ML20214T602
| ML20214T602 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 12/31/1986 |
| From: | LONG ISLAND LIGHTING CO. |
| To: | |
| Shared Package | |
| ML19292G341 | List: |
| References | |
| NUDOCS 8612080645 | |
| Download: ML20214T602 (886) | |
Text
{{#Wiki_filter:. . . _ _ _ _ _ - . - . - - . __ _ _ _ _ _ _ . _ . _ _ . O PLANT DESIGN ASSESSMENT REPORT FOR SRV AND LOCA LOADS i 4 1 SHOREHAM O NUCLEAR POWER STATION l l 1 l ! l i ((h
/Mada/MM REVISION 6 DECEMBER 1986 bd !Db$E $Io$ 2
A LONG ISLAND LIGHTING COM PANY SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY RO AD e WADING RIVE R, N.Y.11792 December 16, 1981 SNRC-645 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Enclosed are fifteen (15) copies of Revision 5 to the Shoreham Plant Design Assessment Report (DAR) for SRV and LOCA Loads. This report is submitted in accordance with our Mark II Containment Closure Program as outlined in cur meeting of 73 March 5, 1981. The DAR Revision 5 contains the completed design l } assessment for the Shoreham Design Basis Loads, (NUREG-0487 & \/ Supp. 1). It also contains the Shoreham confirmatory program results based on the final generic load definitions and accept-ance criteria set forth in NUREG-0802 (DRAFT) and NUREG-08 08. A proprietary supplement to Revision 5 to the Design Assessment Report is being sent under separate cover. Very truly yours, J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station CC:mp Enclosure cc: J. Higgins a FC.893 5
ggg LONG ISLAND LIGHTING COMPANY O Mm** SHOREHAM NUCLEAR POWER STATION P.O. 90X 004. NORTH COUNTRY ROAD e WADMG RNER. N.Y.11792 March 12, 1981 SNRC- 539 Mr. Harold R. Denton, Director * ) Office of Nuclear Reactor Regulation U. G. Nuclear Regulatory Commission Washington, D. C. 20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Enclosed are fifteen (15) copies of Revision 4 to the Shoreham Plant Design Assessment Report (DAR) for SRV and LOCA loads. This report is submitted in accordance with our Mark II Containment Closure Program as outlined in our meeting of March 5, 1981. The DAR Revision 4 contains the completed design assessment for the Shoreham Design Basis Loads, g (NUREG-0487 & Supp. 1) and confirmatory program commitments. A proprietary supplement to Revision 4 to the Design Assessment Report is being sent under separate cover. Very truly yours, [ om J. P. No arro Project Manager Shoreham Nuclear Power Station CC:mp cc: J. Higgins V,O
m LONG ISLAND LIGHTING COM PANY C yk(h SHOREHAM NUCLEAR POWER STATION P.O. BOX 618. NORTH COUNTRY ROAD
- WADING RIVER. N.Y.11792 December 18, 1978 SNRC-347 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Enclosed are fifteen (15) copies of Revision 3 to the Shoreham Plant Design Assessment Report (DAR) for SRV and LOCA loads.
. This report is submitted in accordance with our Mark II
(-} Containment Cloauro Program as outlined in our letter (SNRC-298) dated June 15, 1978. The DAR Revision 3 contains the following: ,
- 1. Incorporation of information previously submitted in our letter (SNRC-309), dated July 28, 1978.
- 2. Information relating to load definition, load combination methods, acceptance criteria and Shoreham plant capability to accept the hydrodynamic loads (inclus'ing annulus pressurization effects) as required by the NRC load evaluation report, NUREG-0487, dated October 1978.
- 3. The final revision to the DAR, scheduled for submittal in June 1979, will contain the confirmatory analyses and demonstrate the capability of all plant structures and equipment to accommodate the Mark II loads.
We have also included an Executive Summary of the Mark II Loads Closure Program for the Shoreham Nuclear Power Station. i r~N 1 l
%J
[ i
i l I ! l j Mr. Harold R. Denton December 18, 1978
) Re Shoreham Plant Design Page 2 i l Assessment Report (DAR) '
I A proprietary supplement to Revision 3 to the Design Assessment I i t Report is being sent under separate cover.
