy consideration and the equations are numerically integraged in small time steps to yield time-history. information of the deformed pipe. 3.9-14 I
LSCS-FSAR . AMENDMENT 39 OCTOBER 1978 3.9.1.2.2.7 Miscellaneous' Analyses (]) 3.9.1.2.2.7.1- WTNOZ Computer Program WTNOZ is a timerhare program for piping weight and nozzle thermal displacement calculations. 3.9.1.2.2.7.2 NOZLOD Computer Program The NOZLOD computer program calculates the resultant shear loads and resultant moment loads on the reactor . pressure vessel nozzles. 3.9.1.2.3 ECCS Pumpsiand Motors An equivalent static computer analysis was performed 'on the ECCS pump motor rotor shafts. The model consisted of lumped masses simulating the distribution of mass in the system, connected by massless elastic menbers, simulating the distribution of shaft. stif fness es. The analysis was performed iteratively to obtain compatibility between the rotor displacements and the magnetic and centrifugal forces acting on the rotor. A _1 other analysis of specific motor components and pump components consisted of hand calculations. (~ RHR Heat Exchangers (/ 3.9.1.2.4 Following are the computer programs used in dynamic and static analysis to determine structural and functional integrity of the , RHR heat exchangers: l
- a. Support Load Seismic Analysis ( ED- 6) -
l
~
This computer program computes the total loads at the upper and 1,ower supports of the RHR heat exchanger. I This computer program takes into account the heat exchanger flooded weight, seismic (either OBE or SSE) 1 and the allowable nozzle loads and sets up the worst combination of these loads. By maximizing seismic , loads together with nozzle loads, maximum conservative moments and forces at the upper and ; lower supports are calculated.
- b. Stress Analysis of supports ( ED- 8)
This program performs a full stress analysis of the i upper and lower supports of the RER heat exchanger. l The stresses in the support (both upper and lower) l caused by loads resulting from seismic and nozzle i loads are computed in the Support Load Program (ED-6) () and are used as input values for this program. This program-computes the membrane stresses on the shell i l 3.9-15 1 1
i L L SC S-FSAR AMENDMENT 39 [ OCTOBER 1978 ;
)- of the heat exchanger - by the .use .of the Bijlaard's.
analysis, as well as the net section stresses (shear, tensile, tearing) . on the lower _ support plate and
- upper lugs. It also computes the stresses on the
. welds which hold.the supports to'the shell of the heat exchanger.
3.9.1.2.5 Computer Programs Used in Analyses of BOP 'Pipinq and' Equipment The PIPSYS Jcomputer program has been used by the A-E in the . dynamic analysesH to -determine the structural and. functional integrity of all . Seismic Category I systems and supports. . A-description of the PIPSYS program is provided in Appendix A of the ~La Salle County. Station - DAR (Reference'1). , Seismic Category I components .and equipment have been analyzed by their manufacturers based on loading conditions given in the i equipment design specifications and the requirements of. ASME Section III. For Code Class 1 components and equipment, a' stress report has been prepared and certified by the manuf acturer and reviewed and approved by the A-E. For Code Class 2 and 3 components. and equipment with extended proportions, a seismic qualification report has been prepared by the. manufacturer and reviewed and approved by the A-E.. 0 3.9.1.3 Experimental Stress Analysis . e Experimental stress analysis ' methods have not been utilized. The analytica1' methods employed for Seismic Category 'I systems, components, equipment, and supports are based on those methods specified by' ASME Section III. 3.9.1.4 Considerations for the Evaluation of the Faulted condition ! Only elastic methods as. described in the ASME BSPV Code Section
'III have been used to evaluate stresses for the Seismic Category I components. Allowable stresses and def ormations are based on those specified lin- the code.
3.9.2 Dynamic System Analysis and Testing 3.9.2.1 Preoperational Vibration and Dynamic Ef fects Testing on Piping-
- Vibration amplitudes of the recirculation system-induced by fluid flow- and recirculation. pump operation are instrumented and measured as a part of the preoperational test program. -The measured amplitudes are compared with the allowable vibration / y- amplitudes, calculated by- an analysis of the system. %sC l'
m , 3.9-16
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 ') The interaction between recirculation pump and the flow control valve will be tested up to full loop flow on a generic basis in a reactor mockup test loop at full operating pressure and temperature before these components are installed in the plant. These tests will include pump starts and pump trips, normal valve flow transient, maximum rate valve opening or closing transients caused by simulated gross malfunction of the control system, and a combin~a tion of pump trips and valve closure transients. In plant system demonstration and preoperational tests will be conducted to determine flow interactions with other components of the reactor and reactor coolant prsssure boundary. These tests will be with cold water, and the maximum flow rate will be set by equipment cavitation limits or by equipment load limits. The tests include pump starts , pump trips, and valve transients. The recirculation piping system will be visually inspected and instrumented to detect vibrations. The vibration measurement, visual inspection, and instrumentation stations will be located according to specified location criteria. If the vibration levels are beyond those allowed by the design stress levels, the pipe hangcrs and supports will be adjusted, relocated or redesigned until postadjustment tests show that the vibration levels have been reduced to within acceptable design limits. A piping dynamics testing program will be performed on the ID wJ recirculation system. The following tests will be performed:
- a. system thermal expansion,
- b. vibration during operation of pump at maximum speed and system cold,
- c. vibration during system startup,
- d. vibration during a recirculation pump trip, and
- e. shakedown of system.
Vibration of the main steam system piping is caused by either steam flow or valve operation. The eff ects of these causes are discussed in the following. 3.9.2.1.1 Steam Flow Flow induced vibration of the main steam piping has been shown to be insignificant by a test program conducted in a prototype plant of the same configuration and flow rates. Therefore, testing or analysis for this condition is not
~
conducted for each plant, since its effect has been found by test (S to be insignificant. U 3.9-17 l
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 3'9.2.1.2
. Turbine Stop Valve closure The effects of. turbine stop valve closure.are evaluated analytically by means of a dynamic analysis of the piping system.
The piping is modeled as a 1mnped mass system. Forcing functions - are applied at points of fluid momentum change, such as elbows. 1 The forcing ' functions are f described by . fluid momentum equations, and 'the 's hock wave velocity. The results of this method .of-analysis are compared with results from actual test measurements-in this plant. 3.9.2.1.3 Relief Valve Operation The effects of relief valve operation on the main steam pipe are evaluated analytically by means of dynamic analysis of the main steam valve and. discharge piping. The main steam and discharge system is- modeled as a lumped mass system. . Forcing functions .are ' applied at points of momentum change in.the system. The forcing f unctions are described by fluid momentum equations and the. shock wave velocity. The results of this method of analysis are compared with results obtained from actual test measurements in this plant. , A piping dynamics testing program is to be. performed on the main
~
steam system. ,
- () - The f ollowing tests . are planned: .
- a. system thermal expansion; 3
- b. flow induced vibration at'various flow rates; I
- c. dynamic system response to relief ' valve L operation, turbine stop valve closure, and main steam icolation ,
j valve closure; and
- d. shakedown of system. i a
3.9.2.2 Seismic Qualification ot_ Safety-Related Mechanical Equipm_ ent This subsection describes the criceria (capability of many of ' the components so noted) for qualification of mechanical safety - related equipment and also describes the qualification testing and/or analysis applicable to thic plant for all the major componer;ts on a component-by-component basis. In some cases, a - module or assembly consisting' of mechanical and electrical equipment is qualified ~as a unit, for example, motor powered pumps. Qualification testing is also discussed in Subsection I 3.9.3.2. Electrical supporting equipment such as control consoles, cabinets, and panels which are part of the NSSS are (]) discussed in Section 3.10. , l 3.9-18 i
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 3.9.2.2.1 Tests and Analysis Criteria ~ and Methods
)
The ability of equipment to perform its seismic Category I
. f unction during and after an' earthquake is . demonstrated by tests i and/or L analysis. Selection of testing, . analysis or a combination l of the two is ~ determined by the type, size, shape, and complexisty of the equipment being considered. When practical, i the Seismic category I operations are performed, simultaneously '
.with vibratory testing. Where this is not practical the-operation'and/or loads are simulated by mathematical analysis _ and l applied Lin addition to physical tests.
Equipment which; is ' large, . simple, and/or_ consumes large amounts of power is usually qualified by analysis or static bend test to show that the loads, stresses and deflections are less than the allowable maximum. Analysis and/or _ static bond testing is also used to show there are no natural frequencies below 33 hertz. If a natural frequency lower than 33 hertz is discovered,. dynamic 1 1 tests may be conducted and used in conjunction with mathematical analysis to verify operability and structural integrity at the-required conditions. Natural frequency may be determined by running a continuous sweep f frequency search using a sinusoidal steady-state input of low magnitude. Vibratory conditions are simulated by testing using
<- vibration input or single frequency input (within equipment
(_)s capability) at frequencies through 33 hertz., Whichever method is used, the input motion during testing envelopes the actual input j motion expected during dynamic conditions. The equipment being dynamically tested is mounted on a fixture which simulates the intended service mounting.and causes no dynamic coupling to the equipment. Equipment having an extended structure, such as a valve operator, l is analyzed by applying static equivalent loads at the center of ; gravity of the extended structure. In cases where the equipment j structural complexity makes mathematical analysis impractical, a i static bend test is used to determine _ spring constant and ! operational capability at maximum equivalent dynamic load j conditions. I 3.9.2.2.1.1 Random Vibration Input l When random vibration input is used, the actual input motion envelopes the appropriate floor input motion at the individual modes. However, single frequency input, such as sine beats, can be used provided one of the following conditions are met:
- a. The characteristics of the required input motion are i dominated by one frequency. l I) i i
3.9-19
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 () b. The anticipated response of the equipment is adequately represented by one mode.
- c. The input has sufficient intensity and duration to excite all modes to the required magnitude, such that the testing response spectra will envelope the corresponding response spectra of the individual modes.
3.9.2.2.1.2 Application of Input Motion When dynamic tests are performed, the input motion .iis applied to-one vertical and one horizontal axis simultaneously. -However, if the equipment response along the vertical direction is not sensitive to the vibratory motion along the horizontal direction, and vice versa, then the input motion is applied to one direction at a time. In the case of single frequency input, the time phasing of the inputs in the vertical and horizontal directions are such that a purely rectilinear resultant input is avoided. 3.9.2.2.1.3 Fixture Design The fixture design simulates the actual. service mounting and causes no dynamic coupling to the equipment. 3.9.2.2.1.4 Prototype Testing Equipment testing is conducted on prototypes of the equipment installed in this plant. 3.9.2.2.2 Seismic Qualification of _ Specific NSSS Mechanical Components The following sections discuss the testing or analytical quali-fication of NSSS equipment. 3.9.2.2.2.1 Jet Pumps A static analysis of the jet pumps was performed assuming 3.0g horizontal acceleration and 1.59 vertical. The stresses , resulting f rom the analysis were below the design allowables. Static analysis with an appropriate amplification factor was used in lieu of dynamic analysis since the jet pump is a simple component with a natural frequency of slightly less than 33 hertz. 3.9.2.2.2.2 CRD and 'CRD Housing The dynamic analysis of the fuel, core support, top guide and control rod drive housing (with contained control rod drive) ,, indicates these components behave essentially in an elastic (,j - manner during the combined loadings. 3.9-20
LSCS-FS AR AMENDMENT 39 l OCTOBER 1978 /~T The housing provides the basic structural member for the drive, \~# so the dynamic load effects on the CRD are evaluated from a drive housing deflection standpoint. Restraints are provided to prevent flange motion, so the housing deflection is limited to a i small midpoint bow. This bow is smaller than the clearance i between drive and housing, and thus will not affect drive' motion. Tests have been conducted on the drive with dynamic deflections of 2 inches peak to peak at the flange (at the natural frequency of drive and housing) . There was no measurable effect. 1 Additional testing of the CRD has been conducted with static ' displacement of the core support and top guide equal to the maximum calculated dynamic deflection. The eff ect on scram performance was negligible. Cnannel bow tests, with varying amounts of fixed channel deflection, indicate very little eff ect on scram with a bow greater than the calculated maximum dynamic deflection under combined loads. Drive performance under dynamic deflections should give even greater margins than the above tests conducted with static deflections. 3.9.2.2.2.3 Core Support, Fuel Support, and CRD Guide Tube () No dynamic testing of the CRD guide tube has been conducted; however, a detailed analysis imposing d i namic effects due to dynamic events has shown that the maximuu stresses developed during these events are much lower than the maximum allowed for the component material. 3.9.2.2.2.4 Hydraulic Control Unit (HCU) The HCR was analyzed for the faulted condition. The maximum stress on the HCU frame was calculated to be below the maximum allowable for the SSE faulted condition. 3.9.2.2.2.5 Fuel Channels GE BWR fuel channel design bases, analytical methods and evaluation results including seismic considerations, are contained in References 13 and 14. 3.9.2.2.2.6 Recirculation Pump and Motor Assembly calculations were made to assure that the recirculation pump and motor assembly are designed to withstand the specific static equivalent forces. The flooded assembly was analyzed as a free body supported by constant support hangers from the brackets on the motor mounting member with hydraulic snubbers attached to (]) brackets located on the pump case and the top of the motor frame. 3.9-21
~
LSCS-FSAR ' AMENDMENT 39'
- OCTOBER 1978
'(}} Primary stresses due to horizontal and vertical forces were: considered to act simultaneously and .are conservatively added , directly. The horizontal and vertical seismic forces were c applied at mass centers and equilibrium reactions determined for - motor and pump-brackets. 3.9.2.2.2.7 ECCS Pump and Motor Assembly This section discusses the ECCS pump and motor ~ assemblies. The t qualification :of these ' pump and motor assemblies as a unit' while operating under dynamic conditions'was provided in the form of a s ta tic ' analysis. - The' maximum specified vertical and horizontal accelerations were . constantly applied simultaneous 1yJin the worst-case combination. The results of the analysis indicate the pump is capable of sustaining the above' loadings without overstressing the pump components. A similar design motor has been qualified via a combination of static analysis and dynamic testing. The complete motor assembly has been qualified via dynamic testing in accordance with IEEE 344-1975. The qualification test program included demonstration of startup and shutdown capabilities as well as no-load operability during. dynamic loading conditions. For static analysis on a similar-design motor, the dynamic forces fg of_each component or assembly _are obtained by concentrating its
\_/ mass at the center of gravity of the component or assembly, and multiplying by the seismic acceleration, (earthquake coefficient)'. The magnitude of the acceleration _ coefficients are 0.14g vertical 'and 1. 5g horizontal .
3.9.2.2.2.8 - RCIC Pump Assembly The RCIC 1 pump construction is a barre 1~ type on a large cross-sec tion - pedes tal. Qualification ~by analysis was performed. The seismic design- analysis is based on 3.0g horizontal and 0.5g vertical accelerations. Results are obtained by using-acceleration forces acting simultaneously in two directions, one vertical and one horizontal. The pump mass, support system and accessory piping have been shown by analysis to have a natural f requency greater _ than 33 hertz. The RCIC pump assembly has been analytically qualified by static , analysis for vibratory loading as well as the design operating s loads for pressure, temperature, and external piping. The results of this analysis confirm that the stresses are substantially less than 90% of the allowable. . 3.9.2.2.2.9 RcIC Turbine Assembly LThe RCIC turbine has been qualified via a combination of static - (]) ' : analysis . and dynamic testing. The turbine assembly consists of I rigid. masses, - wherein static . analysis has been utilized, 3.9-22 )
*c- % ~- e+ e 4ed Pr ^**Pa+1&*-T +c rv- ni?im- 4 -h4 e
.t LSCS-FSAR AMENDMENT,39 -
OCTOBER.1978 r') , interconnected with. control levers. and electronic control
' 'V systems, necessitating . final . qualification via dynamic testing.
Static loading' analysis has been-employed ;to verify the structural integrity of the turbine assembly, and the adequacy of bolting under operating and seismic loading conditions. The complete. turbine assembly has been seismically qualified via dynamic testing in. accordance with IEEE 344-1975. .The qualification test program included demonstration of startup and shutdown capabilities, as well .as' no load operability during. faulted loading conditions. operability under normal load-conditions can be assured by . comparison of the operability of similar turbines in other operating plants. Requirements The specification for qualification of the RCIC turbine and its accessories states that they shall be capable of withstanding the specified accelerations at all frequencies within the range of ' O.25 hertz to 33 hertz. Proper: performance- may be demonstrated -
'by the seller by tests,' analysis, or a combination of both. If all natural. frequencies of the turbine, the- component parts, and the accessories are greater 'than 33 hertz (as defined by test and/or analysis) , a static load analysis may be performed. The seismic f orces. of each component or assembly are obtained by concentrating t its mass at the center of mass of the component or 7, assembly, and' multiplying by the acceleration coefficient. The
( ,j magnitude of the acceleration coefficients are 1.5g y both horizontal and vertical. If component parts and/or accessories have natural frequencies below 33 hertz, these parts must be dynamically analyzed or tested, demonstrating satisfaction of the defined floor. response spectra. If the equipment capability is demonstrated by. test, the equipment must be subjected to simultaneous horizontal and vertical acceleration inpttts of random Lwave-form motion for a minimum duration of 30 seconds. < The random input must envelop the defined floor response spectra. , Test Qualification Results_ The RCIC turbine assembly was subjected to a total of 33 vibratory tests with an accumulated test time of 905 seconds (N15 minutes) .. Input to the equipment was random wave-form motion in two directions, one horizontal and the other vertical, with sine beats superimposed at 1/3-octave intervals as necessary to , envelop the required response spectra. The required response spectra enveloped all postulated dynamic loads including those i f rom seismic and hydrodynamic transients. A 200-psi, 1200-cfm ' air source was used as the operating medium for the turbine. The electronic governor system, the turbine hydraulic system with interconnecting piping and levers, and all turbine instrumentation were under actual operating conditions during the test program. Nozzle loadings were simulated for the turbine inlet-and exhaust piping. ].
, En 3.9-23
?
~ _, ., , . - _ _.. .. . ~ , m. m . . . . , . . . . . . . . . . _ _ _ . - _ . - . ~ . . . . _ - . . . - -
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 r~g The test results proved the operability--startup, no-load (J e eady-state, and shutdown capabilities--of the turbine assembly aad its components during the required loading conditions. Stress levels were not measured during the test program, since the turbine was not subjected to full pressure and temperature conditions. This qualification has been accomplished analytically, as indicated below. Analvtical Qualification Results The rigid components of the RCIC turbine assembly have been analytically qualified via static analysis for vibratory loading, as well a s the design operating loads of pressure, temp (ratare, and external piping loads. The results of this analysis ccnfirm that the stresses in all components are substantially below allowable levels, i.e., less than 85% allowable, with the single exception of the turbine governor valve. 3.9.2.2.2.10 Standby Liould control Pump and Motor Assembly The SLC positive displacement pump and motor mounted on a common base plate has been qualified by static analysis. The design analysis is based on 3.0g horizontal and 0.5g vertical acceleration. Recults are obtained by using acceleration forces 7s acting simultaneously in two directions, one vertical and one (_) horizontal. The pump / motor /tase assembly has been shown by ; static analysis to have a natural frequency greater than 33 hertz. The SLC pump and motor aseembly has been analytically qualified by atatic analysis for dynamic loading as well as the design operating loads of pressure, temperature, and external j piping loads. The results of this analysis confirm that the ; stresses are substantially less than 90% of allowable. 3.9.2.2.2.11 RHR Heat Exchan'gers A dynamic analysis is performed to verify that the RHR heat exchanger will withstand dynamic loads. Testing is an impractical method to verify the equipment adequacy when predictable seismic loads can' be determined by dynamic and static analysis. 1 The heat exchanger, including its appurtenances and supports, is l designed to withstand the static seismic accelerations specified below. The acceleration coefficients are applied at the center of gravity of the heat- exchanger, assuming the heat exchanger to be flooiei. Faulteo r andition acceleration coefficients for the RHR heat exchanger are 1.50g horizcntal and 0.50g vertical. O 3.9-24
l LSCS-FSAR AMENDMENT 39 OCTOBER 1978 I lll 3.9.2.2.2.12 Standby Liquid Control Tank The standby liquid control storage tank is a cylindrical tank 9 feet in diameter and 12 f eet high bolted to the concrete floor. Stresses can be calculated readily by conventional methods. The magnitude of the earthquake coefficients for safe shutdown earthquake (SSE) are 3.0g horizontal and 0.299 vertical. The standby ' liquid control tank has been qualified by analysis for:
- a. stresses in the tank bearing plate,
- b. belt stresses,
- c. sloshing loads imposed by earthquake natural frequency of sloshing = 0.58 hertz,
- d. minimum wall thickness, and
- e. buckling 3.9.2.2.2.13 Main Steam Isolation valves The main steam isolation valve structures have been analyzed and representative models dynamically tested to determine operability at the specified SSE accelerations. Operation of the valve was ggg demonstrated during this test. Dynamic tests completed on this configura tion used a slow sine sweep and sine beat time history f or the 1940 El Centro Earthquake to excite actuator assembly up to 50 hertz. Allowable accelerations measured during the tests were in excess of those calculated from the piping system l analysis. The fundamental requirement of the MSIV following a i saf e shutdown earthquake is to close and remain closed af ter the event which was demonstrated by the dynamic tests.
3.9.2.2.2.14 Main Steam Safety / Relief Valvec Due to the complexity of thic structure and the performance requirements of the valve, the total assembly of the safety / relief valve (including electrical, pneumatic devices) was dynamically tested. Safisfactory operation of the valves were I demonstrated during and after the test. Tests and analysis ! satisfy operability criteria (as defined in the FSAR) . ! 3.9.2.3 Dynamic Response of Regetor Internals Under O,oerational Flow Transients and Steady-State Conditions The major reactor internal components within the vessel are subjected to extensive testing coupled with dynamic system ' analyses to properly describe the resulting flow-induced vibration phenomena incurred from normal reactor operation and f rom anticipated operational transients.
- 3. 9+2 5 :
LSCS-ISAR' . AMENDMENT 39. OCTOBER 1978 TB- In general,.the vibration forcing functions for operational' flow
\"#
transients and steady-state conditions were'not predetermined by detailed analysis. . Special analyses of the response signals, measured ~~ from reactor-internals of similar ' designs are . performed to predict amplitude and' modal contributions. Parametric studies were perf ormed by extrapolating the results from tests of : internals and components of similar designs.' This vibration. predicti'on method is appropriate where standard hydrodynamic theory"cannot be applied due to complexity of the structure and flow conditions. Elements of the vibration prediction method are outlined as.follows:'
- a. Dynamic analysis 'of major components and-subassemblies is' performed to identif y ' natural vibration modes and frequencies. ' The analysis models used for Seismic Category I structures are similar to those outlined in Subsection 3.7. 2. .
- b. Data from: previous plant vibration measurements is assembled and examined to identify. predominant vinration response modes of major components. In
. general, response: modes are similar but response amplitudes vary anong BWR's. of differing size and design.
- c. Parameters are identified which are expected to p)g s influence vibration' response amplitudes among the several referenced plants. These include hydraulic parameters such as velocity and steam flow rates, and structural parameters such as natural frequency' and significant dimensions.
- d. . Correlation f unctions for the various parameters are developed which, multiplied by response amplitudes, tend to minimize the statistical variability between plants. A correlation function is obtained for each major component and response mode. t
- e. Predicted vibration amplitudes for components of the prototype plant' are' obtained from these ' correlation functions, based on particular values' of the '
parameters for that prototype plant. The predicted amplitude for each dominant response mode is stated' , in terms ofr a range, taking into account the degree of st'atistical variability in~ each' of . the correlations. _ The predicted mode and frequency are' obtained from the dynamic analysis of Item a above'. The dynamic model analysis also forms the basis for interpretation of the prototype plant preoperational' and initial startup. test results. Modal stressesLare calculated and j[) relationships - are obtained between sensor response' amplitudes and . peak component stresses for each of the lower normal modes. The i.
'3.9-26 7
.ee.,w+=- ,., ,-m.- . ,- -r-y n d*- -, --m r:-
LSCS-FSAR AMENDMENT 39 - , OCTOBER 1978 O (/ allowable amplitude in each mode is that which produces a peak stress amplitude of i 10,000 psi. 3.9.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals LSCS Unit 2 reactor internals will be tested in accordance with provisions of Regulatory Guide 1.20, Revision 2, for NonPrototype plants. The test procedure will require operation of the recirculation system at or' near rated ~ flow with internals installed (less fuel) , followed by inspection for evidence of vibration, wear, or loose parts. The test duration will be sufficient to subject critical components at least 106 cycles of vibration during two-loop and single-loop operation of the recirculation system. At the completion _of the flow test, the vessel head and shroud head will be removed, the vessel will be drained and major components will be inspected on a selected basis. The inspection will cover all components which were examined on the prototype design, including the shroud, shroud head, and core support structures, the jet pumps, and the peripheral control rod drive and incore guide tubes. Access will be provided to the reactor lower plenum. Reactor internals for LSCS Units 1 and 2 are substantially the same as the internals design configurations which have been () tested in prototype BWR/4 plants. Exceptions are the jet pumps, which are the BWR/5 design. This vibration measurement and inspection program will be compared with the results of the prototype plant, to verify the design of the jet pumps with respect to vibration. Results will be made available for NRC review after completion of the tests. LSCS Unit 1 utilizes an adapter joining the jet pump to the shroud support plant which is of a unique design. Analysis indicates that this adapter might introduce added flexibility to . reduce 'the natural frequencies of LSCS Unit 1 jet pumps below those of the prototype jet pumps. Therefore, vibration instrumentation will be provided in the LSCS Unit 1 jet pumps below those of the prototype jet pumps. Therefore, vibration instrumentation will be provided in the LSCS Unit 1 reactor to evaluate this deviation from the prototype design configuration. Vibration sensors will be installed on the jet pump riser braces, as in the prototype plant, and on the jet pump adapter. One jet i pump pair will be instrumented. Data will be acquired during the preoperational flow test described above, and also at specific flow and power conditions during the startup tests. ' LSCS Unit 1 is designated as a non-prototype plant with reference to i requirements of Regulatory Guide 1.20, only for this particular item. Results of the prototype tests are presented in GE Licensing ('}E L Topical Report, NEDE- 240 57-P (Class III) and NEDO-24057 (Class I) , " Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 3.9-27
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 () Plants," November 1977. This report. also contains additional inf ormation on the jet pump vibration measurement and inspection programs to be performed in the Tokai-2 plant, and on the confirmatory inspection program. 3.9.2.4.1 Preoperational Tests Reactor : 1nternals for LSCS Units 1 and 2 are substantially the same as those of the prototype BWR/4-5 Category I plant except as discussed'below for the LSCS Unit 1 RpV:
- a. Jet pumps for LSCS are the BWR/5 design, which differs from that of the prototype plant. The BWR/5 jet pumps for the 251 size plants are subjected to extensive vibration measurements . and inspection at a foreign plant (Tokai-2) . Results of that program are to be made available for NRC review on a domestic 251-BWR/5 docket.
- b. LSCS Unit 1 utilizes a gusset-type shroud support which differs from-the stilt-type support utilized in earlier 251 size. plants and in the LSCS Unit 2 but which has no significant effect' on the vibratory response of the reactor internals. Vibration measurements have been made for the gusset-type r')
\~/ support in 218 size plants with satisfactory results (Cooper and Hatch-1) and analysis has shown no significant differences between the two support designs in terms of vibration modes and frequencies. LSCS Units 1 and 2 reactor internals are subjected to a flow test and inspection for the purpose of detecting loose parts or damaging vibration amplitudes which could result from assembly errors or undesirable deviations from the prototype plant 1 construction. Tne test conditions and~ duration of this flow test are the same as for the prototype plant (Browns Ferry-1) I preoperational test, which has been successfully completed. At l the completion of the LSCS flow test, the vessels will be drained and all major components, including lower plenum components, will be visually inspected on a selected basis to confirm that there are no loose, damaged, or f ailed parts. 1 3.9.2.5 Dynamic System Analysis of Reactor Internals Under Faulted Conditions l
'3.9.2.5.1 Safety Evaluation The preoperational. and startup testing series are utilized to l authenticate sequentially the adequacy of reactor components and subsystems to respond properly in abnormal and faulted conditions. . Consult Chapter 14.0 for test abstracts and the J
() general pattern of startup tests which demonstrate the integrity of these reactor systems and the reactor internals. i 3.9-28 j l i m.. , ,. . . - -
l L5CS-FSAR AMENDMENT 39 OCTOBER 1978 () 3.9.2.5.2 Evaluation - Methods l To determine that the saf ety design. bases are safisfied, j responses of the reactor vessel internals to loads imposed during l normal, upset, emergency, and faulted conditions are examined. I The effects.on the ability to insert control rods, cool the core, and flood the inner volume of the reactor vessel are determined.- l 3.9.2.5.2.1 Input for Safety Evaluation l The operating' conditions that provide the basis for the design of the reactor internals to sustain normal, upset, emergency, and faulted conditions, as well as combinations of design loadings-that are accounted for in design of the core support structure, are covered. in Tables 3.9-30, 3. 9- 31, and 3. 9-3 2. In addition each combination of operating loads is categorized with respect to either normal, upset, emergency, or faulted conditions as well as the associated design stress intensity or def ormation limits. The bases for the proposed design stress and deformation criteria are also specified in Chapter 3.0. 3.9.2.5.2.2 Events To Be Evaluated Examination of the spectrum of conditions for which the safety design basis must be satisfied reveals three dominating f aulted events:
- a. Recirculation Line Break A break in a recirculation line between the reactor ,
vessel and the recirculation pump suction.
- b. Steamline Break Accident A break in one main steamline between the reactor vessel and the flow restrictor. The accident results in significant pressure differentials across some of the structures within the reactor.
- c. Earthquake An.SSE subjects the core support structures and reactor internals to significant forces as a result of' ground motion.
Analysis of other conditions existing during normal operation, , abnnrmal operational transients, and accidents shows that the-(~g loads affecting the core support structures and reactor internals A/ are less severe than these three postulated events. 3.9-29
LSCS-ESAR AMENDMENT 39 OCTOBER 1978
){) 3.9.2.5.2.3 Pressure Differential During Rapid Depressurization A digital computer code (Reference 4) is used to analyze the transient conditions within the reactor vessel following the recirculation accident. The line break accident and the steamline break analytical model of the vessel consists of nine nodes which are connected to the necessary adjoining nodes by fic < paths having the required resistance and inertial cha;acteristics. The program solves the energy and mass conservation equations for each node to give the depressurization rates and pressure in the various regions of the reactor. ' Figure 3.9-7 shows the nine reactor nodes. ,
3.9.2.5.3 Recirculation Line and Steamline Break 3.9.2.5.3.1 Accident Definition Both a recirculation line break (the largest liquid break) and an inside steamline break (the largest steam break) are considered in determining the design-basis accident for the reactor internals. The recirculation line break is the same as the design-basis loss-of-coolant accident described in Section 6. 3. A sudden, complete circumferential break is assumed to occur in one redirculation loop. The pressure differentials on the reactor internals and core support structures are in all cases ((} lower than for 'the main steamline break. The analysis of the steamline break. assumes a sudden, complete circumferential break of one main steamline between the reactor vessel and the main steamline restrictor. This is not the same accident described in Chapter 15.0 which has greater potential radiological effects. A steamline break upstream of the flow restrictors produces a larger _ blowdown area and thus a f aster depressurization rate than a break downstream of the restrictors. The larger blowdown area results in greater pressure diff erentials across the reactor assembly internal structures. The steamline break accident produces significantly higher pressure differentials across the reactor assembly internal structures than does the recirculation line break. This results from the higher- reactor depressurization' rate associated with the steamline break. Therefore, the steamline break is the design- 3 l basis accident. for internal pressure differentials. 3.9.2.5.3.2 Effects of Initial Reactor Power and Core Flow 1 f For purposes of illustration, the maximum internal pressure loads can be considered to be composed of two parts: steady-state and transient pressure differentials. For a given plant the core flow and power are the two major factors which influence the reactor internal pressure differentials. The core flow ! f-)x
\_ essentially affects only the steady-state part. For a fixed power, the greater the core flow, the larger will be the steady-3.9-30
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 sta te pressure differentials. The core power affects both the (]) steady-state and the transient parts. As the power is decreased there is less voiding in the core and consequently the steady-state core pressure differential is less. H owever, less voiding in the core also means that less steam is generated in the reactc.e pressure vessel and thus the depressurization rate and the transient part of the maximum pressure load are increased. As a result, the total loads on some components are higher at low powe r. To ensure that the calculated pressure differences bound those which could be expecteo if a steamline break should occur, an analysis is conducted at a low power-high recirculation flow condition in addition to the standard saf ety analysis condition (105% steam flow, rated recirculation flow) . The power chosen for analysis is the minimum value permitted by the recirculation system controls at rated or greater recirculation flow. This condition maximizes those loads which are inversely proportional to power. It must be noted that this condition, while possible, is unlikely; first, becau.3e the reactor will generally operate at or near f ull power; second, because high core flow is neither required nor desirable at such a reduced power condition. Table 3.9-1 summarizes the maximum pressure differentials. Condition 1 is the safety analysis condition; Condition 2 is the
~g low power-high flow condition. Comparison of these values (J
s illustrates the statements made in the foregoing paragraphs. 3.9.2.5.3.3 conclusions It is concluded that the maximum pressure loads acting on the reactor internal components result from an inside steamline break, and on some components the loads are greatest with operation at the minimum power associated with the maximum core flow (Table 3. 9-1 Condition 2) . This has been substantiated by the analytical comparison of liquid versus steam breaks and by the investigation of the ef fects of core power and core flow. It has also been pointed out that, although possible, it is not probable that the reactor would be operating at the rather abnormal condition of minimum power and maximum core flow. More realistically, the reactor would be at or near a f ull power condition and thus the maximum pressure loads acting on the internal components would be as listed under Condition 1 in Table 3.9-1. 3.9.2.6 Correlation of Reactor Internals Vibration Tests with the Analytical Results Prior to initiation of the instrumented vibration test program for the prototype plant, extensive dynamic analyses of the () reactor and internals were performed. The results of these analyses were used to generate the allowable vibration levels 3.9-31
LSCS-FSAR AMENDMENT'39 OCTOBER 1978 l(3-
- v--
during' the vibration test. The vibration 1 data obtained during
.the. test were analyzed in detail. The results of the data analysis, vibration ~ amplitudes, natural frequencies, and mode shapes were then compared to. those obtained .f rom the theoretical
~
a nalysis. . Such~ comparisons provided insight into the dynamic behavior of the reactor internals.- ~ The ' additional knowledge gained was.
. utilized in the generation of the dynamic models for seismic and LOCA analyses for LSCS., The models.used for this plant.are the same .as those used' for the vibration analysis of the . prototype plant.
The flow-vibration test data are supplemented by data from forced oscillation tests of reactor. internal components to provide?the analysts . with additional -information. concerning the dynamic ' behavior of the reactor internals. 3.9.2.6.1 Analysis Methods Under LOCA Loadings In order to ensure that no significant dynamic amplification of , load occurs as a result of the oscillatory nature of the blowdown
-f orces, a comparison is made of the periods of the applied forces and the natural periods of the core ' support structures being acted upon by the applied forces. These periods are determined ,
- from a comprehensive dynamic. model of the RPV and internals with -
O, 27 degrees-of-freedom (Figure 3.9-1) . Only motion in the vertical direction is considered here; each structural member (between two mass points) can only have a'n, axial load. Besides the real masses of the reactor pressure vessel (RPV) and core support structures, account is made for the water incide the RPV. Time varying pressure is applied to the dynamic model of the reactor internal described previously. Except for the nature and locations - of the forcing functions and the dynamic model, the dynamic analysis method is- identical to that described _for the analysis- of the seismic and dynamic excitation for suppression pool and annulus pressurization events. 3.9.3. ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3.9.3.1 Case 1' 1 In .the stress analysis .for Seismic Category I equipment- which is '
. part' of the primary. coolant pressure boundary, the reactor vessel isr analyzed according to the requiremerts of the 1968 edition of j ASME- Section 'III' Article 4. -The nain steam piping system is l analyzed 1to comply with ASME Section-III Codes (NB-3600) . The
-(): recirculation piping ' system is analyzed to comply with ANSI J B31.7.- ' Iso the stress' analysis of other Seismic Category I. l l 3.9-32 , ,. - .- - u,, -. - . - . , . - - , . - . _ - . . , , .-
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 f') equipment such as RER heat exchangers and pumps, elastic analysis was used. The maximum allowable stresses are less than the yield stresses of the materials used and these are given in Tables l
- 3. 9-2 through 3. 9-2 2. '
3 . 9. 3 .1.1 Loading combinations and Stress Limits ASME Cod'e Class 2. and 3 components of fluid systems are constructed in accordance with Section III of the ASME Boiler and Pressure Vessel Code. Some components (piping, pumps, and valve s) ordered prior to July 1971 were designed to other industry codes (Table 3.2-2) when the effective section III was not applicable. Tables 3. 9-2 through 3.9-22 list the design loading combinations f or the major components of each safety-related system. 3.9.3.1.1.1 Design Loading combinations The combination of design loadings is categorized with respect to plant conditions identified as normal, upset, emergency, or f aulted as shown in Tables 3.9-2 through 3.9-22 for the major components. This subsection delineates the criteria for selection and r' definition of design limits and loading combinations associated I with normal operation, postulated accidents, and specified seismic events for tne design of safety-related ASME code components, except containment components, which are discussed in Poction 3.8. This section also lists the major AShE Class 1, 2, and 3 pressure parts and associated equipment on a component-by-component basis and 4.dentifies the applicable loadings, calculation methods, calculated stresses, and allowable stresses. Design transients f or ASME Class 2 equipment are covered in Subsection 3. 9.1.1. Seismic related loads are discussed in Subsection 3.9.2.2 and Section 3.7. Tables 3.9-2 through 3.9-22 present the loading combinations, analytical methods (by reference or example) , and the calculated stress or other design values for the most critical areas in the design of each component. These values are also compared to applicable code allowables. 3.9.3.1.1.2 Plant Conditions All events that the plant might credibly experience during a reactor year are evaluated to establish a design basis for plant equipment. These events are divided into four plant conditions. The plant conditions described in the following paragraphs are () based on event probability (i.e. , frequency of occurrence) and 3.9-33
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 , i I) correlated design conditions defined . in the ASME Boiler and Pressure Vessel Code, Section III. 3.9.3.1.1.2.'1 ' Normal Condition Normal conditions are any conditions in the course of system' i startup, operation in the design power range, normal hot standby ; (with co' ndenser available) , and system shutdown other - than Upset, Emergency, Faulted, or Testing. 3.9.3.1.1.2.2- Upset Condition Any- deviations from normal conditions - anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. .The upset conditions include those - transients which result trom and single operator error or control malfunction, transients caused by a f ault in a system component requiring its ' isolation from the system, and ,
-transients due to. loss ' of load or power, vibratory motions due to an operating-basis earthquake are conservatively treated as upset. Hot standby with the main condenser isolated it. an upset condition.
3.9.3.1.1.2.3 Emergency Condition (~' Those deviations from normal. conditions which require shutdown
- for correction of the conditions. or repair of damage in the RCPB.
The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system. Emergency condition events include, but are not limited to, transients caused by one of the f ollowing: a multiple valve blowdown of the reactor vessel; loss of reactor coolant from a small break or crack which does not depressurize the reactor system nor result in leakage beyond normal makeup system capacity, but which requires the safety functions of isolation of containment and reactor shutdown; improper assembly of the core during refueling, and vibratory motions of an OBE in combination with associated system- transients. 3.9.3.1.1.2.4 Faulted Condition-Those combinations of conditions associated with extremely low r probability,' postulated events whose consequences are such that the ' integrity and operability 'of the system may be impaired to .
.the extent that considerations of public health and safety are Linvolved. Faulted ' conditions encompass events that are postulated because their consequences would include the potential for~the release of significant. amounts of radioactive material.
These postulated events are 'the most drastic that must be i; designed against and thus represent' limiting design bases. . s - . Faulted condition events ' include, but are not limited to, one of r the following: a control rod drop accident, a f uel-handling 3.9 ,
-. , , w ,
e- , , , , . . - c.,,-..---r, , , . , ,+,.,~,,-n . . - , . , , - . , , . . - - - . . - w ... i.- - - , - .
- - . . - - ~ . - .
LSCS-FSAR AMENDMENI 39 OCTOBER 1978
- accident, a main steamline break, a recirculation loop braak, . the
(]) combination of any pipe break plus the seismic motion associated with .a safe shutdown earthquake plus a loss of off site power, or the safe shutdown earthquake. 3.9.3.1.1.3 Correlation of Plant Conditions with Event Probability The probability of an event occurring per reactor year associated with the plant conditions is listed below. This correlation can ' be used to identify the appropriate plant condition for any hypothesized event or sequence of events. EVENT
' ENCOUNTERED PROBABILITY PLANT PER REACTOR CONDITIONS YEAR Normal (planned) 1.0 Upset (moderate probability) 1.0 > P > 10-2 Emergency (low probability) 10-2 >p> 10-4
() Faulted (extremely low probability) 10-4 >P> 10-6 3.9.3.1.1.4 Safety Class Functional Criteria For any normal or. upset design condition event, Safety Class 1, 2, and 3 equipment shall be capable of accomplishing its safety ; f unctions as required by the event and shall incur no permanent' changes that could deteriorate its ability to accomplish its safety functions as required by any subsequent design condition event. For any emergency or faulted design condition event, Safety Class ; 1, 2, and 3 equipment shall be capable of accomplishing its safety functions as required by the event, but repairs could be required to ensure its ability to accomplish its saf ety functions 1 as required by any subsequent design condition event. ! 3.9.3.1.2 Design Stress Limits l 3.9.3.1.2.1 Stress Level for seismic Category _I Components 1 Stress analyses were performed for the design basis to determine l structural adequacy of pressure components under the operating l conditions of normal, upset, emergency, or faulted, as I applicable. The stress analyses were performed as appropriate () during the design assessment evaluation. 3.9-35 I
- -- - _ _ . - _ - - - -- .~
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 1 () Significant discontinuities such as nozzles, flanges, etc. were considered. In addition to the design calculations required by the AS4E III code, stress analysis was perf ormed by methods outlined in the code appendices or by other methods by reference to analogous codes or other published literature. Tables 3.9-2 through 3.9-22 also give calculated stress levels, or maxim'um allowable loadings at significant areas of consideration for the major components.. 3.9.3.1.2.2 Field Run Piping Non-Seismic Category I pipe systems (Class C or D) , 2-inch nominal pipe size and less, are field run and are identified in Table 3.2-1. Schematic routing and criteria were provided to the constructor to ensure proper design interface. All non-Seismic Category I piping is anchored where it' interfaces with Seismic Category I piping or interfaces approp.lately controlled by guides. All Seismic Category I pipe systems (Class A, B, and C) up to and including the isolation valves and piping hangers except the main steam, reactor recirculation system, and insert, withdraw, and scram discharge lines, are designed by Sargent S Lundy and identified in Table 3.2-1. Schematic routing and criteria were provided to the constructor ]- (' to interf ace with wall and floor pentrations, shield walls, and equipment access, consistent with the overall plant design. 3.9.3.1.2.3 Stress Levels for ASME Code class 2 and 3 For safety-related ASME Code Class 2 and 3 components, the design stress limits, allowable loads or required dimensions are listed in Tables 3.9-2 through 3.9-22. Inelastic methods as permitted by ASME Section III for Class 1 components were not used for these components. 3.9.3.2 Pump and Valve Operability Assurance ! Active mechanical equipment classified as Seismic category I ar e designed to perform their 5 unction during the life of the plant under postulated plant conditions. Equipment with f aulted condition functional requirements include " active" (active equipment must perform a mechanical motion during the course of accomplishing a saf ety function) pumps and valves in fluid systems such as the residual heat removal system, and core spray systems. Operability is assured by satisfying the requirements of the following programs. Safety-related valves are qualified by (g prototype testing and analysis, and safety-related pumps by s/ analysis with suitable stress limits and nozzle loads. The content of these programs is detailed in the following.
- 3. 9-36
.__ _. __. ~ . _ - _
l LSCS-FSAR AMENDMENT 39 i OCTOBER 1978 l (~)h t ! 3.9.3.2.1 ECCS Pumps i All active pumps as listed in Table 3.9-23 are qualified for operability by first being subjected to rigid tests both prior to installation in the plant and after installation in the plant. The in-shop tests include (1) hydrostatic tests of pressure-retaining parts to 125% of the design pressure times the ratio of material' allowable stress at room temperature to_ the allowable f stress value at the design temperature (2) seal leakage tests; and- (3). performance tests, while the pump is operated with flow, to determine total developed head, minimum and maximum head, net positive ' suction head '(NPSH) requirements, and other pump / motor parameters. Also monitored during these operating _ tests are bearing temperatures ' (except water-cooled l { bearings) and vibration levels. Both will be shown to be below { specified limits. After the pump is installed in the plant, it l undergoes the cold hydro tests, functional tests, and the i required periodic inservice inspection and operation. These ' tests demonstrate reliability of the pump for the design life of ! the plant. .i In addition to these tests, the safety-related active pumps have t been analyzed for operability during a seismic condition by ensuring that (1) the pump will not be damaged during the seismic j (]) event, and (2) the pump will continue to operate af ter the event. 3.9.3.2.1.1 Analysis of Loading, Stress, and Acceleration Conditions In order to avoid damage during the faulted plant condition, the stresses caused by the combination of normal operating loads and dynamic system loads are limited to the material elastic limit. The average membrane stress (o m) . for the faulted condition load l is maintained at 1. 2S, or approximately 0. 75 oy (az yield stre ss) . The maximum stress in local fibers (om + bending stress (ob)) is limited to 1.8S, or apprcximately 1.1 o . y The qualification of the pump. and motor as an integral' unit while operating under dynamic conditions is provided in the form of a static-earthquake-acceleration analysis. Under this criteria, the unit is-considered to be supported as designed, and maximum specified vertical and horizontal accelerations are constantly applied simultaneously in the worst-case combination. The maximum allowable nozzle loads from the attached piping. system were considered in an analysis of the pump support to assure that there would be no geometrical / dimensional deformation on the pump components. 3.9.3.2.1.2 Pump Operation During and Following Vibratorv Loading /"T. S (_) Active . pump / motor rotor combinations are designed to rotate at a constant speed under all conditions. Motors are designed to . withstand short periods of severe overload. The high rotary 3.9-37 I
i 1 l LSCS-ES AR -AMENDMENT 39 OCTOBER 1978 rw l ks inertiat in the operating pump rotor and the nature of the random, short-duration loading characteristics of dynamic events- will prevent the rotor from becoming seized. In actuality, the loading will cause only a slight increase, if any, in the torque (i.e. , motor current) ' necessary to drive the pump at the constant design speed. Therefore the pump will not shut down during a dynamic event and will operate at the design speed af ter the event. 1 The functional ability of the active pumps af ter a faulted condition is assured, since only normal operating lo. ads and steady-state nozzle loads exist. For the active pumps, the-f aulted condition is greater than the normal condition only due , to dynamic loads on the equipment itself. j Faulted events are infrequent and of relatively short duration compared to the design life of the equipment. Since it is demonstrated that the pumps would not be damaged during the f aulted condition, the post-f aulted condition operating loads.. will be no worse than the normal plant. operating limits. This is assured by requiring that the imposed nozzle loads (steady-state loads) for normal conditions and post-faulted conditions are limited ~ by the magnitudes of the normal condition nozzle loads.- The postf aulted condition ability of the pumps to f unction under these applied loads is proven during the normal operating plant Os conditions for active pumps. 3.9.3.2.2 SLC Pump and Motor Assembly and RCIC Pump Assembly These equipment assemblies are small, compact, rigid assemblies with natural frequencies well above 33 hertz. With this fact verified, each equipment assembly has been qualified via static j analysis only. This static qualification verifies operability j under dynamic conditions, and assures structural loading stresses ; within Code limitations. ! 3.9.3.2.3 RCIC Turbine Assembly The RCIC turbine has been seismically qualified via a combination of static analysis and dynamic testing. The turbine assembly consists of rigid masses, wherein static analysis has been utilized, interconnected with control levers and electronic control systems, necessitating final qualification via dynamic t esting. Static loading analysis has been employed to verify the structural integrity of the turbine assembly. and the ddequacy of l bolting under operating and faulted loading conditions. The complete turbine assembly has been qualified via dynamic testing, in accordance. with IEEE 344-1975. The ' qualification test program included demonstration of startup and shutdown capabilities, as well as no-load operability during f aulted loading conditions. 3.9-38
4 1 1 j LSCS-FSAR AMENDMENT'39L
. OCTOBER 11978 . Requirements l
The specification for qualification of the~ RCIC turbine and its accessories states; that they shall be capable mof withstanding the .
'specified1 accelerations at all! frequencies within the range. of 0.25 to 33 hertz.. Proper performance may be demonstrated by the seller by tests, : analysis, or a combination of both. If all natural frequencies of the turbine,' the component' parts, and the accesories are greater than. 33' hertz (as defined by -test and/or analysis) , a static load analysis may be performed.. The forces on each component' or assembly 1are obtained by concentrating its mass at the center of nassiof the component or assembly,-and
. multiplying by the acceleration coefficient. The; magnitude of
- the' acceleration coefficients are 1.5g both horizontal and vertical. If ~ component parts and/or' accessories have : natural- -
frequencies.below 33.~ hertz,-these parts must by dynamically analyzed ' or. tested, demonstrating 1satisf action of the defined floor response spectra.- If the equipment capability is demonstrated by test, the equipment must be subjected to simultaneous: horizontal and vertical acceleration inputs of . random wave-form motion for a minimum duration' of 30 seconds. The randon' input must envelop the defined floor response spectra. ()- Test QuaJ ification Results The RCIC turbine assembly was subjected to a total of 33 vibratory tests with an accumulated' test time of 905 seconds (%15 minutes) . Input to the equipment was random wave-form motion in two directions - one horizontal and the other vertical, with sine beats superimposed at 1/3-octave intervals as necessary to " envelop the required response spectra. ; A 200-psi, 1200-cfm air source was used as the operating medium f or the turbine. The , electronic governor system, the turbine hydraulic system with interconnecting . piping and levers, and all turnine instrumentation were under actual operating conditions during the test program. Nozzle loadings were simulated for the turbine inlet and exhaust piping. t The test results proved the operability-startup, no-load steady- { state, and shutdown capabilities of the turbine . assembly and its ' components during the required loading conditions. Stress levels ," were not measured during the test program, since the turbine was not subjected to fu11' pressure and temperature. conditions. This qualification has been accomplished analytically, as indicated I below. jh 3.9-39 s
, , - - - - , . . , . - -. , , - , . . , , . . , + - a , o n. na , , , - . , , . . -, . .:n. .d ., , , ,v, ,,n- ,. e,.,-, .,,n-r ,-,e -,, , , a
l L SCS-F SAR AMENDMENT 39 OCTOBER 1978 h 1 ? ' Analytical Qualification ~ Results The-rigid components of the RCIC turbine assembly have been analytically qualified .via static analysis for vibratory loading, 1
. as well as. the design operating loads of pressure, temperature, '
and external piping loads.. The results of this analysis confirm
.that the' stresses..in all components are substantially below allowable levels, i.e., less than~ 855 allowable, with the single exception of the ' turbine ' governor . va lve. i 3 . 9 .'3 . 2 . 4 'ECCS Motors-The analysis of the. ECCS motors' is .perf ormed by a computer !'
program 'which consists' of 'the ~ static mechanical . analysis of motor j rotor assembly wheni acted upon by ; externalf forces including
- magnetic-and centrifugal forces at'any' point along the shaft. !
The+ calculation for the seismic condition assumes that the' motor is operating. and: the vibratory, magnetic and centrifuga11 forces-all' act si'multaneouslyfand'in phase on .the rotor shaft ~ assembly. Other components of the1 motor, such f as stator . frame, -lower-end' shield, stator supports, base fasteners, Ltop cap,cand1 conduit - Lbus, are diecked for the combined effects including self-weight-and operational loadings, ' and consideration of bending, shear,
. torsion, anr'63 rect' bearing loads. i The analys) and tests that are used for qualification of ECCS pump motors were performed on an ECCS test motor of very similar mechanical construction.
The type test: was performed on a .1250-hp vertical; motor in ! accordance with IEEE 323-1974, first simulating normal operation during the design lif e, then the motor being eubjected to a ' number of vibrating events, and then to the abnormal l environmental condition possible_ during. and after a loss-of- ' coolant accident .(LOCA) . The test plan f or the type test was as follows:
- a. Thermal aging of the motor electrical. insulation .
' system -(which is a part of the. stator only) was based on< extrapolation.in accordance with:the temperature-j life characteristic curve from IEEE 275-1966 for the
!1nsulation type used on the ECCS motors. The amount 1
Lof. aging equaled the tota 1Jestimated operation days l 1 of maximum insulation surface' temperature.- l
- b. . Radiation aging of the motor.electrica1 ' insulation.
; equals 1the, maximum estimated integrated dose of. gamma during normal and abnormal. conditions.
!n c. The normal operation induced current vibration effect
, (Jf .on the~ insulation system was simulated by 1.5g
'l ,
horizontal' vibration acceleration at current
# 'J ' frequency for .1 hour duration.
01 3.9-40;
- - . _ , m._ _ _ - . _ _ . . . :_ . . . - . --_ ---.-.;._.-_--___.___-.
l l I I LSCS-FSAR' AMENDMENT 39 OCTOBER 1978
/~T l
' 'd. . The deflection analysis perf ormed: on the.' rotor shaft to ensure adequate rotation _ clearance, was. verified by static loading and deflection of the rotor for the '
type test motor.
- e. ' Aging .and testing was _ performed on a biaxial test table in accordance with IEEE 344-1975. During this type ' test,- the. shake table was activated simulating the ' vibration ' design limit of the saf e shutdown .
earthquake with motor starts and operation conditions-which may possibly_.' occur.during a plant life.
- f. An environmental test simulating a LOCA condition' with 100_ days duration tine was perf ormed with the-test motor . fully loaded, ' simulating pump operation.
The test consisted of'startup'and six hours operation at 2120 F.ambientLtemperature and 100% steam environment. Another startup and operation of the; test. motor after 1 hour standstill'in the same environment was followed by sufficient operation- at high humidity and temperature, based _on extrapolation in accordance with the ' temperature flif e - characteristic curve' from IEEE- 275-.1966 tor the insulation type used on the'ECCS' motors. 3.9.3.2.5 NSSS Valves I The Class 1 active valves are the .sain steam isolation valves, saf ety/ relief valves, and the standby liquid control valves. These valves are designed to perform their _ mechanical motion irt conjunction with a design base. accident. Qualification for operability is unique for each valve _ type; therefore, each method of qualification is detailed individually in the 2cllowing.
)
3.9.3.2.5.1 Main __ Steam Isolation-Valve (MSIV)
~
The MSIV's are evaluated for operability during dynamic events by both analysis and-test.
. Analysis - The valve body is designed in accordance with the ASME Boiler and Pressure-Vessel . code, Section III, ' Class 1 (Table 3.9-9).: The code-limits deformation in the operating area of :the valve body'to be within the elastic limit of the material by limiting pressure and pipe reaction input loads (including dynamic) thereby assuring no interference with valve operability.
In order to assure design limits'are not exceeded for both piping l input loads and actuator dynamic loads, the MSIV is mathematically modeled in the' main' steam line system analysis. The' valves' actual ' input ' loads, . amplified accelerations, and
, .l resource frequencies ~ are determined based on site excitation input to the system as a part ' of the overall steamline analysis.
Pipecanchors andLrestraints are employed as required to-limit
'pipeJsystem resonance frequencies and amplified accelerations .to 3.9 41
'LSCS-FSAR AMENDMENT 39 OCTOBER 1978 within acceptable limits for the MSIV's. The MSIV analytical qualification results.are'shown in Table 3.9-9.
Test - A dynamic test is conducted on the MSIV actuator to assure operability at design dynamic loading requirements. A sine wave sweep- test is used to. determine resonance frequencies of .the actuator _ assembly. A sine beat is used to excite the actuator assembly at all frequencies up to 50 hertz with special emphasis at the resonance frequency. Operability is then demonstrated by a response spectrum test which verifies that no significant change in valve closing rate resulted from the test.. It is also demonstrated thatL the valve configuration had sufficient integrity to withstand the required simulated dynamic - event
= without compromise of structure or -electrical f unction.
Faulted conditions such as loss-of-coolant accident (.LOCA) or downstream sine' break have been factored into the valve requirements.. The LOCA wil1 ~ not eff ect. valve closure which is demonstrated by valve qualification test. The valve is also demonstrated to close following: a _ down streamline break by the
" State Sine Test." The main steam isolation valve operability during LOCA~ conditions was demonstrated.as defined in the report APED-5750 (March 1969). The' test specimen was a 20-inch valve of -
a design representative of the LScs MSIV's. (k 3.9.3.2.5.2 Main Steam _Saf etv/ Relief Valves The SRV's are qualified by test for operability during a dynamic event. Structural integrity of the configuration during a dynamic event is demonstrated by both code analysis and test. Analysis - Valves are designed .for maximum moments which may be imposed when installed in service for inlet and outlet conditions of 800,000.in.-lb and 600,000 in.-lb respectively. These. moments are resultants due to dead weight, therral expansion, plus - dynamic loadings (0.59 horizontal and 4.5g vertical) of both valve and the connecting pipe. The safety / relief valve analytical qualification results are shown in Table 3.9-8. A mathematical model of the safety / relief valve is included in the main steamline system analysis to assure that'the equipment- .i design limits are not exceeded. Test - A production saf ety/ relief valve demonstrates operability during a dynamic qualification (shake table) test with moment and "g" loads applied - greater than the specified equipment. design
' limit Joads. Tests include a resonance frequency search and natural frequencies have been determined to be 2 33 hertz. The-test _ qualification results of the safety / relief valve are shown in Table ~ 3. 9-8.
- 3. 9-42
l LSCS-FSAR AMENDMENT 39 l OCTOBER 1978 f} \ 3.9.3.2.5.3 Standbv Liquid control valve (Explosive Valve) The SLC explosive valves have been generically qualified to IEEE 344-1975 by the vendor. The explosive valves are qualified for operability by test firing the detonator under representative acceleration and environmental conditions. The generic qualific'ation test demonstrated ~ the absence of natural frequencies below 33 hertz, and the. ability to remain operable af ter the application of horizontal. dynamic loading equivalent to 6.5g and vertical dynamic loading equivalent to 4.5g at 33 hertz.
~
3.9.3.3 Desian and Installation Details for Mounting of Pressure of Relief Devices safety valves and relief valves with free discharge were analyzed in accordance with the ASME Section III Code,1971 and Summer 1972 Addenda. The method of analysis for safety valves and relief valves suitably accounted for the time-history of loads acting immediately following a valve opening (i.e. , first few milliseconds) . The fluid induced forcing functions were calculated for each safety va3ve and relief valve using one-dimensional equations for the conservation of mass, momentum, and energy. {]) The calculated forcing functions were applied at locations along the associated piping where a change in fluid flow direction occurs. Application of these forcing functions to the associated piping model constituted the dynamic time-history analysis referred to as a hydraulic transient analysis which calculated the dynamic response of the piping system to the forcing functions. Therefore, a dynamic amplification f actor is inherently accounted for in the analyses. These analyses accounted for the actual fact that all valves discharge outwardly away from the reactor vessel simultaneously. This creates maximum energy into the piping system at one time with the lower mode excitation dominating. Thus, the results indicate the maximum response of the associated piping system. It would be noted that the main steam relief valve piping going to the suppression pool has a column of water sitting in the pipe. The hydraulic transient analyses on this piping have accounted for blowing out this water column. Hydraulic snubbers or strut-type restraints are used on all relief valve and safety valve piping to ensure that the stresses resulting from the loads produced by the sudden opening of a relief. or saf ety valve when combined with stress due to other .() upset loads satisfy the ASME Section III code for upset condition s. Also, the analyses show that the loads applied to 3.9-43
LSCS-FS AR AMENDMENT 39 OCTOBER 1978 g
~ the flanges of the safety and relief valves do not exceed the maximum loads specified by the manuf acturer.
3.9.3.4 Component Supports ASME Class 1, 2, and 3 active components are either pumps or valves. Since valves are supported by piping and not tied to building structures, pipe design criteria govern. Seismic category I active pump supports are seismically qualified by testing when the pump supports along with the pump are f ulfilling the following conditions:
- a. simulate actual mounting conditions;
- b. simulate all static and dynamic loadings on the pump;
- c. monitor pump operability during testing;
- d. the normal operation of the pump during and after the test indicates that the supports are adequate; and
- e. supports are inspected for structural integrity after the test.
G k/ Seismic qualification of component supports by analysis is generally accomplished as follows:
- a. Stresses at all support elements and parts such as pump holddown and baseplate holddown bolts, pump support pads, pump pedestal, and foundation are checked to be within the allowable limits as specified in ASME Subsection NF.
- b. For normal and upset plant conditions, the deflections and deformations of the supports are assured to be within the elastic limits and not exceed the values permitted by the designer based on his design verification tests to ensure the operability of.the pump.
- c. For emergency and faulted plant conditions, the deformations must not exceed the values permitted by the designer to ensure the operability of the pump.
Elastic / plastic analysis will be performed if the deflections are above the elastic limits. U 3.9-44
s
. dr '
LSCS-ISAR- AMENDMENT'39 OCTOBER 1978 ' i 3.9.4 Control Rod Drive Systems (CRDS) 3.9.4.1' Descript'ive Information of CRDS' Descriptive Einformation on' the control rod ' drive systems is given in, subsection 4. 6.1.. -In : particular, the design criteria ~'are j discussed in Subsection 4.6.1.1.1, description of the system'is-given in Subsection 4.6.1.1.2, and . method of operation to evaluate adequacy. of ' system is given in subsection 4. 6. 2. 3.9.4.2 Applicable CRDS ' Design Specifications - The quality group classification, code classification, and .; standards applied in- the design, fabrication, and construction of l the CRDS are defined in Subsection 3.'2.2. Regulatory guide ~ conformance'for this' system is! addressed in Appendix B. The portions of the CRDS that form. a part' of the reactor coolant pressure -boundary, .as ~ described in Subsection 4.6.1, have been analyzed in accordance with- the' requirements of AStGE Section III, as s required by the applicable. code . classification. The remainder of the CRDS.has been analyzed.to the requirements of a Group D system, ; as ? defined in Subsection 3. 2.2. 3.9.4.3 Design Loads, Stress Limits, and Allowable U's Deformations-Design loading. combinations are categorized-with respect to plant ! conditions identified as normal, upset, emergency, or faulted as
- shown in Table 3. 9-3 for the ma jor CRDS components. Stress analysis was used to determine the structural adequacy of pressure-retaining components under the design loading combina tions. These analyses utilized allowable stresses and deformation limits established by ASME Section III, and were ;
performed by methods outlined in the Code. Table 3.9-3 also gives the calculated stress levels ~or maximum loadings at significant areas for the major system components. A design
. assessment calculation' was conducted to include dynamic loads.
3.9.4.4 CRDS Performance Assurance Program Quality control (QC) 'of welding, heat treatment, dimensional tolerance , material. verification, and other f actory QC tests are used throughout the manuf acturing process to ensure- reliable performance of the ' mechanical reactivity control components. Acceptance tests- include the -following: (1). control rod absorber tube tests 'to verif y integrity, (2) control rod drive mechanism tests, and (3). hydraulic control unit tests to authenticate . operational performance. !
,n\. 1
(_/? i Afterfinstallation,: all rods and drives are tested. through their i fullistroke for operability. During operation, each' time a contro1D rod is withdrawn a notch, the operator can observe the t 3.9-45
1 l LSCS-FSAR AMENDMENT 39 OCTOBER 1978 t') LJ incore monitor indications to verify that the control rod is , I following the drive mechanism. Hydraulic supply subsystem pressures can be observed from instrumentation in the control room. Scram accumulator pressures can be observed on the nitrogen pressure gauges. Preinstallation specifications define acceptance criteria for characteristics such as seal leakage, friction, and scram perf ormance under fixed test conditions. Normal and scram motions are authenticated in preoperational tests to illustrate proper installation. A surveillance test is made following core alterations to demonstrate adequate shutdown margin. Also, routine rod withdrawal excercises are made by notch motions to authenticate operable control rods during power operations. Coupling and overtravel tests are also a part of rod excercising tests. Scram tests are made at refueling outages to authenticate scram times within acceptable limits. 3.9.5 Reactor Pressure Vessel Internals 3.9.5.1 Design Arrangement (")
The core support structures and reactor vessel internals include (exclusive of fuel, control rods, and incore nuclear instrumentation) the following components:
- a. Core Support Structures
- 1. shroud;
- 2. shroud support;
- 3. core support and holddown bolts;
- 4. top guide (including wedges, bolts, and keepers) ;
- 5. fuel support pieces; and
- 6. control rod guide tubes.
- b. Reactor Internals
- 1. jet pump assemblies and instrumentation;
- 2. shroud head and steam separator assembly (including shroud head bolts) ;
;'~') 3. steam dryers; LJ
- 3. 9-46
LSCS-FSAP AMENDMENT 39 OCTOBER 1978 I) ^' l
- 4. f eedwat er spargers; 1
- 5. vessel head cooling spray nozzle; !
l
- 6. differential pressure and liquid control line;
- 7. incore flux monitor guide tubes and stabilizers; j
- 8. initial startup neutron sources;
- 9. surveillance sample holders; *
- 10. core spray lines; and
- 11. spargers (part of shroud) .
A general assembly drawing of the important reactor components is shown in Figure 3.9-2. The floodable inner volume of the reactor pressure vessel can be seen in Figure 3. 9-2. It is the volume inside the core shroud up to the level of the jet pump suction inlet. 3.9.5.1.1 Core Support structure /') (/ The core support structure consists of the shroud, shroud support, core support, fuel support pieces, control rod guide tubes, and top guide. This structure is used to form partitions within the reactor vessel, to sustain' pressure ~ differentials across the partitions, to direct the flow of the coolant water, and to locate laterally and support the fuel assemblies. Figur e 3.9-3 shows the reactor vessel internal flow paths. 3.9.5.1.2 Core shroud The core shroud is a stainless steel cylindrical assembly that provides a partition to separate the upward flow of coolant through the core from the downward recirculation flow. This partition separates the core region from the downcomer annulus, thus providing a floodable region following a recirculation line break. The volume enclosed by the shroud is characterized by three regions. The upper. shroud surrounds the core discharge plenum, which is ' bounded by the shroud head on top and the top guide below. The central portion of the shroud surrounds the active fuel and forms the longe > aection of the shroud. This section is bounded at the bottom by the core support. The lower shroud, surrounding part of the lower plenum, is welded to the reactor pressure vessel shroud support (see Section 5. 3) . ?) v
- 3. 9-47
~ - -
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 3.9.5.1.3 Shroud Head and Steam Separator Assembly The shroud head and steam separator assembly is bolted to the top of the upper shroud flange to form the top of the core discharge plenum. This plenum provides a mixing chamber f or the steam-water mixture before it enters the steam separators. . Individual stainles's steel . axial flow steam separators, shown in Figure 4.1-5, are . welded to the top of standpipes that are welded into the shroud head. The steam separators have - no moving parts. EIn each separator, the steam-water mixture rising through the standpipe passes vanes that impart a spin to establish a vortex separating the water from the steam. . The separated water flows from the lower portion of the steam separator into the downcomer annulus. 3.9.5.1.4 Core Support Plate The core support plate consists of a circular stainless steel-plate with bored hc3es stiffened with a rim and beam structure. The plate provides lateral support and guidance for the control rod guide tubes, incore flux monitor guide tubes, peripheral fuel supports, and neutron sources. The last two items are also supported vertically by the core support plate. The entire assembly is oolted to a support ledge _ between the (]) central and lower portions of the core shroud. Alignment pins that engage slots and that bear against the shroud are used to correctly position the assembly before it is secured. 3.9.5.1.5 Top Guide The top guide is formed by a series of stainless steel beams joined at right angles to form square openings and fastened to a peripheral rim. Each opening provides lateral support and , guidance for four f uel assemblies or, in the case of peripheral 1 fuel, one fuel assembly. Notches are provided in the bottom of 1 the beam intersections to anchor the incore flux monitors and startup neutron sources. The rim of the top guide rests on a ledge between the upper and central portions of the shroud. The top guide has alignment pins that engage and bear against slots in the shroud which are used to correctly position the assembly before it is secured. Lateral restraint is provided by wedge blocks between the top guide and the shroud wall. 3.9.5.1.6 Fuel Support l The f uel supports, shown in Figure 3.9-4 are of two basic types; ; namely, peripheral supports and four-lobed orificed fuel l supports. The peripheral fuel support is located at the outer edge of the active core and is not adjacent to control rods. O Each . peripheral f uel support will support one fuel assembly and contains a single orifice assembly designed to ensure proper , coolant ' flow to the fuel peripheral assembly. Each four-lobed 1 3.9-48
.___ _ __._ __..____ _. _ ___ __.._ _ _.._,.__ - , _ . _ . _ , _ _ ~ _ _
f LSCS-FSAR AMENDMENT 39 OCTOBER 1978 orificed fuel support will' support four. fuel assemblies and is provided with- orifice plates to ensure proper coolant flow distributionLto each rod-controlled fuel assembly. .The four-lobed- orificed. fuel supports rest' in the top"of the control rod guide tubes which are supported laterally by the core support. The control rods _ pass through slots in the center of the four-lobed orificed fuel-support. A control rod and the four -adjacent ! f uel assemblies represent a core cell (see subsection 4. 2. 2) . 3.9.5.1.7 Control' Rod Guide Tubes [ The control rod guide tubes, located inside the vess' el,. extend f rom the top of -the control rod drive housings up .through holes . in the core support plate. Each tube is designedJas the guide for a control rod and as the vertical support ;for a .four-lobed orificed fuel support piece and the.four fuel assemblies surrounding the control rod. The bottom'of the guide tube is. supported by'the control rod drive housing (see Subsection
- 5. 3. 3) , which in turn transmits the weight of 'the guide tube, fnal support, and fuel assemblies'to the reactor vessel' bottom head. A thermal sleeve is inserted into~ the control rod drive-housing from below and is rotated to lock the control rod guide tube in place. ' A key is inserted into a locking slot in .the bottom of the control rod drive housing to hold the thermal
(~g sleeve in position. J 3.9.5.1.8 Jet Pump Assemblies The jet pump assemblies are located in two semicircular groups in the downcomer annulus between the core shroud and the reactor vessel wall. The. design and performance of the jet pump is covered in detail in References 5 and 6. Each stainless steel
. jet pump consists of driving nozzles, suction inlet, throat or mixing section, and diffuser (see Figure 3. 9-5) . The driving nozzle, suction inlet, and throat are joined together as a removable unit, and the diffuser is permanently installed. High-pressure water' from the recirculation pumps is supplied to each pair of jet pumps through a riser pipe welded to the recirculation' inlet nozzle thermal sleeve. A riser brace consists of cantilever - beams extending from pads on the reactor ,
vessel wall. , The nozzle entry section is connnected to the riser by a metal- , to-metal, spherical-to-conical seal joint. Contact is maintained by. a hold-down clamp. The throat section is supported laterally by a bracket attached to the' riser. There is a slip-fit joint
-between the throat and diffuser. The diff user is a gradual I conical section changing to a straight cylindrical section at the i lower end.
L)g l l 3.9-49
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 3.9.5.1.9 Steam Dryers The steam dryers remove moisture from the wet steam leaving the steam separators The extracted moisture flows down the dryer vanes to the collecting troughs, then flows through tubes into the downcomer annulus (see Figure 3. 9-6) . A skirt extends from the bottom of the dryer vane housing to the steam separator standpi, Delow the water level. . This skirt forms a seal between tne . wet steam plenum and the dry steam flowing from the top of the dryers to the steam outlet nozzles. The steam dryer and shroud head are positioned in the vessel during installation with the aid of vertical guide rods. The dryer assembly rests on steam dryer support brackets attached to the reactor vessel wall. . Upward movement of the dryer assembly, which would occur only under accident conditions, is . restrict ed ^ by steam dryer hold-down brackets attached to the reactor vessel top head. 3.9.5.1.10 Feedwater Spargers The feedwater spargers are stainless steel headers located in the mixing plenum above the downcomer annulus. A separate sparger is welded to each feedwater thermal sleeve which is welded to the vessel nozzle and is shaped to conform to the curve of the vessel gA-wall. Sparger support is undergoing redesign, hence the specific attachment geometry and technique will be reported later. Feedwater flow enters the center of the spargers and is discharged radially inward to mix the cooler fee'dwater with the downcomer flow from the steam separators before it contacts the vessel wall. The f eedwater also serves to condense the steam in the region above the downcomer annulus and to subcool the water flowing to the jet pumps and recirculation pumps. 3.9.5.1.11 Core Spray Lines The core spray lines are the means for directing flow to -the core spray nozzles inside the shroud which distribute coolant so that peak fuel cladding temperatures of 22000 F are not exceeded during accident conditions. Two core spray lines enter the reactor vessel through the two core spray nozzles (s ee S ection 5.4) . The lines divide immediately inside the reactor vessel. The two halves are routed to opposite sides of the reactor vessel and are supported by clamps attached- to the vessel wall. The lines are then routed downward into the downcomer' annulus and pass through the upper shroud immediately below the flange. The flow divides again as it enters the center of the semicircular sparger, which is routed t halfway around the inside of the upper shroud. The ends of the two spargers are supported by brackets designed to accommodate thermal expansion. The line routing and supports are designed to 3.9-50
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 73 \J accommodate differential movement between the shroud and vessel. The other core spray line is identical except that it enters the opposite side of the vessel and the spargers are at a slightly different elevation inside the shroud. The correct spray distribution pattern is provided by a combination of distribution nozzles pointed radially inward and downward from the spargers (see Section 6. 3) . 3.9.5.1.12 Vessel Head Cooling Spray Nozzle When reactor coolant le returned to the reactor vessel, pa .t of the flow can be diverted to a spray nozzle in the reactor head. This spray maintains saturated conditions in the reactor vessel head volume by condensing steam being generated by the hot reactor vessel walls and internals. The spray also decreases thermal stratification in the reactor vessel coolant. This ensures that the water lerel in the reactor vessel can rise. The higher water level provides conduction cooling to more of the mass of metal of the reactor vessel and therefore limits thernal stress in the vessel during cooldown. The vessel head cooling spray nozzle is flange mounted to a , mating flange on the reactor vessel head nozzle (see subsection
- 5. 4. 7) .
r~s i, ) 3.9.5.1.13 Differential Pressure and Liquid Control Line The differential pressure and liquid control line serves a dual f unction within the reactor vessel - to provide a path for the injection of the liquid control solution into the coolant stream and to sense the diff erential pressure across the core support plate (described in Section 5.3) . This line enters the reactor vessel at a point below the core shroud as two concentric pipes. In the lower plenum, the two pipes separate. The inner pipe terminates near the lower shroud with a perforated length below the core support plate. It is used to sense the pressure below the core support plate during normal operation and to inject liquid control solution if required. This location facilitates good mixing and dispersion. The inner pipe also reduces thermal shock to the vessel nozzle should the standby liquid control system be actuated. The outer pipe terminates immediately above the core support plate and senses the pressure in the region outside the fuel assemblies. 3.9.5.1.14 Incore Flux Monitor Guide Tubes Thes'e tubes provide a means of positioning fixed detectors in the core as well as provide a path for calibration monitors (TIP system) . (~)
~
The incore flux monitor guide tubes extend from the top of the incore flux monitor housing (see Section 5.3) in the lower plenum to the top of the core support plate. The power range detectors 3.9-51
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 l e (- for the power range monitoring units and the dry tubes for the source range monitoring and intermediate range monitoring (SRM/IRM) detectors are inserted through the guide tubes. A latticework of clamps, tie bars, and spacers gives lateral support and rigidity to the guide tubes. The bolts and clamps are welded, after assembly, to prevent loosening during reactor operation. 3.9.5.1.15 Surveillance Sample Holders The surveillance sample holders are welded baskets containing impact and tensile specimen capsules (see Subsection 5. 3.1) . The baskets hang from.the brackets that are attached to the inside wall of the reactor vessel and extend to midheight of the active core. The radial positions are chosen.to expose the specimens to the same environment and maximum neutron fluxes experienced by the reactor vessel itself while avoiding jet pump removal interference or damage. 3.9.5.1.16 Low-Pressure Coolant Iniection Lines Two LPCS and one LPCI line penetrata the core vessel through separate nozzles. Coolant is discharged inside the core shroud
,, immediately below the top guide.
t > \s' 3.9.5.2 Design Loading conditions I The spectrum of conditions for which the safety design basis must be satisfied for the three dominating f aulted events is j described in Subsection 3.9.2.5.2.2. 3.9.5.2.1 Pressure Differential During Rapid Depressurization l Pre ssure differentials producea during rapid depressurization are analyzed as shown in Subsection 3.9.2.5.2.3. 3.9.5.2.2 Recirculation Line and Steamline Break l l 3.9.5.2.2.1 Accident Definition Both a recirculation line break and an inside steamline break were previously discussed in Subsection 3.9.2.5. 3. l The steamline break accident produces significantly higher l pressure differentials across the reactor assembly internal l structures than does the recirculation line break. This results from the higher reactor depressurization rate associated with the I steamline break. The depressurization rate is proportional to i the mass flow rate and the excess of fluid escape enthalpy above ' saturated water enthalpy, h f. Mass flow rate is inversely (~)T i N- proportional to escape enthalpy, n e , and therefore the depressurization rate is approximately proportional to 1 - h f/h e* Consequently, the depressurization rate decreases as h e l 3.9-52.
LSC3-FSAR AMENDMENT 39 OCTOBER 1978 O \/ decreases, that is, the depressurization rate is less for mixture flow than for steam flow. Therefore, the steamline break is the design-basis accident for internal pressure diff erentials. 3.9.5.2.2.2 Effects of Initial Reactor Power and Core Flow For purposes of illustration, the maximum internal pressure loads can be considered to te composed of two parts: steady-state and trancient. pressure dif ferentials. For a given plant the core flow and power are the two major factors which influence the reactor internal pressure differentials. The core flow essentially affects only the steady-state part. For a fixed power, the greater the core flow, the larger will be the steady-state pressure diff erentials. The core power affects both the steady-state and the transient part s. As the power is decreased there is less voiding in the core and consequently the steady-state core pressure dif ferential is less. However, less voiding in the core also means that less steam is generated in the reactor pressure vessel and thus the depressurization rate and the tranelent part of the maximum pressure load is increased. Figure 4.4-3 is a power-flow map which defines the permissible operating conditions of the reactor. From this range of operating conditions, it is necessary to determine the gs (,) combination of core power and flow which will result in the maximum internal pressure loads. Consider the condition where steam flow is 105% and the core flow is at 100% of rated conditions. Si nc e , as mentioned previously, a decrease in power will result in higher transient pressure differentials, a more severe initial condition might be the condition of 66.9% power, 110% core flow. In going from 105% steam flow, 100% core flow to the 66.9% power, 110% core flow condition, the steady-state pressure differential has a net decrease. There is an increase due to the slight increase in core flow and a decrease due to the decrease in power (lower core pressure drop) . The transient pressure differential increases due to the decr ease in power. However, the maximum pressure load (steady-state plus transient) has a net increase for the high flow low power condition. If the power is decreased below 66. 9%, the core flow must also be reduced. Analysis has shown that the decrease in core flow and power reduces the steady-state part of the maximum pressure load more than the corresponding increase in the transient part. Hence, the maximum pressure loads (steady-state plus transient) are less if the core flow is reduced from its maximum value. Therefore, the maximum internal pressure loads occur following an inside steamline break from an initial condition in which the reactor is at the minimum power associated with the may1 mum core flow (i. e. , 66. 9 % power, s 1104 core flow) .
- 3. 9-53
LSCS-ISAR AMENDMENT 39 OCTOBER 1978
-Table 3.9-1 lists the maximum pressure loads occurring across the reactor internals during the accident f or two cases. Case 1 is for an initial condition 'of 105% rated steam flow and 100% rated core flow. Case 2 is for the maximum pressure loads and these occur at the initial condition of 66.9% power, 110% flow.
Comparison of Cases 1 and ~ 2 illustrates the generalized statements made previously concerning the relationship between the maximum internal pressure loads and core power and flow. The transient values of the pressure differences for Case 2 are shown in Figure 3.9-8. Realistically, if an inside steamline break were to occur,- the maximum internal pressure loads would probably be closer to Case
- 1. This is because the plant will most probaDly be operating at or near f ull power. Also, the Case 2 condition, although possible, is rather abnormal in that rated core flow is neither required nor desirable at such a reduced power condition.
3.9.5.2.2.3 Response of Structures Within the Reactor Vessel to Pressure Differences The maximum differential pressures are used, in comoination with other structural loads, to determine the total loading on the various structures within the reactor. The structures are then evaluated to assess the extent of deformation and buckling () ins tability, if any. Of particular interest are: (1) the responses of the guide tubes and the metal channels around the fuel bundles, and (2) the potential leakage around the jet pump joints. The guide tube is evaluated for buckling instability caused by externally applied pressure. Two prin.ary modes.of f ailure have been analyzed and are described in Subsection 3. 9. For a guide tube with minimum wall thickness and maximum allowed ovality, the pressure which causes yield stress is 93 psi compared to the service design pressure of 37.5 psi. The design pressure is in all cases greater than any pressure diff erential the guide tube will experience including accident conditions. The stress the guide tube would experience is given in Subsection 3. 9.2. It is concluded that the guide tube will not f ail under the assumed conditiona. The f uel channel load resulting from an internally applied pressure is evaluated, utilizing a fixed-beam analytical model under a uniform ' load. Tests to verify the applicability of the analytical model indicate that the model is conservative. A roller, at the top of the contrcl rod, guides the blade as it is inserted. If the gap between channels is less than the diameter of the roller, the roller deflects the channel walls as it makes its way into the core. The friction force is a small percentage f')'
' of the total force available to the control rod drives for overcoming such friction, and it is concluded that the main
- 3. 9-54
t L SCS-ES AR ' AMENDMENT 39 OCTOBER 1978
~
( k. steamline' break accident does not impede the insertability of the
- control rod.
~!
Jetf pump - joints have 'been. analyzed to ' evaluate the potential leakage f rom within1 theffloodable inner volume of the reactor vessel during. the recirculation line break and subsequent LPCI reflooding. Because the. jet = pump diffuser is welded to the ' shroud, support,' the only < remaining source of leakage from the lowerJplenum.to the downcomer annulus is the jet pump throat to-diffuser joint. : These joints.for-all~ jet pumps leak no more than.
? totallof 225 gpm.
LPCI capacity ' is sized 'to accomodate 500 gpm leakage at these . locations. It is concluded that the reactor vessel structures retain sufficient integrity during the ' recirculation line break accident to allow reflooding of the inner volume of the reactor vessel and in sufficient time to prevent significant -increases in. ' t cladding temperature. 3.9.5.2.2.4 Conclusions t It is concluded that the maximum pressu;e loads acting on the reactor -internal components result .fr9m an inside steamline break occurring while the reactor .is at 6e minimum power associated with the maximum core flow -(Tabir. 3. 9-1, Case 2) .
~
This has been (7) ! substantiated by. the analytica' comparison of liquid versus steam breaks, by the investiaticn sf the effects of core-power and core flow,'and by the break spectrum analysis. It has also been pointed out that, altbm '
'sible, it is not probable that the reactor would be opt , ou the rather abnornal condition of minimum power ano maximum core flow. More ,
realistically, the reactor .would be at or near a- f ull power 'i condition and thus the maximum pressure loads acting on the' ' internal' components would be as listed under Case 1 in Table 3.9-1. 3.9.5.2.3 Dynamic Loads The seismic and dynamic loads acting on ;the structures within the ; reactor vessel are based on a analyses as described in Section 3.7. 3.9.5.2.4 Safety Evaluation 3.9.5.2.4.1 Evaluation Methods
' To. determine that the saf ety design bases are satisfied, responses of the reactor vessel internals to loads imposed during normal, upset, emergency, and faulted conditions are examined.
(~')
. The eff ects on the ability to insert control rods, cool the core,
. and 4to' flood the inner volume of the reactor vessel are
- 3. 9-55
_,_m_ ..,m., . . , _ . J,4w- , . . . . , , , , , . ,, y ., ,,
. . .~.
LSCS-FSAR AMINDMENT 39 OCTOBER 1978 d etermined. The design assessment procedures are included in Subsection 3.9.1.1. 3.9.5.2.4.1.1 Input for Safety -Evaluation The operating conditions that provide the basis for the design of the reactor internals to sustain normal, upset, emergency, and f aulted conditions, as well as load combinations that were used
.for the core support structures, are covered in Table 3.9-25.
In addition, each combination of operating loads is categorized with respect to .either normal, upset, emergency, or f aulted conditions as well as the associated design stress intensity or deformation limits. The bases for the proposed design stress and deformation criteria are also specified in Chapter 3.0. 3.9.5.3 Design Loading Categories 3.9.5.3.1 Stress, Deformation, and Fatique Limits for Reactor Internals (Except Core Support Structure) The stress deformation and fatigue criteria listed in Tables
~
fg 3.9-26 through 3.9-29 were used or the criteria established in ( ,/ applicable codea and standards for sindlar equipment, by i manufacturers' standards, or by empirical methods based on field experience and testing. For the quantity SFmin (minimum saf ety , factor) appearing in those tables, the following values listed
' l were used:
l Design SF min Condition H Normal 2.25 I i Upset 2.25 l Emergency 1. 5 Fault 1.125 l 3.9.5.3.2 Stress, Deformation, and Fatique Limits for Core Support Structures 1 The stress, deformation and f atigue criteria presented in Tables 3.9-3 0- through 3.9-32 were imposed for the original design basis and the design assessment evaluation. These criteria shall be supplemented, where applicable, by the criteria for the reactor internals in the previous paragraph, but in no case shall the 7]) i, criteria presented in Tables 3.9-30 through 3.9-32 be exceeded
- for core support structures.
3.9-56 1
.. . _ . .. . . ~ . _ _
1 LSCS-FSAR - AMENDMENT 39 - OCTOBER 1978 () 3.9.5.4 Design Bases 3.9.5.4.1 Safety Design Bases-
.The reactor core support structures and internals shall meet the following saf ety design bases:
a .- Shall be arranged ;to provide a floodable ' volume in whi'ch the core can be adequately cooled thus limiting fuel damage,
- b. Deformation:shall- be limited to ensure tha't the. ,
control rod movement is not impaired,
- c. Mechanical design of applicable structures shall ensure that safety design bases (a) and (b) are sati' 'ted so that the ~ safe shutdown of the plant and r e' of decay heat are not' impaired.
3.9.5.4.2 3 ration Design Bases The reactor , port structures and internals are designed to the followint . generation design bases:
- a. They provide the proper coolant' distribution during t's all anticipated normal operating conditions to allow "d power operation of the core without' fuel. damage.
- b. They are arranged to facilitate refueling operations.-
- c. They are designed to facilitate planned maintenance ]
and periodic inservice inspection. 3.9.5.4.3 Fuel Assembly Restraints The f uel assembly structural design demonstrates _ sufficient dimensional stability and sufficient fuel rod support to maintain core geometry thus avoiding fuel . damage for both planned j operation and abnormal operational transients.
]
3.9.5.4.4 Material Selection j The material used for fabricating most of the reactor core support and reactor internal ' structures are solution heat-treated, unstabilized Type 304 austenitic stainless steel conforming to ASTM and ASME specifications. Weld procedures and welders are qualified in accordance with the intent of Section IX of the ASME Boiler and Pressure Vessel- Code. Further controls for stainless steel welding are covered in Subsection 5.2.3. All the materials of construction exposed to the reactor coolant
~f are to be resistant to stress corrosion in the BWR coolant. I 3.9-57 i
..._m . _.I
LSC S-FSAR AMENDMENT 39 OCTOBER 1978 ,o k- Conservative corrosion allowances are to be provided for all exposed surfaces of carbon or low alloy steels. Contaminants in the reactor coolant are controlled to very low limits by the reactor water quality specifications. No detrimental effects shall occur on any of the materials from-allowable contaminant levels in the high purity reactor coolant. Radiolytic products in a BWR shall have no adverse eff ects on the construction materials. 3.9.5.4.5 Radiation Ef fects Where f easible, the design is such that irradiation effects on the material properties are minimized. Where irradiation effects cannot be minimized, the design of the reactor vessel internals has provisions for replaceable components, or the design satisfies a set of stress and fatigue design limits that have been arrived at considering the effect of irradiation damage on the f racture toughness, ductility, and tensile properties of the materials. 3.9.5.4.6 Vibration Measurement of Reactor Internals Vibration analysis of reactor components is included in the f3 design to guard against potential problems from hydraulically (_) induced equipment vibrations. The vibration analysis is based on a flow-induced vibration test and inspection program. The prototype vibration testing has been. accomplished at Browns Ferry-1. The flow testing and inspection are to be performed as part of the recirculation system preoperation tests. This test and inspection program is a quality assurance measure intended to reveal loose parts or other assemoly errors which could cause vibration of reactor internals. These internal components are of a design which has previously been verified by a vibration measurement program conducted in accordance with provisions of federal reg ulations for prototype reactor internals. The specified test conditions are intended to produce vibration excitation comparable- to or greater than that experienced in normal modes of reactor operation. The duration of testing is l sufficient to produce at least 10s cycles of vibration at the l minimum significant response frequencies of major components, ' based on the measured vibration response of prototype reactor internals. The inspection is to be performed before and after the flow tests to detect any effects of flow-induced vittation. The established vibration program has been designed to satisfy dederal requirements. Field test data are correlated with the dnalyses to ensure validity of the analytical techniques on a l (__s) continuing basis. l l
\
l 3.9-58 1
LSCS-ESAR- AMENDMENT 39 OCTOBER 1978 '\
.The;GE acceptance : criteria are based upon the.'need- to ensurei that '
vibra tion : will :be ! acceptable for 710: 1 cycles L (40 years) ; therefore, 10T cycles has no specific relevance in.that context. .
-During cold : flow testing on prototype. BWR's, r the reactor:
internals are1 exercised in: excess. of 106 cycles by' reactor coolant flow. The postflow inspection consiots of visual-inspecti'on of selected components, including jet pumps andt components in the: lower plenum.- 3 The reactor: vessel and the componenta. within the vessel are
' designed to . ensure adequate. working' space and access. for inspection of selected components and locations. Criteria.for !
selecting the components and locations to be inspected are based- , on-the probability of a defect occurring or enlarging at a given- ! location and include areas of known stress concentrations' and locations where cyclic strain or thermal ' stress might occur. I The plant is designed to provide.for inservice inspection as- J required by ASME Boiler and Pressure Vessel Code, Section XI. . , 3.9.5.4.7 Accident Conditions. 1 Response analyses of the reactor structures show that ; deformations are sufficiently limited to allow boch ' adequate I p/ control rod insertion and proper operation of the emergency core (- cooling system. Sufficient integrity of the structures is retained during accident-conditions.to allow successful reflooding of the reactor vessel inner volume. The analyses considered various loading combinations, including loads imposed i by external forces. Thus, safety design bases are satisfied. , 3.9.5.4.8 Inspection and Testing Quality control methods are used during the fabrication and assembly of reactor vessel internals to ensure that the design specifications are met.
-1 The reactor coolant system, which includes the core support 1 structures and reactor internals, is thoroughly cleaned and flushed before fuel is loaded initially.
~During the preoperational test program, (Chapter 14.0) operational readiness tests are performed on various systems. In the course of these tests such reactor internals as the feedwater spargers, the core spray lines, the vessel- head cocling spray nozzle, and the standby liquid control system li 'a are f unctionally teste&. '
3.9.6 Inservice Testing of Pumps and Valves () - The LSCS reactor coolant : pressure boundary including the RPV's class 1 piping, and all Class 1, 2, 3, and D+ pressure retaining components,. defined according to. ASME Boiler and Pressure Vessel-3.9-59
. . ~ . - - .. -- . - . - . . . - . - . - ~ . . . . - . - . - - ~ . .!
)
LSC S-FS AR AMENDMENT 39 j OCTOBER 1978 I O \' Code Section III, will be examined in accordance with ASME Section XI, 1974 Edition including the Summer 1975 Addenda. Components exempted from examination are those specified in ASME l Section XI, IWB-1220 and IWC- 1220. One hundred percent of the Class 1, 2, and 3 components shall be examined prior to initial l 1 s ta rtup, except as exempted above. Examination categories for Class 1 and Class 2 components are as specified respectively in ASME Section XI, Tables IWB-2500 and IWC- 2520. Class 1 examination methods, as specified in Table IWB-2600, are the ultrasonic method for volumetric examination and either the liquid penetrant or magnetic particle methods for surf ace examination. Class 2 examination methods, as specified in Table IWC-2600 are these same methods. All Class 3 examinations are visual. Standards for evaluation of the examination are as follows:
- a. Class 1 - As specified in ASME Section X.',
IWA-3000 and IWB-3000
- b. Class 2 - As specified in ASME Section XI, and D+ IWA-3000 and IWC-3000 rw c. Class 3 - As specified in ASME Section XI,
(-) IWA-3000 and IWD-3000 The system pressure test for Class 1, 2, 3, and D+ components shall be conducted in accordance with ASME Section XI, IWA-5000 an6 ASME Section III, NB-6000, NC-6000, and ND-6000, r espectiv ely. Records and inspection reports for Class 1, 2, 3, and D+ components shall be developed and maintained as specified by ASME Section XI, IWA-6000. Personnel performing the nondestructive examinations shall be qualified per procedures prepared in accordance with SNT-TC-1A, June 1975, for the applicable examination method. Inservice Inspection The design of the RPV shield wall and external inservice inspec-tion equipment was complete prior to the publication of the amendment to 10 CFR 50.55a which requires the upgrading of the code commitment. Inasmuch as the LSCS plant might be required to meet the requirements of future editions of Section XI, insofar as practicable, an attempt was made during design to allow more inspection access and added anticipated coverage. The result of mechanical inspection devices and generally allowed piping examinations to be upgraded to the requirements of Summer 1975 () Addenda to ASME Section XI. f or LSCS preservice inspections. 10 CFR 5 0. 55a (g) (2) is applicable 3.9-60
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 O' '~ ' The. initial inservice- examinations conducted during the first 40 months shall comply with ASME Section XI, 1974 Edition-including the Summer 1975 Addenda, or under' the provisions of 10 CFR 50.55a (4) , with requirements in the editions' of the ASME Code and Addenda in effect no more than 6 months prior to the date of 1 start. of commercial operation for LSCS Unit 1. The Class 1, 2, and 3 ' pr' essure-retaining components including the RPV's, -Class 1, 2, and D+ piping, and the. Class .1,. 2, and. 3 pressure-retaining componen ts (as defined in ASME Section III) will be examined. Components exempted from examination are those .specified in ASME Section XI, IWB-1220 and IWC-1.220, respectively for , Classes 1 and 2. Examination Categories are the same as those identified for the preservice examination. Examination methods are also identical to those cited for the perservice examination. The standards for. examination evaluations, . the personnel qualification requirement, and the maintenance of records and reports are accomplished to the same code standards cited for the preservice inspection. The system pressure tests for class 1, 2, and 3 components shall be conducted in accordance with ASME Section XI, IWA-5000 and IWB-5000, IWC-5000, and IWD-5000 respectively. 3.9.6.1 Inservice Testing of Pumpe This inservice test program will be provided later as an i Appendix. I 3.9.6.2 Inservice Testino of Valves - This-inservice test program will be provided later as an 3 Appendix. I O
- 3. 9-61
LSCS-FS AR - AMENDMENT 39 OCTOBER 1978
) 3.9.7 References.
- 1. La Salle County Station, " Mark II - Design Assessment Report (LSCS-DAR) ," Commonwealth Edison Company, Chicago, Ill.' ,is ,
February 1976. -
- 2. E. P'.' ' Quinn, " Vibration of . Fuel Rods in Parallel Flow," USAEC Report GEAP-4059, . General . Electric Co., Atomic Power Equipment Department, July 19 6 2.
3.- E. P. Quinn, " Vibration of SEFOR Fuel Rods in Parallel Flow," USAEC Report GEAP-4966, . Atomic Power Equipment Department,. September 1965. 4 '. General Electric Company, " Analytical Model for Loss of Coolant Analysis in Accordance with 10 CFR 50 Appendix K," Proprietary Document, General Electric Company, NEDE-20566, November-1975. '
- 5. "n9 sign and Performance of GE BWR Jet Pumps," General EI : Company, Atomic Power Equipment Department, APED-5460, J ,_ ,6 8 .
- 6. R. H. Moen, " Testing of Improved Jet Pumps for the BWR/6 Nuclear System," General Electric Co. , Atomic Power Equipment h Department, NEDO- 1062, June 1972.
- 7. Mark II Containment Dynamic Forcing Functions Information Report (DFER) NEDO/NEDE 21061, Septtnber 1975.
- 8. ANSI 176 (Draf t) , " Design Basis f or Protection of Nuclear Power Plants Against Ef fects of Postulated Pipe Ruptures,"
January 1977. ,
- 9. A. Kalning, Journal of Applied Mechanics, Vol. 31 ,
op. 467-476, 'ptembce .19 64.
- 10. E. L. Wilson, "SAPA General Structural Analysis Program,"
Structural Engrg. Lab, Report UCSESM/7020, University of California at Berkeley, September 1970.
- 11. E. L. Wilson, " SOLID / SAP--A Static Analysis Program for Three Dimensional . Solid Structures," Structures and Materials Research, Report SESMUC/7119, Dept. of Civil Engrg. University of California' at Berkely, September 1971 (revised March 1972) . - 1
- 12. K. J. . Bathe, E. L. ' Wilson, and F. E. Peterson, . "S AP IV--A Structural Analysis Program for Static and Dynamic Response of Linear Systems," Report EERC 7311, Earthquake Engrg. Research Center, University of California at Berkeley, June .1973. -
.O 3.9-62
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 h 13. '"BWR Fuel Channel Mechanical Design and Deflection" N EDE- 2135 4-P, September 197d.
- 14. "BWR/6 Fuel Assembly Evaluation of Combined 'Saf e Shutdown Earthquake (SSE) and Loss of Coolant Accident (LOCA) Loadings,"~
NEDE-21175-P,' November 1976. O. O I l 1 l Q l 1 3.9-63
t LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE'3.9-1 ()L ' PRESSURE ' DIFFERENTIALS ACROSS
. REACTOR VESSEL INTERNALS RESULTS -JFAULT CONDITION.
- Main Steamline Break Reactor; Internals and Core Support Structures -' Reactor at.
Power CONDITION.1 CONDITION 2 (PRESSURE (PRESSURE REACTOR' COMPONENT DIFFERENTIAL psid) DIFFERENTIAL psid) Core Plate and Guide Tube 24.l*- 29. l*: Shroud Support. Ring and Lower-Shroud 45.7* 45.3* t Upper. Shroud 25.2 25.4 Shr3ud Head' 25.9* 25.9* O~. Shroud Head to Water 1 Level, Irreversible op 27.8
~
27.2 Shroud Head to Water Level, Elevation op 1.3* -1.9* Average Power Channel Box (Bulge) 15.2* 14.4*- Average Power Channel Box'
. (Collaps e) None* None* ,
Top Guide 2.0* 2.5*- Steam Dryer ** 6.0* 8.l*'
. CONDITION 1: Power-corresponding to.105% rated steam flow; 100%
rated recirculation flow. CONDITION 2: 66.9 %, rated steam. flow /th'rmal e power;1i10% rated 1 recirculation flow. H Design. Basis Event: Main steamline. break inboard of flow ~re- {]); strictors from _ Conditions 1 and 2, ' except * * ' outboard of flow'restrictors.