; Very truly yours, i
67 7~MC J. P. Novarro, j Project Manager ,
- Shoreham Nuclear. Power Station i 2
BRM/cl 4 i i !,O 4 I l I i i l l l i i i !,O [ l i i }
..~. _ -__._ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ . . . . _ , . _ _ _ - - -..._. .,-
1 4 PLANT DESIGN ASSESSMENT <
- FOR SRV AND LOCA LOADS iO i
i i I i i l
- l SHOREHAM NUCLEAR POWER STATION <
O UNIT 1 : ! l i I i ! f i j l gg w/vd4# REVISION 5 DECEMBER 1981 j l I I
l s INSERTION INSTRUCTIONS l t FOR REVISION 5 TO THE DAR The following text, tables, and figures are to be inserted in the l DAR. These pages are either replacement pages or new pages as indicated below. All replacement pages which differ from existing pages are identitled with the revision number and date in the lower right j hand corner. Bars located in the margin of a particular page l indicate material which is new in the revision indicated at the bottom of the page. Remove Old (Pages) Insert New (Pages) yolume 1 Revision 5 Title Page in Front of Revision 4 Title Page EP_1 througn EP-4 EP-1 t hrough EP-4 ' 1-1 through 1-4 1-1 through 1-4 Table 1-1 (page 1 of 4) Table 1-1 (page 1 of 4) i Table 1-1 (page 4 of 4) Table 1-1 (page 4 of 4) Figure 1-5 and Figure 1-5 arid
, s
~'
Figure 1-6 Figure 1-6 Figure 1-8 Figure 1-8 Figure 1-11 Figure 1-11 2-1 through 2-11 2-1 through 2-12 Table 2-1 throu.Jh Table 2-5
. Table 2-1 through Table 2-5 Figure 2-1 Figure 2-1 Figure 2-3 through Figure 2-3 through Figure 2-6 Figure 2-6 Figure 2-10 at.a Figure 2-10 and Figure 2-11 Figure 2-11 Figure 2-14 and Figure 2-14 and Figure 2-15 Figure 2-15 Figure 2-19 and Figure 2-19 and Figure 2-20 Figure 2-20 3-1 through 3-o 3-1 through 3-6 Tablu 3-5 and TaLle 3-:. Table 3-5 and Table 3-b Figure 3-19 and Figure 3-19 and Figure 3-20 Figure 3-20 4-1 through 4-16 4-1 through 4-17 Table 4-3 and Table 4-4 Table 4-3 through Table 4-4 Figure 4-4 anc. Figure 4-4 and Figure 4-5 Figure 4-5 5-3 and 5-4 5-3 and 5-4 5-9 tnrough 5-12 5-11 through 5-12 Figure 5-13 Figure 5-13
(~)
\/
Figure 5-15 through Figure 5-15 through Figure 5-18 Figure 5-1L Figure 6-2 Figure 6-2 1 Revision 5 - December 1981
INSERTION INSTRUCTIONS FCONTegy Remove Old (Pages) Jnsert New (Paces) 7-5 through 7-11 7-5 through 7-11 8-1 through 8-3 8-1 through 8-3 9-3 through 9-20 9-3 through 9-19 Table 9-7 Table 9-7 10-1 through 10-6 10-1 through 10-4 Table 10-1 Table 10-1 Table 10-2 Figure 10-1 and Figure 10-1 and Figure 10-2 Figure 10-2 R-1 through R-3 R-1 through R-3 Volume 2 A-1 through A-14 A-1 through A-14 , Figure B2-5 Figure B2-5 C-3 and C-4 C-3 and C-4 E-1 and E-2 E-1 and E-2 Appendix F Title Page Appendix F Title Page F-1
, F-1 through F-4 G-1 through G-iv G-1 through G-1v G-1 through G-22 G-1 through G-21 Table G-1 and Table G-2 Table G-1 and Table G-2 Table G-4 through Table G-8 Table G-4 thrcugh Table G-8 h Figure G-5 Figure G-5 Figure G-8 through Figure G-8 through Figure G-10 Figure G-10 Figure G-18 through Figure G-18 through Figure G-21 Figure G-21 H-1 and H-2 H-1 and H-2 J-1 and J-2 J-1 through J-3
! Table J-1 Table J-1 i Figure J-1 Figure J-1 , Figure J-2 Figure J-2 j Figure J-3 Figure J-3 Figure J-4 Figure J-4 Figure J-5 Figure J-5 Figure J-6 Figure J-6 Figure J-7 K-1 through K-12 K-1 through K-12 ( Table KJ.1-1 Table K3.1-1 j Table K3.2-2 (2 pages) Table K3.2-2 (2 pages) , Tab Appendix L l - ' Appendix L Tit.le Page L-i through L-v L-1 through L-16 Table L-1 through Table L-3 Tablo L-4 (3 pages) l Table L-5 i 2 Rovision 5 - December 1981
- . . _ . - _ . - - - _ _ --- - =- - - .