*ControlLValue
.. I 3.9-64 .
1
U J %d TABLE 3.9-2 REACTOR PRESSURE VESSEL SUPPORT COMPONENTS ALLOWABLE CALCULATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi) SHROUD SUPPORT ( ASME B and PVC Section III Normal and upset Local membrane 34,950 25,100 (UNIT 1) Primary Local Membrane Plus condition loads: plus bending Primary Bending Limit for SB-168: Later (UNIT 2)
- 1. Dead Weight STRESS CRITERIA / ALLOWABLE 2. Design earthquake (Operating basis Normal and upset condition: earthquake)
S - . = 1.5 x S " = 34,950 psi 1"1 Emergen y condition Local membrane 42,825 42,500 (UNIT 1) loads: plus bending - Emergency condition: Later (UNIT 2) (5) & 7 S limit
= 1.5 x Sy = 42,825 psi- 1. Dead weight $'
e 2. Maximum credible m E (Safe shutdown) $
- Faulted condition: earthquake M . 12, 625 in.-kips (Unit 1)
S 1.5 x S = 42,825 psi Faulted condition Local membrane 12,625(2) 9,597 ,( UNIT 1) limit (Unit 2y 1 ads: plus bending (in.~ kips) (in.-klps)
- 1. Dead weight
- 2. Maximma credible (Safe shutdown) 42,825 psi Later (UNIT 2)(3) earthquake l 3. Jet reaction forces I
- 4. Pressure drop across core support plate and shroud head (1) Unit 1 has gusset supports and Unit 2 has leg supports.
8% eB (2) Allowable based on an evaluation using the critical col) apse moment. @@ (3) The Unit 2 shroud support legs were evaluated for buckling. gg (4) Operability Assurance Demonstration - Required for active components only. 2 (5) For the governing load combination, the faulted, condition loads (which exceed the emergency condition loads) [d were evaluated, and the resulting stresses are compared to the emergency condition allowable stresses. gg (6) The plant unique seismic loads were combined with other design basis loads (weight, pressure, and thermal) and the defined dynamic loads in accordance with the load combinations listed in Table 3.9-25.
,a - ed m
p O V v s TABLE 3.9-2 (Cont'd) ALLOWABLE CALCULATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi) VESSEL SUPPORT SKIRT ASME B and PVC Section'III Primary Stress Limit for SA-516 Grade 70: For normal and upset condition: Normal and upset General membrane 15,477 (UNIT 1) S limit " m = 19,150 psi condition loads: 19,150 Later (UNIT 2}
- 1. Dead weight
- 2. Design earthquake (Operating basis W
earthquake) For emergency condition: Emergency condition General membrane 20,700 (UNIT 1) (3) km m S limit " y = 28,700 psi loads: 28,700 Later l (UNIT 2) (3)
- 1. Dead weight m
- 2. Maximum credible *.
(Safe shutdown) earthquake For faulted condition:( } - Faulted condition General membrane 20,700 (UNIT 1) Slimit " by = 28,700 psi loads: 28,700 Later (UNIT 2)
- 1. Dead weight
- 2. Maximum credible (Safe shutdown) earthquake
- 3. Jet reaction forces o g
oz (1) Operability Assurance Demonstration - Required for active components only. D (2) The vessel support skirt was evaluated for buckling. @$ z (3) For the governing load combination, the faulted condition loads (which exceed the emergency condition r$ loads) were evaluated, and the resulting stresses are compared to the emergercy condition allowable stresses.
- w (4) The plant unique seismic loads were combined with other desigi basis loads (weight, pressure, and thermal) me and the defined dynamic loads in accordance with the load combinations listed in Table 3.9-25.
TABLE 3.9-2 (Cont'd) ALLOWABLE CALCULATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi) , STABILIZER BRACKET - ADJACENT SHELL ASME B and PVC Section III Primary Local Menbrane Plus Primary Bending Limit for SA-533 Grade B,
- Class I
For normal and upset Normal and upset Local membrane 22,500 (UNIT 1) condition: condition loads: plus bending 40,050. S *S m = 40,050. Later (UNIT 2) limit " * ,
- 1. Design earthquake
, (Operating basis earthquake) .
- 2. Design pressure s
For emergency condition: Emergency condition Local membrane- 55,800 (UNIT 1) (2) k C S limit
= 1.5 x S = 63,450. loads: plus bending 63,450 Later (UNIT 2) (2) g' m
- 1. Maximum credible
' (S?fe shutdown) earthquake
- 2. Design pressure For faulted condition:( Faulted condition Local membrane 60,300 (UNIT.1)
S = 1.5 x Sy = 63,450. loads: plus bending 63,450 Later (UNIT'2) limit 4
- 1. Maximum credible (Safe shutdown earthquake
- 2. Jet reaction forces O
- 3. Design pressure I
@f WO z
(1) Operability Assurance Demonstration - Required for active components only. $k (2) For the governing load combination, the faulted condition loads (which exceed the emergency condition gy loads) were evaluated, and the resulting stresses are compared to the emergency condition allowable stresses. * (3) The plant unique seismic loads were combined with other design basis loads (weight, pressure, and- muo thermali and the defined dynamic loads in accordance with the load combinations listed in Table 3 9-25. W _ _~ ~
(~\ p p-y
%) L) L)
TABLE 3.9-3 REACTOR VESSEL AND ASSOCIATED EQUIPMENT ALLOWABLE CALCULATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi) TOP GUIDE - HIGHEST STRESSED BEAM Primary Stress Limit - The allow-able primary membrane stress plus bending stress is based on ASME Boiler and Pressure Vessel Code, Sec. III for Type 304 stainless I. Normal and Upset
- 1. Delta Pressure General membrane 25,388 8,680(2)
- 2. Structure Weight plus bending Stress Intensities 3. Safety Relief Valve Actuation For normal and upset condition: (All)
S . . = 1.5 S = 1.5 x 16,925 w
- p 25,388 p i. II. Emergency Condition ( General membrane 38,081 23,160',I y
, plus bending g /n For emergency condition: 3. Delta Pressure " 2. Structure Weight 4 limit = 1.5 SA S - = 1.5 x 25,388 psi m
- 3. Plant Unique OBE- >
= 38,081 psi. Vertical and #
Horizontal For faulted condition: 4. Safety Relief ' S * * '388 psi Valve Actuation limit " A
= 50,755 psi. IAI III. Faulted Condition II . General membrane 50,775 27,020(2) plus bending
- 1. Delta Pressure
- 2. Structure weight
- 3. Plant Unique SSE o .
Horizontal and
- O Vertical oz
- 4. Safety Relief ED Valve Actuation M (All) 9k (1) Operability Assurance Demonstration - Required for active components only.
Uw (2) The reported stress values are based on the design assessment evaluation loads using the governing load combination for plant operating condition stated. See Table 3.9-25.
1 O O 101 i TABLE 3.9-3 (Cont'd) ALLOWABLE CALCULATED CRITERIA LOADING . PRIMARY STRESS TYPE; STRESS (psi) STRESS (psi) TOP GUIDE - BEAM END CONNECTIONS Primary Stress Limit - ASME Boiler
- and Pressure Vessel Code, Section III, defines material stress limit
.for_ Type 304 stainless steel.
.For normal and upset condition I. Normal and Upset Stress Intensity.
S limit = 0.6 S,= 0.6 x 16,900 1. Delta Pressure Pure Shear - 10,140 .3,420(2)-
- 5. psi = 10*140 psi
- 2. ~ Structure Weight
- 3. Safety Relief
)~ For emergency condition: Y affe. g Actuation limit = 1.5 Sg = 1.5 x'10,140-
- S g~
6 I m psi = 15,210 psi. II. Emergency Condition ( Pure Shear : 15,210 12,930(2) y. e .. .
.m m-For faulted condition: 1. Delta Pressure >
S1 Lait = 2 S3 = 2 x 10,140 psi 2. . Structure Weight , = 20,280 psi. 3.- Plant Unique OBE -
=
Vertical and i Horizontal 4.. Safety Relief Valve Actuation (All) III. Faulted Condition ( } Pure shear' 20,280 ' 15,550(2) 1 1.- Delta Pressure
- 2.- Structure Weight i
- 3. Plant. Unique SSE -
-Horizontal and 4.
Vertical-Safety. Relief gg aM
- Valve Actuation .@$
l -(All) -
.Qg
=
(1) Operability Assurance Demonstration - Required for active components only. $d (2) The reported stress values are based.on the design assessment evaluation loads using the governing load gg combination for plant operating condition stated. 'See Table 3.9-25. 4 4 7 o _ .
, . - , .ry- -' - -
- - e r---- -
ar - - r---w~ >-m=m-._ t____-__-__.2 u_______-._____._________._______.2. _ . . _ _ _ . _ _
f -g c- -.
'% J TABLE 3.9-3 (Cont'd)
ALLOWABLE CALCULATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi) TOP GUIDE ALIGNERS Primary Stress Limit - The allow-able primary membrane stress plus sending stress is based on ASME Boiler and Pressure Vessel Code, Sec. III for Type 304. stainless steel plate. I. Normal and Upset Stress Intensities 1. Delta Pressure General Membrane 25,388 *
- 2. Structure Weight plus bending For normal and upset condition: 3. Safety Relief
= 1.5 x 16,925 Valve Actuation Slimit " 8A IA11I g
psi = 25,388 psi. Q L For emergency condition: II. Emergency Condition General membrane 38,081
- g S
limit = 1.5 Sg = 1.5 x 25,388 psi
- 1. Delta Pressure
= 38,081 psi. 2. Structure Weight
- 3. Plant Unique OBE-For faulted condition: Vertical and S
limit
=2S g = 2 x 25,388 psi Horizontal
= 50,755 psi. 4. Safety Relief Valve Actuation (All)
III. Faulted Condition General membrane 50,775
- plus bending
- 1. Delta Pressure
- 2. Structure Weight
- 3. Plant Unique SSE Horizontal and Vertical Og
- 4. Safety Relief OR Valve Actuation @@
to g (All) "z H s-3 w 24 Wedges eliminates the load on the top guide aligners. These wedges.are installed in the annulus gg between the top guide and shroud and resist the horizontal top guide shear load. (1) Operability A asurance Demonstration - Required for active components only. a
n 9 p TABLE 3.9-3 (Cont'd) ALLOWABLE' CAICUIATED CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi)- ' STRESS (psi)- CORE SUPPORT ALIGNERS Primary Stress Limit - ASME Boiler and Pressure Vessel Code, Section III, define material stress limit
'for Type-304 stainless steel.
For normal and upset condition .I. Normal and Upset-Stress Intensity. . Slimit " 8A = 0.6 S ,= 0.6 x 1. Delta Pressure Pure shear 10,140'
- 16,900 psi = 10,140 psi. 2. Structure Weight
- 3. Safety Relief For emergency condition: Actuation-S * ' = 1.5 x 10,140 limit A _ g --
P psi ='15,210 psi.. II. Emergency' Condition Pure shear 15,210
- W _
Q. b
" For faulted condition: 1.'_' Delta Pressure 5-8 S =2S g = 2 x 10,140 2 Structure Weight-limit
- 3. Plant Unique OBE- '$-
psi =.20,280 psi. Vertical and Horizontal
- 4. -Safety Relief Valve Actuation (All)
III. Faulted Condition Pure shear 20,280 18,975-
- 1. Delta Pressure
- 2. Structure Weight -
. 3. Plant. Unique SSE !' Horizontal and- ! Vertical 4.- Safety Relief O g-Valve Actuation .OM O$- (All)
- 9R-f i
The friction force between core support and core. support flange due to'the preload of the studs is greateri than the shear load induced by the specified dynamic loads. .
.e y
f i,
-(1) .' Operability Assurance Demonstration - Required for active components only.- _
;g;
~
.U..-_
------------_-______.______._____c____. , p - * + , - r o - a -
9 --- -t-m-, e-.. f- , . , _ .
,., g
("T f ) L.) *} TABLE 3.9-3 (Cont'd) CRITERIA ALLOWABLE CALCULATED IDADING PRIMARY STRESS TYPE STRESS (psi) STRESS (psi) CORE SUPPORT - BEAM RIM ( Primary Stress Limit - The allow-able primary membrane stress plus bending stress is based on ASME Boiler and Pressure Vessel Code, Sec. III for Type 304 stainless
. steel plate.
Stress Intensities Normal and Upset I. General membrane 25,388 16,520(2) For normal and upset condition: plus bending
- 1. Delta Pressure S h it = 1.5 S, = 1.5 x 16,925 2. Structure Weight psi = 25,388 psi. 3. Safety Relief Valve Actuation For emergency condition: IA11I limit
= 1.5 Sg = 1.5 x 25,,88 II. Emergency condition (2) General membrane 38,081 18,180(2) psi = 38,081 psi.
y plus bending u>
- 1. Delta Pressure Faulted condition: 2. Structure Weight {
a S . . = 1.5 S = 2 x 25,386 3. Plant Unique OBE- o psi 50,755 p i. Vertical and Y Horizontal y
- 4. Safety Relief >
Valve Actuation (All) III. Faulted Condition General membrane 50,775 18,740 plus bending
- 1. IOCA AP due to steamline break
- 2. Structure Weight
- 3. SSE-Horizontal and Vertical og
- Otu 88 (1) (2) Operability Assurance Demonstration - Required for active components only. "M W The reported stress values are based on the design assessment evaluation loads using the governing z load combination for plant operating condition stated. See Table 3.9-25. gH ww co e
Of O' - C,OI
.s. + w TABLE 3.9-3 (Cont'd)
ALLOWABLE CALCULATED-CRITERIA LOADING PRIMARY STRESS TYPE- STRESS ~ (psi) STRESS-(psil_ CONTROL ROD DRIVE HOUSING Primary Stress' Limit - The allow- Normal and-upset Maximum membrane' 16,660L 13,150(2) able primary membrane stress is condition loads stress intensity based on the ASME Boiler and- occurs at the tube Pressure Vessel Code, Sec. III, 1. Design pressure ~ .to tube weld near the-for Class-1 vessels, for Type- 2. Stuck rod scram. center of the housing: 304 stainless steel. loads for normal, upset,
- 3. Operr_tional basis .and emergency con-For normal'and upset condition: earthquake, with ditions.'
S gg = S ,= 16,660 rsi at housing lateral
.g.
w 575' F Support. installed.
.g au e cn n oa s For faulted condition:
20,000 . U, m w S p__
= 1.2 x 16,660 limit."
- m 7, -Design pressure. f N.'
psi = 20,000 psi. .2. Stuck rod scram loads-
- 3. Safe shutdown earth-quake with housing lateral support
. installed.
(1). Operability Assurance. Demonstration - Required for active components only. (2) The calculated stresses with the dynamic loads included are less than the stress values reported.
-(3) Faulted condition loads were analyzed and compared to emergency condition stress allowables.
kg 39. mo 1
$$:r
'b
- i
+
i .~ . _ . . . .. . , ._ _ _ ._. , _ .: _ . _ __..,u _ _ _ . . . ._ . _ , , ;.
~ % ) )
TABLE 3.9-3 (Cont'd) ALLOWABLE CALCULA2Fu ( } CRITERIA LOADING PRIMARY STRESS TYPE STRESS (psi) STRESS ipeg CONTROL ROD DRIVE Primary Stress Limit - The allow- Normal and upset con- Maximum stress inten- 25,860 20,790 able primary membrane stress plus dition loads Maximum sity occurs at a point bending stress is based on ASME hydraulic pressure on the Y-Y axis of the Boiler and Pressure Vessel Code. from the control rod indicator tube. drive supply pump. NOTE: Accident con-For normal and upset condition: ditions do not in-S A
=
1.5 S" = 1.5 x 17,238 = crease this loading. Dynamic loads are 25,850 psi. ig Ble. n . m y In-Core Housing O $ Primarv Stress Limit - The allow-NM able primary membrane stress is $ based on ASME Boiler and Pressure Vessel Code, Sec. III for Class 1 vessels for Type 304 stainless steel. For normal and upset conditions: S = 16,660 psi at 575' F. Emergency condition Maximum membrane 20,000 Later 1 ads: stress intensity For emergency condition: S limit = 1*2'S~ m
= 1*2 x 1.
2. Design pressure Safe shutdown occurs at the outer surface of the vessel gg gg 16,660 = 20,000 psi. earthquake penetration gg Eh (1) Operability Assurance Demonstration - Required for active components only. g ,] (2) The calculated stresses with the dynamic loads included are less than the stress' values reported. e ww CD W
. - + .} i-
;o; o ;o, -
i. f l' TABLE 3.9-3 (Cont'd) , i. !" . . ALLOWABLE CALCULATED j- CRITERIA LOADING- PRIMARY STRESS TYPEt STRESS (psi) STRESS (psi) '
-CONTROL ROD GUIDE TUBE
~
Primary Stress Limit - The allow-
, able primary membrane stress plus i '
bending' stress is based on the , ASME Boiler and Pressure Vessel -i Code, Sec. III for Type 304: = stainless steel tubing.
'{
i For normal and upset conditions: Slimit " 8 =m 16,925 psi }
. m- 1 O For faulted condition: -Faulted conditior, loads The maximum bending 25,400 L5,701(2) 4 1; u' S . = 1.5 S = 1.5 x 16,295 stress under faulted: $'
- C
! 1. Dead weight- loading conditions
= 00 psi. occurs at.the center,
! 2 Pressure drop.
- across guide tube of the guide tube., ,
due to failure of
- steamline.
- -i
(- (1) Operability Assurance Demonstration' . Required for active components only.. - (2) The Design Assessment Evaluation results with the dynamic. loads. incorporated are less than..the o .
' reported calculated stress values. n ei . O2
- l- tD O !
t=1 *
. :ts tus 2
M *9 1 .e i 4 t.J CD %D i 4 4-- . 4 a K __ . -, , -. - . _ . _ , _ _ _ a, . - - - . . _ _ , . - . .1,.. ._
. ~ - _ ,
en s,s p-s U, U U TABLE 3.9-4 FUEL ASSEMBLY (INCLUDING CIIANNEL) DESIGN BASIS
- COMBIhTD*
LOADING PRIMARY LOAD TYPE ALLOWABLE ACCELERATIONS ACCELERATIONS Normal and upset con- Acceleration profile 1.56g 0.41g dition loads:
- 1. Operating basis earthquake
- 2. Normal pressure load
- 3. Safety relief valve Faulted condition Acceleration profile 3.12g 0.51g load:
- 1. Safe shutdown earthquake
- 2. Normal pressure load &
y 3. LOCA $
. 4. Safety relief ? ?
y valve. M en 8g Uis
*The Design Assessment Evaluation shews that the appropriately combined accelerations are lower than yg design basis accelerations. ,$
U _ - - _ = _ _ _ _
O
~
O 101 TABLE 3.9-5 MAIN STEAM PIPING HIGHEST STRESS INTENSITIES OR STRESS INTENSITY RANGE-CIASS 1.PIPEE LOAD ASME ALLOWABLE COMBINATIONS SECTION III ACTUAL STRESS -' STRESS -STRESS ( CONDITION. NO. EVALUATED ACCEP. CRITERIA OR S.I. (psi) S.I. (psi) RATIO LOCATION LINE A DESIGN 1 . EQ. 9 < l.5 SM 25,172. 27,300' O.92 :SWEEPOLE* AT S/RV - B SERVICE LEVEL A & 2 and 3 EQ. 10 < 3.0 S 69,360 54,600 -SWEEPOLET AT SRV - J'.
. LEVEL B & U < 170 0.31 -1.0 N/AE SWEEPOLET AT SRV - V -
- IF EQ. 10 EXCEEDED en .
E
" EQ. 12 <'3.0 S 13,502 54,600 0.25" SWEEPOLET AT SRV -- J - 4;
,- EQ. 13 3 3.0 g 46,560 54,600 0.85: .SWEEPOLET AT S/RV J g
- =
SERVICE 5 - EQ. 9 < 2.25_S g 33,189' 40,950- -0.81- SWEEPOLET AT S/RV ,B LEVEL C
~~
SERVICE 6 and 10 EQ. 9 < 3.0 Sg 40,930 54,600 0. 75 '. SWEEPOLET AT.S/RV - P.-
. LEVEL D (1) ;See Table 3.9-25 for load combinations.
(2) Stress Ratio = (Actual Stress) divided by (Allowable Stress). _ (3) See-piping diagrams in Sections 3.6-for locations of these points.
-gg
. .$fi k
.5 w
6m . _..__.m. .m_. . .-.- - -U- .
--.r. -,.-y y %' p-- 4 =
9W--'1 m - * * - - - - - - ' - - - - - - - - - - - - - "
- y. <
JO! O: 03 \ t G ,- TABLE 3.9-5 (Cont' d) . t = - LOAD (1) ASME ALLOWABLE $ COMBINATIONS SECTION III ACTUAL STRESS STRESS STRESS (2) , CONDITION. & EVALUATED . ACCEP. CRITERIA OR S.I. (psi) S.I. (psi) -RATIO LOCATION ( LINE B-j DESIGN.: 1 EQ. 9 < l.5 Sg :21,147 27,300'. 0.77- SWEEPOLET AT S/RV - F , 4 . , )- SERVICE ' EQ. 10 < 3.0 S g 69,912 54,600 S N N .AT SRV - S i LEVEL A & 2 and 3 & U < 1. 0 0.51 1.0 N/A ' SWEEPOLET AT SRV - F '
- LEVEL B IF EQ. 10 EXCEEDED .
- EQ. 12 < 3.0 S g 8,888- M,600 0.16' SNOW AT SW - S
' EQ. 13 < 3.0 S3 42,685 54,600 0.78 SWEEPOLET AT S/RV - S j> SERVICE 5 EQ. 9 < 2.25 Sg '.38,446 40,950 0.94 SWEEPOLET AT S/RV - F- .Q
! w LEVEL C o 2
- . T SEmm AT S/W 4 F 39,4% 54,600 0.72-
. co
. SERVICE.
LEVEL D 6 and 10 EQ. 9 < 3.0 Sg _
- :o ,
l
- -(1)- See Table _3.9-25 for load combinations.
(2) Stress Ratio = (Actual Stress) divided by (Allowable Stress). . ,
-(3) _'See piping diagrams in Sections 3.6 for locations of these points.
1-8E
- 3E
' ag
. v i + - V
-u . CD to b
i .
. y -- ..E- , , , - . , - . n * -w7 .uw.-. .i-.. % .s . ,- , c - ~a-:., -
e-.-r--- -- - , y=--- -.,--e- -- -
O O O TABLE 3.9-5'(Cont'd) IDAD ASME ALLOWABLE COMBINATIONS-- SECTION III ACTUAL STRESS STRESS STRESS (2). CONDITION NO. EVALUATED ACCEP. CRITERIA OR S.I. (psi) S.I. (psi) ' RATIO LOCATION-LINE C DESIGN .1 EQ. 9 < 1.5 Sg 22,038 27,300 0.81. SWEEPOLET AT S/RV - R-SERVICE 77,235 54,600 LEVEL A & 2 and 3 EQ.
& U <10 170 < 3.0 S" LEVEL B 0.47 1.0 .N/A IF EQ. 10 EXCEEDED SWEEPOLET AT.S/RV - N
-EQ. 12 <'3.0 S' j 14;136 54,600 0.26~
EQ. 13 ] 3.0 og 41,505 54;600 0.76 w SERVICE: 5 EQ. 9 < 2.25 Sg 39,058 40,950- 0.95 LEVEL C SWEEPOLET AT S/RV - C -E i -Q-a- d SERVICE 6 and 10 EQ. 9 < 3.0 Sg .39,936 54,600 0.73 SWEEPOLET.AT S/RV -lC '$~ LEVEL D
.g (1) See Table 3.9-25 for load combinations.
(2) Stress Ratio = (Actual. Stress) divided- by (Allowable Stress) . (3) See piping diagrams in Sections ~3.6 for locations.of these points. h -. am 25
.g5'.
l l i
. , , -~. ,,c. ~ , -n,- - , - ~ - - . , , . . . , . , - , , --,c, , . , , ~ , .
'N %l .)
TABLE 3,9-5 (Cont'd) 4 LOAD ASME ALLOWABLE gy) COMBINATIONS SECTION III ACTUAL STRESS STRESS STRESS CONDITION NO. EVALUATED ACCEP. CRITERIA OR S.I. (psi) S.I. (psi) RATIO LOCATION l LINE D DE3IGN 1 EQ. 9 < l.5 S g* 22,788 28,725 0.79 LUG NEAR SNUBBER SD - 4 SERVICE EQ. 10 < 3.0 S 73,598 54,600 SWEEPOLET AT SRV - G LEVEL A F 2 and 3 & U < 170 0.38 1.0 N/A SWEEPOLET AT SRV - A LEVEL.B IF EQ. 10 EXCEEDED EQ. 12 < 3.0 S
~
18,405 54,600 0.23 SWEEPOLET AT ERV - G EQ. 13 < 3.0 S'Mg 41,442 54,600 0.76 SWEEPOLET AT S/RV - G Q n y SERVICE 5 EQ. 9 < 2.25 Sg 33,432 40,950 0.82 SWEEPOLET AT S/RV - A Y g LEVEL C .@ SERVICE 6 and 10 EQ. 9 < 3.0 S g 34,816 54,600 0.64 SWEEPOLET AT S/RV - A "~ LEVEL D (1) See Table 3.9-25 for load combinatioas. (2) Stress Ratio = (Actual Stress) divided by (Allowable Stress). (3) See piping diagrams in Sections 3.6 for locations of these points. 88 95
-5 50
_ ___ __________t _ _ % vT y 3 g
Li LQL 0 O - 4 4 - (- 4 TABLE 3.9-6 - RECIRCULATION PIPING SYSTEM
. l-HIGHEST STRESS INTENSITIES OR STRESS INTENSITY RANGE-CLASS 1 PIPE
~ -LOAD ASME ' ALLOWABLE g COMBINATIONS SECTION.III ACTUAL STRESS STRESS STRESS CONDITION & EVALUATED ACCEP. CRITERIA OR S.I. (psi) S.I. (psi) . RATIO LOCATION
. . LOOP A I
. DESIGN 1 EQ. 9 < l.5 S M 24,607 24,975 0.99 RHR TEE AT
- SHUTDOWN SUCTION ,
! SERVICE EQ. 10 < 3.0 S M 79,149' 49,950 . SWEEPOLET AT RE- ! w LEVEL A & 2 and 3 & U < 170 'O.91- -1.0- N/A CIRCULATION DIS . '[- ?c
- LEVEL B IF EQ. 10 EXCEEDED CHARGE RISER (60 .O I" EQ. 12 < 3.0 S 22,391- -49,950 0.45 AZIMUTH) Y-EQ. 13 [< 3.0 S" 36,552 49,950 0.73 L2 .[
SERVICE 5 33,301 '%~ ! l EQ. 9 < 2.25 Sg 37,463 _ 0.89- RHR TEE AT LEVEL C SHUTDOWN SUCTION p 4- 'i SERVICE 6 and 10 EQ. 9 < 3.0 S 44,047 49,950 0.88 RHR TEE AT' LEVEL D SHUTDOWN SUCTION / i i
.9
-(1) See Table 3.9-25 for load combinations. I
'8 i- (2) Stress Ratio = . (Actual Stress) divided by (Allowable Stress).-- H (3) See piping diagrams in Figure 3.6-1 for locations of these points.
_ O$ e5 i' M %s G> W - + _. . p 4 L L e
"---c-
_._m____ _2_ _ _ - _ _ _ _ _ _ _ - _ _ _ - - _ _-_-_____m.*_m . -_1 '* T- __- C- u7" % - 7 ^$- ^giu- t-"-f eM*
L x ' Q O O tq
+
- j. -
.. TABLE 3.9-6 (Cont'd) ' I
' i LOAD (l} ASME ALIDWABLE STRESS I COMBINATIONS SECTION III ACTUAL STRESS STRESS
; CONDITION NO. 1 EVALUATED ACCEP. CRITERIA OR S.I. (psi) S.I. (psi) RATIO . LOCATION LOOP B DESIGN 1 EQ. 9 < l.5 SM 23,754 24,975 0.95 RHR TEE AT
{ _ SHUTDOWN' SUCTION-4 .. } SERVICE 77,276 49,950
-LEVEL A & 2 and 3 EQ.'10
& U < 170 < 3.0 S" 0.95 1.0; -N/A i' LEVEL B IF EQ. 10 EXCEEDED .. RHR TEE AT
; EQ.-12_< 3.0.S M 8,487 49,950 0.17 SHUTDOWN SUCTION EQ. 13 < 3.0 Sg 44,185- 49,950 0.88 .g n
7 -SERVICr 5- EQ. 9 < 2.25 S 30,042 .37,463 0.80. RHR TEE AT ^ Y
- f. .$ LEVEL C ,~ _ SHUTDOWN SUCTION y 1
SERVICE 6 and 10 EQ. 9 < 3.0 S *y 35,287 49,950 0.71 RHR TEE AT *
- LEVEL D SHUTDOWN SUCTION e
L
' (1) See Table 3.9-25 for load combinations.
(2) Stress Ratio = (Actual Stress) divided by (Allowable Stress).- 3G (3) See piping. diagrams in Figure 3.6-2 for locations of these points. .E 1
"r~
d
-p I
a 20 1 e i M 4 _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - - _ . . _ _ - ,-# s. - -- - m a -- . -
- e. . -v -.-,,
O O O TABLE 3.9-7 REACTOR VESSEL SUPPORT EQUIPMEh"r AND CRD HOUSING SUPPORT-ALLOWABLE CALCULATED CRITERIA LOADING LOCATION. STRESS (psi) STRESS (psi) RPV STABILIZER III Primary Stress Limit ATSC specification for the Upset condition Rod P = E4,000 F = 5,900 t t construction, fabrication, 1. Spring preload .4 and erection of structural 2. Operating basis dracket F = Fb == 15,300 F b = . 14,000 22,000 Fy 6,460 steel for buildings. earthquake: For normal and upset condi-tions AISC allowable stresses, but without the usual increase for earthquake loads. For emergency conditions: Emergency condition Eracket- F b = 33,000' F 18,300 [ b o I' S limit = 1.5 x AISC 1. Spring preload e allowable stresses. 2. Safe shutdown F = 213000 Y-m earthquake Rod F[.126,000 F[== 7,710 F 66,300 F 36,000 F = 20,300
- For faulted conditions: Faulted condition Bracket b b S limit = Material yield strength 1. Spring preload
- 2. Safe shutdown earthquake .
F = 21,500 F - = 8,540
- 3. Jet reaction, load _ Rod F 140,000 F = 73,900 t=
RE aa to o (1) operability Assurance Demonstration - Required for active components only. ' E% II The design assessment evaluation stresses with the dynamic loads included are less than the 5 reported stresses. $$ I 'Where Fb = bending, Fy = shear, and F = tension. t
.m -_.----._.~____--e__._-w___.._,_--.--__,_om_-_e-m_,- e_gm-_.- e _
- - - - - -_.m
_ ~
--_-g=__m-es_-.__m_3_._._,___------._em-%-w __
__ .-mw,
) %
TABLE 3.9-7 (Cont'd) ALLOWABLE CALCULATED ' CRITERIA LOADING LOCATION STRESS (psi) STRESS (psi) CRD HOUSING SUPPORT RESTRAINT BEAMIl} . Primary Stress Limit AISC specification for the design, fabrication and erection of structural steel for buildings. For normal and upset conditions: (Dead weights and Fa = 0.60 Fy (tension) earthquake loads are Fb = 0.60 Fy (bending) very small as compared Fy = 0.40 Fy (shear) to jet force.) tn e For faulted condi'. ions: Faulted condition loads Beams (top chord) F = 33,000 F = 12,200 0 Fa limit = 1.5 Fa (tension) 1. Dead weight F# 33,000 F 8 = 16,500 E. FL limit.= 1.5 Fb (bending) 2. Impact force from Beams (bottom chord)Pb == 33,000 F h = 10,300 0' Fy limit = 1.5 Fv (shear) a failure of a CRD F = 33,000 b F*b = 11,700 $ Fy = Material yield strength housing Grid structure 'F = 41,500 P 40,700 Fy = 27,500 Fb == 11,100 0 0 OZ (1) Operability Assurance Jemonstration - required for active components only. tu ry gg (2)The design assessment evaluution stresses with the dynamic loads included are less than the
- reported stresses. 7w
__ _ _._.__ ____ _._m_._ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _-
,-) (s (- ( (1 - TABLE 3.9-7 (Cont'd) (3) ALLOWABLE CALCULATED (2) l CRITERIA LOADING LOCATION STRESS (psi ) STRESS (psi) RPV SUPPORT (RING GIRDER)'- UNIT 1 Ill Primary Stress Limit AISC specification for the Normal and upset condition Top Flange b = 22,'000 P b = 10,000 design, fabrication and 1. Dead loads erection of structural 2. Operating basis earth-steel for buildings. For quake Bottom Flange b= 22,000 F b = 10,000 normal and upset conditions 3. Loads due to scram Vessel to girder F t= 54,000 F t = 35,200 AISC allowable stresses, bolts y a 14,000 Fy = 4,450 but without the usual in-crease for earthquake loads. r* w For emergency conditions: Emergency condition y p y
, S limit = 1.5 x AISC 1. Dead loads' Top Flange b = 33,000 b = 22,000 m i allowable stresses. 2. Design basis earth- p p 4 4 g quake Bot; tom Flange b= 33,000 b = 20,000 to
- 3. Loads due to scram Vessel to girder p 7 $
bolts p t = 81,000 pt == 70,400 v = 21,000 y 8,900 For faulted conditicns: Faulted condition Top Flange b = 36,800 F b = 28,000 S limit = 1.67 x AISC allowable stresses for structural steel members. 1. Dead loads S limit = Yield strenght - 2. Design basis earth- p quake Bottom Flange b= 36,800 P b = 23,400 for high strength bolts (vessels to ring girder) . 3. Jet reaction load Vessel to girderp F t = 94,000 bolts p=125,000 y = 72,000 F = 9,650 y bk en cr o (1) Operability Assurance Demonstration - required for active components only. gg (2)The design assessment evaluation stresses with the dynamic loads included are less than the 0 4:r reported stresses. g Where: P b= bending, F y
= shear, and F = tension.
- - . - - - - - _, -- - a - - _ - - a- -.-- a. - . , , - - - , _ . - - - - - - ~ - - -- , , . _ . __s --a - - - - - , + - - - - - ----,u-
~3 -
() d - C-TABLE 3.9-7 (Cont'd) I ALLOWABLE CALCULATED CRITERIA LOADING LOCATIOP STRESS ( psi) STRESS (Psi) RPV SUPPORT (BEARING PLATE)- UNIT 2 III Primary Stress Limit AISC specification for the Normal and upset condition Bearing plate 22,000 design, fabrication and 1. Dead loads erection of structural 2. Operating basis earth-steel for buildings. . quake
- 3. Loads due to scram
~
Fc; normal and upset conditions AISC allowable stresses, but without the ; usual increase for earth-quake loads. & y For emergency conditions: Emergency condition Bearing plate 33,000 y c) S limit = 1.5 x AISC 1. Dead loads 4,m allowable stresses. 2. Safe shutdown earth- y, quake z
- 3. Loads due to scram For faulted conditions: Faulted condition Bearing plate 36,000 S limit = 1.67 x AISC 1. Dead loads allowable stresses for 2. Safe shutdown structural steel members 3. Jet reaction load a
85 i de ll) Operability Assurance Demonstration'- Required for active components only. NN "Y t;" sa
., . _ e.- - .. .. . _ . . - _ , - .- ,_en - - - , , - . - ~ . ,
7,- (- (/ (- ) ( V. TABLE 3.9-8 g g MAIN STEAM SAFETY / RELIEF VALVES (1) (2) CALCULATED STRESS (psi) ALLOWABLE STRESS OR ACTUAL OR MINIMUM THICKNESS THICKNESS OR. AREA CRITERIA METHOD OF ANALYSIS REQUIRED (in or in 2) Unless otherwise specified, all references are to ASME B&PV, Section III (July 1971).
- 1. Body Inlet and Based on USAS B31.7-1969, Para. 1-704.5.1 Outlet Flange and ASME B&PV Criterion.
Stresses 3H _ fMc + PbB < l . 5 Sm L gcB 491 gR _ (4Btc+1)Mo < l.5 S m Lt2J g , ][Mo - Z SR < l.5 Sm ta2 w b o Where:
. ?
4 Sg = Longitudinal " Hub" Wall Stress, psi y SR = Radial " Flange" (Body Base, Inlet) Stress, psi ST = Tangential " Flange" Stress, psi For Inlet: As SH = 21,390 psi < 27,300 psi (1.5 Smc 27,300 psi 21,390 575'F of A-105 GI) SR= 5,140 psi < 27,300 psi (1.5 Sm) 27,300 psi 5,140 ST = 25,100 psi < 27,300 psi (1.5 Sm) 27,300 psi 25,100 Criterion Satisfied' For Outlet: b$ e to As SH= 15,790 psi < 27,300 psi (1.5 S m) 27,300 psi 15,790 @@ SR = 19,120 psi < 27,300' psi (1.5 Sm) '27,300 psi 19,120 Edz-ST= 5,030 psi < 27,300 psi (1.5 S m) 27,300 psi 5,030 (1) The main steam safety / relief valves were evaluated using the dynamic loads unique to the LSCS project. This evaluation was perforaned as an integral part of the main steam piping analysis. (2) Operability assurance was demonstrated by analysis based on allowable limits established by test see page 3.9-93 of this table. ,
_.p p -g b V TABLE 3.9-B (Cont'd) CALCULATED STRESS (psi) ALLOWADLE STRESS OR ACTUAL OR MINIMUM THICKNESS THICKNESS OR AREA CRITERIA METHOD OF ANALYSIS REQUIRED (in or in2)
- 2. Inlet and Based on USAS B31.7-1969, Para 1-704.5.1 Outlet Stud ASME B&PV Criteria.
Area Require- Total cross-sectional area of bolt shall ments equal or exceed the greater of: AMI = Wdm or AM2 = w]m SBo SBA Where: i Am1 = The total required bolt (stud) area for operating condition, in2 Am2 = Total required bolt (stud) area g for gasket condition, in2 rn . 7 O y For Inlet: 4
$ Am1 = 9.060 in2 > Am2 - 0.807 in2 9.060 in2 14.78.in2 p
- o Am1 = 9.060 in2 > 14.78 in2 (Am, actual stud area) Criterion Satisfied l For Outlet:
Am1 = 4.76 in2 < Am2 = 1.55 in2 Am1 = 4.76 in2 < 9.68 in2 (Am, Actual 4.76 in2 9.68 in2 stud area) Criterion Satisfied
- 3. Nozzle Wall Based on para. NB-3530, NB-3540 and NB-3550 Thickness for nonstandard primary pressure rating
- 1. Valve wall thickness criterion Where: t min > tA Of-
@g t min. = Minimum Calculated Required N Thic? mess With Corrosicn *z Allowable, in. y*i tA = Actual Nozzle Wall Thickness, in. $$
/* ~. * :
;\ -
.. }V; TABLE 3.9-8 (cont'd)
' ~ CALCULATED STRESS '(psi) ALLOWABLE STRESS . OR ACTUAL , OR MINIMUM THICKNESS' THICKNESS OR AREA CRTTERIA METHOD OF ANALYSIS REQUIRED (in or in 2)
- 1. For thinnest section near valve seat' tmin. = 0.467 in. < 0.588 in. 'O.467 in. .0'.582 in. 'l (ta, actual)
- 2. For section about mid-section of ' nozzle tmin. = 0. 46 8 in. < 0.625 in. 0.468'in. 0.625'in. l.
(ta, actual)
, 2. Cycle Rating (NB-3550)
I< .0 (thermal It = 0.032
- 1. 'The. thermal cyclic index criterion t cyclic index . . .
It= g 1 2 + EEM u Ni N1 N2 N3 - not a' stress) [-
'n W - '
4 a
-g:
- 2. The fatigue requirement Criterion ~ 'g-N > 2,000 cycles' N> 106 -.
Spl = 2.Op + Peb + Ot2 + l 3 Otl - (m n mum a que cy es-3 "I~ requirement-. . . I Sp2.'= 0.4 Op + K (Peb + 2_Ot2) not'a stress) 2 l Where: *
- Sp1 = The fatigue stress-intensity.at the inside surface of the crotch, psi-Spy = The fatigue stress intensity at the outside surface of the crotch, psi 8@f am Sp1 = 9130 psi > Sp2 = 545 psi
.O$:
mz mM The permissible number of startup/ shutdown cycles y .] ww. ' me (1) Refer to Table 3.9-33 for thermal transient information. t s y $ - w 4 -W -
O O O TABLE 3.9-8 (Cont'd) CALCULATED STRESS (psi) ALLOWABLE STRESS OR ACTUAL OR MINIMUM THICKNESS THICKNESS OR AREA (in or in 2) CRITERIA METEIOD OF ANALYSIS kEQUIRED
- 4. -Bonnet Based on' Nuclear Power Piping Code USAS B31.7-1969, Para. 1-704.5
- 1. Body to Bonnet Stud Stress Criterion.
Total cross-sectional area of bolt shall equal or exceed the greater of: Am1 = W M or Am2 = wmo SRo SRA Aml = 5.19 in2 > Am2 = 0.549 in2 As Ami = 5.19 in2 < 7.26 in2 (Am, actual stud areal Criterion Satisfied 5.19 in 2 7.2G.in 2 p Y $ e 2. Bonnet flange strength (B31.7-1969, Para. m i 1-704.5.1) Criterion 4 Sg < 1.5 Sm sW SR< l.5 S m ST - 1.5 Sm As SH= 28,400 psi < 29,100 psi (15 Smc 29,100 psi 28,400 psi 500*F of ASTM A105 GI) SR= 2,530 psi < 29,100 psi (1. 5 Sm) 29,100 psi 8,530 psi ST= 8,860 psi <'29,100 psi (1.5.Sm) 29,100 psi 8,860 psi S. DISC Insert Bending stress of disc insert per ?oark's - formulas for stress and strain, 4th Edition, pp. 222-223 auperposition of Case no. 21 and Case no. 22 for flat plates. _ St-Sr + Sr = 3Wi 4a4(m+1)ln a/b-a4(m+3)+b4(m-1)+4a2b2) $ 4t AVG. _ a 2 (m+1) + b2 (m-1) . m
-e
+3WL 232(m+1)1n a/b+a2(m-1)-b2(m-1) $lm 3
lit 4 AVG. ~ aZ(m+1)+b2(m-1) ~
'mw
~'T'9 _ - _ __ _ ___ _ ___ _ _ _ _ - _ _ _ _ _ _ - _ _
r O O O 1 i TABLE 3.9-8 (Cont'd) ALLOWABLE STRESS CALCULATED STRESS OR MINIMUM THICKNESS OR ACTUAL CRITERIA METHOD OF ANALYSIS REQUIRED THICKNESS (psi) Where: Wy = Pressure, psi W2 = Seat Load Lbs. a = 1/2 Disc Insert Outside Dia., 2.72.in. b = Hub Radius 0.592 in. tA vo = Average Thickness, 1.022 in. Sri = Stress Due to Pressure, psi
~
Sr2 = Stress Due to Seat Load, psi e m F 1r = Reciprocal of poisson ratio, 3.333 O I E g 1. The normal operating condition o2 80% of design y pressure: z ST = 32,864 psi < 43,500 psi (Smc 600'F of 43,500 psi 32,864 psi ASTM A461 Type 630) Criterion Satisfied Wnere: W1 = 1000 psi W2 = 5422 lb. , 2. The maximum possible full flow pressure at 10% above design pressure (the stress is entirely due to pressure): Og_ ST = 33,354 psi < 43,500 psi (Smi Criterion 43,500 psi _ -33,354 psi Q@ Satisfied @y. Where: $ W1 =-1375 psi _ [H W2=0 $$ y + y w 4o3%,, 3 , - - - ,. mq + . , c--3 ' + ---w-f,
TABLE 3.9-8 (Cont ' d)
- ' AILOWABLE STRESS ~ -CALCULATED STRESS OR MINIMUM THICKNESS- OR ACTUAL-3 CRITERIA METHOD OF ANALYSIS REQUIRED- THICKNESS (psi) 3 .' The set spring load, zero inlet pressure, load temperature (The stress is entirely
[ due to seat load, zero inlet pressure.) 4 St= 43,038 psi < 46,700 psi (Sm @ Room 46,700 psi 43,038 psi
' Temperature of ASTM A461 Type 630)-
Where: , W1 = -. O W2= 27.110 psi . .[ w .n-
. (A i- ;g g.
w a t 4 a O 1 i
'hOZ h
4 we z P >3 e CD w
- < ~ .-. , ,
1-
1-TABLE - 3. 9-8 (Cont'd) 1 Discharge Safety / Relief Valve' Operability Assurance LOAD GOVERNING . - QUALI-COMBINATIONS -LOAD FICATION LOAD EVALUATED ALLOWABLE- CALCULATED -. C O M B I N A T I O N .- METHOD t Inlet Flange - l ', 2,n3,.4, 8 x 10 5 in-lb !559,'338 in-lb - 6 : Test Moment 5, 6, : 8' static plus 3.85 x 105 in-lb dynamic 'i l Outlet 1, 2 , '3 , 4, 6 x'10 5 in-lb 459,910 in-lb -6 Test , Flange 5,'6, 8- static _ j Moment tv , en w Horizontal 1, 2, 3, 4, 5.0g. -3.46g 6 . Test- . g-e Accelera- -5, 6, 8 i tion ' @l w >- i Vertical 1, 2, 3, - 4 , 4.2g 2.92g: 6' Test' -*
- . Accelera- 5, 6,
- 8 i tion L
i O$, t a to . (1) See Table 3.9-25. - Z-2 -(2) The safety / relief valve' operability.'is required for theiload combinations- ' @ E'-
~. to l evaluated. . Instant depressurization occurs for: load combinations 7 and 9. #$
.i
. ,H e MW 00 W i
.-. - . .. ~_ _ - _ . . . . . _ . - . _ _ . , - - , , - - - - - , - - , - . .n -
-l
J 1 TABLE 3.9-9 MAIN STEAM ISOLATION VALVE I} CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS All references are made to ASME B&PV Code Section III Nuclear Power Plant Ccmponents, 1971 ed as addended by summer 1971 and winter 1971, unless otherwise specified. Reference the same code for explanation of the symbols used.
- 1. . Body Minimum Wall Reference Paragraph NB-3543 Non-Standard Thickness Pressure Vented Valve, Table NB-3542-1 For design conditions of 1250 psig and 4
, 575* F the primary rating = 495 g O
Based on a core diameter of 23.90 inches y
^
, tm = 1.58 inches .l.~58 in. 2.12 in. l $ (including 0.120 inch for corrosion M allowance)
- l. Body Shape Rules Reference Paragraph NB-3544 2.1 Radius of Crotch Criterion r2>.3t,
- Where r 2= 1.0. inch t,- 1.58 inches a s 1. 0 > .3 x 1. 58 = . 4 7 criterion satisfied 2.2 Corner Radii on Internal Reference Paragraph NB-3544.1 (b)
Surface og Criterion r 4 <-r y g eo M3 Where r = 0.69 inch; r = 1.0 inch 2g 4 2 wn as 0.69 < l.0 criterion satisF _d - mw
*L 2.3 Ovality Referenced Paragraph 3544.5, Figure NB-3545.1-2 Criterion b + 3 (3b -2 2ab-a2)
,y< 1.5 Sm t 4 t 2 P b b s (1) The main steam isolation valves.were evaluated using the dynamic loads _ unique to the LSCS project.
This evaluation was performed as an integral part of the main steam piping analysis. (2) . Operability assurance was demonstrated by analysis. See page 3.9-95 of this table.
g. g,- i _() ^QJ rY 4 TABLE 3.9-9 (Cont'd) CALCULATED ALLOWABLE STRESS STRESS OR
-OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM '1HICKNESS THICKNESS Where a = 10.20 in. , b = 15.75,
., t = 4.13 in., S, = 19,400 psi at 5h0* P for ASME SA 216 WCB (Reference Table A)
- As 18.83 < 21.56 criterion' satisfied 2.4 Longitudinal Curvature Reference Paragraph NB-3544.6
, Criterion 1 +1 >4 r long r lat. Wm 4 Where r iong
= 35.31 in., r ,= 15.75 in.,
t u d,= 23.90 in. _
- en As-0.09 > 0.06 criterion satisfied 4
en
' 2. 5 Flat Wall Limitation Reference Paragraph NB-3544.7 $ ,
d_ 1 3, dm t 2 rm i Where d = 23.90 in., t, = 1.58 in., d = 35.76 in., t = 4.06 in. As 8.81 < 22.69 criterion satisfied 2.6 Minimum Wall at Weld End Reference Paragraph NB-3544.8 1.58 in. 3.80 in. Actual thickness at 1 x 1m (i.e. 1.58 in.)- measured along the'run. direction is 3.80 in. 8g eH
- 3. Primary Crotch Stress Due to Reference Paragraph NB-3545.1' @$
Interr.al Pressure yy 7. Criterion: P m= (Al + 0.5) P s.
<S
- m. Ud qw
_ m ** 4 h ~ . - __~
O O O TABLE 3.9-9 (Cont'd) CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL ' CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS Where A g = 591.8 in.2 A = 128.4 in. , 19,400 psi 6897 psi P,= 1350 psi, S,= 19,400 psi l P,= 6897 psi As 6897 < 19,400 criterion satisfied
- 4. Valve Body Secondary Stress 4.1 Primary Plus Secondary Reference Paragraph Stress Due to Internal
' Pressure- Op =c ( + 0.5). P,
. e o e m q
A Where C =3, r.i = 11.80, P s= 1350, te' = 3.80 4 cn p m Q = 14,601 psi-l For wye type valve O' = C, O p Whe re C, = 1. 3 3 i O'p = 19,420 psi 4.2 Secondary Stress'Due to Reference Paragraph NB-3545.2 (b), Pipe Reactor Figures NB-3545.2-3, NB-3545.2-5, NB-3545.2-6 4.2.1 Direct or Axial Load.Effect 8 B . P =FS d o~ ed G d N Where S = 41,000 psi, F d = 345 in.2 , ~ e G " ^ 4493 psi 29,100 psi 4493 psi $$ d" ed - b
. . - ._ ,,w. -, y -- _
._3.~, _,
O O O TABLE 3.9-9 (Cont'3) s CALCULATED ALLOWABLE STRESS -STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS 4.2.2 Bending Load Effect P
- S eb b b Where bF = 380 in.3 , ID = 23.65, r g=
11.80 in., te = 3.80, r = 13.70 in. S = 41,000 Sincerte = 0.257 > .19 ' Cb"1 G b
= I ; I = 30,453 in.4 , rg =-11.83 in.,
r g+te b-w T,= 3.80 in. +
- e M
,L G b
= 1948 in.3 +
g w-x- P = 7998 psi 29,100 psi ' 7,998 psi. eb 4.2.3 Torsion Load Effect Reference Paragraphs NB-3'5 45.2(b) (1) , 3545. 2 (b) (6) (c) P = 2FSb Vt Where Fb = 380 in , G = C t A Where E = 3.43 in., t A = 596 in2, S = 41,000 C = 2 (2 + 3t )/ (1 + t,) (1 + af) a ag by gg f E is - Where t, = 2.45 in., gy e
. ay = 9.25 in.a 6 1 =.8.8 in. MM
- r. ,, -
- h. -w-- - --m- .++ w- .
-Y, _a #-g w y _ _______u- --
,y
f w, .g ,, TABLE 3. 9-9 (Cont'd) CALCULATED ALLOWABLE' STRESS STRESS OR OR ACTUAL CRITERIA MET!!OD OF ANALYSIS MINIMUM THICKNESS THICKNESS C t = 1.78 in. => G = 3E39.in 29,100 psi '8,563 psi
=> P, = 8563 psi 4.3 Thermal Secondary Stress at Reference Paragraph NB-3545.2(c),
Crotch Region Figure NB-3545.2(c)-2, NB-?545.2 (c) (2) , NB-3 54 5. 2 (c)-3, NB-354 5.2 (c )-4 d Te y= 5.20 in., Q 3200, T1 O T2 = C62 C AT2 C 2 = 0.48.'AT 2 = 1.6* F, C 6- 210
> => QT2 = 161 psi e m m QT =OTi + OT2 = 3361 psi 4 m
Criterion S n *O p. ed+ T2 1 m Where O ' = 19,420, P =
,OT2 = 161' 58,200 psi 24,235 psi ed as 24235 < 58,200
=> criterion satisfied
- 5. Normal Duty Valve Fatigue Reference Paragraphs NB-3545.3, Requirements NB-3545.3(a), NB-3545.3a, Figure 1.9-1.
Criterion N > 2000 cycles S =2 Op +Peb + OT3 + *
~
py 3 Tl @ 2- e - oz
- OC
'S * ~4 0 + 20T3I p2 *h(P b.
Sf 4 Where Q* = 19,420, P eb
= 7,998, k = 2, gg Q T1 = 3200, 03T = 175 psi k
TABLE 3.9-9 (Cont'd) CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICFNESS
=> S = 21,267 psi, S = 16,116 py 2 S, equal to the larger of S py
& S o = 21,267 p2
=> N = 7N .tKi cycles > 2000 criterion satisfied
- 6. Cyclic Loading Requirements Reference Paragraph NB-3550 at Valve Crotch For the largest temperature change range a CriterionQg+ Ped + C C C624 AT g max 1 35, Where Q' = 19,420 psi, P ed " '
m C = 210 at AT max g of 342* F,'C = .48, 6 2 e, C 4= 15, S = 19,400 m E o
* => 29084 1 58,200 criterion satisfied 58,200 29,084 y Thermal transient not excluded by code g N . y~
Criterion I Eh N
<1 M i
Calculate the usage factor (I ) as follows: Sg max = O ' + Peb + C C C634 ATpmax = > S n max = 33,343 psi. . Since S . max < 35, (=58,200), the following equation is used: Sg = 0 O'+Peb + C6 (C C34+C5 fi ri o ri O fi i i NL Ogm oz wa 120 90 23,000 56,808 0.0050 gg' z 8 342 2,100 120,973 0.0038 Gd 10 122 18,000 64,956 0.0005 a- --wu . . ,,+--~-- w m -na.w- .sn- msu - ---- ae-a- .- >-m-- -e-m - - - --.=m- , . -n -
- --,,au-- cu,,s.-----.--..~ - air,a --.-r- s ---e-. .-s--m- - =-,a
r- cg ,. s (
,3 I 4 (,. i ml %/ )
TABLE 3.9-9 (Cont'd) CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS N . as I = I [ 1= .0093 < 1 criterion satisfied
- 7. Disk
Reference:
Paragraph NB-3546.3, Table I-l.1, Roark, 4th Edition, pp.220 and 222. Disk design conditions: P = 1350 psi @ 500* F, Sm = 20,800 psi at*500* F Case number 13: 5=3Wn 4mt2+ (a2-b2 ) Where Q = a 4 (3m+1) +b 4 (m-1)-4ma22 b -4 (m+1) a2b21n (g) f e
- Where W = 1350 psi, m =3.33, t= 1 in. $
o Y o a= 11.93, b = 2.28 y S t
= 424,848 psi Case number 14: S t =3W e 2T 2 Where e = 2a2(m+1) In" g + (m-1) a -b W= 61,072, S = 150,633 Hence the required Tg=
m where c13 = 424,828,c14 = 150,633, S, = 20,800 = >T R
= 5.26 in. O T=T R
+ 2 x .120 = 5.50 in. ORO 2:
WD ( 120 is the corrosion allowance for each surface) $ we Case No. 21 5.50 in. 6.60 in. co e 3 _ 3WB 4t _ _ _ _ _ _ . - . _ __- _ m_#m.__--m_-. ____mua._ . . , _ .m.,m. ~ . _ _ ___.__.u____,._ . ___a _ m __-m__._,,,,.mm- m , .__.-__-.-._.m _ _ m , . _ . , _ _ _ _ _ - , _ m gg _g , - --_3
' O, .
O,,, . . V TABLE 3. 9-9 ' (Cont ' d)
- CAICULATED
. ALLOWABLE STRESS STRESS OR
-OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS-Where S = 4a4 (m+1) ln"- - a4(m+3) + b (m+1) + 4a b 2 a (m+1) +b (m-1) w = 1350 psi, m = 3.33, a = 11.93, b = 9.88 t = 1.80 in.
=>S = 2643 psi r
Case No. 22 _ .- Sr = 3W 2a2(m+1) In*g + a2(m-1) -b 2g,_y)
^
2 wt2 _ a 2 (m+1) +b 2g ,_y) _
~
o m
"* Where W = 448,998 lbs, m ='3.33, t= 2.67, a = 11.93, b = 9.88 h'
n , S = 11,996 r 4 S =S (Case 21) + Sr (Case 22) = 14,638 psi S < S ,= 20,800 psi .20,800 psi 14,638 psi Shear at inner edge disk shear = F = 61072'lbs =L623 psi S 12,480. psi
^ 623 psi A 98.08 ind Shear at seat bore S
shear = F = 620458_lbs = 4630 psi g A 134 Ind o Allowable shear = .6Sm = 12,480 > 4630 12,480 psi 4630 psi 5 Hub tensile stress xM g, $ e S = F = 440813 lbs.= 2276 psi . 5% ___ A 193.7 in' s v.
~
O '%) \,.,) TABLE 3.9-9 (Cont'd) CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA MET!!OD OF ANALYSIS MINIMUM THICKNESS THICKNESS Allowable tensile stress 20,800 psi 2276 psi S,= 20,800 > 2276
- 8. Stem Disk
Reference:
Roark 4th Edition, P. 216, Table I-1.1 8.1 Tensile and Shear Design conditions 1350 psi at 500* F, allowable stress 20,800 psi Case No. I w S =S = 3W (3m+1)
. r t __p y 8nmt2 g s
O O W = PA = 26,085 lb, m = 3.33, t= 1.0 Y P g S = 10,285 psi > w Case No. 3 2 h(m-1) + (m+1) In *o - (m-1) 3W #O g ,g r t ,2nnt2 2aI
~
W= 35,210 lb., a = 2.48, r = 0.94, m = 3.33 g S = 26,190 psi t R= 1 + 3 where 01 = 10,285, c2= 26,190, Sm = 20,800 i Sm OD e Fi t = 1.32 on trequired = :r + 2 x .12 = 1.56 in. $8 (corrosion allowance is 0.12 air surface) 1.56 in. 1.85 in. "5 Shear stress above seat g4 uw mw S shen ,Ps = 292010.65m = 12,480 psi 12,480 psi 2920 psi A where Fg = 65,535 psi A= 22.2
4 d%
, t
- +
r TABLE 3.9-9 (Cont'd) CALCULATED ALLORABLE STRESS STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS . MINIMUM THICKNESS THICKNESS 8.2 Thread Strength
Reference:
Federal Thread Standard Part No. 1, p.5 (1957 Ed.) 1-7/8 - 12UN-2 thd
'l -
ASn = 3.1416nL eD smin + 0.57735 (D smin ~ n max Where AS n = shear area of internal thread L e = minimum length of. engagement required = 1.62 in. E &- n max = maximum pitch diameter of internal y ' thread = 1.8287 - < i . m. g Dsmin = minimum major diameter of external m w thread = 1.8613 y j' n = thread per inch = 12 m ASn = 6.91 in 2 x Allowable shear stress =~0.6S,= 12,480 S = [--Sn = 5096 psi; where F = 35210 lbs. 12,480 psi- 5,096 psi Design-
- 9. Piston Design Conditions: 1350 psig at 500* F, S ,=_19.400 psi 4
Ultimate Tensile Stress = 70,000 psi 5 9.1 Thread Strength Thread Shear Area AS =YEL e where E = 12.3125 in, L = 2.38 in. oB e Qg-l 2~ Oo ASn = 46.03 in.2 "*
"E.
U When F = 449,197 S shear =_F = 97 59 < 0. 6S, = 11,64 0 ~ 11,640 psi' 9,759 psi AS _., n
+"rve*W"W'
,~
/~x ym TABLE 3.9-9 (Cont'd)
CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS 9.2 Hoop Stress
Reference:
Roark 4th Edition, p. 308, Case No. 34 2b2 S=P b2 -a'2 b = 10.25 in. (minimum allow-Where a' =
- h (2b ) able inside radius) b = external radius = 11,046 in. , P = 19,400 S = 19,400 T = b - a = 0.796 in., where R
T = 0.796 + (2 x .120) = 1.04 in. (corrosion allowance is 0.120 in. per surface) 1.04 in. 1.486 in. E w 9.3 Tensile Stress T actual = 1.436 in. Q
- at Thread
, F kta g Relief Sm
- ZX-t >
a 5 s Where F = piston load = 449,424 lb. tiAt = w/4 (12.162 -c,2), g = 19,400 lbs+ c' = 10.88 in. m TR" 2
= .64 in + t = 0.64 + 2 x .12 = 0. 8 8 in .
(corrosion allowance is 1.205 0.88 0.12 per surface)
- 10. Bonnet Design
Reference:
Para. NB-3 6 4 7. l (a ) , Para. UG-34 (k) (2) of Calculations Section VIII, Div. 1, 1971 edition.
+ E O to Pfd " eq OZ D7 0 Where P 16M 4F eq " 7wG+ -G7, M = 834,415 in-lb, FG==39,400,1bs, gg w 24,72 in. z P 364 psi + P fd = 1714 Psi gg eq 4
- n, p ,- ~.
! l t 1 i j v v v TABLE 3. 9- 9 (Cont 'd ) CALCULATFE ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA METHOD OF ANALYSIS MINIMU*4 THICKNESS THICKNESS Minimum thickness t = d CP 1.78Wh 4 R 1-+ YS Sd3 Where C = 0.3, P = 1714, S = 19,400, hg = 3.05, d = 24.72 W= 1,077,640
= 5.329 + 0.12 = 5.45 5.45 in. 8.56 in.
tR = 5.329 = >t 10.2 Reinforcement
Reference:
Paragraph UG-39 (d) (2) of Section VIII, Div. 1, 1971 edition t= (/J)d CP
-+
1.78Whg = 7.54 in. S Sd3 With a corrosion allowance of 0.120 in. Total thickness = 7.66 7.66 in. 8.56 in.
- 11. Bonnet Stud Design
Reference:
Paragraph NB-3232.1 and Article E-1000 ta w Bolt used- 24 pieces of 1-5/8 - 8 UN bolts, $ Total bolt area = 42.72 in.2 y a m 5 11.1 Normal Operation 1. Pressure stress at operating condition $
- o vi S1 = Wml = 25226 psi; where ml = 1,077,640 lbs,'Ab = 42.72 in.2 Ab
- 2. Gasket load at ambient condition with no internal pressure S2=Wm2 = 2616 psi; where Wm2 = 111,774 lbF, A b " 42*72 i"*
Ab Maximum tensile stress = 25,226 psi Thermal stress = 0 as the coefficients of 8D thermal expansion of stud Qh and bonnet material are to o the same. E@ The higher load is the standard preload y e of 45,000 psi.which is less than the allowable of 69,400 psi 69,400 psi 45,000 osi dd m 7
+
g _. ,- V T TABLE 3.9-9 (Cont'd) CALCULATED ALLOWABLE STRESS STRESS OR OR ACTUAL CRITERIA- METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS
- 12. Body Flange Design
Reference:
Paragraph NB-3647.1 and Section VIII Div.1 of Pressure Vessel Code - Total flange moment under operating conditions Mo=MD+MG+MT MD -Hp hD HD= .785 B2P, hD = R + 0.5g1 = 1.94 Where B = 24.14 in., P.= 1714 psi,
=>MD = 1,521,096 in-lbs MG=Hg gh Hg = W -H, h = C -G U 2 Where W is the higher of WM1 & WM2 n
b W'1 g = .785 G2P + (2b x 3.14 GMp) = 1,077,640 lbs 8 S Y WM2 = 3.14 Gby = 111,774 lbs m m G = 24.72, b = .32, M = 3, y = 4500
=> Hg = 255,440; hg = 3.02 => MG = 771,429 in. lbs MT HT hT R+g1+hG C-B -g HT = H -Ho, hT ,R = y- = .575 in., 91 = 2.73 in., hc = 3.02 in.
2
=> HT= 38,130 lbf, h T =
3.16'in. =>MT = 120,586 in-lbf Total flange moment under operating conditions M g
= 2,413,111 in -lbf, where MD'= 1,521,096 in lbs, MG = 771,429 in Ibs,
& MT = 120,586 in.lbs g oh OZ Total flange moment under gasket seating conditions S
- @N w
M = W (C'-G) - e f5 0 o 7 ,aw W= (A, + Ab) S a Where C = 30.25, Ab = 42.72 in, G = 24.72 in, A ,= 31.06, Sa" '
=> W = 1,475,600 lbs, Mo = 4,448,934 in. lbf
r D,em d u-TABLE 3.9-9 (Cont'd) CALCUIATED ALLOWABLE STRESS STRESS OR DR. ACTUAL CRITERIA METHOD OF ANALYSIS MINIMUM THICKNESS THICKNESS 12.1 Longitudinal Hub Stress
Reference:
Paragraph NB-3647.l(c) S = o + H 2 g-- Lg yB o At operr 'ing condition S
- 14'484-psi 29,000 psi 14,484 psi H
At atmospheric condition SH = 23,507 psi 34,950 psi- 23,507 psi 12.2 Radial Stress
Reference:
UA-51 (1) equation (7) of Section VIII ASME B&PV Code, 1971 edition 3 , (1.33 t,-+ 1) Mg R 2 Lt B At operating condition SR = 5144 psi 29,100 psi 5144 psi n. w m o At atnospheric condition SR = 9483 psi 34,950 psi 9483 psi ~- m. hH m 8 12.3 Tangential Stress
Reference:
UA-51 (1) equation (8) of Section y' VIII ASME B&PV Code, 1971 edition - w S = YM -ZS t R Where Y = 5.7, t = 5.56-in, Z = 3.0, L = 24.14 At operating condition ST = 3015 psi 29,100 psi 3015 psi -; At atmospheric condition ST = 5533 psi 34,950 psi 5533 psi 12.4 Flange Stress. Criteria
Reference:
Paragraph UA-52 of Section VIII of ASME B&PV Code, 1971 edition Criteria SH+bR <S.m; H+ T <Sa 2 y At operating condition 19,400 psi 9,814 psi 8& 8H+ R = 9,814 H+ 7 = 8,750 19,400 psi 8,750 psi criteria satisfied E$z At atmospheric condition 16,495 psi [6 SH+ 8 R = 16,495 23,300 psi gw SH+ST = 14,.,,95 23,300 psi 14,295 psi
--~ 2 2- cri' aria satisfied
TABLE 3.9- 9 (Cont'd) CALCULATED
;- M-LOWABLE STPESS STRESS OR OR ACTUAL l METHOD O? ANALYSIS MINIMUM THIC. ESS. THICKNESS CRITERIA 13 Sten Design 13.1 Backseat Stress S=[
Where F = net upward force, 9916 lbf A = the sinallest cross-secti,on area of valve stem, 2.268 in' 2&'i00 psi 4372_ psi S = 4372 psi S = 11,501 psi 26,700 psi 11,501 psi 13.4 Stem Thread Strength
Reference:
Fedcral Thread Standard $ W O
, For Thread Mating with stem disk. i e 1.E75 in. - 12 UN - 2 Thread @
S ASy= nL e K n max g + .57735 (E s nin P Where n = 12, L g= 1.62, E *
- s min 2
s K = 1.8030 => ASy = 0.' 2 3 in I
- Where F = 39,450 lbf => i = 7543 psi 16,020 psi 7,S43 psi
- For Thread Mating with Pneumatic Actuator -
2" - 12 UN-2 og n_= 12, S s min = 1.9380, L, = L.14 in., g_ K = 1.928 => ASy = 3.93 in 2 p W - T = v HH A"2 Where F = 39,450 lbf, => r = 10,038 16,020 psi 10,038 psi $$ m W i w - ^_ _ _ u -_
(. ) TABLE 3.9-9 (Cont'd) MSIV Operability Assurance LOAD GOVERNING QUALI-COMBINATIONS ALLOWABLE CALCULATED LOAD FICATION LOAD EVALUATED LOADS LOADS COMBINATION METHOD 6 962,099 9 Analysis Bonnet Combination 2.03 x 10 Moment L through 9 in.-lb in.-lb Combination 31,400 lb 7600 lb 9 Analysis Bonnet Vertical 1 through 9 Load ~ t* g Actuator Combination Less than 0.375 in. 9- Analysis , g. u2
- Deflection 1 through 9 1.6 in.
b $ o
* :o J.
o> - 03: o$ tu ts (1) See Table 3.9-25. yy , wb L mm
, gr vp prw
..~ . .2 TABLE 3.9-10 RECIRC'TLATION PUMPS ALLOWABLE STPS.SS CRITERIA METHOD OF ANALYSIS ANALYTICAL RESULTS OR ACTUAL THICKNESS p p'
+ t = 2.855 inches S = 15,075 psi
- 1. Casing Minimum Wall Thickness "Il "
t = SE - O.6P , . t act.
= 4.8 8 inches A. Leads: where:
.Norral and Upset Condition Design pressure and t = min. req'd thickness, in.
. tempereture P = design pressure, psig R = max. internal radius, inc. B. Primary Membrane Stress S = allowable working stress, psi Limit: E = joint efticiency C = corrosion allowanca, in. t* i w Allowable working stress 8 1
, per ASME Sec. III, -
rn
- i Class C .g
- H zn H 3* O 2 6 1 ' O O O es o z , H8 u) Q LJ . 0340 e t 5 _ __m_.
< , _ _ . _ . - - - - $ _, -,rr - i- - --- -
O -O O TABLE 3.9-10 (Cont'd) ALLOWABLE STRESS , ANALYTICAL OR ACTUAL CRITERIA METHOD OF ANALYSIS RESULTS THICKNESS.
- 2. Casing Cover Minimum Thickness N and Upset S "
+ ^
Condition r " 4t a (m-1) +b (m+1) - '5S = 22,607
= .
m 2 Design pressure and , 3W y _ 2mb - 2b2 (m+1) 1 (a/b) temperature 2t 2 3 2(m-1) + b2 (,,y) - B. Primary Bending _ S = 4,927 psi Stress Limit: t 2 ,4 -b 4 22 g , -3w(m _y) - 4a b In (a/b) . SA = 15.075 psi - 1.5 Sm per ASME Code t for pumps and valves 4mt a (m-1) +b (m+1)
. a _
for nuclear power 2 2 Cla ss I. 3W y , ma 2(m-1) - mb2 (m+1) - 2(m -1) a In (a/b). U 2rmt a (m-1) +b (m+1s - T - - 0 0 w 4 us .
.:o where:
S r = radial stress-'at outer edge, psi S t = tangential stress at inner edge, psi w = pressure load, psi W = uniform load along inner edge, Ib. t = disc' thickness, in. m = reciprocal of poisson's ratio o a = radius of disc, in. O oz b' = radius of disc hole, in. g
*z tu l'
l I
LO O O TABLE 3.9-10 (Cont'd).
'i ALLOWABLE STRESS CRITERIA hiETHOD OF ASIALYSIS ANALYTICAL RESULTS OR ACTUAL THICKNESS
- 3. Pump Discharge Nozzle Stress Pressure P P
= 1644 psi 1.5 S ,= 29400 psi (Pressure Bending, Axial SET and Torsional) P = R + 0.6t Pg = 7232 psi p
A. : Loads: P = 3605 psi ed
' Normal and Upset Bending P = 7233 psi et Condition CFS bb Pg =G b
P c
=
19714 Psi Design pressure and temperature piping reactions during Axial normal operation. F8 d & w P ed
=Id Q" '
w b B. Combined Stress U Limit: @ Torsional 1.5 S ,per ASME code for Pumps and Valves Pet "2 b' ' for Nuclear Power G Class I. t Combined P =P c p +Peb + Ped + Pet
- 4. Cover and Seal Flange Bolt Areas Bolting loads, areas Cover Flange Bolts 'S ggy = 25,325 psi and stresses shall be A. Loads: calculated in accor- S Act. = 16772 psi 'A . = 78.6 sq. in.
uance with Rules for min Og Normal and Upset Bolted Flange'Connec- eM
- Condition tiens" ASME Sec. A, = 118.8 sq. in. @@
VIII, Appendix II, yg Design pressure and Temperature $g Design gasket load gu J._..m._-_m_,awa,,., .m _ .m _. n.m.-, m msa-&--_ ,e__--.-e._m-a2 .a_.,.~ ..m,-: .m., - - - - -
-,a. .i---
_ --__1.2_. _ e.gg - .,%
vg f,,g :4 ( ) 't i
. %_/ %./
TABLE 3.9-10 (Cont'd) I ALLOWABLE STRESS
-OR CRITERIA METHOD OF ANALYSIS ANALYTICAL RESULTS ACTUAL-THICKNESS, B. Bolting Stress Limit:
Allowable working 4 stress per ASME Sec. III Class C
- 5. Cover Clamp Flange Flange thickness and stress Flange Thickness and Thickness shall be calculated in accor- Stress dance with " Rules.for Bolted Flange Connections" - ASME t = 7.08 inches t Act
= 8 3/8 inches
- . A. Loads: Sec. VIII, Appendix II.
Normal and Upset Act. = 12,497 psi S ggy " = 15,000 psi' Condition @ w @ i
- Design pressure and I temperature @
[ Design gasket load g w Design bolting load B. Tangential Flange Stress Limit 4 Allowable working stress ASME Sec. III Class C PR t = 1.071-in. S" = 15075 psi
- 6. Seal Compartment Wall Thickness t = SE - 0.6P A. Loads:
where: t = 1.854 in. Act Normal and Up3et Condition t = min. reg'd thickness, in. P = design pressure, psig og Design pressure and tempera- R = max. internal radius, in. OR ture S = allowable working stress, psi @@ E = joint efficiency Mg Primary Membrane Stress C = corrosion allowance, in. z B. Limit $8
-ww co e 2
Allowable . working stress per A3ME Sec. III Class C.
>-r-
J TABLE 3.9-10 (Cont'd) ALLOT?ABLE . STRESS CRITERIA- OR FETHOD OF ANALYSIS- ANALYTICAL RESULTS ACTUAL THICKNESS-
- 7. Seal Gland Retainer S,= q Ss = $486 psi S = '9480 psi 3
A. Loads:. Normal and Upset Condition w = load imposed d = diameter at shear resistarice t = thickness at shear resistance Design pressure and tempera-ture B. Allowable working stress per ASME Sec. VIII-
- 8. Shock Supressor Lug Combined Loads shall be applied in the Combined Stress F Stress normal direction simultaneously (Shear plus Tensile)
S" = 19,435 psi to determine tensil, shear and E-H A. Loads: -bending stresses in the brack-S Y
= 21,600 psi g U ets. Tensile, shear, and bend-Lug Il SC = 21,430 psi :
DBE horizontal seismic ing stresses shall be combined to Lug #2 Sg = 20,915 psi 5 force = 1.5g determine max. combined stresses. g Lug #3 SC= ,540 psi B. Combined Stress Limit: Yield stress per'ASME Sec. III
- 9. Hanger Bracket Combined Bracket vertical loads shall be SC = 8,327 psi S = 12,600 psi Stress determined by summing the equip- m ment and fluid weights and A. Loads: vertical seismic forces.
Flooded weight of equipment DBE vertical seismic Load = (WB* C+ D 8D force = 0.14g gQ Note: the multiplier (0.33) is @g added as a safety factor specified m* on the Purchase Part Drawing. Wh w >5 B. Combined Stress Limit: Uw co w Yield stress per ASME WB * "" " # Sec. VIII WC = weight of motor mount WD = weight of pump case -_ _ .___m -__ - _ - T'
(40 s. (b* b%
%/ \~ ,'
TABLE 3.9-10 (CSnt'd) ALLOWABLE STRESS OR CRITERIA METIf0D OF ANALYSIS ANALYTICAL RESULTS ACTUAL THICKNESS
- 10. Stresses Due to Seismic The flooded pump-motor assembly Motor Bolt Tensile Loads shall be analyzed as a free body Stress:
supported by constant support A. Loads: hangers from the pump brackets. Horizontal and vertical seismic SAct." 1 '9 1. psi Sali,= 30,000 psi Operation pressure and forces shall be applied at mass Pump Cover Bolt. temperature center of assembly and equili- Tensile Stress: brium reactions shall be deter-DBE horizontal seismic mined for the motor and pump SAct.= 18,542 psi force - 1.5g brackets. Load, shear, and Sggy,= 25,375 psi moment diagrams shall be con-DBE vertical seismic structed using live loads, dead Motor Support Barrel force - 0.14g ' loads, and calculated snubber Combined Stress: W reactions. Combined bending, B. Combined Stress Limit: tension and shear stresses S [ e' Act. = 729 psi SAll.= 12,600 psi n shall be determined for each y
- d. Yield stress per ASME major component of the assembly H
Sec. VIII including motor support barrel, @ bolting and pump casing. The
- maximum combined tensile stress in the cover bolting shall be calculated using tensile stresses determined from loading diagram plus tensile stress from operating pressure.
6 n O we W
~?,
w MW C3 w e
- lO O O TABLE 3.9-11 STRUCTURAL AND MECHANICAL LOADING CRITERIA (Reactor Recirculation Gate Valve 24-Inch Suction) 4 COMPONENT / REQUIRED ACTUAL LOADS / DESIGN DESIGN PROCEDURE DESIGN VALUE DESIGN VALUE 1.0 Body and Bonnet 1.1 Loads Design Pressure, Design Temperature, Pipe Reac-tion, Thermal Effects 1.2 Pressure Rating, psi Used NB-3531 Table Pr = 678 psi Pr = 678 psi NB-3531-4, and NB-3531-5 of Section III 1.3 Minimum Wall Thickness, Used NB-3543, and Table tm >_ l.747 in. tm = 1.75 in.
, inches NB-3542-1 g
- o
? 1.4 Primary Membrane Stress, Used Section NB-3545.1 P, <_Sm(500*F) = 19,600 psi P,= 11,075 psi y
[ psi g m > 5 1.5 Secondary Stress Due to Used NB-3545.2(b) Ped i 1.5 Sm = 29,400 psi Ped = 6,400 psi Pipe Reaction (S = 30,000 psi) Pe < 1.5 Sm = 29,400 psi Peb = 13,150 psi Pet i 1.5 Sm = 29,400 psi Pet = 13,150. psi 1.6 Primary Plus Secondary Used NB-3545-2(a) See 1.8 below Op = 21,730 psi Stress Due to Internal Pressure 1.7 Thermal Secondary Stress Used NB-3545-2(c) See 1.8 below QT1 = 2000 psi QT2 = 830 psi O$ e tn OT3 = 960 psi @g t:5
-a w a> w Notes: Section III = ASME Boiler and Pressure Vessel Code, Section III, 1971 " Nuclear Power Plant Components".
_--__u-___.___- ____a _ 1 e
y-. ,- U k D TABLE 3.9-11 (Cont'd) COMPONENT / REQUIRED ACTUAL LOADS / DESIGN DESIGN PROCEDURE DESIGN VALUE DESIGN VALUE 1.8 Sum of Primary Plus Used NB-3545.2 S Secondary Stress n *O p + Ped + T2 S = 29,790 n Sn5 3Sm(500*F) = 58,800 psi-1.9 Fatigue Requirement Used NB-3545.3 Na 1 2000 cycles Na a 106 c/cles 1.10 Cyclic Rating Used NB-3550 I, 3 1.0 Ie = .003 2.0 Body to Bonnet Bolting
2.1 Loads
Design Pressure, Used NB-3546.1, NB-3647.1 Design Temperature, and Section VIII Gasket Loads. Stem Operational Load, Seismic Load (Design Basis Earthquake) e w $
, 2.2 Bolt Area Used NB-3546.1, NB-3647.1 -Ab 1 31.53 sq.in. Ab Y-e e
and Section VIII = 47.73.in.2-- y M Sb 1 27,975 psi Sb = 18,480 psi $ a 2.3 Body Bonnet Flange Used NB-3546.1, NB-3647.1 Stresses anl Section VIII 2.3.1 Gasket Seating Used NB-3546.1, NB-3647.1 Sgi 1.5 Sm (575'F) = 28,837 psi SH = 25,245 psi Condition and Section VIII SR1 1.5 Sm(575'F) = 28,837 psi Sg = 9425 psi ST1 1.5 Sm(575*F) = 28,837 psi ST = 10,140 psi 2.3.2 Operating Condition Used NB-3546.1, NB-3647.1 and Section VIII Sn 5 1.5 Sm (150 *F) = 30,000 psi SH = 17,170 psi SR 1 1.5 Sm(150'F) = 30,000 psi SR = 5,735 psi 'og ST 1 1.5 Sm (150 *F) = 30,000 psi ST = 6,120 psi gg to a 3.0 Stresses in Stem
- o $
g
3.1 Loads
Operator H +5 Thrust and Torque. ww 3.2 Buckling on Stem Calculated slenderness Maximum allowable load = ' Slenderness ratio = ratio, If greater than 30, 100,000 lbs 42.5. Actual calculate allowable load thrust load on stem from Rankine's formula = 18,100 no using safety factor of 4 buckling. e-
'O; O O-TABLE 3.9-11 (Cont'd) i COMPONENT / REQUIRED ACTUAL LOADS / DESIGN DESIGN PROCEDURE DESIGN VALUE PESIGN VALUE 3.3 Stem Thrust Stress Calculate stress due S to operator thrust T,C 1 8m = 30,800 psi ST,C = 5,340 psi-in critical cross-1 section.
i 13 . 4 Stem Torque Calculate shear stress SC 1 6S m= 18,480 - SC = 4,200 psi' t due to operator torque , in critical cross-section
~
4.0 Disc Analysia-4.1 Load: Maximum Differential Pressure 4.2 Maximum Stress Calculated maximum Smax 1 Sm = 15,800-psi- Smax = 11,389 psi stress according to w chapter of R.J. Roark
" Formulas 'for Stress
'$n I.
r and Strain" Y p m m
# 5.0 .. Yoke and Yoke Connections >
5.1. Load: Stem Operational Calculated stresses in Load, Design Basis the yoke connections to Ear thquake acceptable structural analysis methods. 5.2 Maximum Stress Smax 1 1.25Sm = 21,000 psi Smax = 17,750 psi O>- og Ok WO 2 b CO W s 9 ansamamme A 7
U V~ u) TABLE 3.9-11 (Cont'd) COMPONEM1/ REQUIRED ACTUAL LOADS / DESIGN DESIGN PROCEDURE DESIGN VALUE DESIGN VALUE
- 1. Body Bonnet
1.1 Loads
Design Used NB-3513 of pressure, design Section III* temperature 1.2 Pressure-Rating, psi Used ND-3531 of Pr
- 879 psi Pr = 900 psi-Section III 1.3 Minimum Wall Thickness, inches Used Table NB-3542.1 of Section III tm> .28 inch tm = .5625 inch W
=
b 0- 0:
- 2
- tr o>
0$ O 2: t: ts 03
*Section III = ASME Boiler and Pressure Vessel Code, Section III, 1971, " Nuclear Power Plant Components."
e
't.,
- p. j :g. OL TABLE 3. 9-11. (Cont' d) .
Recirculation: Discharge Valve Structural I'ntegrity - ALLOWABLE GOVERNING CALCULATED ' ALLOWABLE GOVERNING : CALCULATED: ELOADING HORZONTAL LOAD. HORIZONTAL VERTICAL -LOAD: VERTICAL. , CATEGORY - ACCELERATION COMBINATION ACCELERATION ACCELERATION COMBINATION' COMBINATION Normal- ., and Upset .. Condition 9.50- 2 4.86' 4.0 2 .l.59- . (Service l~ jLevel 'A' - and 'B') Emergency - r-
' Condition (Service 9.50 '3 5U3 4.0 3 1.69 8' 9-
,e ~ Level 'C') %.
I u) - H M Faulted > W-
' Condition 9.50 6- 6.16 4.0 6 .1.81
- i. (Service. -
U . LevelD')
- O$
e to oz (1) The calculated accelerations with; dynamic loads.' included are lesslthan.the -E E. , ~. . allowable' accelerations,.which demonstrates design. assessment.- M y-(2)' Discharge-valve is designed'.to' accommodate loads-transmitted by piping-such -- H e ! that maximum stress:in the pipe,.atythe. point of attachment.'to the valve, is w , .' <.30,000. psi. ** -
'(3) 5ee' Table 3.9-25.
- w. . . . ~ . , ,, . :__ -
___.-_: . _ = _ - .
, yn U
TABLE 3.9-11 (Cont ' d) Recirculation Suction Valve Structural Integrity ( }( } ALLOWABLE GOVERNING CALCULATED ALLOWABLE GOVERNING CALCULATED LOADING HORIZONTAL LOAD HORIZONTAL VERTICAL LOAD VERTICAL CATEGORY ACCELERATION COMBINATION ACCELERATION ACCELERATION COMBINATION ACCELERATION Normal and Upset Condition 11.06 2 6.61 4.0 2 2.55 (Service Level 'A' and 'B') m Emergency '@ Condition 11.06 3 6.70 4.0 3 2.67 o 7 (Service Y [ Level 'C') ] w > Faulted # Condition 11.06 6 6.81 4.0 6 2.82 (Service Level 'D') O> o ;r e tn (1) oz The calculated accelerations with dynamic loads included are less than the WD (2) allowable accelerations, which demonstrates design assessment. $k Suction valve is designed to accommodate loads transmitted by piping.such that maximum stress in the pipe, at the point of attachment to the valve, is s$
< 30,000 psi. co e (3) See Table 3.9-25.
TABLE 3. 9-12 HYDRAULIC CONTROL UNIT PIPING ALLOWABLE CALCULATED CRITERIA LOADING LOCATION STRESS (psi) STRESS (psi) Hydraulic Control Unit Piping From USAS B31.1.0 - Code for Normal condition load 3/4-inch drive Power Pressure Piping Maximum normal. hydraulic withdraw piping 15,000 14,596 system pump pressure ' For Normal Conditions: S = 15,000 psi h For Upset and Emergency Upset condition load 3/4-inch drive Condition: 1. Shutoff pump pressure withdraw piping 18,000 16,950 w . 2. Maximum probable earth-e quake (negligible load) C When upset or emergency: -Q
-h- condition exists for less y than 1% of the time,.the 4 code allows 20% increase y_
- o -
in stress. S a
=
1.2 Sh = 18,000 psi . Emergency condition 3/4-inch drive
- 1. Shutoff pump pressure withdraw piping 13,000 16,950
- 2. Maximum.possible earth-quake , (negligible load)
'eO 2: -
tv t1 - M 3: N su u_________.__-______-___________._1..______._.__ _ _ ___________ _ _ _ _ _ _ _ _ _ _ _ - _ _ . _
s O O LO- . TABLE 3.9-13 RHR PUMP 1 CALCULarION i CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS (osi) (psi)
- 1. Closure Bolting Bolting loads and stresses Maximum Allowable ~ Stress (2) Maximum Calculated (2) calculated.per " Rules for for bolts Stresses for bolts -
',- Loads: Normal and Upset. Bolted Flance Connections" ASME Section VIII, App. II Motor Mounting 25,000 9,501 Design Pressure and Pump Mounting 25,000 4,368-De gn G sket Load (SSE) Seismic Acceleration, . Pumping Element 25,000 16,267 Nozzle Forces and/or Momer.ts Static Mass Forces, Safety Relief valve and LOCA Bolting Stress Limit Allowable Working Stress per ASME Section VII
" 2. Wall Thickness Per rules of Part UG Sec. VIII Maximwn Allowable Stress Maximum Calculated $
e 17,500 10,370 .Q
- r. Loads: Normal and Upset b!
S$ Design Pressure and .g Temperature Stress Limit ASME Section VIII ,
- 3. Nozzle Loads I4) For the maximum stresses due I to pipe reaction by maximum I force or moment Loads: Normal Plus Upset Design Pressure and Suction Suction.
Temperature Dead Weight, Maximum 26,250 20,190 Force and Moment, Operational. Basis Discharge Discharge og' ' Earthquake, and Safety- em Relief Valve 26,250 23,200 @$.
~5
. .m
v J U*s-TABLE 3.9- 13 (Cont'd) CALCULATION
' CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS (psi) (psi)
Loads: Faulted f-Design Pressure and Suction Suction Temperature Dead Weight, Maximum 31,500 20,970 Force and Moment, and Safe Shutdown Earthquake, Discharge Discharce Safety Relief Valve and LOCA 31,500 29,480 Stres:- ulmit aSME Section VIII Primary general local membrane 1.5 for Normal and stress Upset, 1.8 of allowable stress for emergency 4
.u M vs
- S A 4 u m .
- o (1) Operability demonstrated by analysis.
(2) The stresses calculated for the faulted operating condition loads are less the allowable stress limits for the upset condition.
'(3) The design assessment evaluation results are equal to or lower than the reported values. 8$-
(4) The maximum allowable. nozzle force and moments were imposed with other appropriate loads to @ demonstrate acceptance. es o
$b=
H a-3 CD W -
" -' " ~ "
_ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ ___ac- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
m- y -- (- V. (_) ' TABLE 3.9-14 L RilR HEAT EXCHANGER ALLOWABLE STRESS OR MINIMUM-CRITERIA METHOD OF ANALYSIS THICKNESS REQUIRED (psi) ACTUAL (psi)
- 1. Closure Bolti j Bolting loads and stresses calculated per " Rules for-Loads: Normal and Upset Bolted Flange Connections" ASME Section VIII, App. II Design Pressure and Temperature Design Basket Load Bolting Stress Limit a. Shell to tube sheet bolts 25,000 24,083 Allowable Working'- b. Channel cover bolts 25,000 24,711 Stress per ASME Section VIII
- 2. Wall Thickness Shell side ASME Section III, A TEMA Class C Design Pressure and p
. Temperature m 'f O g Stress Limit 4
sn M ASME Section VIII a. Shell 0.8844 inch 0.9375 inch $
- b. Shell cover 1.3124 inch. 1.3125' inch
- c. Channel ring 0.4999 inch l0.500 inch
- d. Tubes 0.049 inch 18 Bh'G
- e. Channel cover 4.6938 inch 4.75. inch
- f. Tube sheet 7.0439 inch 7.0625 inch O>
Oh OZ t3 D d 20
. . . ~ _ m .- - _ - . ,
.l_.; _. z__._._m=..-_ =m._m.m ... .-eem5
O O O TABLE 3.9-14 (Cont'd) ALLOWABLE STRESS OR MINIMUM CRITERIA METHOD OF ANALYSIS THICKNESS REQUIRED (psi) - ACTUAL (psi)
- 3. Nozzle Loads The maximum moments due to (See belott) pipe reaction and the (see below).
Design Pressure and' maximum forces shall.not Teraerature exceed the allowable limits. Dead Weight, Thermal Expansion,-Safe Primary stress less than 1.5 Shutdown Earthquake ASME Section VIII allowable. Allowable Limits (Design Basis) N1 N2 N3 N4 M = 27,557 in-lbs 27,557 in-lbs 204,000 in-lbs 83,550 in-lbs g = 27,557 in-lbs 27,557 in-lbs 204,000 in-lbs 83,550 in-lbs y Fp = 1,949 lbs 1,949 lbs 13,600 lbs 13,100 lbs y M = 35,291 in-lbs T 35,291 in-lbs 261,000 in-lbs 252,000 in-lbs 5 w a so m m. 00 - O O
"E gn.
at
-w- . . _ _ . - - - <-_m_--- m --- - - - - - - ~ - - - - - - - - -. ' -- a.i. -- - - - - --*---.&
---_g------- m-- , ---%---.a- - - - ~ - - _ . .
i
-.[ .
i I-i i TABLE 3.9-15 I l LOW-PRESSURE CORE SPRAY PUMP (lI ): CRITERIA METHOD OF ANALYSIS- ALLOWABLE STRESS (psi) ~ ' CALCULATION (psi)(2)
- l. Closurc Bolting, Bolting loads and 3 tresses Maximum Allowable- Maximum calculated calculated per." Rules for' ' Stress 25,000 13,954
' Loads: Normal and Upset Bolted Flange Connections"' .l- ,
ASME Section VIII, App. II ; i Design Pressure and Temperature i Design Gasket Load,'(SSE) l Seismic Acceleration, Nozzle Forces and/or Moments, Static Mass Porces, SRV, LOCA l, I Bolting Stress Limit w Allowable Working l e- Stress per ASME- 5 Q~
$- Section VIII M
4 TJ) 1
- 2. Wall Thickness Per rules of' Part UG Sec VIII. Maximum Allowable- Maximum' calculated Stress-17,500- 9,580 Loads: Normal and Upset Design Pressure and
, Temperature Stress Limit: ' ASME Section VIII I 8{ ' gg :; ng' '
"E .
y (1) Operability of this component under the above~ loading conditions has been-demonstrated'by analysis.
' (2) This component has been. evaluated using dynamic loads unique to the LaSalle Project. _ The governing-I
-load combination listed in Table 3.9-25 has been applied appropriately. The stresses calculated in' this evaluation are listed-in'the appropriate column.
- t
...y. -g . . . , , _.
yy.. _ . _ _ . - _ . . _ _ ,_y_.,y_.- . _ , . , . , . . , . - .<om.. g 7 "9, ,,,..3 y a. __.. .
i:
- O ;O Joi TABLE 3.9-15 (Cont'd)
METHOD OF ANALYSIS . ALLOWABLE STRESS (psi) CALCULATION (psi) CRITERIA
- 3. Nozzle Loads For the maximum stresses' ).
- due to pipe reaction by maximum force or moment l
Loads: Normal Plus Upset Design Pressure and Temperature. . Suction Dead Weight, Maximum Force and Moment, and- 26,250 8,530
-Operating Basis Earthquake-Discharge-26,250 24,140 Loads: Emergency Suction w Design Pressure and. 31,500 9,050 g:
Temperature o e Dead Weight, Maximum Discharge y
.5 Force and Moment, M
u ' " Safe Shutdown 31,500 24,140 $ Earthquake, SRV, and LOCA Stress Limit ASME Section VIII Primary local general'mem-1.5 for normal and brane stress upset, 1.8 of allow-able stress for-emergency 8E-em OZ' to O M3
'*E
, ed w wu mm (3) The maximum allowable nozzle forces and moments were-imposed with other appropriate loads to demonstrate acceptance. 1 amenemme y-9. _ . . , . . u. . 4.~p_,- _,.y,._m,y-4. - s- ,,..y,.r.4 , v
~ V ,J v TABLE 3.9-16 RCIC TURBINEIII ALLOWABLE ( l ST ESS LIMITS CRITERIA METHOD OF ANALYSIS (psi) CALCULATION (psi) l 1
- 1. Closure Bolting Bolting loads and stresses Maximum Allowable calculated per " Rules for Maximum Calculated Stress 25,000 (Normal Stress 20,100 Loads: Faulted Bolted Flange Connections" Operation Limit)
ASME Section VIII, App. II Design Pressure and Temperature, SSE, g SRV, and LOCA I. Bolting Stress Limit Allowable Working Stress per ASME Section VIII
- 2. Casing Wall Thickness Per rules of Part UG Sec VIII Maximum Allowable y Maximum Calculated E Stress 21,000 Stress 18,000
- Loads: Faulted (Normal Operation @
& s g Design Pressure and Limit) $
Temperature, SSE, SRV g and LOCA Stress Limit ASME Section VIII O> Oh oz NE "E t;"
-a ta co w (1)
Operability of the RCIC Turbine under faulted conditions has been demonstrated by Dynamic Test in accordance with IEEE - 344 - 1975. (2) The stresses calculated for the faulted operating condition loads are less the allowable stress limits for the upset condition. e 9 .A
TO LO :OL TABLE - 3. 9-16 . (Cont'd) . CRITERIA METHOD OF ANALYSIS CALCULATED FORCE (lb) ( 3) FORCES AND MOMENTS ( I AND MOMENT-(ft-lb) '
- 3. Nozzle Loads For the resultant moment' Force in Ibs,, Moment in-due to' pipe reaction the ft-lbs-Loads: Normal resultant force shall not
- Design Pressure and exceed the allowable. Inlet ' Inlet Temperature Detailed design analys Dead weight and Thermal Force has demonstrated the ac-Expansion F= (2,620-M)/3 Moment:
ceptability of.these values. Exhaust. Exhaust
~ Force F= (6,000-M)/3 Moment Load: Normal Plus Upset-Inlet' Inlet Design Pressure and Temperature Force w F =- (3,500-M)/3 Moment
- Dead Weight,' Thermal e
- e. Expansion and the Exhaust 8' 4 envelop of Operating Exhaust'- rn -
w Basis' Earthquake,'SRV
.-F= 4-5 (7,000-M)/3. Force Moment-
- o Loads: Faulted Inlet Inlet-Design Pressure and Temperature Force F= (4.2 00-M) /3 Moment Dead Weight, Thermal Expansion,. Safe Shut-down Earthquake, SRV, ..
and LOCA', Stress Limits Exhaust Exhaust' oy
' Specified by Vendor for F= 03:
(8,400-M)/3. . Force Norma 1, ASME Section ~ oE' VIII for Upset, Moment @U Increased 20% for Emergency M U
-a w co e . ~
(3) The maximum demonstrate allowable nozzle. forces and moments were imposed with other appropriate loads to acceptance. r,-,< , -3 e .,y. ,- w - s- - - * , , - , yy. . . ,- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
p a G a TABLE 3.9-16 (Cont'd) ALLOWABLE STRESS LIMITS CRITERIA METHOD OF ANALYSIS (psi) CALCULATION (psi)
.: . Turbine Mounting Bolts Vertical and horizontal forces T'furbine to Biseplate) on mounting bolts calculated as the sum of combined dynamic accele- .l Loads: Normal and Upset rations on the turbine and the pipe reaction forces and moments Operating Basis Earth- on the nozzles. Tensile and shear stress By meeting the quake for bolting materials nozzle load crite-Nozzle Loads for Normal as specified in ASME ria of 3 above, l and Upset Dead Weight and Section VIII the detailed seis- I Thermal Expansion mic analysis indi-cates the mounting Tensile stress less than bolts satisfy the Loads: Faulted 0.9 yield and shear stress allowable stress Safe Shutdown Earth- less than twice allowable . requirements quake, maximum allowable of ASME Section VIII Nozzle Loads for faulted condition, Dead Weight, P Thermal Expansion, and $
y LOCA g'
~ ,
U Stress Limits @ ASME Section VIII Allowables for Normal and Upset. For Faulted 0.9 yield and twice allowable shear O> Oh 85 M2 P 20 ._-_____________.___m. _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ . _
7 () (3 t')
%-) %]
TABI-E 3. 9-17 RCIC PUMP $1) ALLOdadLE 2) 373g33 L73773 CRITERIA METHOD OF ANALYSIS (psi) CALCULATION (psi)
- 1. Closure Bolting Bolting loads and stresses Maximum Allowable Stress Maximum Calculated calculated per " Rules for 25,000 (Normal Operation- 20,740 Loads: Faulted Bolted Flange Connections" Limit)
ASME Section VIII, App. II Design Pressure and Temperature, SSE, SRV, and LOCA Bolting Stress Limit Allowable Working Stress per ASME Section VIII
- 2. Wall Thickness Per rules of Part UG Maximum Allowable Stress Maximum Calculated Sect. VIII Main pump 17,500 (Normal 8,988
. Loads: Faulted Barrel Stress Operation Limit) E Y O e Design Pressure and a U Temperature, SSE, 5 SRV, and LOCA >
Stress Limit " ASME Section III 84 e 5i 05 Di5 d (1) Operability demonstrated by Computer analysis using 2.25g acceleration load. . Shaft deflection Minimum clearance 0.006 in. Maximum deflection 0.002 in. Shaft shear stress Allowable stress 25,500 psi calculated stress 3,089 psi Bearing Static load capacity Drive end allowable force 7,670 lbs- Drive end calculated force 98 lbs Thrust end allowable force 17,200 lbs Thrust.end calculated force 766 lbs (2) The stresses calculated for the faulted operating condition loads are less the allowable stress limits for the upset condition. _ _ _ . - .-- . _ - - _ - - - - . . - - - - - - - ~- -- - - - - -
(. t i' e t wi %J TABLE 3.9-17 (Cont'd) CALCULATED FORCE ( CRITERIA METHOD OF ANALYSIS ALLOWABLE FORCEF AND MOMENTd3I AND MOMENT (f t-lb) I
- 3. Nozzle Loads For the maximum moment due Force in Ibs, Moment in to pipe reaction the maxi- ft-lbs.
mum force shall not exceed Loads: Normal Plus Upset the allowable. Design Pressure and Total nozzle stress with Suction . Temperature this criteria does not Dead Weight, Themal exceed stress limits. Force 1700 Force Expansion, Operating Moment 4500 Moment Basis Earthquake, and SRV Discharge Force 2750 Force Moment 7800 Moment Loads: Faulted Suction w Design Pressure and Force 2L40 Force b n , Temperature Moment 5400 Moment y a Dead Weight, Themal m U W Expansion, Safe Shut- Discharge: $ down Earthquake, SRV W and LOCA Force 3301 Force Stress Limit Moment 936u Moment ASME Section VIII for Normal and Upset, 1.5 of allowable stress for faulted 8e e EG oz cs a (3) The maximum allowable nozzle forces and moments were imposed with other appropriate loads ng to demonstrate acceptance. "z we o MW 03 W
7.- g ,.3 U) v) 1
'v)-
TABLE 3.9-18 RECIRCULATING PIPE AND PUMP RESTRAINTS COMPONENT IDADINGS IDCATION DESIGN LIMITS CALCULATED LIMITS Restraint Reaction force Multiple on 50% of uniform < 50% for all Frame from pipe break recirculation piping ultimate strain restraints Connecting Members Reaction force Multiple on Primary membrane and Fasteners from pipe break recirculation piping stress, at at <97 (Gu) or at 1 0.7 (ou) or oy + (o oy + j (ou-oy), whichever which}ever isu + Oy)e is higher higher Shear Stress, osh Osh < 0.6 (ot) osh < 0.6 (og) Bearigg Stress, ob 2 ob 1 7 (oul ob < y (oul Pump Restraint Reaction force from pipe break One on each pump Primary membrane stress at i 1.0 (ay)-
'ot i 1.0 (ay) p w 'u> $
a Attachment Reaction force At piping and pump Weld shear stress oSH <- l.5 (oAws) 02 C Welds from pipe break restraint locations oSH 1 1.5 (oAws) 4 s tn
- D 1 NOTES: o u = min. ultimate strength by testing or f rom ASTM spec.
gk o
$E
~
oy = min. yield strength by testing or f rom ASTM spec. to o Uniform ultimate strain determined by testing or from ASTM specified min. $$ wk oAWS = allowabla weld stress from AWS velding code or AISC structural code. g 5-me N
__ ^
/-
9'- TABLE 3.9-19
^
HPCS PUMPIII l CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS (psi) CALCULATION (psi)(2)f
- 1. Closure Bolting Bolting loads and stresses calculated per " Rules for Maximum allowable stress Maximum calculated 25,000 19,522 Loads: Normal and Upset e ange onnecdons" l ASME Section VIII, App. II Design pressure and
' temperature Design gasket load, (SSE) Seismic acceleration, nozzle forces and/or -f moments, SRV, and LOCA Bolting Stress Limit - w Allowable working stress
-o per ASME Section VIII p 8
g C 2. Wall Thickness Per rules of Part UG Sect. VIII Maximum allowable stress g m Maximum calculated . y 17,500 8,180 >
- o Loads: Normal and Upset Design pressure and temperature Stress Limit ASME Section VIII oz NE (1)
Operability of this component under the above loading conditions has been demonstrated by analysis. "E (2) 58 This component has been evaluated using dynamic loads unique to the LSCS project. The governing NU
?oad combination listed in Table 3.9-25 has been applied appropriately The stresses calculated in this evaluation are listed in the appropriate column. .