INSERTION INSTRUCTIONS (CONTO p) O Remove Old (Paces) Insert New (Pages) Table L-6 (2 pages) Table L-7 and Table L-8 Table L-9 (2 pages) Table L-10 and Table L-11 Figure L-1 through Figure L-39 l Tab Appendix M Appendix M Title Page M-1 through M-3 Table M-1 Figure M-1 THESE INSTRUCTIONS ARE TO BE FILED IN VOLUME 1 FOLLOWING THE INSIDE TITLE PAGE. 1
- O i
4 i O 3 Revision 5 - December 1981
~ _ - - _ - _ - _ . - _ . _ _ - -
PLANT DESIGN ASSESSMENT O' s FOR SRV AND LOCA LOADS SHOREHAM O NUCLEAR POWER STATION
- i UNIT 1 l
1 l O [d(O REVISION 4
,as<e.n:7.9 ztr FEBRUARY 1981 l
V'
. .- _ . ~ _ - .
1 _ INSERTION INSTRUCTIONS FOR REVISION 4 TO THE DAR The DAR. following text, tables, and figures are to be inserted in the These pages are either replacement pages or new pages as indicated below. All replacement pages which differ from existing pages are identified hand with corner . the revision number and date in the lower Bars located right indicate material which is new in the revision indicatedin the margin otthe a particular bottom of the page. at Remove Old (Pages) _ Insert New (Pages) Volume 1 Outside Title Page Outside Title Page and Spine and Spine Revision 4 ' Title Page in front of Revision 3 Title Page EF-1 through EF-3 (,s 1 through xvi EP-1 through EP-4 i through xv 1-1 through 1-4 Table 1-1 (2 pages) 1-1 through 1-4 Figure 1-1 through Figure 1-11 Table 1-1 (4 pages) 2-1 through 2-6 Figure 1-1 through Figure 1-11 2-1 through 2-11 Table 2-1 through Table 2-4 Table 2-1 through Table 2-5 Figure 2-1 through Figure 2-18 Figure 2-1 through Figure 2-20 3-1 through 3-6 3-1 through 3-6 Table 3-1 through Table 3-7 Table 3-1 through Table 3-6 Figure 3-1 through Figure 3-22 Figure 3-1 through Figure 3-22 4-1 through 4-20 4-1 through 4-1b Table 4-1 through Table 4-4 Table 4-1 through Table 4-4 Figure 4-1 through Figure 4-16 Figure 4-1 through Figure 4-9 5-1 through 5-14 5-1 through 5-12 Table 5-1 through Table 5-11 Table 5-1 through Table 5-11 Figure 5-1 through Figure 5-62 Figure 5-1 through 5-4 7 6-1 through 6-5 6-1 through 6-5 Table 6-1 through Table 6-6 Table 6-1 through Table 6-b Figure 6-1 through Figure 6-8 Figure 6-1 through Figure 6-8 7-1 through 7-11 7-1 through 7-11 Table 7-1 Table 7-1 Figure 7-1 through Figure 7-5 Figure 7-1 through Figure 7-5 8-1 through 8-3 8-1 through 8-3 ' 9-1 through 9-23 9-1 through 9-20 E9.1-1 through E9.1-5 - Table 9-1 tnrough Table 9-6 O Table 9-7 (2 pages) Table 9-1 through Table 9-b Table 9-7 1 Revision 4 - February 1981
Remove Old (Pages) Insert New (P&qes) Volume 1 (Cont) Table 9-8 Table 9-8 (3 pages) Table 9-9 through Table 9-23 Table 9-9 through Table 9-22 Figure 9-1 through Figure 9-35 Figure 9-1 through Figure 9-13 10-1 through 10-5 10-1 through 10-6 Table 10-1 Figure 10-1 and Figure 10-2 R-1 through R-3 R-1 through R-3 The accompanying DAR Revision 4, Volume 2, contains revised Appendix A through revised Appendix E, and newly created Appendix F through newly created Appendix K. Therefore, remove all remaining material f rom Volume 1 and insert Tab Appendix A through Tab Appendix E in Volume 2 as follows:
- 1. Insert Tab Appendix A in front of Appendix A title page.