' - ~ * - "- N F " " -* 7
_ m -__. -w _ _ _ -~f A.
l O O O ._
+
.t TABI E 3. 9-19 (Cont'd)
CRITERIA METHOD OF ANALYSIS ALLOWABLE STRESS (psi) CALCULATION (psi)
- 3. Nozzle Loads -For the maximum stresses due to pipe reactioh by maximum force or moment Loads: Normal Plus M et '
Design pressure and temperature H Dead weight, maximum Suction force and moment, Suction operating basis 26,250 17,920 3 I earthquake, and SRV Discharge Discharge 26,250 10,919 f f Loads: Emergency E Y w 7esign pressure and y ' temperature i Dead weight, maximum Suction
.5 force or moment, Suction .$
safe shutdown i 31, 17,920 i earthquake, SRV, and LOCA Dist.l. ec . Discharge 31,500 10,919 Stress Limit ASME Section VIII Primary general local membrane stress 1.5 for normal and upset- 1.8 of allowuole. stress for og emergency. Qq O2 CD O M3
. lC M Z
<m 03 @
e- . - - . - - . . .__ _ __ , . . . , _ _, , s ,,
. - %./ . w TABLE 3.9-20 STANDBY LIQUID CONTROL PUMP 4 ALIDWABLE CALCULATED CRITERIA ME. D OF ANALYSIS STRESS LIMIT 9 (psi) STRESS (psi)
- 1. Closure solting Bolting Loads and Stresses Stuffing Box Bolts- 24,750 calculated per " Rules for 25,000 (Normal Loads: formal and Upset Bolted Flange Connections" Operation Limits)
ASME Section III, App. XI. Design Pressure and Tem- Cylinder Head Bolts . 18,830 perature, safe shutdown 25,000 (Normal Operation earthquake, safety re- Limits) lief .ralve , LOCA, and Design Gasket Ioad.
-Bolting Stress Limit Allowable Working Stress per ASME p
Section III g .. w m
- 2. Wall Thickness Pressure area metnod maximum. 17,800 10,394 u> stress point on fluid cylinder 4 l, Loads: Faulted ASME Section III, Subsection (Normal _ Operation Limits) y W m NC, NC - 3482 & 3324.3.
Design Pressure.and Tem-perature, Safety relief valve, safe shutdown earthquake, LOCA Stress Limit ASME Section III
- 3. Motor Mount Bolts ( Seismic forces acting on motor Tension ~ 4,523 I' ~
subject bolts to tension and 37,500 Loads: Emergency shear H = 1.75g V = 1.75g Shear SAE Handbook, 1972 Edition 30,000 3,028(3) S *e Shutdown Earthquake. (Normal Operation o> Sofety relief valve, LOCA Limits) O@ OM' Stress Limit 0.9 yield for tension .[8 and twice allowable ww shear ASME VIII "*
a Y VS. '. .- g
; g) :
[Qf - _.
~
,. ,, 4 m.
['__ o
* /
h TABLE 3.9-20 (Cont'd) 4 ALLOWABLE CAlfULATED ' f' CRITERIA METHOD OF ANALYSIS . STRESS LIMITS fpsi) STRESS (psi) , 4. Nozzle' Ioads Nozzle forces and moments are Suction i determined on the basis that a F = 770 lb- Force
' Loads: Normal Plus' Upset 6000 FSI fiber stress is im- Moments
^
posed by the londs. The 6000 Mo = 490 ft-lb Design Pressure and Tem-' PSI fiber stress is related to 3 _ perature. -Dead weight, nozzle forces and moments bys Discharge Force Thermal Expansion,-and OBE S = F/A + M/Z where "S" is the 'F =~370 lb Moments fiber stress, F and. R are the 4 Loads: Faulted resultant forces-and moments, '"o = 110 ft-lb i and "A" and "Z" are the metal I Design Pressure and Tem- area and section nodulus,.re- Suction . , perature. .Dcad Weight, spectively, of-the suction F = 910 lb , Force t-y Thermal: Expansion, SSE, and discharge piping. Moments 8
- e: SRV, LOCA- o = 590 5t-lb CA I
5
. g N g' Stress Limit- Discharge $
F =.440 lb Force M ASME'Section VIII
~
Moments
= 3.30 ft-lb
' ~ (1) The sum of the plunger and rod assembly, pounds mass times 1.75g accelerated is much_less-than the trust loads i_ encountered during normal operating conditions. Therefore, loads during the faulted condition have no effect 4 on pump operability. (2) The stresses calculated for.the faulted operating condition loads-are less than'the allowable stress limits-for the upset condition.
. (3) The results are based on existing-calculations for 1.75g. loads (horizontal and v'ertical). This "G" loading 8$
conservatively exceeds and envelopes the plant unique seismic combined with other appropriate dynamic ?aads. $$'
~ (4) The'naximum allowable nozzle forces and moments were imposed with other appropriate loads to demonstrate gg acceptance. m rn 2:
HH w
. _ _ _ _ _ _ _ - . _ _ _ _ _ _ . _ _ . . . _ - _ _ _ . . - _.i~__. . ~ , ~ . , . - _
~ . .J . , , -a , , c , -,- , ,
~ ,sq e, _ ,.
i TABLE 3.9-21 gs . STANDBY LIQUID CONTROL TANK ALLOWABLE STRESS LIMITS (psi) CALCULATED CRITERIA METHOD OF ANALYSIS .CR MINIMUM THICKNESS REQUIRED (in.). VALUES
- 1. Shell Thickness Minimum Thickness 0.015 in. 3/16 in.
t ' Loads: Normal and Upset ,2.6 D (H-1)G in. SE Design Pressure and Temperature D = Nom. ID H = Tank Height [ Stress' Limit G = Specif. Gravity-S = Allowable Stress w Allowable Working Sti 2s E = Joint Efficiency [ 7 -o
- per ASME Section VIII i- 'T Not less than 3/16 in. ?
4 e [ 2. Shell Stress -]>
* . Loads will not produce Tensile *
- Loads: Emergency excessive tensile or compressive (buckling) 10,000 psi 685 psi 9afe shutdown stresses.
1 Earthquake Nozzle Load. Compressive Stress Limit 5,190 psi 2,188 psi , ASME Section VIII , Compression 1/3. yield 8!E
' . e en O2
' es o (1) Operability Assurance Demonstration - Required-for active ~ components only. . . .
' y@
2 (2) This component has been evaluated using dynamic loads unique to the La Salle Project. The governing 58 load conbination listed in Table 3.9-25 has been applied-appropriately. The_ stresses calculated in this s evaluation are listed in the appropriate column. y$ .) a 6
,- .. 3, ,gu -w--,,.. - , , , ,e--.f~. ~.,w--.-, ..- - . - - - - - , ~ . , , , . . <w , . - - - < - 3..,,,:,--v-. g .
- n pn.
%s) (Jl f
TABLE 3.9-21 (Cont'd) CALCULATED CRITERIA METHOD OF ANALYSIS AIJ,0WABLE FORCES AND MOMENTS (3) VALUES
- 3. Nozzle Inads For the maximum moment Force in Ib moment in ft-lb due to pipe reaction, the maximum force shall not exceed the allowable.
Loads: Normal and Upset. Overflev Force: less than allowables
-Design Pressure and F< 770 - 1.57 M
-Temperature Z Moment: less Dead Weight, Thermal Outlet than allowables Expansion, and OBE Earthquake F< 770 - 1.57 M ta Y - $.
y Loads: Emergency Overflow 5' -
"1 w
$ Design Pressure and F< 925 - 1.57 M Force: less $
Temperature - than allowables Dead Weight, Thermal Outlet Expansion, and SSE ~ Moment: less Earthquake- F< 925 - 1.57 M than allowables Stress Limit ASME Section VIII for - Normal, Upset, and Emergency o a H OZ (3) The maximum allowable nozzle forces and moments were imposed with other appropriate loads to ED demonstrate acceptance. :n h
~n e
W r
- -------.:_.. --___A--.__-----_.--.-_--_----------.____a- _ - _ - - - . - - - - - _ _ - _ _ - - - -
- - - _ . _ _ - - _ -
- v __2
y -- p j"5 N) 'Y U TABLE 3.9-22 JET PUMP LOAD COMBINATION PRIMARY ALLOWABLE CALCULATED ACCEPTANCE CRITERIA' (See Table I) STRESS TYPE STRESS (psi) STRESS (psi) ASME B & PV Code g Section III. Primary Local Membrane plus bending for:
- 1. ASTMAS3/2 or ASTMA 463 (304SS) S,= 16,900 psi at 550' F, Table I - 1.2 ,
- 2. ASTM 9 166 (inconel 600)
S,= 23,300 psi at 550' F Table I - 1.2 t+ W Nortaal and Upset (1) 1. Appropriate Loading Local Membrane 25,400 1191 m~ e Combinations from. plus bending. O
/. (Service Levels A&B) Table 3.9-A were 4_
O 8 allowable-m used. p
= 25,400 psi (304SS) 5 Emergency Appropriate Loading Local' Membrane 38,000 2640 Combinations from plus bending.
(Service Level C) Table 3.9-A were used. b allowable = 2.25 S,
= 38,000 psi (304SS) e Faulted (2) Appropriate Loading Local Membrane 69,900 42,415 o combinations from plus bending.
(Service Level D) Table 3.9-A were o :e
= 69,900 used. -E E .
8 allowable " m psi (inconel 600) o
$w (1) Highest stressed location is in the 304SS material.
(2) Highest stressed location is in the inconel 600 material. _ _ _ - . _ _______.___.1. _= _ - _ ~_ - _ . .
(
^-- d
-\
.(j TABLE 3.9-23 ACTIVE ASME CLASS 1, 2, AND 3 PUMPS EQUIPMENT NUMBER EQUIPMENT DESCRIPTION ASME CODE CLASS C41-C001 Standby Liquid Control Pump 2 E12-C002 Residual Heat Removal Pump 2 E12-C003 , Residual Heat Removal Water Leg Pamp 2 E12-C300 _ Residual Heat Removal Service Water Pump 3 [
- y
+ O y E21-C001 Low-Pressure Core Spray Pump 2 7 5 E to E21-C002 Low-Pressure Core Spray Water Leg Pump 2 >
lc E22-C001 High-Pressure Core Spray Pump 2 E22-C002 HPCS Diesel Cooling Water Pump 3 E22-C003 HPCS Water Leg Pump 2 E51-c001 Reactor Core Isolation Coolant Pump 2 o>- E51-C003 RCIC Water Leg Pump 2 gg' o 2: DG0lP Diesel Cooling Water Pump -3 @C
- o $
D00lP Diesel-Generator Fuel Transfer Pump 3 $ ww 3 co ' ' - FC03P , Fuel Pool Emergency Makeup Pump
-ss- - -,,
-m-
Y" LSCS-FS'AR AMENDMENT 39 OCTOBER 1978 (] TABLE 3.9-24' LIST OF SAFETY-RELATED ACTIVE VALVES
- r
.i 1B21-F032A,B 1B21-F013U' lE12-F315' ' lE32-F001B J1833-F060A,B 1B21-F013S lE12-F317 1E32-F002A lE12-F041B 1B21-F013R lE12-F319 . 1E32-F002B c1E12-F050A . lB21-F013P ;lE12-F359B- lE32-F002D a lE12-F050B' 1921-F040U: 'lE12-F359A lE32-F002C
~
lE21-F006- 1B21-F040VD 1E12-F360B lE51-F013 J lE22-F005 1B21-F040D lE12-F360A - lE51-F063-lE51-F065' 1B21-F040C lE21-F003- . lE51-F064 "7 1E51'-F06 6 - 1B33-F078Bf lE21-F033 1G33-F001 ilB33-F019 1B33-F078A lE21-F304 1G33-F004:
'lB33-F020 ' lCll-F120A- 'lE22-F002 1G33- F040 '
1Cil-F010 1C11-F120B -lE22-F007- 1G33 F100~ . 1Cll-F011 - lE12-F331C- lE22-F016' 1G35-F101 lE12-F051A lE12-F331B- lE22-F024 1G33-F102 lE12-F051BL lE12-F331D .1E22-F304 1G33-F106-lE12-F065A - lE12-F331A ;1E12-F341- 1C41-F001B lE12-F065B lE21-F009 lE12-F031C 1C41-F001A-
.1E51-F004' lE22-F025 1E12-F031B lE12-F003A '
lE51-F005 lE22-F028 'lE12-F031A lE12-F003B 1E51-F025 . lF21-F040E- lE12-F046A lE12-F004A 4 p\_/-.
'lE51-F026 1B21-F040R' lE12-F046B . lE12-F004C 1B21-F013F 1B21-F040S 11E12-F046C lE12-F004B 1821-F013h 1B21-F024B: lE12-F054A lE12-F006B 1B21-F013L 1B21-F029D lE12-F054B lE12-F006A 1B21-F013K 1B21-F029C .lE12-F084C lE12-F0llB 1B21-F013J 1B21-F029B lE12-F084B 1E12-F0llA 1821-F013H 1B21-F029A 1E12-F084A lE12-F016B 1B21-F013G 1G33-F309 1B33-F307C lE12-F016A 1821-F013F 1G33-F312A 1B33-F309B 1E12-F017A 1B21-F013E 1G33-F312B 1B33-F309D 1E12-F017B 1821-F013D 1821-F024A 1B33-F309A lE12-F021 1B21-F013C 1B21-F024D: 1B33-F309C lE12-F024A 1B21-F022D 1B21-F024C 1B33-F319A. = lE12-F024B 1B21-F022C lE51-F021 1B33-F319B lE12-F026B 1D21-F022B lE51-F028 1C41-F033B -lE12-F026A 1821-F022A' lE51-F030 1C41-F033A lE12-F027B 1B21-F028D. 1E51-F04 0 1E12-F042A lE12-F027A 1B21-F028C 'lE51-F047 lE12-F042C 1E12-F040B 1B21-F028B lE51-F061 1E12-F053B 1E12-F040.A 1B21-F028A. 1E22-F342 lE12-F053A 1E12-F047B 1Cll-F121 lE32-F010 lE12-F099A lE12-F047A lE12-F041A lE32-F0ll lE21-F005 lE12-F048B.
lE12-F041C lE51-F0ll lE226F004 1E12-F048A' 1821-F013Br 1E51 "015 lE32-F001D lE12- F049 A 1B21-F013A lE51-F017 lE32-F001A lE12-F049B 1821-F013V 1E51-F018 .lE32-F001C- - lE12-F052A ~ 3.9-143 e v ,s. m._.,-p ,,aN..- ~~ ,. a -sm,r. , , . ,,a , ,y --,-~wi.-,,
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE 3.9-24'(Cont'd)- . lE12-F052B lE12-F336B 1B33-F013A 1B21-F355 lE12-F064C lE12-F336A 1B33-F013B 1B 21- F 35 7 1E12-F064A lE22-F319 1B 3 3-F 017A 1B 21- F 3 59 lE12-F064B 1B33-F023A 1B33-F017B 1B21-F361 lE12-F073A 1B33-F023B 1B33-F301A 1B21-F363 lE12-F073B 1B33-F067B 1B33-F301B 1B21-F365 1E12-F074A 1B 3 3-F0 6 7A 1B33-F305B 1821-F367 lE12-F074B 1B33-F074A 1B33-F305C 1B21-F370 lE12-F087A 1B33-F074B 1B33-F305D 1B21-F372 lE12-F087B 1Cll-F082 1B33-F305A ODOOO4 1DG036 lE12-F00B 1B33-F307D 1D0004 ODC002 lE12-F009 1833-F307A- 1DOO 14 - lHG007 lE12-F023 1B33-F307B 1DOO24 1HG008 lE12-F042B 1821-F374 OVG002B 1HG016 'lB21-F001 1B21-F376 OVG003A 1DG002 1B21-F002 1B21-F378 OVG003B 1E12-F093 1B21-F005 IB21-F380 lVG004A lE21-F001 1821-F016 1B21-F382 OVG004B lE21-F0ll 1B 21- F019 1B21-F437 .lVG001 lE21-F012 1B21-F067D 1B21-F439 lVG002 lE22-F001 1B21-F067C 1B21-F441 lVQO 37 - gm lE22-FC10 1B21-F067B 1B21-F443 lVQO38 IVP063B (') lE22-F0ll lE22-F012 1B21 '0( 7A 1B21-F065B 1B21-F445B 1B21-F445A lVP063A 1E22-F015 1B21-F065A 1B21-F447 IVQO21 lE22-F023 3E51-F076 1B21-F449 IVQ026A' lE3 2-F0 0 3D S:1-F045 1821-F451 lVQO26B lE32-F003C lE31-F005F4 1B21-F453 lVQO27B lE32-F003B lE31-F005C4 1821-F455B IVQO27A lE32-F003A 1E31-F005B4 1B21-F455A 1VQO28 lE32-F004A lE31-F005D1 1821-F457 1VQ029 ' lE32-F006 lE31-F005B3 1821-F459 1VQO30 lE32-F007 lE31-F005C3 1B21-F461 lVQ031 lE32-F008 lE31-F005B7 1B21-F463 lVQ033 1E32-F009 1E31-F005C2 1B21-F465B IVQD34 lE51-F010 lE31-F005G1 1B21-F327B IVQO36 lE51-F019 lE31-F005B2 1B21-F327C 1VQO39 lE51-F022' lE31-F005C6 18 21-F 32 7D LWR 029 lE51-F031 lE31-F005C1 1821-F327A LWR 040 lE51-F046 1B21-F465A 1B21-F328A ODG009 lE51-F059: 1B21-F467 1B21-F328D 1DG011 lE51-F068 1B21-F469 1B21-F328C 1DG035 lE51-F069 1B21-F471. 1B21-F328B OVG002A lE51-F080 1B21-F473 1B21-F344 1XX085 lE51-F086 1B21-F475A 1B21-F346 lHG001A 1E12-F068B 1B21-F475B 1B21-F348 lHGdOlB 1 lE12-F068A 1B21-F506A 1B21-F350 lHG002A (~) .lE12-F094 1B21-F506B 1B21-F353 1.HG002B \# lHG003A 3.9-144
LSCS-FSAR' AMENDMENT 39
. OCTOBER 1978 v3 I
\_) TABLE 3.9-24 (Cont'd) i
'lHG0038 lN62-F008A l lHG004A IN62-F015B '
lHG004B 1N62-F015A 1HG005B IN62-F030B
'lHG005A IN62-F030A 1HG006A .lN62-F085B 1HG006B IN62-F085A lHG009 1C41-F006 lVP053B 1C41-F007' lVP053A 1B21-F325D lVQOll 1321-F325C lVQ012 1821-F325A.
IVQ019 1B21-F325B i 1B21-F326D' IVQ023 lVQO25 1821-F326C IIN001B 1B21-F326A lIN001A 1B21-F326B IIN017A lE12-F019 1RE024' ; 1RE025-
-lRE026 1RE029
() 1RE030 1RE031 1RF005 1RF006 1FC033 1FC044A 1FC044B lE31-F005C5 lE31-F0 0 5F 3 lE31-F00EF2 lE31-F005BA lE31-F005F" lE12-F060B lE12-F060A lE12-F075A lE12-F075B lE12-F097 > lE31-F005F1 lE31-F005A5 lE31-F005A3-lE31-F005A4
.lE31-F005El lE31-F005A2 lE31-F005Al 1N62-F007B Q.-
1N62-F007A 1N62-F008B 3.9-145 a
.q
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE 3. 9-2 5 l O'"' DESIGN LOADING CONDITIONS AND COMBINATIONS OPERATING CONDITIONS DESIGN LOADING A_Np STRESS LIMITS
- CONDITIONS AND COMBINATIONS Normal and Upset N and Ap or N and U Emergency N and R or other conditions which have a 40-year encounter probability from 10-1 to 10-3 Fault N and Am and R or other conditions which have a 40-year encounter probability from 10-3 to 10-6 where: N = normal loads, U = upset loads excluding earthquake,
~
Ap = safe shutdown earthquake /2 includ-(]} ing any associated transients, EfE Am = safe shutdown earthquake (SSE) including any associated transients, R = automatic blowdown or equivalent auxil-iary pipe rupture loading including any associated transients - pipe rupture loadings are not directly considered on piping itself because this is handled by a fat. lure mode analysis, and R = primary loadings which result from rup-ture of a main steamline or a recircula-tion line. i
*The design stress, deformation, and' fatigue limits are:
- a. for RPV and appurtenances - ASME Section III; I b. for core support structures - Tables 3.9-27, 3.9-28, and 3,9-29; and r"' c. for reactor internal structures - Tables 3.9-23, k-} 3.9-24, 3.9-25, and 3.9-26.
3.9-146 (
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 ' TABLE 3.9-25
~ - LOAD COMBINATIONS AND ACCEPTANCE CRITERIA L/
- a. NSSS EQUIPMENT-Operating Number Load _ Combinations
- Condition Categories 1 N + SRV (ALL) Upset.
2 N + OBE Upset 3 N + SSE Faulted 4 N+ (OBE + SRV (ALL)) Emergency 5 N+ (SSE + SRV (ALL)) Faulted 6 N+ (SBA + SRV (2))~ Emergency 7 N+ (IBA + SRV (2)) Faulted 8 N+ (SBA + SRV (ADS))
. Emergency 9 N+ (SBA/IBA ' + SSE + SRV ~ (ADS) ) Faulted 10 N+ Faulted , (LOCA(1-6) + SSE)***
I\T 11 N+ (LOCA gy_7))** Faulted The design stress, deformation, and fatigue limits'are:
- a. for RPV and appurtenances - ASME Section III;
- b. for core support structures - Tables 3.9-30, 3.9-31, and 3.9-32; and
- c. for reactor internal structures - Tables 3.9-26, 3.9-27, 3.9-28, and 3.9-29.
- See Legend on the following pages for definition of terms and criteria for combining loads, rm Prom all initial conditions.
U *** From rated power initial conditions. 3.9-147
LSCS-FSAR AMENDMENT'39 OCTOBER 197c f TABLE 3.9-25 (Cont'd) i'
- b. NSSS PIPING AND PIPE MOUNTED EQUIPMENT NO. LOAD COMBINATIONS SERVICE LEVEL 1 Normal Operating + Operating Basis Design Condition l Earthquake j 2 Normal Operating,. Operating Basis Earthquake, Operating Transients A,B 3 Normal Operating + Operating Transients
+ Operating Basis Earthquake C 4 Normal Operating + Small Break Loss of Coolant Accident and Associated' Operating Transients C 5 Normal Operating + Infrequent Operating Transients C 6 Normal Operating + Operating Transients +
Safe Shutdown Earthquake D 7 Normal Operating + Large Break Loss of Coolant Accident + Safe Shutdown () Earthquake D 8 Normal Operating + Intermediate Break Loss of Coolant Accident and Associated Operating Transients + Safe Shutdown Earthquake D 9 Normal Operating + Large Break Loss of Coolant Accident (Annulus Pressurization) D 10 Normal Operating + Safe Shutdown Earthquake
+ Large Break Loss of Coolant Accident (Annulus Pressurization) D rh 3.9-148
LSCS-FSAR AMENDMENT 39 OCTOBER 1978
<s TABLE 3.9-25 (Cont'd)'
V LOAD DEFINITION LEGEND Normal (N) - Normal and/or abnormal loads depending on acceptance criteria. OBE - Operatinal basis earthquake loads. SSE - Loads due to vibratory motion from safe shutdown earthquake loads. SRV (2) - Safety / relief valve discharge induced loads from 2 adjacent valves. SRV (ALL) - The loads induced by actuation of all safety / relief valves which activate within milliseconds of each other (e .g. , turbine trip operational transient).
-SRV (ADS) - The loads induced by.the actuation of safety /
relief valves associated with automatic depres-surication system which actucte within milli-seconds of each other during the postulated small or intermediate size pipe rupture. LOCA - The loss of coolant accident associated with the postulated pipe rupture of large pipes (e . g . , main steam, feedwater, recirculation piping). LOCA y - Pool swell drag / fallout loads on piping and compo- j nents located between the main vent discharge outlet l and the suppression pool water upper uurface. LOCA - Pool swell impact loads acting piping and components l 2 located above the suppression pool water upper surface. l LOCA - Oscillating pressure induced loads on submerged piping 3 and components during condensation oscillations, i.e., chugging. LOCA - Building motion induced loads from chugging l 4 (condensation oscillation). l l LOCA - Building motion induced loads from main vent air S clearing. LOCA - Vertical and horizontal loads on main vent piping. 6 LOCA - Annulus pressurization loads. 7 p/-
'- SBA - Small break accident.
IBA - Intermediate break accident. 3.9-149
1 LSCS-FSAR- AMENDMENT 39 OCTOBER 1978 (~g TABLE'3.9-26 l u DEFORMATION LIMIT
'(For Reactor Internal Structures Only)
EITHER ONE OF (NOT BOTH) GENERAL LIMIT
~ ~
- a. Permissible deformation,'DP- , _ 0.9 Analyzed deformation - SF
_ causing loss of function, DL. - min b.." Permissible deformation, DP ** 1.0 Experiment deformation - SF _ causing loss ofL function, DE. min t where: DP' = permissible deformation under stated condi-tions of normal, upset, emergency, or fault DL =-analyzed deformation which could cause a-system . loss of function; **
, DE ,= experimentally determined deformation which
()_ could cause a system loss of function; and ' SF = minimum safety factor min
- Equation b was not used because equation a criterion was met.
** Loss of function can only be defined quite generally until f attention is focused on the component of interest. In cases of interest, where deformation limits can affect the. function !
of equipment and components,:they will be specifically de- l lineated. .FromLa practical viewpoint, itfis convenient:to 1 interchange some deformation' condition at which function-is assured with the loss of function condition if the' required safety margins from the functioning. conditions can be achieved. f Therefore, it is often unnecessary tcr determine the actual '
. loss of-function condition because this interchange procedure produces conservative and' safe' designs. ' Examples where de- !
' formation limits' apply are: control rod drive alignment and. j clearances for proper' insertion', core support deformation i g- causing fuel disarrangement, or excess leakage of any com- i qs)s , ponent, j 5
- 1 3.9-150 l s
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 PRIMARY ST ESS LIMIT (For Reactor Internal Structures Only) ANY ONE OF (NO MORE THAN ONE REQUIRED) GENERAL LIMIT
- a. " Elastic evaluated primary stresses, PE' 1 2.25
, Permissible primary stresses, PN SF min
- b. ' Permissible load, LP -
. , 1. 5
_ Largest lower bound limit load, CL, - SF min
- c. " Elastic evaluated primary stress, PE 1 0.75 Conventional ultimate strength SF
_at temperature, US m'in < d.F Elastic-plastic evaluated nominal primary stress, EP 10.9 Conventional ultimate strength SF
.at temperature, US min
- e. ' Permissible load, LP ~* 10.9
(~g , Plastic instability load, PL , SF min V '
- f. Permissible load, LP *
< 0. 9 Ultimate load from fracture - SF
_ analysis, UF min
~
- g.1" Permissible load, LP 11.0 Ultimate load or loss of function SF 1 _ load from test, LE mn where:
PE = primary stresses evaluated on an elastic basis. The effective membrane stresses are to be averaged through the load carrying section of interest. The simplest average bending, shear, or torsion stress distribution which will sup-port the external loading will be added to the membrane stresses at the section of interest. PN = permissible primary stress levels unSer normal or upset conditions under ASME Boiler and Pres-sure Vessel Code, Section III.
-s Equations e, f, and g were not'used because criteria "a, b, j
) and c" were met.
3.9-151
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 cs TABLE 3.9-27 (Cont'd) (_) LP = permissible load under stated conditions of normal, upset, emergency, or fault. CL = lower bound limit load with yield point equal to 1.5 S where Sm is the tabulated value of allowable stress atm temperature of the ASME-III code or its equivalent. The lower' bound limit load is here defined as that produced from the analysis of an ideally plastic (nonstrain hardening) material where deformations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined material yield strength using either a shear theory or a strain , energy of distortion theory to relate multiaxial yield to.the uniaxial case. US = conventional ultimate strength at temperature or loading which would cause a system malfunction, whichever is more limiting. EP = elastic plastic evaluated nominal primary stress. Strain hardening of the material may be'used for the actual mono-tonic stress strain curve at the temperature of loading or any approximation to the actual stress strain curve (~')
%- which everywhere has a lower stress for the aamo strain .
as the actual monotonic curve may be used. Either the shear or strain energy of distortion flow rule may be used. PL = plastic instability load. The plastic instability load is defined here as the load at which any load bearing section begins to diminish its cross-sectional area at a faster rate than the strain hardening can accommodate the loss in area. .This type analysis requires a true stress-true strain curve or a close approximation based on monotonic loading at the temperature of loading. UF = ultimate load from fracture analyses. For components which involve sharp discontinuities (local theoretical stress concentration <3) the use of a fracture mechanics analysis where applicable, utilizing measurements of plane strain fracture toughness may be applied to compute c fracture' loads. Correction for finite plastic zones and thickness' effects as well as gross yieldini may be necessary. The methods of linear elastic stre,a analysis may. be used in the fracture analysis where its use is clearly conservat.ive-or supported by experimental' evidence. _ Examples where fracture mechanics may be fm applied are for fillet welds ~or end of fatigue life (_) . crack propagation. 2 3.9-152 n e
- o. . _ . . _ . _ - _
LSCS-FSAR ~ AMEN 9 MENT 39 OCTOBER 11978
/~Y TABLE 3. 9 (Cont ' d)'-
u
' LE = ultimatelload: or -loss of function load as determined.
from experiment. .In-using this. method, account shall be taken of the dimensional-tolerances,which'may exist between.the actual partLand the tested part or parts as well-as differences which may exist:in.the. ultimate
' tensile strength of the actual part and the tested parts.
The' guide to be used in each of'these areas is that the experimentally' determined-load'shall use. adjusted values to account:for material property and dimension varia-' tions, each of which has no greater probabilityLthan 0.1-of'being. exceeded in the actual part. r i O i I I i
? . A. '
t 3.9-153 n';
; -- ,- .m--. . . . , - - , . .. ~ , , , , . , - . . - ,, . - - . . . - - - - - - - -m .- . .-
Y LSCS-FSAR ' AMENDMENT 39
- OCTOBER:1978 s
TABLE'3.9-28 l
! '/'Vd' BUCKLING STABILITY LIMIT' (For Reactor' Internal Structures'Only)
ANY ONE OF (NO MORE THAN ONE REQUIRED) GENERAL LIMIT
~ -
- a. Permissible: load, LP 2.25 .
Code. normal event permissible load, PN 1 SFmin,
~
- b. ~ Permissible load, LP- 0.9 Stability analysia load, SL ,
1 SF f Permissible load,..LP "
- c. .l.0 Ultimate buckling collapse load from: test, SE -.SF
- min
-where:
LP = permissible. load under stated conditions of normal,
. upset, emergency,.or fault.
PN = applicable' code normal event permissible-load. () SL'=-stability. analysis load. The' ideal buckling analy-sis is often sensitive to otherwise minor deviations from. ideal geometry and; boundary conditions. 'These-effects shall be-accounted for in the analysis of the buckling stability loads. Examples of.this are ! ovality.in externally pressurized shells or eccen-tricity on column members.. SE = ultimate buckl'ing. collapse load as determined from , experiment. In.using this. method, account shall be )
.taken of the dimensional toleraness which may exist I between the~ actual part and the tested part. The guide to be used in each of these areas is'that the experimentally-determined load sha]1 be adjusted to account for material' property and dimension varia-tions, each of which has no greater probability than 0.1 of being exceeded in the actual part.
l J ff'1 Y
- Equation c was not used because criteria "a and b" were met.
3 - L 3.9-154 l o
. c 'LSCS-FSAR- -AMENDMENT 39 OCTOBER 1978 TABLE 3.-9-29 e FATIGUEiLIMIT:
1
- (For uReactor. Internal Structures' Only) .;
1 Summation of~fabigue damage usage with design and. operation loads following, Miner hypotheses . . .
. LIMIT FOR NORMAL ~-
AND UPSET DESIGN. ANY ONE OF (NO MORE THAN ONE - REQUIRED)' CONDITIONS
- a. .Mean . fatigue **,t.. cycle usage 1
'from analyses .
< 0.05' l
l b.- Mean' fatigue **,'t? cycle usage' . .
'from test _.0.33
- c. ' Design' fatigue-cycle usage.from analysis:using the method of-Table: 3.9-30
< 11.0 I )
l l 1
- M.A. Miner, " Cumulative Damage'in Fatigue'," Journal of -
Applied Mechanics,'Vol. 12, Vol. 67, pp A159-A164, ASME, September 1945.
** Fatigue failure is defined here~as a~2'%5 area reduction for a load carrying member which is-required'to function, L
or excess leakage, whichever is more limiting. t Equations a and b were not used.. .(~(
\
..N J l l 3.9-155 q 1
'1
, ,,.+.e. , . , ~ . . , ,-a,, -. , . . . . < . . . . - . . - , . . ~c - + . r~, -c -
n- - ~ ~ ~ - - - -- - n~
-+-- Y
f .
') I.
'~
. TABLE.3.9-30.
CORE SUPPORT STPUCTURES STRESS CATEGORIES AND' LIMITS OF-STRESS' INTENSITY'FOR NORMAL'AND UPSET CONDITIONS
' STRESS PRIMARY STRESSES SECONDARY STRESSES PEAK STRESSES. ~
CATEGORY.- MEMBRANE, P m
' BENDING.Pb MEMBRANE & BENDING .
PEAK,'F (NOTES - 4, 7, AND 8) (NOTES 14, - 7, . AND 8) . SECONDARY, Q (NOTES.2,'.4, AND.6) (NOTES 2,L4, AND 6) P P +P w m b P m +Pb + Q. P
+Pb+Q+F -
t_ ~ o j w ~
~3 S, Y
m 1.55m ' g l- . to NORMAL
- AND -- ELASTIC- ELASTIC ELASTIC . ELASTIC UPSET or : ANALYSIS or ANALYSIS 'or ANALYSIS : FATIGUE' E
(NOTE 6)- (NOTE : 6) (NOTE 1) D 9)- L 0.67L S L g ( LIMIT LIMIT- . _ PLASTIC
'or ANALYSIS or ANALYSIS' orf~ ANALYSIS (NOTE 10) (NOTE 10) '(NOTE i) 0.44L u 0.44L P+PD+Q+F S
O
- FOR CYCLES o
'LESS THAN' ELASTIC- oz=
TEST- TEST 1000,'.USE PLASTIC @y 1 (NOTE-11) - (NOTE ' ll) PEAK FATIGUE N g-l : (NOTE 12) - (NOTES 3', se i 9,.AND,12)
- yg i'
co e .
, _ - _ _ _ - - - - , ' ~ . -
P k
-LSCS-FSAR - AMENDMENT 39 OCTOBER 1978 ); TABLE 3. 9-30 .'(Cont ' d) l NORMAL AND UPSET CONDITIONS-NOTE l- - This. limitation applies to the range of stress :
intensity. When the secondary stress is due to ! a te erature excursion at the point at which the stre es are being analyzed, the value of Sm shall 6e taken as the average of the Sm values tabulated in Tables I-1.1, I-1.2, and I-1.3 of ASME Boiler and Pressure Vessel Code, Section III (ASME . III) for.the highest and the lowest tempera-ture_of the metal during the transient. When part of.the secondary stress is due to mechanical load, the value of Sm shall be taken as the S m value ! for the highest temperature of the metal during the transient. NOTE 2 - The stresses in Category 0 are those parts of the total stress which are produced by thermal gradi-ents, structural discontinuities, etc., and do not include primary stresses which may also exist at the same point. It should.be noted, however, that s a detailed stress analysis frequently.givee the
/') combination of primary and secondary stresses direct-
'\~/ ly and~, when' appropriate, this calculated value rep-resents the total of Pm + Pb + 0 and not 0 alone. Similarly, if the stress in Category F is produced by a stress concentration, the quantity F is the additional stress produced by the notch, over and above the nominal stress. For_ example, if a plate has a nominal stress intensity, Pm=S, Pb=0, 0 = 0, and a notch with a stress concentration K is introduced, then F = Pm (K-1) and the peak stress M intensity equals Pn, + Pm (K-1) = KPm* NOTE 3 - Sa is obtained from the fatigue curves, Figures I-9.1 and I-9.2 of ASME III. The allowable stress intensity for the full range of fluctuation is 2 Sa* NOTE 4 - The symbols Pm, Pb, 0, and F do not represent single quantities, but rather sets of six quantities rep-resenting the six stress components otr oli c r' Ttl' T1r' Trt' NOTE 5 - S n denotes,the structural action of shakedown load as defined in Paragraph NB-3213.18 of ASME III cal-culated on a plastic basis as applied to a specific'
- (~) location on the structure.
u N' 3.9-157
- . - - _ . . - .. . , . . . - . - . _-_ _- ~
l 1 LSCS-FSAR AMENDMENT 39 l OCTOBER 1978 l l /^ (] TABLE 3.9-30 (Cont'd) l NOTE 6 - The triaxial stresses represent the algebraic sum of the three primary principal stressec (al + a2 +
- 03) for the combination of stress components.
Where uniform tension loading is present, triaxial stresses are limited to 4 S m-NOTE 7 - For configurations where compressive stresses occur, the stress limits shall be revised to take into account critical btz.ekling stresses (see paragraph NB-3211 (c ) of ASME III). For external pressure, the permissible equivalent static external pressure shall be as specified by the rules of Paragreph NB-3133 of ASME III. Where dynamic pressures are involved, the permissible external pressure shall be limited to 25% of the dynamic instability pres-sure. NOTE 8 - When loads are transiently applied, consideration Ghould be given to the use of dynamic load ampli-fication, and possible change in modulus of elasticity. (3 NOTE 9 - In the fatigue data curves, where the number of (_/ operating cycles are less than 10, use the sa value for 10 cycles; where the number of operating cycles are greater than 106 , use the Sa value for 106 cycles. NOTE 10 - L L is the lower bound limit load with yield point equal to 1.5 S m (where Sm is the tabulated value of allowable stress at temperature as contained in ASME III). The lower bound limit load is here defined as that produced from the analysis of an . ideally plastic (nonstrain hardening) material where deformations increase with no further in-crease in applied load. The. lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined mate-rial yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the uniaxial case. NOTE 11 - For normal and upset conditions, the limits on < primary membrane plus primary bending need not be satisfied in a component if it can be shown from the test of a prototype or model that the specified loads (dynamic or static equivalent) do not exceed 44% of Lu, where Lu is the ultimate load or the I) maximum load or load combination used in the test. 3.9-158
.w
t t
.LSCS-FSAR LAMENDMENT 39 LOCTOBER 1978-l i i
, Y! '
k#' TABLE 3.9-30 (Cont,d) . f In.usingLthis method, account shallLbe taken ofL the = size ef fect and dimensional. tolerances which may!. exist;between the. actual part'and'theTtest
~
part, or. parts, as well as differences which~may , exist in-the. ultimate'strengtn or other governing material. properties _of the actual'part'and-the tested'part!to assure.that the loads ~obtained from thel test: are a conservative . representation ofl.the -
' load' carrying < capability of the actual' component under the1 postulated loading for normal and upset conditions.
' NOTE 12 The allowable value for the maximum 1 range of this .
- stress lintensityfie 3Sm excep0 for.. cyclic events
~
l which occur less than 1000ctimes during the design-life of the' plant. ' For this exception,'inflieu of meeting the 3Sm limit, an. elastic-plastic fatigue-analysis in'accordance with'ASME.III'may.be.per-
~
formed to demonstrate that-the cumulative l fatigue-usage-attributable to'the combination of these low i events, plus all other. cyclic events, does not ex-ceed a. fatigue usage _value of 1.0 ; . i t t i i 1 r h
.-t
?~./ :' ,
t I-t I s
- cd 3.9-159 1
00; O O TABLE 3.9-31 CORE SUPPORT STRUCTURES STRESS CATEGORIES.AND LIMITS OF' STRESS-INTENSITY FOR EMERGENCY CONDITIONS
,I l PRIMARY STRESSES SECONDARY STRESSES PEAK STRESSES 1,.- STRESS' i .
MEMBRANE,.Pm- BENDING, PB- MEMBRANE-AND BENDING PEAK-CATEGORY 'tNOTES l', 2, AND 10) (NOTES 1, 2, AND 10) SECONDARY, Q F P+P m B ELASTIC ELASTIC
-ANALYSIS. 2.25 Sm ANALYSIS 1.5S m (NOTE 3) (NOTE 3)
. 'OR OR b
-LIMIT LIMIT 0
' :f e -- L ANP1YSIS Lt ANALYSIS 4
.. 4 (NOTE 4)
OR (NOTE 4) cn
-o OR EMERGENCY PLASTIC PLASTIC 1.SS ANALYSIS 2.25 Sm ANALYSIS EVALUATION EVALUATION (NOTE 9)
(NOTE 6) . (NOTES - 5 :NOT REQUIRED NOT.' REQUIRED [ OR AND.'6) ' O TEST
- 0.5 S u (NOTE 5)
(NOTE 7) . OR- . - OR STRESS- TEST. S T -8 $ ~ i E 0.6 Le ~ (NOTE 7) 8M ANALYSIS @g j (NOTE 8) OR trj 3: ' STRESS- W$ .i , RATIO i*8 l KS E ANALYSIS $G (NOTE 8) z_ _ ._ - ._. . _ - _. _. _ . - _ _ _ _ __ _ ._ _ _ - - 1
LSCS-FSAR AMENDMENT 39 OCTOBER 1978
) TABLE 3.9-31 (Cont'd) l EMERGENCY CONDITIONS NOTE 1 -
The symbols Pm, P 3, Q, and F do not represent single quantities, but rather sets of six quan-tities representing the six stress components Utr 01, a rr Ttli Tlrr T rt* NOTE 2 - For configurations where compressive stresses occur, stress limits shall be revised to take into account critical buchling stresses. For external pressure, the permissible equivalent static external pressure shall be taken as 150% of that permitted by the rules of Paragraph NB-3133 of ASME Boiler and Pressure Vessel Code, Section III (ASME III). Where dynamic pressures are in-volved, the permissible external pressure shall satisfy the preceding requirements or be limited to 50% of the dynamic instability pressure. NOTE 3 - The triaxial stresses represent the algebraic sum of the three primary principal stresses (al + 0 2 + a3) for the combination of stress components. t,_) Where uniform tension loading is present, triaxial
~'
stresses should be limited to 6 Sm-NOTE 4 - L7 is the lower bound limit load with yield point equal to 1.5 Sm (where S m is the tabulated value of allowable stress at temperature as contained in ASME III). The lower bound limit load is here defined as that produced from the analysis of an ideally plastic (nonstrain hardening) material where deformations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere ex-ceeds the defined material yield strength using either a shear theory or a strain energy of distor-tion theory to relate multiaxial yielding to the uniaxial case. NOTE 5 - S u is the ultimate strength at temperature. Multi-axial effects on ultimate strength shall be con-sidered. NOTE 6 - This plastic analysis uses an elastic-plastic evaluated nominal primary stress. Strain harden-ing of the material may be used for the actual monotonic stress-strain curve at the temperature jg of loading or any approximation to the actual stress-strain curve which everywhere has a lower l 3.9-161
LSCS-FSAR AMENDMENT 39 / OCTOPER 1978
- lll TABL: '3.9-31 (Cont'd) stress for the same strain as the actual monotonic ,
curve may be used. Either the shear or strain energy of distortion flow rule shall be used to account for multiaxial effects. NOTE 7 - For emergency conditions, the stress limits need not be saticfied if it can be shown from the test of a prototype or model that the specified loads (dynamic or static equivalent) do not exceed 60% of Le, where L e is the ultimate load or the maximum load or load combination used in the test. In us-ing this method, account shall be taken of the size effect and dimensional tolerances which may exist be-tween the actual part and the tested part or parts as well as differences which may exist in the ulti-mate strength or other governing material proper-ties of the actual.part and the tested parts to assure that the loads obtained from the test are a conservative representation of the load carrying capability of the actual component under postu-lated loading for emergency conditions. NOTE 8 - Stress ratio is a method of plastic analysis which uses the stress ratio combinations (combination or stresses that' consider the ratio of the actual stress to the allowable plastic or elastic etross) co compute the maximum load a strain hardening s material can carry. K is defined as the section / factor; Se < 2Sm for primary membrane loading.
' NOTE 9 -
Where deformation is of concern in a component, the deformation shall be limited to two-thirds the value given for emergency conditions in the design specification. NOTE 10 - When loads are transiently applied, consideration should be given tc 'he use of dynamic load ampli-fication and possible change in modulus of elas-ticity. O 3.9-162
?,A ..
.c.
- .r- ,
. s ., , ..
'T
-TABLE.3.9-32 ,
~
- CORE 1 SUPPORT STRUCTURES STRESS CATEGORIES'AND-~ LIMITS- ,
. OF STRESS INTENSITY FOR FAULT CONDITIONS-PRIMARY STRESSES- SECONDARY STRESSES- PEAK STRESSES SiTRESS MEMBRANE,.P m .
BENDING, PB- MEMBRANE & BENDING PEAK , CATEGORIES (NOTES'1, 2, and 3) -(NOTES 1, 2,.and 3)' SECONDARY,: Q F Pm Pm+PB 2.4 Sm ELASTIC 3.0 S m M STIC
-ANALYSIS ANALYSIS s
OR 0.75 S u (NOTE 5) : LIMIT' l.33 L ANALYSIS OR' .(NOTE 4) M v1 - y LIMIT OR O
$ FAULT L AN M SIS . EVALUATION EVALUATION- k
" (NOTE.4) S NOT REQUIRED NOT REQUIRED- '
(NOTE 5) _ (NOTES 5 A 6)
-PLASTIC 0.67 S u ANALYSIS OR (NOTES 5 AND 6)
[0.8 Lp TE- S-OR -(NOTE 7) OR TEST 0.8 h (NOTE D STRESS- 8 OR KS RATIO @2 . F ANALYSIS mo STRESS- (NOTE 8) $$
- i. / g RATIO g$
F ANALYSIS e (NOTE 8) '$$
. , . , ., -3 f .
.. .Ep- _ , . . _, ,,u ,.[,_,_,., -
LSCS-FSAR AMENDMENT'39 OCTOBER 1978 l ('} s-
. TABLE 3. 9-32 (Cont ' d)
NOTE 1 - The symbols Pm, Pb, Q, and F do not represent single quantities,'but ratner sets of six quanti-
. ties representing the six stress components, y t' 01, Or' T lr, and Trt' NOTE 2 -
When loads are transiently applied, consideration should be given to the use of dynamic load ampli-fication and possible changes in modulus of elas-ticity. NOTE 3 - For configurations where compressive stresses occur, stress limits take into account critical buckling stresses. For external pressure, the permissible equivalent static external pressure shall be taken as 2.5 times that given by the. rules of paragraph NB-3133 of ASME Boiler and Pressure Vessel Codes Section III (ASME III). Where dynamic pressures are involved, the per-missible external pressure shall satisfy the pre-ceding requirements or shall be limited to 75% of the dynamic instability pressure. NOTE 4 - LL is the lower bound limit load with yield point () equal to 1.5 Sm (where Sm is the tabulated value of allowable stress at temperature as contained in ASME III). The lower bound limit load is here defined as that produced from the analysis of an ideally plastic (nonstrain hardening) material where deformations increase with no further in-crease in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined mate-rial yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the uniaxial case. NOTE 5 - Su is the ultimate strength at temperature. Multiaxial effects on ultimate strength shall be considered. NOTE 6 - This plastic _ analysis uses an elastic-plastic evaluated nominal primary stress. Strain hard-ening of the material may be used for the actual ! monotonic stress-strain curve at the temperature l of loading, or any approximation to the actual stress-strain curve which everywhere has'a lower stress for the same strain as the actual curve may be used: either the maximum shear stress or (% strain energy of distortion flow rule shall be
\~) used to account for multiaxial effects.
3.9-164 l
LSCS-FSAR AMENDMENT 39 OCTOBER 19 78 (~T ~ BLE 3.9-32 (Cont'd) s/ i NOTE 7 - For fault conditions, the stress limits need not be satisfied if it can be shown from the test of a prototype or model that the specified loads (dynamic or static equivalent) do not exceed 80% of Lp, where Lp is the ultimate load or load combination used in the test. In using this method, account shall be taken of the size effect and dimensional tolerances as well as differences which may exist in the ultimate strength or other governing material properties of the actual part and the tested parts to assure that.the loads obtained from the test are a conservative repre-sentation of the load carrying capability of the actual component under postulated loading for fault condition. NOTE 8 - Stress ratio is a me'thod of plastic analysis which uses the stress ratio combinations (combination of stresses that consider the ratio of the actual stress to the allowable plastic or elastic stress) to compute the maximum load a strain hardening material can carry. K is defined as the section (,) factor; Sf is the lesser of 2.4 Sm or 0.75 Su for primary membrane loading. NOTE 9 - Where deformation is of concern in a component, the deformation shall be limited to 80% of the value for fault. conditions in the design speci-fications. O 3.9-165
. . - . . - -. . - ~. _- .-_. . _ . . . .. . ..
LSCS-FSAR AMENDMENT 39 I OCTOBER 1978 TABLE 3.9-33 APPLICABLE THERMAL TRANSIENTS (Prestartup Hydro Test 130 Cycles Condition - Test) INITIAL FINAL TEMP. RATE' ATEMP. -
-PIPELINE TEMP. *F -
TEMP. *F
- F/ HOUR *F Main steamline 70 100 60 30 i
Head spray (RHR) 70 100 60 30' i Feedwater 70 100 60 30 i j Recirculation suction 70 100 .60 30 'l Recirculation discharge 70 100 60 30 ~! i r
' Core spray 70 100 60 30 CRDHS return 70 100 60 30 Standby liquid control 70 100 60 30 10-minute duration 100 50 Step 50 50 100 Step 50 Bottom drain 70 100 60 30 1
i Note':- Aft'er. temperature is raised to~100 F reactor pressure is increased to 1250 psig
,'j {) . and then-decreased to O psig. i l
l 3.9-166- !
LSCS-FSAR' AMENDMENT 39 l OCTOBER 1978 ' j"3 TABLE 3.9-33'(Cont'd)
%).
(Applicable Thermal Transients Startup l 120 Cycles Condition - Normal) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. *F TEMP. *F
- F/ HOUR 'F Main steamline 100 552 100 452 l
Head spray (RHR) 100 552 100 452 Feedwater 100 552 100 452 552 50 Step 462 i
.50 420 660 330. "
Recirculation suction 100 552 100 452 552 544 Step 8 544 528 32 16 Recirculation discharge 100 552-() 552 544 100 Step 452 i 8 ; 544 528 32 16 Core spray 100 552 100 452 552 544 Step 8 544 528 32 16 CRDHS return 100 50 Step 50 1 Standby liquid control 100 400 100 300 f i 400 544 Step 144
^
544 528 32 16' Bottom drain 100 400 100 300 400 544 Step 144 544 528 32 16 O 3.9-167
- - - . -.-. . - . . ._. - . - = - . -
LSCS-FSAR AMENDMENT 39 OCTOBF 1978 TABLE 3.9-33 (Cont'd) (Applicable Thermal Transients Daily Power Reduction and Rod Pattern Change 10,400 Cycles Condition - Normal) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. 'F TEMP. *F
- F/ HOUR 'F Main steamline 552 552 No Change Head spray 552 552 No Change Feedwater 420 354 264 66 354 420 264 66 Clean return 436 436 No Change Recirculation suction 528 528 No Change Recirculation discharge 528 528 No Change Core spray 528 528 No Change CRDHS return 50 50 No Change O Standby liquid control 528 528 No Change Bottom drain 528 528 No Change Note: Reactor pressure remains at 1000 psig.
() 3.9-168
LSCS-FSAR AMENDMENT 39 - OCTOBER'1978 l
;(]~) . TABLE.3.9-33 (Cont'd)
'(Applicable Thermal Transients' Weekly Power Reduction 2000 Cycles Condition - Normal) i INITIAL FINAL TEMP. ; RATE 'ATEMP. !
PIPELINE TEMP. *F TEMP. *F :* F/ HOUR
- F-Main steamline' 552 552 No Change '
Head spray (RHR) 552 552. 5:o Change ; Feedwater 420 324 192 96: 324 420 192 96-Cleanup return- 436 436 No Change Recirculation suction 528 528 No Change I Recirculation discharge 528 528 No Change Core spray 528 528 No Change CRDHS return - 50 50 No Change Standby liquid control 523 528 No Change. Bottom drain 528 528 No Change i Note: Reactor pressure remains at 1000 psig.- ( )- i 3.9-169
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 () TABLE 3.9-33 (Con t ' d) (Applicable Thermal Transients Turbine Trip _ 100% Bypass 10 Cyc?.es Condition - Upset) INITIAL FINAL TEMP. RATE ATEMP . PIPELINE TEMP. F TEMP. *F F/ HOUR F Main steamline 552 552 No Change Head spray (RHR) 552 552 No Change Feedwater 420 100 Step 320 100 420 4800 320 Recirculation suction 528 496 1280 32 496 528 480 32 Core spray 528 496 1280 32 496 528 480 32 CRDHS return 50 50 No Change ,} { Standby liquid control 528 496 1280 32 496 528 480 32 Bottom dr ain 528 496 1280 32 496 528 480 32 i Note: Reactor pressure remains at 1000 psig. 3.9-170
4
' LSCS-FSAR AMENDMENT 39 OCTOBER 19 78
. TABLE 3.9-33 (Cont'd)
(Applicable Thermal Transients Feedwater Heater Loss Partial Heater-Bypass 70 Cycles Condition - Upset) INITIAL FINAL TEMP. RATE ATEMP. l PIPELINE TEMP. *F TEMP. *'F
- F/ HOUR *F Main steamline 552 552 0 0 1 Head spray' (RHR) 552 552 0 0 Feedwater 420 265 6200 155 265 420 3100 155 Recirculation suction 528- 518 Step 10 518 528 Step 10 Recirculation discharge 528 518 Step 10 518 528 Step 10 Core spray 528. 518 Step 10 O 518 528 Step 10 CRDHS return 50 50 0 0 Standby liquid control 528 518 Step 10 518 528 Step 10 Bottom drain 528 518 Step 10 518 528 Step 10 l
l Note: Reactor pressure remains {} at 1000 psig. ! l l 3.9-171
LSCS-FSAR AMENDMENT 39 i OCTOBER 1978 ! j} TABLE 3.9-33 (Cont'd) (Applicable Thermal' Transients Scram G Trip
.Feedwater On MSIV Open 40 Cycles Condition - Upset)
INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. *F TEMP. *F
- F/ HOUR *F Mainisteamline 552 565 4680 13 565 538 6500 27 538 400 100 138 400 552 100 152-Head spray (RHR) 552 565 4680 13 565 538 6500 27 538 400 100 138 400 552 100 152 Feedwater 420 275 8700 145 275 100- 700 175 100 250 Step 150 250 420 340 170 Recirculation suction 528 400 100 128 400 552 100
- 0. 552 544 Step 152 8
544 528 32 16 Recirculation discharge 528 400 100 128 400 552 100 152 552 544 Step 8 544 528 32 16 Core spray 528 400 100 128 400 552 100 152 552 544 Step 8 544 528 32 16 CRDHS return- 50 50 No Change 10 Cycles only ( 50 528 Step 478 ( 528 50 Step 478 Standby liquid control 528 250 200 278 250 400 100 150 400 544 Step 144 544 528 32 16 e Bottom drain 528 250 200 278
.250 400 100 150 :
400 544 Step 144 ! -() 544 528 32 16
-Note:
' Reactor pressure increases to 1125 psig all relief valves open; pressure decreases to 240 psig and then increases to 1000 psig.
3.9-172
LSCS-FSAR- AMENDMENT 39 OCTOBER 1978
/"Y TABLE 3.9-33-(Cont'd).
Q; (Applicable Thermal Transients' All Other Scrams 140 Cycles Condition - Upset) INITIAL FINAL' TEMP. RATE ATEMP. PIPELINE- TEMP. 'F TEMP. *F
- F/ HOUR *F Main steamline 552 538 3360. 14- '
538 400 100 138 400 552 100 152 Head spray (RHR) 552 538 3360 14 538 400 100- 138 400 552 1100- 152 Feedwater 420 275 8700 145 275 100 700 175 100 -250 Step 150
'250 420 340 170 Recirculation suction 528 400 100 128 ,
400 .552 100 152 552 544 Step 8 544 528 32 16 () Recirculation discharge 528 400 100 128 400 552 100 152 552 544 Step '8 544 528 32 16 Core spray 528 400 100 128 400 552 100 152 552 544 Step 8 544 528 32 16 CRDHS return 50 50 No Change 4 30 Cycles only ( 50 528 Step 478 ( 528 50 Step 478 Standby liquid control. 528 250 200 278 l 250 400 100 150 400 544 Step 144 544 528 32 16 Bottom drain 528 250 200 278 l 250 400 100 150 400 544 Step 144 544 528 32 16 (m As Note: Reactor pressure decreases to 240 psig and then increases to 1000 psig. 3.9-173
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 o TABLE 3.9-33 (Cont'd): (Applicable Thermal Transients Rated Power I See Below for Events Condition - Normal) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. *F TEMP. *F
- F/ HOUR F f Main steamline 552 552 No Change 0 r Head spray (RHR) 552 552 No Change 0 Feedwater 420 420 No Change 0 Recirculation suction 528 52'8' No Change O Recirculation discharge 528 528 No Change 0
. Core spray 528 528 No Change O Core spray high pressure
*30 Cycles 528 440 Step 488 CRDHS return. 50 50 No Change 0 Standby liquid control 528 60 468
*10 Cycles- 60 528 462 468 Bottom drain 528 528 No Change 0 l
Note:" Reactor pressure remains at 1000 psig. O 3.9-174
. . . - . ._. ~. - _.. . .
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 {} TABLE 3.9-33 (Con t ' d) (Applicable Thermal Transients Reduction to 0% Power 111 Cycles Condition - Normal) INITIAL FINAL TEMP. RATE 6 TEMP. PIPELINE TEMP. *F TEMP. F ' F/ HOUR- *F Main steam 552 552 0 0 Head spray 552 552 0 0 Feedwater 420 265 310 155 Recirculation suction 528 552 32 24 Recirculation discharge 528 552 32 24 Core spray 528 552 32 24 CRDHS return 50 50 0 0
-Standby liquid control 528 552 48 -24 552 400 200 152 L
Bottom drain 528 552 48 24 552 400 200 152 ( l I i Note: Reactor pressure remains at 1000 psig. O. 3.9-175 . - . _ . . . - . . . - . _ . _ _ _ . . . - - . . . _ . - . _ . . . . _ . _ , _ - _ . . _ . _ _ . . . _ . _ . - - . _ . . - . . . . _ , ~ , . . . . _ _
l LSCS-FSAR AMENDMENT 39 ! OCTOBER 1978 l f-TABLE 3.9-33 (Cont'd) l (Applicable Thermal Transients Hot Standby 111 Cycles Condition - Normal) INITIAL FINAL TEMP. RATE ATEMP . PIPELINE TEMP. *F TEMP. *F F/ HOUR F l Main steam 552 552 0 0 l Head spray 552 552 0 0 Feedwater 265 552 Step 287 552 100 Step 452 100 552 Step 452 1 Recirculation suction 552 552 0 0 Recirculation discharge 552 552 0 0 Core spray 552 552 0 0 CRDHS return 50 50 0 0 (~) Standby liquid control 400 400 0 0 Bottom drain 400 400 0 0 l l l l l l 1 1 Notes Reactor pressure remains r at 1000 psig. t l l 3.9-176 l
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 ,s TABLE 3.9-33 (Cont'd) V (Applicable Thermal Transients' Shutdown 111 Cycles Condition - Normal) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP.
- F TEMP. *F
- F/ HOUR F Main steam 552 375 100 177 375 330 270 45 330 100 100 230 Head spray 552 375 100 177 375 50) 15 sec Step 325 50 300) duration Step 250 300 100 100 200 Feedwater 552 100 5 cycles 100 452 552 100)of 3 min Step 452 100 552) duration Step 452 during 1st 2 hrs Recirculation suction 552 375 100 177 375 330 270 45 l{]) 330 100 100 230 Recirculation discharge 552 375 100 177 375 300 Step 75 300 260 240 40 260 100 100 260 Core spray 552 375 100 177 375 330 270 45 330 100 100 230 CRDHS return 50 50 0 0 Standby liquid control 400 375 100 25 375 330 270 45 330 100 100 230 Bottom drain 400 375 100 25 375 330 270 45 330 100 100 230 Note: Reactor pressure decreases l from 1000 psig to O psig.
I') v 3.9-177
1 I LSCS-FSAR AMENDMENT 39 ;
' OCTOBER 1978 i i
,r g TABLE 3.9-33'(Cont'd) ' 's/ (Applicable Thermal Transients Unbolt 'l
~ 123 Cycles Condition - Normal) j i
. j
. INITIAL FINAL TEMP. RATE ATEMP.
-PIPELINE TEMP. *F TEMP. 'F *F/ HOUR 'F -
Main steam 100 70 Step 30 t Head spray 100 70 Step 30 Feedwater 100 70 Step 30 l Recirculation suction 100 70 Step '30 I Recirculation discharge 100 70 Step 30 Core spray 100 70 Step 30
-3 CRDHS return 100 70 Step 30 ;
Standby, liquid-control 100 70 Step 30 Bottom drain 100 70 Step 30 I 4 I l i l i Note: Reactor pressure remains at 0.psig. ("~)\
- s. a 3.9-178 I
I l LSCS-FSAR AMENDMENT-39
' OCTOBER 1978 7 '(^g TABLE 3.9-33 (Cont'd) l 1 %)
(Applicable Thermal Transients Loss of Feedwater Pumps MSIV Close 10 Cycles Condition - Upset) l l
. INITIAL FINAL TEMP. RATE ATEMP.
PIPELINE TEMP. F TEMP.
- F. *F/ HOUR F Main steam 552 573 Step 21 573 561 4320 12 561 525 240 36 525 573 '480 48 573 561 4320 12 561 490 610 71 490 573 625 83 573 561 4320 12 ,
561 485 650 76 485 400 100*/hr 85 400 552 100*/hr 152 Head spray 561 40 Step 521 3 times cycle 40 525 Step 485 Feedwater 552 573 Step 21 573 561 4320 12 O 561 525 525 573 240 480 3e 48 573 561 4320 12 561 490 610 71 490 573 625 83 573 561 4320 12 561 485 650 76 485 400 100 85 400 552 100 152 552 100 Step 452 100 250 Step 150 250 420 340 170 Recirculation suction 528 525 20 3 and discharge 525 573 480 48 573 561 4320 12 561 490 610 71 490 573 625 83 573 561 4320 12 561 485 650 76 485 400 100*/hr 85 400 552 100 /hr 152 552 544 Step 8 544 528 32 16 O 3.9-179 ,
. . _ - - - _ , _ . - _ _ _ . . _ _ . ~ . . _ _ _ . - . - . . _ _ _ _ _ . . _ _ - - . _ _ _ . _
l LSCS-FSAR AMENDMENT 39 OCTOBER 1978 i em TABLE 3.9-33 (Cont'd) b (Applicable Thermal Transients Loss of Feedwater Pumps MSIV Close 10 Cycles Condition - Upset) l l 1 INTIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. 'F TEMP. 'F 'F/ HOUR F Core spray 528 40 Step 488 40 528 Step 488 528 40 Step 488 40 528 Step 488 528 40 Step 488 40 528 Step 488 528 400 100 /hr 128 400 552 100 /hr 152 552 544 Step 8 544 528 32 16 CRDHS system 50 528 Step 478 528 250 200*/hr 278 250 400 100'/lir 150 400 544 Step 144 544 528 32 16 () Standby liquid control 528 250 250 400 200 100 278 150 400 544 Step 144 544 528 32 16 Bottom drain 528 250 200 278 250 400 100 150 400 544 Step 144 544 528 32 16 Note: Reactor pressure increases to 1180 psig. All relief valves open. Pressure decreases to 1125 psig and relief valves close. RCIC initiates and pressure decreases to 875 psig. (]' RCIC trips off on high level and pressure increases to 1125 and one relief valve opens and then closes as pressure decreases at rate of 100' F/hr. 3.9-180 l
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE 3.9-33 (Con t ' d) (Applicable Thermal Transients Reactor Overpressure Delayed Scram 1 Cycle Condition - Emergency) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. 'F TEMP. F 'F/ HOUR p. Main steam 552 583 31 583 538 5400 45 538 400 100- 138 400 552 100 138 Head spray 552 583 -31 583 538 5400 45 538 400 100 138 400 552 100 138 Feedwater 420 276 8640 244 276 100 704 176 100 250 Step 150 250 420 340 170 Recirculation suction 528 400 100 128 400 552 100 152 f]' 552 544 544 528 Step 32 16 8 Recirculation discharge 528 400 100 128 400 552 100 152 552 j 544 Step 8 544 528 32 16 3 l 1 Core spray 528 400 100 128 l 400 552 100 152 552 544 Step 8 544 528 32 16 CRDHS return 50 50 0 0 Standby liquid control 528 250 200 278 250 400 100 150 l 400 544 Step 144 544 528 32 16 Bottom drain 528 250 200 278 250 400 100 150 400 544 Step 144 544 528 32 16 rw Note: Reactor pressure increases to (-) 1350 psig. All relief valves and safety valves open. Pres-l sure decreases to 240 psig. l 3.9-181
LSCS-FSAR AMENDMENT 39 i OCTOBER 1978 {} TABLE 3.9-33 (Cont ' d) (Applicable Thermal Transients Single Safety or Relief Valve Blowdown 8 Cycles Condition - Emergency) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. 'F TEMP. 'F *F/ HOUR 'F Main steam 552 375 1062 177 375 100 100 275 - Head spray 552 375 1062 177 375 100 100 275 Feedwater 420 276 8640 144 276 100 704 176 Recirculation suction 528 375 918 153 375 100 100 275 Recirculation discharge 528 375 918 153 375 100 100 275 Core spray 528 375 918 153 f- 375 100 100 275 b CRDHS return 50 50 0 0 Standby liquid control 528 100 100 428 Bottom drain 528 100 100 428 Note: Reactor pressure decreases to (]) 0 psig with one relief valve or safety valve open. 3.9-182
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 f- g TABLE 3.9-33 (Cont'd; v (Applicable Thermal Transients Automatic Depressurization 1 Cycle Condition - Emergency) INITIAL FINAL TEMP. RATE ATEMP . 1 I PIPELINE TEMP. F TEMP.
- F. F/ HOUR 'F ,
Main steam 552 375 3218 177 375 281 300 94 Head spray 552 375 3218 177 375 281 300 94 Feedwater 420 276 8640 144 276 100 704 176 Recirculation suction 528 375 2780 153 375 281 300 94 Recirculation discharge 528 375 2780 153 375 281 300 94
. Core spray 528 375 2780 153 s 375 281 300 94 CRDHS return 50 50 0 0 Standby liquid control 528 375 2780 153 375 281 300 94 281 130 200 151 Bottom drain 528 375 2780 153 375 281 300 94 281 130 200 151 Note: Reactor pressure decreases with autodepressurization relief valves open to 35 psig.
3.9-183
. -- .-~. . . . . -
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE 3.9-33 (Cont'd) O (Applicable Thermal Transients Improper Start of Cold Recirculation Loop 1 Cycle Condition - Emergency) INITIAL FINAL TEMP. RATE ATEMP. PIPELINE TEMP. *F. TEMP. *F F/ HOUR *F Main steam 552 552 0 0 Head spray 552 552 0 0 Feedwater 420 420 0 'O Recirculation suction 528 130)26 see Step 398 130 528) duration Step 398 Recirculation discharge 528 528 0 0 Core spray 528 528 0 0 CRDHS return 50 50 0 0 Standby liquid control 528 528 0 0 Bottom drain 528 (]) 528 0 0 Note: Reactor pressure remains at 1000 psig. O 3.9-184
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 ,S TABLE 3.9-33 (Cont'd) V (Applicable Thermal Transients Sudden Pump'Scart in Cold Loop 1 Cycle Condition - Emergency) INITIAL FINAL TEMP RATE. ATEMP. PIPELINE TEMP. 'F TEMP. F F/HOUL F Main steam 552 552 0 0 Head spray 552 552 0 0 Feedwater 420 420 0 0 Recirculation suction 528 528 0 0 Recirculation discharge 528 130) 34 sec Step 398 130 528) duration Step 398 Core spray 528 528 0 0 CRDHS return 50 50 0 0 Standby liquid control 528 528 0 0 (]) Bottom drain 528 528 0 0 Note: Reactor pressure remains at 1000 psig. (~J 3.9-185 i
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 t'"3 TABLE 3.9-33-(Cont'd) U (Applicable Thermal Transients Pipe Rupture and Blowdown 1 Cycle Condition - Faulted) INITIAL FINAL TEMP. RATE 6 TEMP . PIPELINE TEMP. 'F TEMP. F ' F/ HOUR ' F Main steam , 552 281 Step 271 Head spray 552 281 Step 271 Feedwater 420 281 Step 139 Recirculation suction 528 281 Step 247 Recirculation discharge 528 281 Step 247 Core spray 528 40 Step 488 40 130 Step 90 CRDHS return 50 281 Step 231 Standby liquid control 528 281 Step 247 281 223 Step 58 () 223 50 Step 173 50 130 Step 80 l Note: Reactor pressure decreases from 1000 psig to 35 psig in (~'i y.) 15 seconds. 3.9-186
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 An examination of Table 3.11-5 shows that the integrated (]) radiation exposures are insufficient to cause damage to the equipment organic components. Thus, the class 1E instruments were qualified by analysis for use in applicable radiation environments. 3.11.2.1.4 NSSS Valve-Mounted Instrumentation and Electrical Equipment This paragraph discusses the testing and analysis of the safety-related instrumentation and electrical equipment supplied with NSSS valve assemblies. Safety-related l valve actuators are qualified based on performance require-ments using IEEE 323-1971, and IEEE 344-1971 as guidelines. In the Generic qualification, one of a family of actuators is type tested for the following conditions per IEEE 382-1972:
- a. Aging Simulation
- 1. thermal aging,
- 2. radiation aging, and
- 3. operational aging.
- b. Seismic Simulation Aging
- c. Accident and Postaccident Simulation
- 1. radiation,
- 2. temperature,
- 3. pressure,
- 4. humidity / steam condition, and
- 5. chemical exposure.
The aging simulation is designed to put the test specimen in the end-of-life or 40 year plant service condition, whichever is earlier. The seismic simulation is to put the test spec-imen through the expected OBE events and SSE event estimated l for the service life. The accident and postaccident test is to put the test specimen through the dasign-basis event. The duration of the last test will be related to the length of time that the actuator has to perform its safety function. 1 In the specific qualification, an actuator will be selected l from the qualified family to meet the specific requirements. It shall be demonstrated that the selected actuator is indeed similar to the test specimen, that no further type test is (]) required. Also, the selected actuator is manufactured and 3.11-4
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 Q _t ' FSAR SECTION/NRC QUESTION CROSS REFERENCE GUIDE: CHAPTER 6.0* FSAR SECTION QUESTION NUMBER 6.0 031.8 031.15 040.74 6.1.2 312.21 6.1.3 .312.22 6.2 031.154 031.225 l 6.2.1 021.1- 021.3 021.4 021.5 021.6 021.7 021.8 021.9 021.15 021.16. 021.17 021.18 021.36 021.38 021.39 021.40 021.42 021.43 021.44 021.45 021.46 021.47 021.48 021.49 021.50 021.51 021.52 021.62 021.69. 031.155 021.65 021.70 l 6.2.2 021.10- 021.19 021.72 021.21. 021.22' /~l s/ 031.37 212.26 021.63- -021.64 021.71 6.2.3 021.11 021.23 021.24 021.25 021.26 021.66 021.68 6.2.4 021.12 021.27 021.28 021.41 031.1 031.7 031.19 031.83 021.67 031.160 021.54 031.229 031.230 031.231 031.235 6.2.4.2 031.229 031.230 031.231 031.235 6.2.5 021.13 021.14 021.29 021.30 021.31-021.32 423.41 021.56 021.55 021.57 021.58 021.73 l l 6.2.6 021.33 021.34 021.35 021.59- 021.60 l 021.61 021.53 l 6.3 040.72 040~.80 212.15 212.26 212.29 212.30 212.31 212.32 212.33 212.34 212.35 212.36 212.37 212.38 212.39 212.40 212.41 212.44 212.52 212.70 212.87 212.111 212.112 212.124 212.91 , 212.92 212.93 212.96 212.99 031.168 l 212.94 212.100 212.123 212.101 212.95 ,e 212.121 040.105 031.211 031.212 L. yI l
*The complete text of the questions and responses is given in the:FSAR volumes entitled " Responses to NRC Questions."
l 6.0-00
f , 1 LSCS-FSAR. . AMEN" ENT'39
, OCTOt R:1978 .
\ t'
~
JV; ' '
,. y 1 ' CHAPTER 6 '. 0" 7(Cont ' d)
FSAR.- .
-SECTION' l ! QUESTION' NUMBER
~6.3.1 031.9 031.66 ,..
l 6.3.2 031.11 - 031.8s . 212.43 212.44 212'.'45'
- 212.97: L212.102 212.98. '
{ 6.3.3- .031.83- 1212.46 212.47-
\
6.3.4' .212.48 '212.103i 6.4 '031.126' + 312.'7 312.8' 312.9- '312;10' y: 312.111 312.12' 312.23:: ' 312.31 ~
.312.30- . . ..
031.174'-
~
312.31 031.238 c;
.l
- 6. S '- 312.13 312.14' --312.15. .312L24- .031.172-6.5.1 -031.100 321.70 3 6 '. 7 : 031.151 031'152 .
; .-6.711.-
- 6. 7. 2 -
.l 6.7.3-6.7.4 6.7.5 031.153'
- 6. A. ' 2 2] .'19.
a
+
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a i g i
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( . 6'.0-01
.... - . -. , , , a .. . - , , . . . .. w , , . . . - . . . . .. .-. - .. . . . - .
LSCS-FSAR' AMENDMENT 39 OCTOBER 1978 , TABLE OF CONTENTS (Cont ' d) PAGE 6.7.3.1 Functional Protection Features 6.7-7 6.7.3.2 Effects of Single Active Failures 6.7-7 6.7.3.3 Effects of Seismic Induced Failures 6.7-7 6.7.3.4 Isolation Provisions 6.7-7 6.7.3.5 Leakage Protection Evaluation 6.7-8' 6.7.3.6 Failure Mode and. Effects Analysis 6.7-9 6.7.3.7 Influence on Other Safety Features 6.7-9 6.7.3.8 Radiological Evaluation. 6.7-9 6.7.4 Instrumentation Requirements 6.7-9 6.7.5 Inspection and Testing 6.7-10 t ATTACHMENT 6.A ANNULUS PRESSURIZATION 6.A-i O o t.a 6.0-vi
t, LSCS-FSlal AMENDMENT 39 OCTOBER'1978 LOL , r e 1 5 t ATTACHMENT 6.A ? t x ANNULUS PRESSURIZATION O i 1 i f s e t
;01 .
- V w*w w ree,w-ye e:- ogg y em^9, er r,-y --wyo-y-w y gesu ,ey+ 'ya vsev , dye +y esey m =mee y r y= -ge,,,w evepww ew+ . _ w ow e ew- we n- w emm-ew
-=8-+---'mw=-ve+6=--* ee- --==*
LSCS-FSAR- AMENDMENT 39 l OCTOBER 1978 ATTACHMENT 6.A d('S TABLE OF CONTENTS PAGE l 6.A ANNULUS PRESSURIZATION 6.A-1 6.A.1 Introduction 6.A-1 6.A.2 Short-Term Mass Energy Release 6. A- 2 6.A.2.1 Instantaneous Guillotine Break 6.A-3 6.A.2.2 Break Opening Flow Rate 6.A-4 6.A.2.3 Combined Break Flow 6.A-5 6.A.2.4 Determination of the Mass Flux, G 6.A-5 6.A.2.5 Biological Shield Wall 6.A-5 6.A.2.6 Comparison of the GE Model with the Henry /Fauske Correlation 6.A-6 6.A.3 Load Determination 6.A-11 6.A.3.1 Acoustic Loads 6.A-11 6.A.3.2 Pressure Loads 6.A-11 ( 6.A.3.3 Jet Loads 6.A-11 6.A.3.4 Dynamic and Seismic Ana. lysis (DYSEA) Computer Program 6.A-12 6.A.4 Pressure to Force Conversion 6.A-15
- 6. A. 5 Sacrificial Shield Annulus Pressurization and RPV Loading Data 6.A-17 6.A.6 Jet Load Forces 6.A-20 6.A.7 Recirculation and Feedwater Line ' Break Forcing Function 6.A-21 (a~h .
6.A-i
LSCS-FSAR AMENDMENT 39. OCTOBER 1978 (). . ATTACHMENT 6.A LIST OF TABLES NUMBER - TI TLE ' PAGE 6.A-1 Time ' History for ' Postulated Recircu-lation Suction Pipe Rupture 6. A-2 2 .; 6.A-2 Acoustic Loading on Reactor Pressure
. Vessel Shroud 6.A-27' 6.A-3 RPV Coordinates'of Nodal Points 6.A-28 6.A-4 Maximum Member Forces Due to Annulus Pressurization 6.A-30 6.A-5 Maximum Acceleration Due to Annulus '
' Press urization 6.A-31 !
6 . A- 6 RELAP4 Input' Data, Recirculation Line Outlet Break '6.A-32 6.A-7 RELAP4 Input Data, ,Feedwater Line Break 6.A-35 6.A-8 Force Constants and Load Centers For Recirculation Line Outlet Break 6.A-38 6.A-9 ' Force Constants and Load Centers For Feedwater Line Break 6.A-40
- 6. A-10 DYSEA01 Program Input For Jet Load
/'l Forces 6. A-4 2 I 6.A-11 Time Force Histories - Recirculation >
Line Break 6.A-44 6.A-12 Time Force Histories - Feedwater Line Break 6. A-7 6 A 6.A-il
- m. = _ .
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 l 1
/~\ ATTACHMENT 6.A' I
_ (_) ) LIST 'OF FIGURES j NUMBER TITLE 6.A ' Safe End Break Lo'c ation
' 6 . A-2 Break Flow Vs. Time - Feedwater Line Break 6.A-3 Geometry 6 . A-4 Wave Speed 6.A-5 Mass Flux, Moody Steady ~ Slip ' Flow 6.A-6 Break Flow Vs. Time 6.A-7 Nomenclature' for Time History Computer Printout 6.A-8 Feedwater Line System Nodalization - Leg EA 6.A-9 Feedwater Line System Nodalization - Leg EB 6.A-10 Recircirculation Line System Nodalization 6.A-11 . Comparison of the GE and RELAP4/ MOD 5 Methods -
Feedwater Line Break, Leg EA 6.A-12 Comparison of the GE and RELAP4/ MOD 5 Methods - Feedwater Line Break, Leg EB _ 6.A-13 Comparison of the GE and RELAP4/ MOD 5 Methods - Recirculation Line Break, Finite' Opening time 6.A-14 Horizontal Model for Annulus Pressurization 6.A-15 Annulus Pressurization Loading Description-6.A-16 Annular Space Nodalization For Recirculation Line q(_s Break 6.A-17 Annular Space Nodalization For Feedwater Line Break l I
\
U
- 6. A-lii
. _ _ _ _ _ _ . _, _m
.LSCS-FSAR AMENDMENT 39 OCTOBER 1978
() 6.A ANNULUS' PRESSURIZATION i 6.A.1 Introduction
- Annulus pressurization refers to the 1 ading on the shield wall and reactor vessel caused by a postulated pipe rupture at. the reactor pressure . vessel' nozzle safe-end to pipe weld. .The pipe break . assumed -is an -instantaneous guillotine rupture which allows mass / energy release into the drywell.and annular region between the. biological shield wall and the reactor pressure vessel (RPV) '. e The mass and energy released during the postulated pipe rupture cause:
- a. A rapid asymmetric decompression acoustic loading of r the annular region between the : vessel and shroud from the pipe break at or beyond the vessel nozzle safe-end weld,
- b. A transient asymmetric different *.a1 pressure within the annular region between the biological shield wall and the reactor pressure vessel (annulus pressuriz ation) .
- c. A jet-stream release of the reactor pressure vessel inventory and the impact of the ruptured pipe against O- the whip restraint attached to the biological shield wall.
I The results of the iss and energy release evaluation are then used to produce a d) amic structural analysis (force-time history) of the RPV and shield wall. . The force time history output fro'm the dynamic analysis is subsequently used to compute loads on the reactor components. The following is more detailed description of the annulus pressurization calculation performed- ( for the La Salle County Station. l I I l l
,4 .Q 6.A-1
LSCS-FSAR AMENDMENT 39 OCTOBER 1973
\" 6.A.2 Short-Term Mass Energy Release The postulated pipe rupture at the weld between recirculation or f eedwater piping and the reactor nozzle safe end leads to a high rate of water and steam mixture into the annulus between the RPV and the shield wall. Figure 6. A-1 illustrates the' location of this break. Calculation of the mass / energy release is performed using the generic method for short-term mass releases. This method and a sample calculation are described below. Figure 6.A-2 illustrates a typical mass flux vs. time for a feedwater line break.
The purpose of this procedure is to document the method by which-short-term mass release rates are calculated. The flow rates which could be produced by a primary system line break for the first 5 seconds include the effects of. inventory and subcooling. Optionally, credit may be taken for a finite break opening time. ASSUMPTIONS The assumptions are as follows:
- a. The initial velocity of the fluid in the pipe is z ero. When considering both sides of the break, the -
ry effects of initial velocities would tend to cancel
's_/ out.
- b. Constant reservoir pressure.
- c. Initial fluid conditions inside the pipe on both sides of the break are similar.
- d. Wall thickness of the pipe is small compared to the diameter.
- e. Subcompartment pressure = 0.
- f. Mass flux is calculated using the Moody steady slip flow model with subcooling.
- g. For steamline breaks, level swell occurs at 1 second af ter the break with a quality of 7%.
NOMENCLATUR E (See Fiqur e 6. A-3) A bR Break. area. An Minimum cross-sectional area between the vessel and the break. This can be the sum of the areas of
,_ parallel flow paths.
V, s
- 6. A- 2
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 L(). c -
' Sonic velocity (see Figure 6. A-4) .
D - Pipe inside diameter at the. break location. Fy - Inventory flow multiplier. F7 = 0.75 for saturated steam. F y. = 0.50 ' for liquid and saturated steam-liquid mixtures. gc - Proportionality-constant (=32.172 lbm-f t/lbf-seca' ; . G - Mass flux. Gc - Maximum mass flux - (see Figure 6. A-5) . ho - Reservoir or vessel enthalpy. hp - Initial enthalpy of the fluid in the pipe. h7 - Enthalpy at P g and a quality of 7%. Ly - Inventory length. The distance between the break and the nearest area increase of AL whichever is less. ( M - Mass flow rate.
$7 -
Mass flow rate during the inventory period. P - Reservoir or vessel pressure. o P - Saturation pressure for liquid with an enthalpy of SAT h.p t - Time. ty - Length of the inventory period. v - Specific volume of. the fluid initially in the pipe. Vy - Volume of th . pipe between the break and A L. X - Separation distance of the break. 6.A.2.1z Instantaneous Guillotine Break The f ollowing method should be applied to each side of the break
. and - the results summed to determine the total flow.
,. Inventory Period 6.A-3
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 Prior.to a pipe break, the fluid-in the pipe is moving at a relatively low velocity. After the break occurs, a finite time is required to accelerate the fluid to steady-state velocities. The length of this time period is conservatively estimated as follows:
- a. If A A BR I' ty = 2L 7
l ( 6. A- 1)
- b. If A A <F y, BR ty = V A GF v (6. A-2) 33 I where G is calculated as shown in Subsection 6. A.2.4 for a large separation distance and t < ty.
During this thme period, the mass flowrate is calculated as
$ =GA ( *^~ }
7 BR I Steady-State Period 'I Following the inventory period, the flow is assumed to be choked at the limiting cross-sectional flow area. 1 For t < t < 5. 0 s econds, (6. A- 4 )
$=A n G 6.A.2.2 Break Opening Flow Rate l 1
See Table 6.A-1 for the pipe displacement time history for postulated recirculation suction pipe rupture and Figure 6. A-7 I f or the nomenclature . used. Inventory Period The inventory period is determined as described in Subsection 6.A.2.2. The flow rate as a function of pipe separation distance is given by b = G Tr D X (6. A-5) (') u 6.A-4
LSCS-FSAR AMENDMENT.39 OCTOBER 1978 h : where G is obtained. by using the methods of Subsection 6. A.2.4 (a or b) . Determining Flow Rate Following the inventory period, equation 6. A-5 is used to deter-mine.the flow rate where the mass flux, G, is determined from ! Subsection 6. A. 2. 4 (a, . c, or d) . 6.A.2.3 combined Break-Flow To determine the total flowrate released from the break, the results of .Subsectior.s 6. A. 2.1 and 6. A.2. 2 are compared and whichever produces the smallest flowrate at any time is used (see Figure 6. A- 6) . Brch methods produce maximum flow rates based on different limiting areas. The transfer from one curve to the other represents.a change in the point where the flow is choked. 6.A.2.4 Determination of the Mass Flux, G Depending on the time period, fluid conditions, and break separa-tion distance, the mass - flux is determined as follows: XB = /1 - (PSAT/Pg) (D/2) ( 6. A- 6)
- a. If X < X Be G= /2gg Pg /y
- b. If X > X B and t < t y G=Go (Pg , h p) from Figure 6.A-5
- c. If X > X B and t > t y G=G c (P g, h) g f rom Figure 6. A-5
- d. If the break is a steamline and t > 1.0, level swell occurs.
G=G c (Po, h) 7 f rom Figure 6. A-5 Note that for complete break separation (subsection 6. A. 2.1) , A is always greater than XB, and for saturated water, XB is equal to zeor. j 6.A.2.5 Biological Shield Wall For the purpose of analyzing the biological shield wall pressuri- ! I zation, credit may be taken.for flow which escapes through the wall penetration. If the initial break location is in the
- O ennu1ue reeion between the we11 and the vesee1, no f10 ie assumed to escape through the penetration. If, however, it is I
6.A-5
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 h~ located within the? penetration itself, some'.of the. flow-may be assumed to escape. .It is recommended that the fraction of-the flow which escapes ~ be . calculated based 'on the ratio of; the minimum annular flow area between .the penetration and pipe surface and between the penetracion and pipe surface and between the . penetration and the saf e-end nozzle.
'6.A.2.6 Comparison of the GE model with the Henry /Fauske ;
Correlation-1 The GE methodology for calculating the ' mass. energy. release from a . l recirculation. line break which results in an annulus pressuriza-tion event was provided the NRC's Mr. Denwood. F. Ross, Assistant
. Director 'for Reactor Safety, via GE letter dated May 2, '1978,
'from Mr. E. D. Fuller of BWR Licensing. This methodology'was used in the adequacy assessment made for LSCS.
The definition of the annulus pressurization lis given in the ) introduction (Subsection' 6. A.1) . A description of the' time aspects of the calculated mass and energy flow rates followed by a description of the modeling for the feedwater line and i separately for the recirculation line is provided below. A' ; comparison .is then made between GE's analytical method and the l method used in RELAP4/ MODS.. Finally, both graphical and ! numerical results of this comparison are provided to substantiate the ' conclusion that the resulting break. flows using the GE (d' methods are much more . conservative than those predicted by the j use of RELAP for the La Salle plant. l l Timing Aspects of Mass and Energy Flow Rates The GE method for calculating the short-term mass / energy releasef i ascumes that the overall. time for mass release may be divided into two periods, the inventory period and. the quasi-steady period. The inventory period is defined as the time required to accelerate the pipe fluid to steady-state velocities, at which. time the flow is assumed to choke at~ minimum flow cross sections.
.During-this-time, the mass flux is based on initial thermodynamic conditions existing within the pipe. In the quasi-steady period, during which' the ' flow is choked, the mass flux. is based on
' thermodynamic conditions upstream from the choke points. For both time periods the mass flux is determined from a . graph of critical mass flux versus enthalpy, as calculated by the Moody. j Slip Flow Method. Each side of the break is analyzed separately and the lresults summed to give the total mass release rate.
Method for Feedwater Line Modelinq The feedwater system for la Salle County station consists of'the pumps, heaters, valves,: and piping necessary- for the' transmission ypy . of 1hotwell condensate to the reactor vessel- as part of the closed
. q,/ cycle cooling loop. LSCS.has three.feedwater pumps, two steam-driven 'and one electric-driven. During normal operation, the 6.A-6' k
r LSCS-FSAR. AMENDMENT 39 OCTOBER 1978'
'q i kJ electric . pump is in standby. . - The. fl w passes through a complex seriesf of pipes 'and components . from the feedwater pumps to the reactor vessel. .
.t The break location for the'feedwater line break-is the safe end to the pipe weld housed within the vessel to shield wall >
subcompartment. For the feedwater.line break, instantaneous
. break opening is assumed. . Fow for the vessel side passes through the,feedwater nozzles of the broken line'and out the break. Flow- '
f rom the system side ~ pass es through the' feedwater piping: network' t and' out the break. The nodalization of the ' eedwater f system is .shown in Figures: ' 6.A-8 and 6.A-9. A series of: 24 modes was selected af ter sensit .vity studies were completed which _ demonstrated that a 24-ncJe model was . conservative relative to higher-noded systems. The-broken feedwater leg to be analyzed was chosen by multiple RELAP runs to determine the limiting break . location. The critical assumptions inLthe analysis.are as follows:
- a. The .feedwater pumps are simulated as- (constant)1 mass flow sources.
- b. The reactor pressure vessel- (RPV) is an infinite O reservoir. at . constant temperature and pressure. ,
- c. The temperature of the. pump-nide hydraulic: network remains constant.
- d. Appropriate sections of -the hydraulic network are combined by means1 of " Ohm' t. Law a expressions for series and parallel circuits, assuming constant fanning friction actions, The RPV thermodynamics state 'is subcooled at' the
~
e. prevailing temperature in the lower plenum - (5320 F) . The break is modeled .as an instantaneous guillotine pipe break ' l with complete pipe of fset. Before the break occurs, a fully open L valve connects, Volumes 18 and 19. Closed valves connect those l volumes to volume 1, an infinite sink at constant pressure and temperature (atmospheric conditions) . .The break.is initiated at time zero by closure of the valve between Volumes 18 and 19 simultaneous opening of the valves.to volume 1. Method of - Recirculation Line Modeling' .t The recirculation system ' for La Salle county station 'is similar to : the recirc ulation ' system . of other BWR's. Flow is taken from L h the lower jet pump diffuser region, passed through 21-inch lines l . to . a constant-speed pump, and then through a flow control valve 6.A-7
l l LSCS-FSAR AMENDMENT 39 OCTOBER 1978 to a header which feedw flow to five risers which provdie ' flow to two jet pump nozzles each. { The nodalization for the recirculation line leak is shown in Figure 6.A-10. The system has been modeled using 21 nodes. The I break is located at the vessel nozzle safe-end to pipe weld on the recirculation pump suction side. The type of break considered here has a finite break opening time. For this case the break opening is complete after 30 milliseconds, at which time the pipe of f set longitudinal distance is 5.8 inches. The break area is modeled as the surface area of an imaginary volume having a length of 5.8 inches and a diameter equal to that of the recirculation pipe ID. This volumn (#18) .is connected by a valve (Type 3) to an infinite reservoir (volume #19) , and also by valves (Type 2) - to the vessel side volume (#27) and pump side-volume (#21). Both valves (Type 1) also connect Volumes 17 and
- 21. It is normally open .before the break, and at the initiation of the break, closes at the same rate as the other valves open.
The sum of the areas of the Type 2 valves equals the pipe area. This network of valves best represents the break with finite opening time. Valves of Type 2 are opened at the same rate as Type 3 to ensure that choking occurs at Junctions 21 and/or 22.. Junction 23 (having valve Type 3) is in reality a fluid surface, and choking cannot physically occur there. Choking must at least occur at Junctions 21 and/or 22, where the fluid is constrained (]) by the pipe diameter. Other assumptions in the analysis include:
- a. Negligible effects of core reactor kinetics on rated heat transfer to the core volume (Volume 2) .
- b. Constant flow of steam from the steam dome (Volume 5) at rated conditions,
- c. Constant flow of feedwater at rated conditions.
- d. Recirculation pumps trip at the time zero and are modeled via pump characteristic curves for coastdown.
- e. Jet pump hydraulics were modeled as one equivalent pump per recirculation loop.
Comparison of General Electric Analysis to RELAP4/ MODS For the annulus pressurization event, the NRC has questioned General Electric's method for computing mass and energy flow rates following a postulated LOCA~from long lines containing subcooled fluid. A program was developed to expedite the licensing of the La Salle County Station to perform RELAP (]) analyses using appropriate assumptions and to compare the results with those obtained using General Electric's method. 6.A-8
I LSCS-FSAR AMENDMENT 39
. OCTOBER 1978-RELAP4/ mod 5 is a general computer program which can be used to
- analyze L the thermal hydraulic transient behavior of a. water-cooled nuclear. reactor . subjected to postulated _ accidents Lsuch as loss-of-coolant accidents. The program simultaneously solves the fluid _ flow, heat transf er, 'the reactor kinetics equations describing the behavior of-the reactor.- ,
Numerical input data is utilized to describe the initial-conditions and geometry of the system being analyzed. .This ' data . includes fluid volume, geometry,. pump characteristics, Lpower. ' generation, heat exchanger properties, and nodalization of fluid flow paths. once the system has been described with initial flow, pressure, temperature, and power level boundary conditions,. -' transients such as loss-of-coolant accident can. be simulated by . control action inputs. RELAP then computes fluid conditions such as flow, pressure, mass inventory and quality as a function of , time.- For the brief transients considered here (t 5 0.5 seconds) , appreciable siuplification of the overall thermal-hydraulic system, including the reactor pressure vessel, is justified owing to the relatively longer time constants which apply-for. heat transfer. Brief summaries of-the modeling approaches for feedwater and recirculation line breaks are given below. - The assumptions applied to these analyses .are_ as follows: (])
-s~ .a. Feedwater line:
1.. La Salle RELAP deck as basic.
- 2. Henry-FAuske-Moody flow model is used. ;
- 3. Instant break ~ opening.
- 4. Mass flux terms between vessel and break (short (
side) are eliminated. i
- b. Recirculation line:
- 1. La Salle RELAP deck as basis.
- 2. Finite break opening time'is allowed for.
- 3. Henry-Fauske-Moody flow model is used.