- 2. Insert Tab Appendix B following page A-14.
- 3. Insert Tab Appendix C following Figure B7-4
- 4. Insert Tab Appendix D following Figure C3-15.
l l i 5. Insert Tab Appendix E following page 130-15. l The remaining material that was removed from Volume 1 may be discarded. THESE INSTRUCTIONS ARE TO BE FILED IN VOLUME 1 FOLLOWING THE INSIDE TITLE P30E. 1 0 2 Revision 4 - February 1981
1 Q PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS l O SHOREHAM NUCLEAR POWER STATION UNIT 1 [d[h
/MEM/MX REVISION 3 NOVEMBER 1978
i INSERTION INSTRUCTIONS FOR REVISION 3 TO THE DAR The following text, tables, and figures are to be inserted in the DAR. These pa'es g are either replacerent pages or new pages as indicated below. All replacement pages which differ from existing pages are identified with the revision number and date in the lower right hand corner. Bars located in the margin of a particular page indicate material which is new in the revision indicated at the bottom of the page. Remove Old (Pages) Insert New (Pages) Revision 3 Title Page (in i front of Revision 2 Title Page Preface - Matrix Table - Required Information (from letter - of April 18, 1975) Request for Additional Information - (from letter of April 21, 1975) Tab LIST OF EFFECTIVE PAGES O. - EF-1 through EF-3 i through xiv i through xiv xv and xvi 1-1 through 1-4 1-1 through 1-4 Table 1-1 (2 pages) Table 1-1 (2 pages) Figures 1-1 through 1-4 Figures 1-1 through 1-4 Figures 1-6 through 1-8 Figures 1-6 through 1-8 Figure 1-11 Figure 1-11 2-1 through 2-4 2-1 through 2-4 2-5 and 2-b Table 2-2 Table 2-2 Tables 2-3~and 2-4 Figures 2-2 through 2-9 Figures 2-2 through 2-9 Figures 2-12 and 2-13 Figures 2-12 and 2-13 Figures 2-15 through 2-17 Figures 2-15 through 2-17 3-1 through 3-5. 3-1 through 3-5 3-6 Tables 3-1 through 3-5 Tables 3-1 through 3-5 Tables 3-6 and 3-7 Figures 3-1 through 3-12 Figures 3-1 through 3-12 Figures 3-13 through 3-22 4-1'through 4-17 4-1 through 4-17 4-18 through 4-20 i Tables 4-1 through 4-3 Tables 4-1 through 4-3 1 Figure 4-1 Figure 4-1
! Figure 4-3 Figure 4-3 Figures 4-7 through 4-9 Figures 4-7 through 4-9 5-1 through 5-13 5-1 through 5-13 1 Revision 3 - November 1978
i Remove Old (Pages) Insert New (Feqes I 5-14 I Tables 5-1 through 5-5 Tables 5-1 through 5-5 1 Table 5-7 Table 5-7 l Table 5-9 Table 5-9 Table 5-11 Table 5-11 Figures 5-1 through 5-7 Figures 5-1 through 5-7 1 Figures 5-28 through 5-57 Figures 5-28 through 5-57 Figures 5-58 through 5-62 6-1 through 6-5 6-1 through 6-5 6-6 and 6-7 - Tables 6-1 through 6-4 Tables 6-1 through 6-4 l Tables 6-5 and 6-6 Figures 6-1 through 6-7 Figures 6-1 through 6-7 Figure 6-8 1 7-1 through 7-11 7-1 through 7-11 i 8-1 through 8-3 8-1 through 8-3 9-1 through 9-4 9-1 through 9-4 9-5 through 9-23 ;