- 4. Momentum flux terms 'in RELAP between vessel and >
break (short side)_ are eliminated. e O 6.A-9
-_,-,,J,..m .L.,- ,m.,..._,.~,,,-.-m.,m ........~..;...__ . - . . . . . . . - . . - . . - - - - - -
1
-LSCS-FSAR AMENDMENT 39 OCTOBER 1978 O
Reeu1te of ene Ane1veie The resulting~ break flows Ang the GE- methods are much more conservative than. those. obtained by the use of RELAP. This is indicated - graphically in Figures 6. A-11 through 6. A-13.- conclusions The mass release' result for the GE mass energy release method and the RELAP4/ Mod 5 calculations are fcompared in Figures 6. A-11 through 6.A-13'for the postulated feedwater line break'and recirculation line break respectively. The analyses show.that the GE method is conservative relative to RELAP 4/ Mod 5 ' for both cases. The ration (r) of the GE method flow rates to those from RELAP/ MOD 5 is as follows: Break Location r(t = 0.1 secl ~ r(t = 0.5 sec) Feedwater (Leg EA) 2.300 1.70 Feedwater (Leg EB) 2.200 1.60 Recirculation Line 1.065 1.21 0 3 O. 6.A-10
l LSCS-FSAR AMENDMENT 39' l OCTOBER 1978 1 6.A.3 Load Determination 6.A.3.1 Acoustic Loads Because the boiling water reactor (BWR) is a two phase system that operates at or close to saturation pressure (1000 psi) , . the differential pressure across the reactor shroud is of short duration, and the BWR system is not subjected to a significant shock-type load with respect to structural supports. .This short-duration acoustic -load is confined to a bending moment and shear force on the reactor pressure vessel and reactor shroud support. r Typical results of the integrated force acting on the reactor pressure vessel shroud are given in Table 6. A-2. 6.A.3.2 Pressure Loads The pressure responses of the RPV-shield wall _ annulus for a recirculation suction line and a feedwater line were investigated using the RELAP4 computer code. An asymmetric model using several nodes and flow paths was developed for the analysis of the recirculation and feedwater line breaks. Further description of these analytical models and detailed discussion of the analyses may be found in LSCS-ESAR, Section 6.2. The pressure histories generated by the RELAP4~ code were in turn (3 used to calculate the loads on the sacrificial wall and the U reactor pressure vessel. The annulus was divided into seven zones and an eighth-order Fourier fit to the output pressure histories made for each zone to produce the Fourier coefficients required for the structural analysis of the shield wall. The specific loading data consisted of the time-pressure (psia) histories for each node within the annulus. Time-force histories representing the resultant loads on the RPV for each node through its geometric center were generated by taking the product of the node pressure and its nef fectiven surface area. A sample pressure-to-force calculation is shown in Subsection 6.A.4. Subsection 6. A. 5 shows the nodalization schemes and pressure areas used in this calculation. The time-f orce . j histories (forcing functions) calculated at each nodal point for t both a recirculation and a feedwater line break are shown in Subsection 6.A.6. The nodal points are illustrated in Figure 6.A-14. 6.A.3.3 Jet Loads To address structural loads on the vessel and internals completely, jet thrust, jet impingement, and pipe whip restraint loads must be considered in conjunction with the above mentioned pressure loads. Jet thrust refers to the vessel reaction force , with results as the jet stream of liquid is released from the () break. Jet impingement refers to the jet stream force which leaves the broken pipe and impacts the vessel. The pipe whip 6.A-11
LSC S-FSAR AMENDMENT 39 OCTOBER 1978 ; 1 /*% (_) restraint J oad is the force which results when the energy-absorbing pipe whip restraint restricts the pipe separation to less than one full pipe diameter. This restricted separation is accounted for as a finite break opening time in the mass / energy release calculation. These jet loads are calculated as described in AN SI 176 (draft), " Design Basis For Protection Of Nuclear Power Plants Against Eff ects Of Postulated Pipe Ruptures", Janua ry 1977. The jet load forces used in this analysis are shown in Subsection 6.A.6. Although these values have been calculated for a recirculation line break only, they are also conservatively used for the feedwater load evaluation. This is conservative because the calculation of these jet effects depends largely on the area of the Dreak, and the recirculation line is about 2.5 times larger in area. Figure 6.A-15 illustrates the location of the pressure loads and jet loads with respect to the RPV and shield wall. The pressure loads and jet loads described above are then com-bined to perf orm a structural dynamic analysis. Both of these loads are appropriately distributed along a horizontal beam model, which is shown in Figure 6. A-14. The vessel coordinates of these nodal points are described in Table 6. A-3. The force time histories are then applied to a composite lumped-(]) mass model of the pedestal, shield wall, and a detailed representation of the reactor pressure vessel and internals. The DYSEA01 computer program is used for this analysis. This computer program is described in Subsection 6. A.3.4. The dnalysis produces acceleration time histories at all nodes for use in evaluating the reactor pressure vessel and internal components. Response spectra at all nodes are also computed. The peak loading on the major componnets used to establish the adequacy of the component design is shown in Tables 6. A-4 and 6.A-5. 6.A.3.4 Dynamic and Seismic Analysis (DYSEA) Computer Program The DYSEA (Dynamic and Seismic Analysis) program is a GE pro-prietary program developed specifically for seismic and dynamic analysis of RPV and internals / building systems. It calculates the dynamic response of linear structural systems by either temporal modal superposition or response spectrum method. Fluid-structure interaction effect in the RPV is taken into account by way of hydrodynamic mass. The DYSEA program was based on the program SAP-IV with added capability to handle the hydrodynamic mass ef fect. Structural stif f ness and mass matrices are formulated similar to SAP-IV. r's Solution is obtained in the time domain by calculating the k# dynamic response mode by mode. Time integration is performed by 6.A-12 i
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 '(()- iusing Newmark's . S-method. Response spectrum solution is also available as an option. Program Version and Computer The DYSEA version now operating on the Honeywell 6000 computer of GE, Nuclear Energy Systems Division, was developed at.GE by modifying the SAP-IV program. Capability was added to handle the hydrodynamic mass eff ect due to fluid-structure interaction in the reactor. The program can handle three-dimensional dynamic problems with beam, trusses, and springs. Both acceleration time histories and response spectra may be used as input. History of Use The DYSEA program was developed in the summer of 1976. It has been adopted as a standard production program since 1977 and it has been used extensively in all dynamic and seismic- analysis of the RPV and internals / building systems. Extent of Application The current' version of DYSEA has been used in all dynamic and seismic analysis since-its development. Results from test problems were found to be in close agreement with those obtained from either verified programs or analytic solutions.
)
Test Problems Problem 1: r The first test problem involves finding the eigenvalues and eigenvectors from the following characteristic equation: (W2 [ M ]-[ K ]) {x} = 0 where w is the circular frequency, x is the eignevector, and [K) and [M] are the stiffness and the mass matrices given by: 4 4 4 1 2 2 2 n n qn [M] = 14 2 7 4 gn n 3 3 4
- Symmetric 1- l 25n 2- (6.A-7) t )
6.A-13
LSCS-FSAR AMENDMENT 39 OCTOBER-1978 (~\ U - 2 5 3 1 .+ E7 15 [K] = 1+9} 2 Symmetric 1 + 25 (6. A- 8 ) Ihe analytical solution and the solution from DYSEA are: a) Eigenvalues e f - i DYSEA SOLUTION ANALYTIC SOLUTION 1 5.7835 5.7837 2 30.4889 30.4878 3 75.0493 75.0751 b) Eigenvectors $i:
- 1. DYSEA SOLUTION ANALYTIC SOLUTION
(]) _ 1.000 1.000 1.000 1.000 1.000 1.000 0.0319 -0.0319 - 1.5536 -1.2105 -0.0319 -1.554 -1.211
-0.0072 -0.0666 2.0271 . -
-0.0072 0.0666 2.027 -
Problem 2: The second test problem compares the dynamic responses of the reactor pressure vessel, internals and reactor building subjected to earthquake ground motion. The mathematical model of the reactor pressure vessel, internals and reactor building is given in Figure B-1. The inputs in the f orm of ground spectra are applied at the basement level. Response spectrum analysis was used in the analysis. Natural frequencies of the system and the maximum responses at key locations have been calculated by both DYSEA and SAMIS. Result comparison are given in B-1 and B-2. It can be seen that the results calculated by DYSEA agree closely with those obtained by SAMIS. A N. 6.A-14
-1 LSCS-FSAR ' AMENDMENT 39 OCTOBER 1978 r 6.A.4 Pressure-to__ Force Conversion I
(_h)~' l The RELAP4 pressure diFtribution output is Converted to equiva-lent forces which are input into the DYSEA01 computer program. , Each pressure. is represented by a' force acting normal to the RPV or shield wall at the center of the given pressue' surface area. These forces are then converted into resulting forces (x component) acting on the respective DYSEA01 RPV beam nodal points. Mathematially, this is described as: FR = PA cos 0-where:
.F R = resultant force (1b),
P = RELAP4 node pressure (psia) , A = RELAP4 node _ surface area (in2), and 8 = Component angle. These parameters are illustrated in Figure 6. A-4. As an example, the pressure to force conversion at DYSEA01 node points 31 and 32 is shown_below: O Time = 0.0800 seconds I ELEV PRESSURE AREA
- ANGLE FORCE NODE (inches) (lb/in2) (in a) (degrees) (1b) 6 1089.14 43.61 5828.44 15 245516 7 1089.14 35.34 5828.44- 45 145660 8 1089.14 39.24 5828.44 75 59188 a 9 1089.14 41.40 8617.79 112.5 -
136539 l 10 1089.14 39.99 8617.79 157.5 - 318367 4543 For 3600, the resultant force is 2 times 4543 lb or an inward (positive) force of 9086 lb. Since DYSEA nodal points 31 and 32 are at Elevations 1065.2 inches and 1125.7 inches respectively, the RELAP4 pressure / force at Elevation 1089.14 inches is distributed accordingly. { Consequently:
*See Table 6.A-8 ,
6.A-15
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 () F 31 = [ 1125.7 - 1089.14 } (9086) = 5491 lb, and ( 1125.7 - 1065.2 / F32.= 1065.2 - 1089.14 (9086) = 3595 lb. 1065.2 - 1125.7 These values' can be compared to the computer-calculated DYSEA01 resul ts, which are 5632.6 lb and 3252.7 lb respectively. In the matrix displacement method of structural analysis, exter-
. nally applied nodal forces and moments are required to produce nodal displacements . equivalent to those that would be produced by . forces or pressures applied between nodes. GE considers the external moment effects for La Salle AP to be negligible because of the close nodal-spacing of the La Salle RPV mathematical !
model. O v C') v 6.A-16
- _ _ _ _ . . .~ . _ _._ - _.
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 r5 6.A.5 Sacrificial shield, Annulus Pressurization, and RPV V Loading Data This subsection provides a brief description of the analyses
- performed and the nodalization schemes, force constants, and load centers for the recirculation and feedwater line breaks. These data are used as inpu to the pressure to force conversion calculation.
The pressure. responses of the RPV-sacrificial shield wall annulus - to postulated pipe breaks at the RPV nozzle safe-end to pipe weld in a recirculation outlet line and a feedwater line were investigated using the RELAP4 computer code. Throughout the' analyses the following assumptions were made:
- a. RPV thermal insulation displaces to the shield wall while retaining its original volume and leaving its support structure in place.
- b. Insulation above the shield wall yields to elevated pressures and blows out into the drywell allowing venting of' annulus at the summit of the shield wall.
- c. Sacrificial shield penetration doors remain . closed, allowing for limited venting of the annulus through all nozzle penetrations.
The nodalization schemes for both studies remain consistent with the guidelines cited above, with the exception of the region directly above the break, where it was anticipated that a finer mesh would be necessary to properly account for the highly localized pressure gradients expected there (see Figures 6. A-16 a nd 6 . A- 17 ) . The final nodalization was determined on the basis . of available sensitivity studies for similar analyses. The mass and energy release rates were derived with the methods outlined in Subsection 6. A. 2. The blowdown rates for the. recirculation outlet line break analysis account for actual pipe displacement, while those for the feedwater line reflect an assumption of instantaneous pipe displacement (see RELAP4 input listings, Tables 6. A-6 and 6. A-7) . The specific loading data complied for the NSSS adequacy evaluation for postulated pipe breaks within the annulus consists of the time-pressure history (psia) and two time-force (1bf) histories for each node within the annulus. The latter two histories represent integrated forces acting through the center of each node on the RPV and the sacrificial shield Wall respectively. The time-f orce histories were generated by taking the product of the node pressure and a predetermined constant, n y or nss, which accounts for the curved surface of the RPV and the sacrificial shield respectively (see Tables 6. A-8 and 6. A-9) . (]) 6.A-17
, LSCS-FS AR AMENDMENT 39 OCTOBER 1978
' O N/ The two loading histories, one for the RPV and one for the shield wall, are defined below.
+A0/2 1 D2 (6. A-9)
F y, _ Pg E g R cos Ode - (EP g n Ej ' 1 3 1 4 1
~60/2 1
D*
= P g 2E g Ry - sin - ( A0/2) -P g (E n Ej
= P.
1 ny where: Fy, E nodal resultant force on RPV (lbf) , 1 E node absolute pressure (poia) , Pi 1 E node height (inches) , 1 Ry E RPV radius (inches) , A0 E azimuthal width of node (degrees) , and D g, E pipe OD (in.). J (6. A- 10)
+A0/2 2 F gg, , P g A g R gg cos Od6 - (EP g nD ss g 1 3 \4 1
-60/2 I j :
l s 1
= Pg 2E g R (E nD 2 ss j ss sin (60/2 - P g l
3 I4 1
\
= Pg n ss l
l l l where: F E nodal resultant force on shield wall (lbf) , ss. () P. 1 E node absolute pressure (psia), 6.A-18
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 1- E node height (inches), f] 1 R shield wall inner radius -(inches) , ) ss 60 E azimuthal width of node (degrees) , Dss 5 penetration ID (inches) , and , (60) s 2 E E proportionality factor = , 2 O 5
.- O-
- 6. A-19 i- - -
~
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 p. i 6.A.6 Jet Load Forces This subsection provides the jet load forces which result from pipe separation during the postulated accident.- The pipe whip schematic is shown in Figure 6. A-7, and the resulting loads are listed in Table 6.A-1. These loads are applied to the appropriate nodal points for input to the DYSEA01 computer program. The DYSEA01 program input is provided in Table 6.A-11.
'( )
l l l m, 6.A-20
.LSCS-FSAR AMENDMENT 39 OCTOBER 1978 O 6^7 necircu1eeto= e#a reea eter tt#e areex rorcioa r=#ceioa The time force histor.tes provided in Tables 6. A-11 and 6. A-12 are those values converted from the time-pressure histories which were calculated with the RELAP4 computer program. These time forces histories are used as input to the DYSEA01 computer program.
O O 6.A-21
-s p }
V V TABLE 6.A-1 TIME HISTORY FOR POSTULATED RECIRCULATION SUCTION PIPE RUPTURE * , ** Pipe Displ. Pipe Velocity Pipe Acceler. Rel. Displ. Total Displ. Restr. Lcad Restr. Load Blowdowq Time At Restraint At Restraint At Restraint Cf End Of End Comp. PD1 Comp. PD2 Force , (sec) (in.) (ft/sec) (ft/sec2) (in.)' (in.) (lb) (lb) (lb) 0.00153 4.147E-02 3.547E 00 1.679E 03 0. 4.64BE-02 0. O. 346919. 0.00233 8.294E-02 4.889E 00 1.655E 0 3 0. 9.295E-02 0. O. 346919. L 0.00297 1.244E-01 5.932E 00 1.645E 0 3 0. 1.394E-01 0. O. 346919. 0.00351 1.659E-01 6.816E 00 1.640E 03 9. 1.859E-01 0. O. 346919. l 0.00398 2.074E-01 7.597E 00 1.635E 03 0. 2.324E-01 0. O. 346919. 0.004#1 2.488E-01 8.304E 00 1.632E 03 0. 2.789E-01 0. O. 346919. -$ m o i 5 0.00481 2.903E-01 8.955E 00 1.630E 03 0. 3.253E-01 0. O. 346919. -Y L- 2 4 u 0.00519 3.318E-01 9.561E 00 1.628E 03 0. 3.718E-01 0. O. 346919. > w 0.00554 3.732E-01 1.013E 01 1.626E 03 0. 4.183E-01 0. O. 346919. 0.00587 4.147E-01 1.067E 01 1.624E 03 0. 4.648E-01 0. O. 346919. 0.00(.87 5.427E-01 1.077E 01 3.194E 02 2.689E-02 6.351E-01 50588. O. 346919. 0.00787 6.742E-01 1.ll7E 01 4.35sE 02 9.147E-02 8.471E-01 108204. O. 346919. 0.00887 8.108E-01 1.159E 01 3.863E 02 1.808E-01 1.089E 00 168037. O. 346919. 346919. 0 5 O.00987 9.519E-01 1.190E 01 2.419E 02 2.875E-01 1.354E 00 229892. O. ' iliR O.01087 1.096E 00 1.203E 01 3.532E 01 4.076E-01 1.636E 00 293042. O. 346919. g5 o 0.01187 1.240E 00 1.19 4 E 01 -2.099E 02 5.388E-01 1.928E 00 356421. O. 346919. M $ .
- Output Parameters are listed at the end of this table.
** Except for the restraint load components PD1 and PD2, all variables l below are in a direction parallel to the blowdown force.
O O O TABLE 6.A-1 (Cont'd) Pipe Displ. Pipe Velocity Pipe Acceler. Rel. Displ. Total Displ. Restr. Load Restr. Load Blowdown Time At Restraint At Restraint At Restraint Of End Of End Comp. PD1 Comp. PD2 - Po rce (sec) (in.) (ft/sec) (ft/sec 2) (in.) (in.) (lb) (lb) ' (lb) 0.01287 1.381E 00 1.158E 01 -4.744E 02 6.802E-01 2.228E 00 418752. O. 346919. 0.01387 1.517E 00 1.096E 01 -7.414E 02 8.316E-01. 2.531E 00 478650. O. 346919. 0.01487 1.643E 00 1.007E 01 -1.027E 03 9.934E-01 2.835E 00 538908. . 0. 346919. 0.01587 1.757E 00 8.948E 00 -1.197E 03 1.166E 00 3.136E 00 581800. O. :346919. 0.01687 1.857E 00 7.672E 00 -1,335E 0 3 1.350E 00 3.431E 00 618871. O. 346919. 0.01787 1.941E 00 6.278E 00 -1.438E 03 1.543E 00 3.719E 00 649762. O. 346919. 0.01887 2.008E 00 4.801E 00 -1.504E 03 1.746E 00 3.996E 00- 674226. O. 346919. e m 0.01987 2.056E 00 3.279E 00 -1.531E 03 1.956E 00 4.261E 00 692131. O. 346919. 8
~
> Y g 0.02087 2.086E 00 1.751E 00 -1.519E 03 2.172E 00 4.510E 00 703465. O. '346919. g w >
0.02187 2.098E 00 2.567E-01 -1. 469E 0 3 2.392E 00 4.744E 00 708338. O. 346919. 5 0.02222 2.098E 00 0. O. 2.470E 00 4.822E 00 708572. O. '346919. 0.02242 2.098E 00 0. O. 2.513E 00 4.865E 00 708572. O. 346919. 0.02262 2.098E 00 0. O. 2.555E 00 4.907E 00 708572. O. 346919. 0.02283 2.098E 00 0. O. 2.598E 00 4.950E 00 708572. O. 346919. op 0.02304 2.098E 00 0. O. 2.640E 00 4.992E 00. 703572. O.
.QR 346919. oz mo 0.02325 2.098E 00 0. O. 2.683E 00 5.035E 00 708572. O. 346919. N"Eg
.W 0.02347 '2.098E 00 0. O. 2.725E 00 5.077E 00 708572. O. 346919.
- w 0.02370 2.098E 00 0. O. 2.768E 00 5.120E 00 708572. O. 346919.
0.02393 2.098E 00 0. O. 2.810E 00 5.162E 00 708572. O. 346919. _ _ _ - - . -- .g..,sa,-- y --
. -:w- -.y ,---:--.----.-----,_-_- - - - - - . ._--,__..-m.,..n- _x. ,.- -
z R r D V k- k/
' FABLE 6. A-1 (Cont'd) l Pipe Displ. Pipe Velocity Pipe Acceler. Rel. Displ. Total Displ. Restr. Load Restr. Load Blowdows Time At Restraint At Restraint At(ft/sec Restrgint Of End -Of End Comp. PD1 Comp. PD2 Force (sec) (in.) (ft/sec) ) (in.) (in.) (lb) (1b) (lb) 0.02417 2.098E 00 0. O. 2.853E 00 5.205E 00 708572. O. 346919.
0.02442 2.098E 00 0. O. 2.895E 00 5.247E 00 708572. O. 346919. 0.02467~ 2.098E 00 0. O. 2.938E 00 5.290E 00 708572. O. 346919. 0.02494 2.098E 00 0. O. 2.980E 00 5.332E 00 708572. O. 346919. 0.02522 2.098E 00 - O. 3. 3.023E 00 5.375E 00 708572. O. 346919. 0.02551 2.098E 00 0. O. 3.065E 00 5.417E 00 708572. O. 346919. 0.02582 2.098E 00 0. O. 3.108E 00 5.460E 00 708572. O. 346919. 7' O.02614 2.098E 00 0. O. 3.150E 00 5.502E 00 708572. O. 346919. E'
> Qa g 0.02649 2.098E 00 0. O. 3.193E 00 5.545r 00 708572. O. 346919. M 0.02687 2.098E 00 0. O. 3.235E 00 5.587E 00 708572. O. 346919. '"
0.02728 2.09 8E 00 0. O. 3.278E 00 5.630E 00 708572. O. 346919. 0.02774 2.098E 00 0. O. 3.320E 00 5.672E 00 708572. O. 346919. 0.02827 2.098E 00 0. O. 3.363E 00 5.715E 00 708572. O. 346919. 0.02893 2.098E 00 0. O. 3.405E 00 5.757E 00 708572. O. 346919. p 0.02992 2.098E 00 0. O. 3.448E 00 5.800E 00 - 708572. O. 346919. d go-xE 4
$a t
5 0 i m ' < - - g g g
.1 LSCS-FSAR ' AMENDMENT 39 R OCTOBER 1978 1
( TABLE 6.A-l ~ (Cont'd) cOutput Parameters Summary
' Effective. ' Length from Restraint clearance restraint to loading (inches) break (ft) direction 0.415~ 3.542; 0 degrees Pipe bending ~
Pipe rotation Max. allowable- i strain stability. bending moment. limit (in/in) limit (degr.) (ft-lbs) 9,054E-02 7.7815 1417307 Impact Velocity = 10.67 ft/sec Impact Time = 0.0059 seconds Number of bars Defl. of struc. Defl. of restr.. composing:the. in direction of in direction of restraint thrust.(in.) thrust'(in.) 2 0.7086 0.9754 Force on restr. Force'on struc. Time at peak
-As.(') in direction of in direction of dynamic load thrust (lb) thrust (lb) (seconds) 708572 708572 0.0221 Total energy Energy absorbed Energy absorbed absorbed by by the by the the restraint structure bottom hinge (ft-lb) (ft-lb) (ft-lb) 30522 20920 1956 Energy absorbed Restraint Restraint by the top . load (peak) load. (static) top hinge components (1b) components (lb)
(ft-lb) PD1 PD2 PS1 PS2
- 0. 708572 0. 138258 0.
C 4 6A-25 , i
, ,, - . - . _ , , - - . . - , , . . . _ , . . . - - . , _ - . , ~ . _ , . . . . - . . _
. =.- .. - .. .-
LSCS-FSAR AMENDMENT 39 i OCTOBER 1978 j ). TABLE 6.A-1 (Cont'd) Relative defl. Total defl. of pipe end of the pipe end - in the direction of the thrust (in.) 3.4649 5.8168 Defl. time Total , for pipe end time of (seconds af ter ' impact) movement 0.0250 0.0309 Energy absorbed Total absorbed by the energy. restraint hinge (ft-lb) -
~ ( f t-lb) 115445 168843 Pipe'defl. Pipe defl.
at. restraint at the break components'(in.) l components (in.) XR1 XR2 XP1 XP2 c- 2.0986 0. 5.8168 0. 1 6 1 _
- 6. A-2 6 >
. .. . - ...-. . _. . . . . . . .._-.-.. ~ - . - . .. .. _ . .
,4
.LSCS-FSAR AMENDF;42 39 ;
OCTOBER'1978 TABLE.6.A-2 ACOUSTIC' LOADING ON REACTOR PRESSURE VESSEL SHROUD
. ACOUSTIC.
TIME LOAD (msec) (kips) i 0- 0 ,
. l'. 2 0 !
I 1.6 150 2.0 320 2.5: 650 2.8 250 3.0 100 3.2 0 O b 1 O- l' 6.A-27
,,..n-..mm.-=.-e.mewes,.===eme=e-a
i LSCS-FSAR AMENDMENT-39 OCTOBER 1978 I TABLE 6.A-3 7
)
kPV COORDINATES OF NODAL. POINTS
'N'ODAL COORDINATES NODE- ,
NUMBER X ORDINATE Y- ORDINATE Z-ORDINATE 1 -912.000 774.000 ~1563.0001 i 2' '-912.000 774.000 1556.000 l 3 -912.000 774.000 981.200 4 -912.000 774.000 740.000 5 -912.000 774.000 1356.000 6 -912.000 774.000 1316.000 7 -912.000 774.000 1279.200 8 -912.000 774.000 1240.400 9 ' 912.000 774.000 1201.600 10 -912.000 774.000 1163.600 11 -912.000 774.000 1141.700
- 12. -912.000 774.000 '1125.700 13 ' -912.000' 774.000 1065.200-14 -912.000 774.000 1035.200 '
15 -912.000 774.000 1021.300 16 -912.000 774.000' 994.200 17 -912.000 774.000 1601.700 l 18 -912.000 774.000 1559.700 () 19 -912.000 774.000 1499.700 20 -912.000 774.000 1436.900 21 -912.000 774.000 1398.500 22 -911.000 774.000 1318.000 ! 23 -912,000 774.000 1279.200 ' 24 -912,000 774.000 1240.400-25 -912.000 774.000 1201.600 l 26 -912.000 774.000 1163.600 ; 27 -912.000 774.000 1141.700 28 -912.000 774.000 1125.700 ; 29 -912.000 774.000 1021.300 , 30 -912.000 774.000 1935.200 31 -912.000 774.000 1065.200
.32 -912.000 774.000 1125.700- i 33 -912'.000 774.000 1141.700 34 -912.000 774.000 1163.600 35 -912.000 774.000 1201.600 i 36 -912.000 774.000 1240.400 37 774.000 l
--912.000 1279.200 i 38 -912.000 774.000 1318.000 39 -912.000 774.000 1356.600 40 -912.000 774.000 1398.500 41 -912.000 774.000 1436.900 42 -912.000 774.000 1499.700-43 -912.0,0 774.000 1559.700 l 44 -912.000 774.000 1563.600
' {])
6.A-28.
LSCS-FSAR'- AMENDMENT.39 OCTOBER 1978 j ( )- TABLE 6.A-3 (Cont'd) NODE NODAL COORDINATES NUMBEW. X- URDINATM Y-UND1NATE Z - ORDINATE 45 -912.000 774.000 1601.700 46 -912.000 774.000 1619.800 47 ' -912.000 774.000 1724.200 48 -912.000 774.000 1743.600 49 -912.000 774.000 1768.200 50- -912.000 774.000 1817.100 51 -912.000 774.000 1866.000
- 52. -912.000 774.000 1563.000.
53 300.000 774.000 886.000 54 -912.000 774.000 446.000 55 -912.000 774.000 318.000 56 -912.000 774.000 0. 57 -912.000 774.000 740.000 o a 6.A-29
,_ . . _ . . . . . . , . , - . . , - - , _ . . _ , . ~ . _ . . . _ . , _ . _ . _ . . . ~ . . _ - . . _ _ -
w/ (- v v TABLE 6.A-4 MAXIMUM MEMBER FORCES DUE TO ANNULUS PRESSURIZATION COMPONENT DESCRIPTION ELEMENT NUMBER FEEDWATER RECIRC. JET REACTION Top guide (L)* 4 22.20 38.00 29.0 Core plate (L) 7 20.80 42.00 30.0 Fuel support (L) 8 19.00 69.00 74.0 CRO housing (L) 9.10 22.00 70.0 CRD housing (M) .24 .56 1.9 [ b' n. Shroud head (L) 19 59.80 78.00 133.0 ? m Shroud head (M) 19 6.40 5.90 6.1 $ lc Shroud support (L) 26 184.00 296.00 246.0 Shroud support ~(M) 26 19.80 40.00 22.0 Vessel skirt (L) 50 1220.00 3204.00 1858.0 Vessel skirt (M) 50 216.00 221.00 130.0 o Pedestal cont. (L) 3 486.00 2325.00 859.0 0 [g;i Pedestal cont. (M) 3 326.00 85 680.00 206.0 gg Z Stabilizer (L) III 1722.00 1949.00 746.0 [8 ww CRD support beam (L) 4.50 27.00 50.0 **
*(L) Load - 10 x lb (M) Moment - 106 x in. x lb All loads incorporate appropriate factor to account for shell behavior
. ~ _ _ _ _ _ - - ___ _ - - _ _ = _ _ - _ - - - _ - - _ ~ _ _ _ _ - _ - . _ .
0; 0: O . TABLE 6.A-5 MAXIMUM ACCELERATION
- DUE TO ANNULUS' PRESSURIZATION (in./sec )
i RECIRC. LINE-COMPONENT DESCRIPTION NODE NUMBER FEEDWATER- BREAK- JJET LOAD ~ AP line 9 80 283- '675-4 CRD-guide tube 11 86 '298 309-Separators 17 155 342 306_.
-Head spray 51 178 416 898 g Steam dryer 46 118- 200
# 451. @
O Feedwater_sparger 43 109 157 s 538' $ Jet pump 38 N :
'133' ~362 406 !
, :RPV 30 62 253 '514s 1-j RPV (bottom) 16 61-- 254 598- , Shield wall' 2 ~282 398 229 Top of shield wall -1 190 326 254- @ @( to lr Fuel 5 .. 74 198' ~ 394- #$. He Fuel -e 7 gg. 27 '51 .77 Fuel 9 80 283 675
*All accelerations incorporate'a factor to account ~for shell behavior.
4 T
. r.y, . . - _ . . _ - . , , ,,
j .g LSCS=PSAR AMENDMENT 39 OCTOBER 1978 TABLE 6 A-6 RELAP 4 INPUT DATA, RECIRCULATION LINE OUTLET BREAK I eLA5&LLE RPV=5MIELO ANNULU5 PRESSURl!ATION Sfu0y == h5LD CALC N0'3C7*0976*001 2 ,* PROJECT NO 9264 00 R.M. HOGAN == DeLeR08th50N == NUCLEAR ANALV5TS-3- *RECIRCULAil0N OUILET LINE #REAK
-9 '*
5
- CASE .A= BASE LI$7tNG 12/27/76
- 6. * .
i 7' . *2395678904239567890123456789012395678906239567890123956789082395678901239567890
'8
- PROBLEM O!MENSIONS 9 '* CARD - LOMP*NE0l=NTC==NINP=NVOL=NBU8=Nf 0V=NJuh=N0hE=NF LL=h0hE (
10' 010001 =2 0 3 6 38 0 0' 86 09 0 1 00000 t 11
- 12
- PROBLEM CON 59 ANTS 13 010002- 00 10 19 +
15
- TIME STEPS . ..
le 030080 I 'l 10 'O. 0 0001 I E=04 - 0 025 17 030020 ;l. I 5 0 0 001 IE=06 ' 02 .
'88- 030030 I '
l' O .0eol .lE=0e 10 19
- 20 - -e1Ri? CONTRC~5 s 28- 090060. I O 0 De2' Oe0 - *EhD PROBLEM'Oh ELAP5ED TIME' 22 090020 2 1 0 0 00 00
- ACTION s2 ON ELAP6ED *,lME IFILL) 23 .090030' 3 9 -30 .36 30 00
- ACTION 83 ON DP 10PEh VALVEB 29 090090 9 9 31 36 30 00 *ACiloh e9 ON DP 106EN VALVEl 25' 090050 5 9 32 34 3eo 00
- ACTION e5 CN OP 10 PEN v&LVEl 26 -090060 6 9 33 36 30 00
- ACTION e6 CN DP 10 PEN VALVEl'
'27
- 28' *BE6tN VOLUME DATA 29 *23956789012345678901239567890l2395678901239b67890123956789082395678901239567890 '
30 ev0LUME 8==R== PRE 55.=fEMP=GUAL=====v0LUME==M1===== Mis ===fP==FLonA===0lAMvaELEv== 38 050011 0 0 15.95 -le 0 996 600 6 5 07 5 07 ' 0 . 18.90 00 755.29 32 050028 0 0 15e95 -1. 0 996 100.6 5.07 5 07 0 48.90 00' 755 29 33 050031 0' 01 15e95 =le De996 100.6 5 07 5 07 0' .58e90 Geo - 755 29 39 050091 0- D lle95 -le 04996 150,9 6.07 5 07- 0 23 36 l00 .755 29
. ,0 35 36 050058 0 0 15.95' al. 0 996 050061 0 0 85e95 '=le 0 996-150.9 121 0 5.07 7.97 5 07. 0 - 23e36 7 97 0 00 20 98' Dec 755 29 760 36 e 37 050071 0 0 15e95 =le 0 996 121 0 7.97 7 97'.0 '20 98 00 760 36 38 050081 0 0 15e95 =le 0*996 128 0 7e97 7.97 0 20 98 Dec 760 36 39 050091 0 0 15.45 =le De996 181 5 7.97 7 97 0 25.69 00 740 36 90 050101 0 0 15.95 ' -l e 0 996 181 5 7.97 7 97 0 25.69 Geo 760 36 98 050181 0 0 15.95 ' ale 0*996 39 87 6.92 6.92 0 30,02 00- 767 83 42 050123 0 0' 15.95 =le- De996 59.28 9.90 9.90 0 40 50 00 767.83 93 050131 0 0 15.95 =le 0 996 61 99 9.90 9.90 0 10 50 00 747.83
- 99 050891 0 0 15,95 =le' Oe996 88 93 9.90 9.90 0 13 97 00 767.83 95 050150 0 0 lbe95 =1.. 0*996 .80 59 9.90 9.90 ' O 43 97 00- 767.83.
96 050165 0 0 15 95 ale 0 996 26.77 2.47 2 67 0 8.93 Dec 779 75 97 050871 0 0 15.95 =le 0*994 52el8 9.69 9.69 0 10 30 00 772.73 90 050180 0 0 15.95 =le De996 .52.l8 geet 9 69 0 10 30- 00 772 73 99 050198 0 0 15.95' =le 0 996 78 28 9.69 9 69 0 13 27 00 772 73' 5 50 0$0201 0 0 15. 9 5 ' a l ,' De996 ' 77 39 9.69 9 69 0 13 27 00 772.73 58 050218 0 .0 15.95 =le 0 996 67e98 6.98 6 91 0 12.99 00 777.92 52 050224 0 0 15.95 =le be996 67.98 ee91' 6 91 0 12.99 00 777.92 53 050231 0 0 15.95 =1. 0 996 - 67e98 ee98 6 98 0 12.99 Dec 777.92 59 050291' O -0 15.95 =le 0 996 801 2 6.98 6e91 0 15 5h 00 777.42 55 050254 0 'O 45.95 =I. 0*996 101,2 6.91' 6 91 0 15 52 00 777.92 56 '- 05C261 0-0 15.95 =le 0 996 171 8 9.59 9 59 'O 88 61 .00 783 83 57 050278 0 0 15,95 ' = l e 0 996 155.8 9.59 9 59 0 '88 61 Dec 783.83
'58 050288 0 0 15.95 -=le 0 996. 155 8 '9.59 9 59 0 .18.61 00 783 93 59 .050291 0 0. 15 45' =le 05996 171.8 9.59 9.59 'O 88 61 00 785.a3 60 050301 0 0 15.95 =le 0i996 155.8 8.81 8.81 0 17.86 . 00 793.92 el 050381 .0 O' 15 95- al. 04996 153.9 '8.88 8.81 0 87.86 00 793 92 62 .050321 0 0 15.95 =le 0 996. 193 9 8.81 8.88' O 17.86 00 793.92 63 .050338 0 0 15.95'.=le 0 996- 169.1 Se8l '8.81" 0 47 86 00 793 92 69 050391 0 0 15.95 =le. Di996 19 76 6.92 6 92' O 10 02 00 747.83 .
65 050358.'0 0 15 95- ele De996 19 52- 9.92 9 92- 0 7 09' ^00- 769 54 i 66 050364 0 0.'lle95 -le 0 557 16345e 98 0 9t*0 0 900* 00 793 42 67 050378 0. 0 15 95'-le 0i557 Ileebe 12el 12el .0 965. 00 788 32 48 050386 0 0 15 95 -I.' De557 82776. 99 7 99 7 0 1850. 00 736 62 49 ' ev0LUME=Ba=R== PRE 55*=fEMP=GUAL**==*v0LUME==MT===== MIX ===TP==FLO9A===01AMvaELEv=* 70 . *2345678901239567890123956789012345678901239567890123956789082395678908234567890 78 *END VOLUME DATA 72 e-73 *BE6th MOR120NTAL FLOW PA[M5 klTMih 5 5eAhhuLUS 79' *2395478901234567890123956789012395678908239567890823956789082395678908239567890
- j L 75 *JUNCT====lh=0f=P=V=FLO.=AJuha*=2JUN===*]N=====FJUF =FJUR==v=C=l=EG=0M===CC==C=E 6.A=32
-- -. ,, , a_.., __, ~,.-.-..,_.4-_._,.-,_.,_,%.,,_._...__,~_... _ _ _ . a,-.-.-
LSCS-FSAR' AMENDMENT 39 OCTOBER 1978 i TABLE 6.A-6 (Cont'd)
-j'"g4 ina l. L 76 080011 1 2 0 0 De0 19 86 757 82 0 90 0 29 0.00 0000 00 0.6 40
- 77. 080021- 2 30000 19 86 757.82 0 90 0 29 0.00 .0 0 0 0 00 0.6 I J 78 080031- 3 9 0000 19 86 757 82 0 50- 0 90 0.00 0000 Dec 0.6 1 (
79 . OP9091 9 50000- l9386 757.82 0 60 0 92 0.00 0000 00 0.6 30 80 080058 'a .7 0 0 0 0 20i19 769 10 0 30- 0 22 0.00 0000 00 0.6 1 0 81 080061 7 8 0000 20*19 769 40 0 30 0.22 0.00 0000 00 0.6 1 0 82 080071 8 9 0 0 Dec 20 59 769*10 0 38 0 39 0.00 0000 00 0.6 10 83 080003 9 10 D 0 Dec 20*19 769 10 0 95 0 91 0.00 0003 00 0.6 1 0 89 080091 35 39 0000 7409 '772 02 0 30 0 85 0.00 0000 00 0.6 10 85' 080808 39 11 0000 10 02 771 29 0 32 0 35 0.03 0000 Dec 0.6 1 0 86 080181 !! 120000 7.e7 770 28 0 69 0 56 0.00 0003 00 0.6 1 0 87 080821 12 130000 7*09 770 28 0 90 0.89 0.00 0000 0e0 0.6 10 88 080531 13 19 0 0 Dec 7*09 770 28 1 13 0 85 0.00 0003 00 0.6 3 0 89 080891 19 150000 7109 770 28 a.35 le69 0.00 000 3 00 0.6 1 0 90 080858 Il 170000 2 ell '.773 79 2 26 0 05 0.00 0000 00 0.6 1 0 91 .080163 16 170000 3*87 776 09 le96 0 38 0.00 0003- 00 0.6 1 0 92 '080371 17 18 0000. 6*79 775 07 0 99 0 83 0.00 000 3 0e0 0.6 10 93 083481 38 19 0000 6379 775 07 let? 0 85 0.00 0003 00 0.6 1 0 99 080191 19 200000 6 79 776 07 l.98 le63 0 00 000 3 00 0.6 1 0 95 080201 28 22 0 0 0 0 9i83 780 62 0 65 0.36 0.00 0003 00 0.6 1 0' 96 080211 22230000 9*83 700 62 0 65 0 36 0.00 000 3 00 0.6 1 0 97 080221 23 29 0000 9 83 780 62 0e8I 0 67 0.00 000 3 00 0.6 10 98 .080238 29 25 0 0 0 0 '9*83 780 62 0*97 0 68 0 00 0003'00 0.6 10 99 080291 26 27 0 0 0 0 19*68 788 62 0 65- 1 28 0.00 000 3 00 0.6 10 100 080258 27 28 0000 19*68 788 62 0 65 0*68 0.00 000 3 00 0.6 1 0 108 080261 28 29 0000 19*68 788 62 0 65 1 28 0.00 000 3 00 0.6 1 0 102 080278 30 31 0000 -13 99 797 83 'De?! l.27 0.00' O O O 3 00 0.6.1 0 103 080281 38 320000 13.g9 797.83 0 71 tel3 0.00 0003 00 0.6 1 0 109 080298 32 330000 13*49 797 83 0 78 1 27 0.00 000 3 00 0.6 10 105 *JUNCT. ..IN=0T=P=v=FLv.*AJUN..*2JUN==.=lN=*===FJUFa=FJUR*=W* Cal-E0=0M===CC**C=E 106 +2395678901239567890123956789012395678908239567890123456789012395678908239567890 807 *END HORI2ONTAL FLou PATH 5 mlTHIN 5 5* ANNULUS 108
- 809 *BEGIN VERTICAL FL0m PATHS stTHIN.5 5eANNULUS 110 823956789012395678901239567890123g5678901234567890823956789012395678901239567890 til *JUNCT====lN=0T*P=V-FLO.*AJUN===2Jbh====lN=****FJUFe*FJUP==v= Cal *EGe0M*==CC*=C.L 312 080301 6 1 0000 18 90 760 36 0 33 0 03 0.03 1 0 0 3 'O.0 0.6 10 O ll3 119 185 080311 080321 080338 7
8 9 20000 18*90 760 36 0 33 30000 18e90 760 36 0 33 9 0000 23e36 760 36 0 22 0 03 0 03 0 03 0.03 0.03 0.03 'I 1 003 00 300 3 OO3 00 0*O 0.6 t0 0.6 6 0 0.6 1 0 lie 080194 IO 50000 23A36 760e36 0 22 0 03 0.03 1000 00 0.6 1 0 117 080351 39 60000 3 61 767e83 le90 tel3 0.90 1 003 00 use 1 0 188 080363 Il 6 0000 3kel 767 83 8 90 tel3 0.90 1003 00 0.6 10 lit 080371 12 7 0 0 Dec 7 22 767 83 0 62 8 13 0.90 1 003 00 0.6 10 120 080381 13 b0000 7 22 767 83 0 62 1 90 tel7 1 00 3 00 0.6 1 0 121 080398 19 9 0 0 0e0 10e89 767 83 0 98 1 13 0.90 1 000.00 0.6 10 122 080908 15 to 0 0 0 0 10 89 767 83 De98 lel3 0.90 1 000 00 0.6 10 123 080911 12170000 8 56 772 73 0 56 0 96 0.00 1 003 00 0.6 10 129 080921 43 18 0 0 Dec 8 56 772.73 0 56 0 96 0.00 1 000 00 0.6 10 425 080931 19 19 0000 19e50 772 73 0 33 0 59 0.00 1 000 0e0 0.6 1 0-126 080991 15 200000 19 50 772 73 0 33 0 68 0.00 1 000 00 0.6 10 127 080958 39 le0000 5*99 779 75 0 99 0 33 0.00 1 0 0 .3 00 0.6 i fJ 120 080961 Il 16 0 0 De0 Se9g 779 75 0 99 0 88 0.00 1 00 3 00 0.6 1 0 129 080971 to 21 0000 7*72 777 92 0 99 0 67 0.00 1 000 00 0.6 10 1 130 080988 17 22 0 0 0 0 7572 777*92 0 59 0 68 0.00 1 000 00 0.6 1 0 131 080998 88 230000 7 72 777 92 0 59 0*68 0.00 1 000 00 0.6 1 0 132 080504 19 29 0000 18 57 777e92 0 40 0 68 0.00 4 000 00 0.6 1 0 1 133 08051l 20 25 0 0 0 0 lle57.777.92 0 90 0 68 DeJO 1 000 00 0.6 30 I 139 08052l 21 26 n000 7372 783e83 0 80 0 96 0.00 1 000 00 0.6 10 i 135 080531 22 26 0 0 0 0 3*86 783e83 le60 1 09 0.00 1 003 00 0.6 1 0 j 136 080698 22 27 0 0 0 0 3k86 783 83 1 60 1 09 0.00 3 000 00 0.6 1 0 l 137 080551 23 27 0000 7*72 783 83 0 80 0 69 0.00 4003 00 0.6 n0 i ! 134 080568 29 28 0000 11 57 783 83 0 59 0 96 0.00 1 000 Dec 0.6 1 0 i
- l. 139 08057) 25 29 0000 11 57 7B3e83 0 59 0 97 0.00 1 000 00 0.6 1 0 l 190 080581 26 30 0 0 0 0 18*57 793*92 0 60 l*00 0.00 i OOO 0 0- 0.6 l- 0 .
198 080691 27 31 0000 11557 793e92 0 60 le09 0.00 1 000 Geo 0.6 1 0 I 192 080605 28 32 0 0 0 0 18 57 793e92 0 60 0 97 0.00 1 000 00 Dee1 0 l 143 08C#al 29 330000 11 57 793e92 0 60 1 00 0.00 1000 00 0.6 I O 199 'Ju1CT*=**lN=0T=P=W-FLO==AJUN-*=2JUN====lN=**==FJUFa=FJUR**v= Cal =EG=0M*==CC==C.E [ 695 +23"5678908239567890123956789012395678908239567890123956789012395678901239567890 l- 896 *END WERTICAL FL0m PATHS WITHIN Se5* ANNULUS 197 . . 198 *BEGIN. F10m PATHS TO CONTAINMENT = PENETRAfl0N5 mlTH 5HIELDING 000R5 149 *2395678906234b67890123966789012395676901239567890123956789082595678908239567890 150 ' JUN C T = = == l N*0 T = P= V-f L O. = A JUN= =* 2 JUN = = = = l N= = == =F JUF e *F JUR == W *C a l = E e = 0M=== C C = *C = E 158 080628 30 36 0 i Dec 9 27 7 9 7 e 'A 3 1 06 0 75 0.00 0 000 00 0.6 1 0 152 08Gb31 38 36 0200 13 90 797 83 0 70 1 69 0.00 0000 D*O 0.6 3 0 6.A=33
. . - . - ,, , , . . . . ,, . , - - .- . . , . . -. , - . .- __ ~ - --.__.
- -~ . . . - - .
LSCS-FSAR AMENDMENT 39
' OCTOBER 1978 s
e TABLE 6eA-6 (Cont'd) 1 1 l 153 080691 32 36 0300 13 90 '797.83 0 70 le69 0.00 0 000 00 0.6 8 0 159 '080658 33 36 0 9 00 9 27- 797 83 1 05 0 75 0.00 0 0.0 0 0+0 0.6 1 0 ISS 080663 33 36 0000 2 09 797 83 leos 1 72 0.00 0003 00 0.6 1 0 lb6 080471 32 36 0000 0+68 '797.83 3 39 te71 0.00 0003 00 0.6 80 157' 080688 31 36 0000 2elo 797.83 1 81 le71 0.00 000 3,00 0.6 10 155 080691 30 36 0000 le77- 797 83 1 26 8 72 0.00 000 3 00 0.6 i 0 l 159 080701 36 370000 900e 793.92 0 06 0 05 0.00 1 003 00 0.6 1 0- j 160 080711 29 370000 1 39 788 62 1 50 8 73 0.00 000 3 00 0.6 10 ' lol 080721 28 37 0000 0 75 788+62 3 30 1 71 0.00 0003 00 0.6 i D i 162 080731 27 37 0000 0*71 788.62 3 30 left 0.00 000 3 00 .De6 1 0 l 163 080795 26 37 0000 1*39 788.62 I.50 8 78 0+00 0003 00 0.4 10 169 080751 37 38 0 0 0+0 965. 788 32 0 03 0.05 0.00 1 00 3 00 0.6 i 0 165 080768 20 38 0000 1 25 775 07 1 97 8 71 0+00 0003 00 0.6 I O l 166 080771 19 38 0 000 1 07 775 07 2 20 le78 0.00 000 3 00 0.6 1 0 ' 167 080781 18 38 0000 0 71 775 07 3 30 1 71 0.00 0 0 0'3 00 0.6 1 0 168 080791 17 38 0000 0i78 775 07 3 30 le76 0.00-0 0 0'3 00 0.6 10 169 080801 45 38 0 000 8 25 770 28 let? 1 71 0.00 0003 00 0.6 1 0 170 0806): 19 38 0000 1 07 '770 28 2 20 8 71 0.00 0003 Dec 0.6 8 0 171 080821 13 36 0000 te97 770 28 1 50 17I 0 00 0003 00 U.6 10 172 080831 12 38 0000 0 78 770 28 3 30 le78 0.00 0003 00 0.6 10 , 173 080898 II.38 0 0 0 0 0 71 772 02 3e30 1+78. 0.00 0003 Dec 0.6 1 0 179 080858 35 38 0 0 Dec. Lice 772 02 2e93 le71 0.00 0000 00 0.6 1 0 , 175 .JUNCT*=.alN=0T=P=V FLO==AJUN===ZJUN====lN=====FJUF==FJUR==V* Cal.E0 0M-==CC==C=E I 176 *2395678901239567890823456789012345678901239567890123956789012395678901239567890 177 .END FLOW-PATH 5 TO CONT &!NMENT = PENETRAfl0NS ntTM 5MIELOING 000R5 l 178
- l 179 8EGIN FILL PATH j 180 *2345678901239567890123956789012395678908234567890123956789062395678901239567890 '
181 eJUNCT====lNa0T*P=V=FLO==AJUN===2JUN*=ealN*====FJUF*=FJUR==V=C*I-E0=0M===CC==C=E 182 080861 0 35 1 000 1 00 772 02 0 00 0 00 0.00 0003 00 1.0 10 183 .JUNCT**=alN*0T=P*V=FLO.=AJUN==e2JUN===*lN=*==*FJUFe*FJURe*V=C*[=EGe0Me==CC**C.E 189 82395678901239567890123956789012395678901234567890823956789012395678901239567890 185 .END FILL PATH 186 e 187 . VALVE DATA CARDS 188 110080 *3 0 0 0 0 0 0 0.0 189 !!0020 -9 0000000.0 / 190 180030 5 0 0 0 0 Ce0 0.0 l 191 ll0040 -6 0 0 0 0 0 0 0.0 192 e 193 eFILL TA8LE DATA CARDS 199
- FILL CONTN0L 195 130100 16 2 0 0 1060. 533e 196
- CARD TIME Flow TIME FL0m TIME FLon <
'197 130101 00 00 0 002 373e 0 009 1899e l 198 130102 0 006 2976. 0 000 9963. 0 010 7088. '
199 130103 0 0173 18092. 0 019395 18092. 0 019905 Vl62e 200 l30109 0 022 10573. 0 029 18995. 0 026 12897. 20I 130105 0 028 82618. 0 030 12865. 0 038 12885. 202 130106 50 12885. 203 + 209 20h e2395678901239567890123956789012395678901239567890.123956789012395678901239567890
* * *
- 8 * *
- e . e s . e
- e . . .e s . .e e e e . e e . . e e . e e. . e .eeeeees ..ee.eeeeeeeeeeeeee.e.ege.ee. <
206 .MODEL REVISIONS j 20,- ..... ...ee......e...ees....e......e.ee.ee..e........e......e..ee.eee..e.e.ee..e l 208 e i
'1 1
1 J'~\ 'l I l l 6.A-34
n +
, LSCS-FSAR' AMENDMENT 39
. OCTOBER 1978 r*N g TABLE 6.A=7 N'
RELAP. 4 INPUT DATA, FEEDWATER LINE BREAK I eLA$6LLE RPV=5HIELO ANNULUS PRES $URl2ATION $fypy ** NSLO CALr N0 3C7 0976-006 2
- PROJECT NO 9266 00 R.M. HOGAN == 0.L. ROBINSON **- NUCLEAR ANALy$f3 J
- FEE 0 WATER LINE BPEAK 9 e 5 eCASE 'C' BASE Ll5flNG I/3/77 6 e 7
8
*2395678901239567890123*56789012395678901239567890823956789012395678908239547890 ePR08LEM O!MENSIONS 9 eCARD LOMP=NE0l*NTC**NTRP=NVOL*NBUB=NTOV*NJUN*N0NE NFLL=uCNE 10 080004 -2 0 3 8. 32 0 0 70 060 1 00000 Il e
'12 ePR08LEM CONSTANTS 13 010002 00 10 19 e 15 eflME STEP 5 le 030010 t i 50 0 0 0003 IE*06 0 025 17 030020 1 1 25 0 0 001 IE*06 02 18 030030 I I I O 0 01 IE*06 I.0 19 e 20 ofRIP CONTROLS 21 090080 8. 1 0 0 02 00 *END PROBLEM ON ELAPSED flME 22 090020 2 'l 0 0 00 00
- ACTION s2 ON ELAP9E0 TIME (FILL) 23 090030 3 9 23 30 30 00 eACTION 83 DN DP IOPEN VALVEl 29 090090 9 9 :29 30 '30 00 eACil0N #9 ON DP (OPEN VALVEl 25 090050 5 g 25 30 30 00 eACTION sl ON DP (OPEN VALVE) 26 090060 6 9 26 30 30 00 eACfl0N se CN DP 10 PEN VALVE) 27 090070 7 9 27 30 30 00
- ACTION e7 ON DP 10 PEN VALVEl 28 090000 8 9 28 30 30 00
- ACTION 88 DN DP 10 ped WALVE) 29 e 30 e8EGIN VOLUME DATA Ji e2395678901234567890123956789012395678908239567890123956789012395678908239567890 32 eVOLUME=8-=R== PRESS.=fEMP=8UAL=*=== VOLUME **HT***==MIXe==TP=*Ft0mAa==0IAMV-ELEV**
33 0500ll O' 0 15.95 .l. 0 994 150.9 5.07 5 07 0 73.36 00 755 29 39 050021 0 0 15.45 0 996 450.9 5.07 765 29
.l. 5 07 0 23.36 00 j' 35 050038 0 0 15.95 .l. 0 996 150.9 5.07 5 07 0 23.36 00 755 29
- 36 050098 0 0 15.95 .l. 0 996 150.9 5.07 5 07- 0 23.34 0.0 755 29 N 37 050051 0 0 15.95 =l. 0 996 188.5 7.97 7.97 0 23.80 00 740 36 38 050068 0 0 15.95 .l. 0 994 188.5 7.g7 7.97 0 23.80 00 760 36 39 050078 0 0 15.95 -3 0 996 188 5 7.g7 7.97 0 73.80 00 760 36 90 050084 0 0 15.95 .l. 0 996 188.5 7.97 7 97 0 23.80 00 760 36 98 050091 0 0 15.95 *l. 0 996 159.7 9.59 9 59 0 17.83 00 767 83 92 0$0301 0 0 15.95 =1 0 996 157.9 9.59 9 59 0 17.83 0.0 767.83 93 050811 0 0 15.95 =l. 0 996 157.9 9.59 9 59 0 17.83 00 767 83 99 050521 0 0 15.45 .l. 0 996 167.9 9.59 9 59 0 87.83 00 767 83 95 050138 0 0 15.95 .l. 0 996 67.98 4.98 4.91 0 12.99 00~ 777.92 96 050198 0 0 15.95 0 996 6.98 97
=1. 67.48 6 91 0 12.99 00 777.92 050151 0 0 15.95 *l. 0 996 67.98 6.98 4 94 0 12.99 00 777 92 98 050861 0 0 15.45 .l. 0 996 101.2 6.91 6 98 0 15.79 00 777 92 99 050871 0 0 15.95 *l. 0 996 101.2 6.98 6.91 0 15.79 00 777e92 50 050188 0 0 15.95 *l. 0 996 100.8 9.59 9.59 0 15.52 00 783 83 El 050191 0 0 15.95 *l. 0 996 110 0 9.59 9 59 0 15.52 00 783 83 52 050201 0 0 85.95 =1. 0 996 116.1 9.59 9.59 0 15.52 00 783 83 53 050218 0 0 15.95 .l. 0 996 171.8 9.59 9 59 0 18.64 00 783 83 59 050228 0 0 85.95 -1. 0 9g6 155.8 9.59 9 59 0 18.61 00 783 83 l 55 050231 0 0 15.95 .l. 0 996 95.22 10.58 10 58 3 13.39 00 793 92 i 56 050294 0 0 15.95 =1. 0 996 b5.63 10.58 10 58 0 13.39 00 793 92 I 57 050251 0 0 15.95 *l. 0 996 lie.2 .50.50 10 58 0 16.98 00 793 92 48 050261 0 0 15.95 .l. 0 996 138.5 10.58 10 58 0 16.98 00 7t3 92 59 050278 0 0 15.95 -1. 0 996 176.7 10.58 10 58 0 19.57 00 793 92 '
60 050281 0 0 15.95 .l. 0 996 171.8 10.58 10 58 0 19.57 Dec 793 92 el 050291 0 0 15.95 en. 0 996 16.12 9 00 9 00 0 5.92 00 796.75 62 050308 0 0 15.95 .3 7 557 16315. 98.00 91 00 0 900. 00 793 92 63 050318 0 0 15.95 =.. 0 557, 18465. 42.40 12 40 0 965 0.0 781 32 49 050324 0 0 15,95 *l. 0 557 82775. 99.70. 99 70 0 1850. 0.0 736.62 ' 45 eVOLUME*8**R== PRESS **TfMP=0UAL*=.== VOLUME ==HT*==== MIX ===TPa=FL0pA=*=01AMV*ELEV** 66 e2395678908239567890123956789082396678901239567890123956789012395678908239567890 67 eEND VOLUME DATA 48 e 69 e8EGIN HORI2ONTAL Flow PATMS hlTHIN SeS. ANNULUS 70 +2395678901239567890123956789012395678901239567890623g56789082395678901234567890 ; 73 eJUNCT*===lN=0T*P=W*FLO**AJUN===2JUN=**=lN=====FJUF=.FJUR=*V-C=l=E8=0M*==CC**C=L 1 72 080011 1 2 0 0 0.0 19.86 757 82 0.60 0.29 0.00 0000 00 0.6 30 73 j 080023 2 30000 19.86 757.82 0 40 0.93 0.00 0000 00 0.6 80 ' 79 080031 3 9 0000 14.86 757.82 0.60 0.29 0.00 0000 0.0 0.6 1 0 / ; 75 080094 5 6 0000 20 19 769 30 0 95 0.25 a00 0000 00 0.6 10 'A I 6.A-35 I
, _ - ~ _ _ . __ . _,_ . . - _ _ _ _ __ _
LSCS=FSAR AMENDHENT 39 OCTOBEP, 1978 TABLE 6.A=7 (Cont'd) V) 76 080054 6 7 0 0 0.0 77 20*l9 769e10 0 95 0.91 0.00 0000 00 0.6 1 0 78 080061 080071 7 8 0 0 Coo 20 19 769 10 0 95 0.25 0,00 0000 00 0.4 10 9 to 0 0 0.0 13 88 772 63 0 69 1.31 De00 0000 00 0.6 1 0 79 080088 to 11 0000 13 88 772.63 0 69 80 1.27 0.00 000 3 00 0.6 10 81 C80094 080801 11 120000 13 84 772e63 0 69 leal 0.00 000 3 00 0.6 10 13'89 0000 9.83 780 62 0 45 0.51 0.00 0000 00 0.6 1 0 82 0808 1 19 150000 9.83 780 62 0 65 83 080821 0.51 0.00 000 3 00 0.6 10 89 15 le0000 9.83 780 62 0 81 0.38 0.00 000 3 00 0.6 4 0 080131 le 170000 9.83 780 62 0 97 0.39 0.00 000 3 00 85 0.6 1 0 080898 18 19 0000 19 68 788 62 0 99 0.79 0.00 0003 00 0.4 1 0 84 08085l It 20 0 0 0 0 ' ig.68 87 788 62 0 99 0.83 0.00 000 3 00 0.6 8 0 080161 20 21 0000 1g.68 788 62 0 59 Dell 0.00 000 3 00 Ba 0.6 1 0 080371 28 220000 19.68 788.62 0 65 0.85 0.00 0003 00 0.6 1 0 89 08088I 29 23 0 0 0 0 5 92 798.75 0 90 0.85 0.00 0000 00 0.6 40 90 080191 23 29 0 0 0.0 16 19 798.75 0 20 0.33 0.00 0003 00 0.4 I O 98 92 080208 29 25 0 0 0 0 14e19 '798.75'O.30 0.05 0.00 000 3 00 0.6 3 0 080281 25 26 0000 leelt 798.75 0 90 le33 0.00 000 3 00 0.6 40 93 080221 26 270000 16 19 798 75 0.50 99 8.39 0.00 0003 00 0.6 40 080231 27 28 0000 16 19 798.75 0 60 0.39 0.00 0003 00 0.6 1 0 95 96 #JUNCT===*lN*0T=P=v=FLO==4JUN**=2JUN====lN=**=*FJUF==FJUR==V= Cal =E0=0M*==CC=*C*E 97 *2395678908239567890823956789012395678901234567890823956789012395678901239567890 eENO HOR 21 0NTAL FLOW PATHS WITHIN 5.S* ANNULUS 98 e 99 eBEGIN VERTICAL FLOW PATHS wlTHIN 5 5. ANNULUS 100 101 *2345678901239567890823956789012395678901234567890123956789012395678901239547890 802 eJUNCT*===lN*0T*P*V-FL0==AJUN***2JUN=***lN==***FJUF*=FJUR*=VoC=l=(G*0M**=CC==C.E 080298 5 l 0 0 0.0 23 80 760 36 0 26 0.03 0.00 1 00 3 00 0.6 1 0 103 080251 6 2 0 0 0.0 23 00 760 36 0 26 0.03 0.00 8003 00 109 0.4 1 0. 080261 7 30000 23.80 760 36 0.26 0.03 0.00 1003 0.0 0.6 10 105 080273 8 9 0000 23 80 760 36 0 26 0.03 106 080283 0.00 1 000 00 0.6 4 0 9 50000 10 89 767 83 0.59 1.83 1.28 1 000 00 0.4 1 0 107 080298 10 6 0000 Ig.89 767.83 0.59 1.83 1.24 108 100 3 00 0.6 40 080301 11 70000 10*89 747 83 0 59 1.13 1.28 80 00 00 0.6 40 109 080331 12 80000 10 89 767 83 0 59 1.28 180 lel3 1000 00 0.6 I O 080321 13 9 0000 7 22 777.92 0 83 0.96 0.00 1003 00 0.4 10 ill 080331 19 9 0000 3 68 777.42 8 66 0.96 0.00 1.0 0 3 00 0.6 1 0 112 080391 19 to0000 3 61 777.g2 1 66 0.96 0.00 100 3 00 0.6 10 ("*j 113 080351 35 100000 7 22 777.92 0 83 0.96 0.00 1 000 00 One 1 0 ( ,/ 189 080361 16 Il 0 0 Dec 'Ig 89 777.92 0 56 0.96 0.00 1 000 00 0.6 l0 115 080371 17 120000 10 89 777.92 0 56 1.01 0.00 1000 00 0.6 1 0 lie 080383 18 130000 7 75 783.83 0e80 0.68 0.00 1000 00 0.6 3 0 117 080398 19 19 0000 7e78 118 783.83 0 80 1.03 0.00 1000 00 0.6 10 119 080901 20 150000 7e71 783 83 0 30 0.96 0.00 1000 00 0.6 10 120 080918 21 16 0000 11 57 783e83 Def i 0.97 0.00 8000 00 0.0 10 080921 22 I7OO00 18 57 783.83 0.59 0.94 0.00 1000 00 0.6 3 0 121 080931 23 18 0000 3 86 1.99 122 793.92 0.70 0.00 1003 00 0.6 1 0 080998 29 le0000 3.86 793e92 1.99 0.70 De00 1 000 00 0.6 1 0 323 080951 25 19 0 0 0.0 7.78 793.92 0 97 0.98 0.00 1 000 00 0.6 1 0 129 080961 26 20 0 0 0 0 7.71 793.92 0.97 1.00 0.00 1 000 00 0.6 1 0 125 080971 27 21 0 0 0.0 11 57 793.92 0.65 0.99 0.00 1 000 0.0 0.6 1 0 i2e 090988 28 22 0 0 0 0 89 57 793e92 0 65 0.97 0.00 127 1 000 00 0.6 10 I28 *JUNCT*===lN=0T=P=v=FLO =AJUN***!JUN===*lN=***=FJUFa*FJUR*=V.CoI*EG*0M===CC=*C*E 129 *2395678901239567890123956789082345678901239567890123456789012395678908239567890
*END WERTICAL FLOW PATHS WITHlh 5.$. ANNULUS 430 e 131 eBEGIN FLOW PATHS TO CONTalNMENT = PENETRAfl0NS WITH SMlEL0lwG D00R$
132 133 *2395678901239567890423956789082395478901239567890123956789012395678908239567890 139 eJUNCT*===lN=0T*P*V.FLO==AJUN===2JUN====lN=*=**FJUF**FJUR==V* Cal =CO*0M===CCa* Cat 080991 23 30 0 l 00 1 59 798.75 3.60 1.61 0.00 0000 00 0.6 40 135 080508 29 300200 3 86 798.75 I.30 1.07 0.00 0000 0.0 0.6 40 136 080934 25 300300 7.78 798.75 1 06 1.99 0.00 0000 0.0 0.6 1 0 137 080521 26 30 0 9 00 7.71 798.75 3 06 138 1.99 0.00 0000 00 0.6 80 080531 27 30 0 5 10 9.27 798.75 0.79 2.90 0.00 0000 00 0.6 30 139 080698 28 30 0 6 00 11 57 798 75 0.65 8.82 0.00 0000 00 0.6 1 0 190 080551 29 30 0 0 0 0 0 68 798.75 3.96 le71 0.00 0000 00 194 0.6 1 0 080561 2b 300000 0 68 798 75 3.96 1.71 0.00 0003 00 0.6 10 192 080578 27 30 0 0 0 0 3 36 798 75 B.98 1.78 0.00 000 3 0.0 0.6 1 0 193 080581 . 26 30 0 0 0 0 !.36 798.75 1.70 1.73 0.00 000 3 00 0.6 10 199 080591 25 300000 0 48 798.75 3.96 1.75 0.00 0003 00 195 0.6 3 0 080608 30 38 0000 900. 793.92 0 06 0.05 0.00 3 00 3 00 0.6 1 0 196 080613 22 31 0000 0 78 788.62 3.86 4.78 0.00 0003 00 0.6 80 197 OM0621 28 34 0 0 0 0 1e39 788.62 1.70 1.73 0.00 0003 00 0.6 1 0 198 080631 20 31 0000 0 68 788.62 2.98 1.79 0.00 000 3 00 899 0.6 1 0 080698 19 al 0000 1 92 788.62 4.93 8.78 0.00 0003 0.0 0.6 40 ISO 080651 38320000 965. 788 32 0 03 0.05 0 00 8003 0.0 0.6 8 0 lbl 080661 12 32 0 0 Dec 2 89 772.63 0e90 1.71 0.00 000 3 00 0.6 10 152 080671 Il 32 0 0 0 0 2.50 772 63 1.17 8.71 0.00 000 3 0.0 0.6 1 0 l 6.A-36 l l 1
- -- , , .n-. -
-n ,+ . ..n.w .,n.. . , ,,, e
. LSCS-FSAR AMENDMENT 39 OCTOBER 1978 f"+s TABLE 6.A-7 (Cont'd).
\/
}-
153 080688 10 32 0 0 0*O 2 50 772.63 tel? le71 0.00 000 3 00 De6 1 0 159 080698 ~ 9 32 0 0 0 0 2 49 772.63 1 29 8.74 0.00 000 3 0 0' O.6 10 ISS *JUNCT***=lN*0T=P=V=FLO=*AJUNa**2JUN=**=lh'a===FJUFa*FJUR==V= Cal *Eg*DM==*CC=*C=E 156 *2395678'0123456789p12395678901239567890123 %67890123956789012395678901239567890 157 *END Flow PATMS TO CONTAINMENT
- p(NETRAfl0N) 4 ?fH SHIELDING 000R5-158 .
159 e8EGIN FlLL PATH 160 tel *2395678901239567890824956789012395678908239567890123956789012395678901239567890
*JUNCT====lN*0f*P*V*FLO**AJUN***!JUN=***lN'***=FJUF**FJUR**V= Cal *EQ'OM*=*CC*=C*E 162- 080701 0 29 1 000 'le0 798 75 De0 Geo Dec 000 3 00 8.0 80 163 eJUNCY**=*lN=0T*P*V=FL0==AJUN*==2JUN===*lN==*==FJUFa*FJUR**V*C=l*EG=DM*==CC==C.E-169 *2395678901239567890123956789012395678908239567890123456789012395678908239567890 165 *END FILL PATH lee e 167
- VALVE DATA CARDS 168 180010 ~3 De0 De0 Dec 0 0 169 800020 -9 00000*O00 170 180030 =5 Deo 0 0 0 0 0 0 171 '880090 *6 Dec 0 0 0 0 0 0 172 180050 *7 Deo 0 0 0 0 0 0 173 180060 *8 0 0 0 0 0 0 0 0 179 .
175 . FILL TABLE DATA CARDS 176
- FILL CONTROL 177 130100 9 . 2 0 0 1095 920.
17A . CARD TIME Flow flME Flow 179 130105-' 0*O 19200.' Oe001050 19200e l et' 130102. 0 001060 21600e le00 24600e 181
- 182 *2395678901239567890823956789012395678901239567890123956789012395678901239567890 183 ************.*************.****************************************************M 889 *MODEL REVISIONS 185 130108 00 7100e 0 008050 7800 CARD A80VF 15 REPLACEMENT CARDe 186 130802 0 001060 10800e 1 00 10000.
CARD A80Vf 1
,8, ........... 5 REPLACEMENT CARDe (v- y G.A-37
LSCS-FSAR AMENDMENT 39 OCTOBER 1978 TABLE 6.A-8 FORCE CONSTANTS AND LOAD CENTERS
-FOR RECIRCULATION LINE OUTLET BREAK ^
NODE U v U ss ELEVATION 0 1 -3696.03 4948.35 757.82 15.0*, 345.0' 2 3696.03 -4948.35 757.825 45.0*, 315.0* 3 3696.03 4948.35 757.825 75.0 , 285.0 4 5464.86 7316.51 757.825 112.5*, 247.5* 5 5464.86 7316.51 757.825 157.5*, 202.5* 6 5828.44 7290.77 764.095 15.0 , 345.0 7 5828.44 7290.77 '764.095 45.0*, 315.0 8 5828.44 7290.77 764.095 75.0*, 285.0 () 9 8617.79 10779.95 764.095 112.5*, 247.5* 10 8617.79 10779.95 764.095 157.5*, 202.5 11 2857.42 2503.87 771.290 22.5*,.337.5* 12 4038.29 3887.97 770.280 45.0*, 315.0 13 4022.57 2990.40 770.280 75.0*, 285.0* 14 5970.91 5748.63 770.280 112.5*, 247.5* 15 5891.80 5523.42 770.280 157.5 , 202.5 16 2234.85 2605.94 '776.085 15.0*, 345.0 17 '3862.52 3683.01 775.075 45.0*, 315.0* 18 3862.52 3683.01 775.075 75.0. , 285.0 19 5711.02 5445.58 775.075 112.5 , 247.5* 20 5631.91 5220.37 775.075 157.5*, 202.5* 21 5325.49 6256.20 780.625 15.0*, 345.0" 6256.20
~
22 5325.49 780.625 45.0*,.315.0* 6,A-38 L
- - . .- . ~. . -. -- . .
'LSCS-FSAR- -AMENDMENT-39
[
' OCTOBER 1978 TABLE 6. A (Cont'd)
UV U ss-
~ NODE ELEVATION '8' 23 s5325.49 6256.20 780.625 75.0 , 285.0 24 7874.13 '9250.27 780.625 112.5*, 247.5*-
25 7874.'13' 9250.27 780.625 '157.5*, 202.5*- 26 11713.96 11338.09 788.625 22.5*, 337.5
- 27. '11713.28 12957.61 788.625 67.5 ,1297.5 '
28 11713.28 12957.61 788.625 112.5*, 247.5* 29 11713.96 11338.09 788.625 157.5 , 202.5 - - 30 12864.45_ 12694.81 798.710 22.5 , 337.5*'
.31 12809.98 12622.87 798.710 67.5 , 297.5*
32 12934.41' 14386.28 798.710 112.5*, 247.5* (} 33 12867.88 11885.05 798.710- 157.5 , 202.5* 34 1557.92 2042.96 771.290 7.5*, 352.5 35 1140.80 0.00 772.020 0.0*, 360.0 i. o l i Q? 6.A-39
- .- .-. . . . - , . - . - , ~ . - - , - . . . . . - , . , . , . , ..- - , , . . ..-,4
LSCS-FSAR- AMENDMENT 39 OCTOBER 1978 1 TABLE 6.A-9 ,
-O FORCE CONSTANTS AND LOAD CENTERS
-FOR FEEDWATER LINE BREAK U U ss NODE V ELEVATION O 1 5464.86 7316.51 757.825 22.5*,.337.5 l
2 5464.86 7316.51 757.825 67.5 , 292.5 3 5464.86 7316.51 757.825 112.5 , 147.5 4 5464.86 7315.51 757.825 157.5*, 202.5. 5 8617.79 10779.95 764.095 22.5*, 337.5* 1 6 8617.79 10779.95 764.095 67.5*, 292.5* 7' 8617.79 10779.95 764.095 112.5*, 247.5* 8 8617.79 10779.95 764.095 157.5*, 202.5 9 11681.94 11194.20 772.625 22.5', 337.5 10 11523.72 10743.78 772.625 67.5*, 292.5
)
11 11523.72 10743.78 772.625 112.5 , 247.5* 12 11666.44 10309.43 772.625 157.5 , 202.5 13 5325.49 6256.20 780.625 15.0 , 345.0* 14 5325.49 6256.20 730.625 45.0*, 315.0* 15 5325.49 6256.20 780.625 75.0*, 285.0* 16 7874.13 9250.27 780.625 112.5*, 247.5* 17 7874.13 9250u27 780.625 157.5 , 202.5*
.18 ' 7967.46 9359.90 788.625 15.0*, 345.0*
19 7841.24. 7570.97 788.625 45.0*, 315.0* 20 7963.08- 7716.94 788.625- 75.0 ,-285.0* 21 11713.96 11338.09 788.625 112.'5 , 247.5* 22 11718.28 12957.61 788.625 157.5*, 202.5*
;() 23 3530.66 4305.38 798.710 7.5*, 352.5 6.A-40
\' . . . - . . . . . . - - - . - . . , . . . . . . - . - - . - - -
i LSCS-FSAR AMENDMENT 39 OCTOBER 1978
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INSTRUMENTATION 3/4.3.4 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION o V LIMITING CONDITION FOR OPERATION 3.3.4 The reactor core isolation cooling (RCIC') system actuation instru-mentation shown in Table 3.3.4-1 shall be OPERABLE with their trip set-points set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2. APPLICABILITY: CONDITIONS 1, 2 and 3 with reactor steam dome pressure
> 150 psig.
ACTION:
- a. With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allow-able Values column of Table 3.3.4-2, declare the channel in-operable and place the inoperable channel in the tripped con-dition until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the requirements for the minimum number of OPERABLE chan-nels not satisfied for one trip system, place the inoperable (s) channel in the tripped condition or declare the RCIC system inoperable within one hour.
- c. With the requirements for the minimum number of OPERABLE chan-nels not satisfied for both trip systems, declare the RCIC system inoperable within one hour.
SURVEILLANCE REQUIREMENTS 4.3.4.1 Each RCIC system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4-1, 4.3.4.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system. 3 (/ - LSCS-1 3/4 3-39 October 1978
G TABLE 3.3.4-1 O l REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM NUMBER OPERABLE CHANNELS FUNCTIONAL UNITS PER TRIP SYSTEM
- a. Reactor Vessel Water Level - Low Low, Level 2 2 00 O
l? n s,i M O O e
,,, (- , ,-
(. U (> 1 r-g TABLE 3.3.4-2 F REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT VALUE
- a. Reactor Vessel Water Level - Low Low, Level 2 > -50 inches * > -57 inches
- I 1
s b Y' O I 1 i R G Si
@ *See Bases Figure B 3/4 3-1.
a> i 4
0; TABLE 4.3.4-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL FUNCTIONAL UNITS CHECK TEST C.ALIBRATION
- a. Reactor Vessel Water Level -
Low Low, Level 2 D M Q w Y e R G S,i a m O O O
INSTRUMENTATION 3/4.3.5 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION
- Q LIMITING CONDITION FOR OPERATION
'3.3.5. The control rod withdrawal block instrumentation shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2.
APPLICABILITY: As shown in Table 3.3.5-1. ACTION:
- a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of. Table 3.3.5-2, declare the channel inoperable until-the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Set-point value.
- b. With the rerequirements for the minimum number of OPERABLE chan-nels not satisfied'for any trip function, place that trip function in the tripped condition within one hour.
() c. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5. SURVEILLANCE REQUIREMENTS 4.3.5. Each of the above required control rod withdrawal block instru-ment channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations - during the OPERATIONAL CONDITIONS and at the frequencies shown in. Table 4.3.5-1. r
.LSCS-1 3/4 3-43 October 1978
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. g; TABLE 3.3.5-1
- S?
4 jb- CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION 2 MINIMUM NUMBER OF- -APPLICABLE OPERABLE CHANNELS OPERATIONAL 1 TRIP FUNCTION PER TRIP FUNCTION CONDITIONS
- l. APRM
! a. Flow Biased Neutron Flux-High 4 1-
- b. Inoperative 4 -1, 2, 5
- c. Downscale . _ 4 1
- d. Neutron Flux-High, 12% 4 2; 5 j{ 2. ROD BLOCK MONITOR (b)
}[ a. Upscale 1 1(a) se b. Inoperative 1 1(a)
- c. Downscale 1 1(a)
- 3. . SOURCE RANGE MONITORS
- a. Detect (djot full in(c) 3 (i) 2, 5 4
- b. Upscale 3 (i) 2, 5
- c. Inoperati 3 (i)
- d. Downscale{g(d) y 2, 5 3 (i) 2, 5 l g7
- 4. ~ INTERMEDIATE RANGE MONITORS ( } i hh a. Detector not full in(9) 6 2,~ 5 Q b. Upscale 6 2, 5
! _. c. Inoperati{g) 6 2, 5 9] d. Downscale 6 2 ca l J t-1 a
- q e: 9 -
. . _ , _ _ . . , _ - - - - - . -- _ ~ _ - - - - - - - - - - - - - - - - - - - - - - -
_%.(.\) sJ 'LJ G TABLE 3.3.5-1 (Continued) en i CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION MINIMUM NUMBER OF APPLICABLE CPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS
- 5. SCRAM DISCHARGE VOLUME
- a. Water Level-High 1 -1,2,5(h)
- 6. RECIRCULATION FLOW UNIT m a. Upscale 1 1 2 b. Inoperative 1 1 w c. Comparator 1 1 b
O O e 1 cn l
k n TABLE 3.3.5-1 (Continued). . CONTROL R00 WITHDRAWAL' BLOCK INSTRUMENTATION
' NOTE
- a. When THERMAL POWER exceedsithe' preset power level of the RWM'and RSCS.
- b. The RBM is bypassed when a peripheral control rod is selected. .
- c. This function is. bypassed if detector is reading > 100' cps or the IRM channels are on range 3 or higher.
- d. This function is' bypassed when the associated IRM channels are on ,
range 8 or higher..
- e. This function is bypassed when the IRM channels are on range 3 or .
higher, j
- f. A total of 6 IRM instruments must be OPERABLE.
- g. This function is bypassed when the IRM channels are'on range .l.
- h. Wito any control rod withdrawn. Not applicable to control. rods:
removed per Specification 3.9.10.1 or 3.9.10.2. ggg
- i. A total of 2 SRM instruments must be operable in OPERATIONAL CONDITION 5 (see Specification 3.9.2).
O' LSCS-1> -3/4 3-46 October 1978
TABLE 3.3.5-2 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1 1. APRM
- a. Flow Biased Neutron Flux-High < 0.66 W + 42%* < 0.66 W + 45%
- b. Inoperative NA NA
- c. Downscale > 5.0% of RATED THERMAL POWER > 3% of RATED THERMAL POWER
- d. Neutron Flux-High, 12% < 12% of RATED THERMAL POWER < 14% of RATED THERMAL POWER
- 2. R0D BLOCK MONITOR
- a. Upscale 1 0.66 W + 40% 1 0.66 W + 43%
w b. Inoperative NA NA 2 c. Downscale > 5% of Full ' Scale > 3% of Full Scale w d g 3. SOURCE RANGE MONITORS
- a. Detector not full in NA NA 5 5
- b. Upscale < 2 x 10 cps < 5 x 10 cps c.
Inoperative NA NA
- d. Downscale 1 3 cps > 2 cps
- 4. INTERMEDIATE RANGE MONITORS
- a. Detector not full in NA NA
- b. Upscale < 108/125 full scale < 110/125 of full scale
!? c. Inoperative NA NA
{ d. Downscale > 5/125 full scale
> 3/125 of full scale G
.g.
TABLE 3.3.5-2 (Continued) > .e,
' }'f CONTROL ROD' WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS -
TRIP FUNCTION - TRIP SETPOINT ALLOWABLE VALUE
- 5. SCRAM DISCHARGE VOLUME
- a. Water Level High __<( ) gallons ** NA.
- 6. RECIRCULATION FLOW UNIT
- a. Upscale < 108/125 full scale
- b. Inoperative NA
- c. Downscale < 10/125 full scale
- 4e i
ZD
?'
- $E-4 4
i 4 E? '
*2 51 >1 ; pp = 1.0 for all values of 'TP i Gs i gj ** Measured at installation of level switches.
U-4 S: O O
~
i
i o o o
.m l ' 0;: TABLE 4.3.5-1 .Q:'
- ,~
CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS i i 1-CHANNEL - OPERATIONAL-CHANNEL FUNCTIONAL CHANNEL.- CONDITIONS IN WHICH
. TRIP FUNCTION CHECK TEST CALIBRATION (a) SURVEILLANCE-REQUIRED
~
[ 1. APRM ! a. Flow Biased Neutron Flux-High NA S/U ,M R 1
- b. Inoperative NA NA -1, 2, 5
- c. S/U(b),M '.
Downscale -NA- S/U M .. NA 1 l- d. Upscale (Fixed) NA- S/U(b),M , R 2,5 I E* 2. ROD BLOCK MONITOR i ! ? a. Upscale- NA b) 1(d)
-S/U(b),M R 3 -b. Inoperative NA .S/U( ,M NA 1
- c. Downscale NA S/U ,M R 1
- 3. SOURCE RANGE MONITORS
- a. Detector not full in NA S/U ,W NA 2, 5
- b. Upscale NA R 2, 5 l- c. Inoperative S/U(b),W NA NA 2, 5 .
l d. Downscale NA S/U(b),W-S/U , R 2, ~ 5 1 g 4. INTERMEDIATE RANGE MONITORS 2
- h. .a. Detector not full in NA S/U ,fc NA 2,. 5 i % .b. Upscale NA R 2,- 5
_ c. Inoperative NA S/U(b)' S/U (c) NA 2, 5
- d. Downscale NA S/U(b)'., (c) R '2, 5 t- c:>
1 i:
-. - ---u-.-v. ,,-..m , A. - . - . - - -- ~.g.. m - , --~ -
G TABLE 4.3.5-1 (Continued) D 1 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED
- 5. SCRAM DISCHARGE VOLUME
- a. Water Level-High NA Q R 1,2,5(e)
- 6. RECIRCULATION FLOW UNIT w a. Upscale NA S/U R 1 2 b. Inoperative NA S/U NA 1 m c. Comparator NA S/U R 1 O
O r+ O O .O O
TABLE 4.3.5-1 (Continued)' (O-
' CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4
f NOTES:
- a. Neutron detectors may be excluded from CHANNEL CALIBRATION.
b; Within-24. hours prior to startup, if not performed within the' previous 7 days.
- c. When changing from COND_ITION'1 to CONDITION _2, perform the required surveillance within 12 hours after entering CONDITION 2.
- d. When THERMAL POWER exceeds the preset power level'of the RWM and RSCS.
- e. With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
O O LSCS 3/4 3-51 October 1978
,.-,..._......-.......,....-...__.-..--........:._i__.,,_....L,-_.,_ .- .,-. _,.. ,_...,_.,,..~.5
INSTRUMENTATION 3/4.3.6 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION g LIMITING CONDITION FOR OPERATION 3.3.6.1 The radiation monitoring instrumentation channels shown in Table 3.3.6.1-1 shall be OPERABLE with their alarm / trip setpoints within the specified limits APPLICABILITY: As shown in Table 3.3.6.1-1. ACTION:
- a. With a radiation monitoring instrumentation channel alarm / trip setpoint exceeding the value shown in Table 3.3.6.1-1, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.6.1-1.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS l 4.3.6.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL ' CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations during l the OPERATIONAL CONDITIONS and at the frequencies shown in Table 1 4.3.6.1-1. ' O LSCS-1 3/4 3-52 October 1978
0- O O . i i s- ' W TABLE 3.3.6.1-1 T RADIATION MONITORING INSTRUMENTATION MINIMUM APPLICABLE
- CHANNELS OPERATIONAL ALARM / TRIP MEASUREMENT INSTRUMENTATION OPERABLE CONDITIONS SETPOINT RANGE ACTION Liquid Radwaste' *. -1 6
- 1. 1 10 -10 cps- 50 Effluent Monitor
- 2. Service Water 1 All 10-1-106cps 51 1
Effluent Monitor
- 3. Reactor Building Closed 1 All 10-1-106cps 51 Cooling Water Monitor
-1 0 4 '. RHR Service Water 2 10 -10 cps y Effluent Monitor.
- 5. Station Vent Stack 1 All -1 0 10 cps 51
- Radiation Monitor
- 6. Standby Gas Treatment All ,
System Rad. Monitor
- 7. Reactor Building Vent -2 2 3 All 10 -10 mR/hr-- 51 Monitor
- 8. Fuel Pool Vent' Monitor -2 2
! R 3 All 10 -10 mR/hr 51~ ci ni M- )
h TABLE 3.6.1-1 (Continued) Y'
~
- RADIATION MONITORING INSTRUMENTATION MINIMUM APPLICABLE l CHANNELS OPERATIONAL ALARM / TRIP MEASUREMENT INSTRUMENTATION OPERABLE CONDITIONS SETPOINT RANGE ACTION l
! 9. Off Gas Post 2 1,2 10-1-106cps 52 Treatment Rod Monitor
- 10. Control Room HVAC 3 All 53 Intake Rod Monitor
- 11. Area Monitors 2 6 R a. Refuel Floor High 1 ** 10 -10 mR/hr 54
[ Range Monitor
-1 b b. Refuel Floor Low 1 ** 10 -10 mR/hr 54 Range Monitor
** -1 2
- c. New Fuel Storage 1 10 -10 mR/hr 54 Vault Monitor
-2
- d. Control Room Rod 1 All 10 -10 mR/hr 54 Monitor o
3 e
"I k
*With radioactive waste stored in or being discharged from the radioactive liquid waste storage system.
**With fuel in the new fuel storage area or the spent fuel storage pool.
O O O
1 TABLE 3.3.6.1-1 (Continued) RADIATION MONITORING INSTRUMENTATION ACTION ACTION 50 - With the required monitor inoperable, take and analyze two independent samples of each radwaste discharge tank to be discharged prior to release. With the required monitor inoperable for > 72 hours, suspend release of lijuid radwaste. ACTION 51 - With the required process monitor inoperable, obtain and analyze grab samples of the monitored parameter at least once per 24 hours. ACTION 52 - With one of the required monitors inoperable, operation may continue provided that the inoperable monitor is placed in the tripped condition within one hour. With both of the required monitors inoperable, be in at least HOT SHUTDOWN within 12 hours. ACTION 53 - With one of the required monitors inoperable, operation may continue provided that the inoperable channel is O placed in the tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours, initiate and maintain operation of the control room emergency filtration system in the (pressurization) mode of operation. With both of the required monitors inoperable, initiate and maintain operation of the control room emergency fil-tration system in the (pressurization) mode of operation within one hour. ACTION 54 - With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instru- , mentation at least once per 24 hours. l l 1 l A y I l 1 LSCS-1 3/4 3-55 October 1978
c-
.M ,
TABLE 4.3.6.1-1
~
RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL' CHANNEL CONDITIONS IN CHANNEL FUNCTIONAL CHANNEL- WHICH SURVEILLANCE
. FUNCTIONAL TEST CHECK TEST CALIBRATION REQUIRED
- 1. Liquid Radwaste D M R
- Effluent Monitor 2.- Service Water D M R Al1-Effluent Monitor
- 3. Reactor Building Closed D M R All
{ Cooling Water Monitor Y' 4. RHR Service Water D M R All E Effluent Monitor
- 5. Station Vent Stack D M R All Radiation Monitor
- 6. Standby Gas Treatment All System Rad. Monitor
- 7. Reactor Building Vent D M R~ All Monitor n
E 8. - Fuel Pool Vent Monitor D M R All II s W m e . . _ O O
i-(u/ us kJ TABLE 4.3.6.1-1 (Continued) ~ RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS IN CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE FUNCTIONAL TEST CHECK TEST CALIBRATION REQU' RED
- 9. Off Gas Post D M R 1,2 Treatment Rad. Monitor
- 10. Control Room HVAC D M R All Intake Rad. Monitor
- 11. Area Monitors
- a. Refuel Floor High D M R **
y Range Monitor w
- b. Refuel ficar Low D M R **
Range Monitor
- c. New Fuel Storage D M R **
Vault Monitor
- d. Control Room Rad. D M R Monitor R
e g *Witn radiaoctive waste stored in or being discharge from the radioactive liquid waste storage system. - **With fuel in the new fuel storage area or the spent fuel storage pool. s
1NSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION O LIMITING CONDITION FOR OPERATION ,, 3.3.6.2 lhe seismic monitoring instrumentation shown in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more seismic monitoring instruments inoperable for more than 30 days, in lieu of any other .'eport required by Specification 6.6.B, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 14 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS O 4.3.6.2.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the fre-quencies shown in Table 4.3.6.2-1. 4.3.6.2.2 Each of the :bove required seismic monitoring in. s actuated during a seismic event shall be restored to OPERAP' is within 24 hours and a CHANNEL CALIBRATION performed withir 'ol-lowing the seismic event. Data shall be retrieved from acts stru-ments and analyzed to determine the magnitude of the vibrate d motion. In lieu of any other report required by Specificatic s.B, a Special Report shall be prepared and submitted to the Commis. on pur-suant to Specification 6.6.C within 14 days describing the magnitude, frequency spectrum and resultant effect upon f acility features important to safety. l Ol l I LSCS-1 3/4 3-58 October 1978
x TABLE 3.3.6.2-1 i < SEISMIC MONIT ' dNG INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE-
- 1. Triaxial Time-History Accelerographs '
- a. Containment Foundation (El. 673'-4") lg 1
- b. Containment Structure (El. 843'-6") lg 1
- c. Free Field lg 1
- d. Aux. Elec. Equip. Room 1g 1
- 2. Triaxial Peak Accelerographs
- a. SGTS lg 20:1 Dynamic Range 0-20 Hz 1
- b. Main Steam Tunnel lg 20:1 Dynamic Range 0-20 Hz 1 ;
- c. Diesel Generator 19 20:1 Dynamic Range O
V - 0-20 Hz 1
- d. Main Ct 19 20:1 Dynamic Range 0-20 Hz 1
- 3. Triaxial Seisc
- a. Containment Foundation (E1. 675'-4") 0.005 to 0.15g adjustable 1*
- b. Also has Internal Trigger for each Accelerometer which can be individually set adjustable 1*
- 4. Triaxial Response-Spectrum Recorders
- a. Terra Tech. Digital Cassette with playback feature 1*
(~) V *With reactor control room indication and annunciation. LSCS-1 3/4 3-59 October 1978
TABLE 4.3.6.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS _ CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHECK TEST CALIBRATION
- 1. Triaxial Time-History Accelerographs
- a. Containment Foundations M SA R
- b. Containment Structure M SA R
- c. Free Field M SA R
- d. Aux. Elec. Equip. Room M SA R
- 2. Triaxial Peak Accelerographs
- a. SGTS NA NA R
- b. Main Steam Tunnel NA NA R
- c. Diesel Generator NA NA R
- d. Main Control Board NA NA R
- 3. Triaxial Seismic Switches
- a. Containment Foundation M* SA R
- b. Also has Internal Trigger M SA R for each Accelerometer which can be individually set h
- 4. Triaxial Response-Spectrum Recorders
- a. Terra Tech. Digitel Cassette M SA R with playback
*Except seismic trigger. h LSCS-1 3/4 3-60 October 1978
l l INSTRUMENTATION l l METEOROLOGICAL MONITORING INSTRUMENTATION ' (] . LIMITING CONDITION FOR OPERATION 3.3.6.3 The meteorological monitoring instrumentation channels shown in Table 3.3.6.3-1 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more meteorological monitoring channels inoperable for more than 7 days, in lieu of any other report required by Spe-cification 6.6.8, prepare and submit a special Report to the Com-mission pursuant to Specification 6.6.C within the next 14 days outlining the cause of the malfunction and the plans for restoring the system to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENT 3 (u.,) 4.3.6.3 Each of the above required meteorological monitoring instru-mentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequen-cies shown in Table 4.3.6.3-1. C'i v LSCS-1 3/4 3-61 October 1978
TABLE 3.3.6.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE INSTRUMENT
- 1. Wind Speed 1
- 1. Elev. 200 ft.
- 2. Elev. 375 ft.
- 2. Wind Direction 1
- 1. Elev. 200 ft.
- 2. Elev. 375 ft.
- 3. Air Temperature 0
- 1. Elev. 33 ft.
- 4. Air Temperature Difference 1
- 1. Elev. 33/200 ft.
- 2. Elev. 33/375 ft.
O 9 LSCS-1 3/4 3-62 October 1978
, TABLE 4.3.6.3-1 METEOROLOGICAL fj0NITORING INSTRllMENTATION' SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Wind Speed
- 1. Elev. 200 ft. W SA
- 2. Elev. 375 ft. W SA
- 2. Wind Direction
- 1. Elev. 200 ft. W SA
- 2. Elev. 375 ft. W SA
- 3. Air Temperature
- 1. Elev. 33 ft. W SA
- 4. Air Temperature Difference
- 1. Elev. 33/200 ft. W SA
(] 2. Elev. 33/375 ft. W SA r% V LSCS-1 3/4 3-63 October 1978
INSTRUMENTATION REMOTE SHUTDOWN MONITORING INSTRUMENTATION , LIMITING CONDITION FOR OPERATION 3.3.6.4 The remote shutdown monitoring instrumentation channels shown in_ Table 3.3.6.4-1 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY: CONDITIONS 1, 2, and 3.
+
ACTION:
- a. With the number of OPERABLE remote shutdown monitoring channels less than the requirements of Table 3.3.6.4-1, either restore the inoperable channel to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. The provisions of Specification 3.0.4 are not applicable, ,.
SURVEILLANCE REQUIREMENTS 4.3.6.4 Each of the above required remote shutdown monitoring instru-h- ' mentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.6.4-1. t 4 O LSCS-1 - 3/4 3-64 October 1978 -e-ee-. ,-g- w - 9-,,wyv v r-ee.- g-c--w---e,--+ 1r ,, re y -e-w , - - - - . . , + , -g-er-=w<-,e.. ,-<w,wn-w., y *,e-,-,+.,- + .- e e e- .w,s w *m- s-
O O O C;. TABLE 3.3.6.4-1 g . . , . REMOTE SHUTDOWN INSTRUMENTATION MINIMUM READOUT CHANNELS FUNCTION UNIT LOCATION . OPERABLE-
- 1. Reactor Vessel Pressure Remote Shutdown Panel- 1
- 2. Reactor Vessel Water Level Remote.-Shutdown Panel il-
- 3. Safety / Relief Valve Position ( 3 valves) Remote. Shutdown Panel 1
- 4. Suppression Chamber Water Level Remote Shutdown Panel 'l
{ 5. Suppression Chamber Water Temperature Remote Shutdown Panel 1
- 6. Suppression Chamber Air Temperature Remote Shutdown Panel 1
- 7. Drywell Pressure Remote ~ Shutdown Panel 1 -
- 8. Drywell Temperature Remote Shutdown Panel 1
- 9. RHR Flow Remote Shutdown Panel 1
- 10. RHR Service Water Flow Remote Shutdown Panel 1
- 11. RHR Service Water Temperature Remote Shutdown Panel 1:
R 12. RCIC Turbine Speed Remote Shutdown Panel 1 G . .
- 13. RCIC Flow : Remote Shutdown' Panel. 1
{ e __ __ . _ _ _. ._ . . _ _ . . - ~ . _ _ ;_,___ ,_ .. , _ _ . . - - , _ - - .
'R TABLE 4.3.6.4-1 G-1~ REMOTE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL . CHANNEL-FUNCTIONAL UNIT CHECK ' CALIBRATION
- 1. .ReacEorVesselPressure- M Q 2._ Reactor Vessel Water Level M Q
- 3. Safety / Relief Valve Position- R NA
- 4. Suppression Chamt.r Water Level M R
$ 5. Suppression Chamber Water Temperature M. R
- 6. Suppression Chamber Air Temperature M R
- 7. Primary Containment Pressure M Q
- 8. Drywell Temperature M R
- 9. RHR F1ow- M R -
- 10. 'RHR Service Water Flow M R
- 11. RHR' Service' Water Temperature M R R 12. RCIC Turbine Speed M' R-E ,
- 13. RCIC Flow. M R E,.
M O O O
- u INSTRUMENTATION 7 POST-ACCIDENT MONITORING INSTRUMENTATION *
(V LIMITING CONDITION FOR OPERATION 3.3.6.5 The post accident monitoring instrumentation channels shown in Table 3.3.6.5-1 shall be OPERABLE. APPLICABILITY: CONDITIONS 1 and 2. ACTION:
- a. With the number of OPERABLE post-accident monitoring channels less than recJired by Table 3.3.6.5-1, either restore the inoperable channel to OPERABLE status within 30 days or submit a special report to the NRC outlining planned corrective action.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS f 4.3.6.5 Each of the above required post-accident monitoring instrumenta-
- i. j3 tion channels shall be demonstrated OPERABLE by performance of the CHAN-NEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown )
in Table 4.3.6.5-1. j i i i 1 l l l
- Imposition of this Specification is entirely out of order and premature t f')
v because Regulatory Guide 1.97 is non-resolved issue. I I LSCS-l' ' '4 3-67 October 1978 a
s G G TABLE 3.3.6.5-1 POST-ACCIDENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE
/
i 1
- 1. Reactor Vessel Pressure >
- 2. Reactor Vessel Water Level 1 Suppression Chamber Water Level 1 3.
Suppression Chamber Water Temperature 1 4. Suppression Chamber Air Temperature 1
$ 5. i w 1 a 6. Drywell Pressure co
- 7. Drywell Temperature 1
- 8. Containment Hydrogen Concentration 1
- 9. Containment Gross y Radiation
- 1 1
i O l o ni
- Manually activated following seismic event.
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