- E9.1-1 through E9.1-5 '
Tables 9-1 through 9-6 Tables 9-1 through 9-6 Table 9-7 (2 pages) ;
- Tables 9-8 through 9-23 Figures 9-1 through 9-4 Figures 9-1 through 9-4 )
Figures 9-5 through 9-35 10-1 through 10-5 10-1 through 10-5 1 Table 10-1 ' l Figures 10-1 through 10-8 - - - l R-1 R-1 R-2 and R-3 i (following Appendix A Title Page) A-i Table A-3 (tollowing l Table A-1 page 3 of 3) 020-44 020-44 020-64 through 020-71 Figures 020.71-1 through 020.71-4 l 020-72 through 020-75 l Figure 020.75-1 B-i through B-iv (following Appendix B Title Page) B-1 through B-4 B-1 through B-4 b-5 through B-9 B1-1 - B2-1 through B2-3 - B3-1 through B3-3 - 1 B4-1 - B5-1 - j Figures B2-1 through B2-3 Figures B2-1 through B2-3 Figure B3-10C Figure B3-10C Figure B3-11C Figure B3-11C Figures B3-15 and B3-16 Figures B4-3 and B4-4 l Tab APPENDIX C 2 Revision 3 - Nover.ber 1978
t Remove Old (Pages) Insert New (Pages
- Appendix C Title Page
- . C-i through C-iii
- C-1 through C-9
- Figures C3-1 through C3-15
- Tab A7PENDIX D
- Appendix D Title Page
- D-i through D-iii
- D-1 through D-13
- ED-1 and ED-2
- Tables D2-1 through D2-8
- Table D3-1
- Table D3-2 (2 pages)
- Table D3-3 (5 pages)
- Table D3-4
- Figures D2-1 through D2-3
- Tab APPENDIX E
- Appendix E Title Page
- E-i through E-lii
- E-1 through E-17
- Tables E4-1 and E4-2
- Table E4-3 (2 pages)
- Table E4-4
- Figures E4-1 through E4-8 THESE INSTRUCTIONS AhE TO BE FILED BEHIND THE REVISION 3 TITLE PAGE l
a O 4 3 Revision 3 - November 1978
A m , ,.._2 _. .-.a-
,,A-.- m _m_ . - .
1
'Q PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS l l
e i i O SHOREHAM NUCLEAR POWER STATION UNIT 1 J l [ REVISION 2 SEPTEMBER 1977 i
INSERTION INSTRUCTIONS 70R REVISION 2 TO THE DAR The following text, tables, and figures are to be inserted in the DAR. These pages are either replacement pages or new pages as indicated below. All replacement pages which differ from existing pages are identified with the revision number and date in the lower right hand corner. Bara located in the margin of a particular page indicate material which is new in the revision indicated at the bottom of the page. Remove Old fPages) Insert New (Pages) ' Revision 2 Title Page (in front of Revision 1 Title Page) vi vi APPENDIX A i i Table A-1 (2 pages) Table A-1 (2 pages) Table A-2 (3 pages) 020-27 through 020-57 O - 020-58 and 020-58a 020-59 through 020-63 130-8 through 130-15 APPENDIX B Tab Appendix B
, Appendix B Title Page B-1 through B-4 B1-1 B2-1 through B2-3 B3-1 through B3-3 B4-1 B5-1 Table B2-1 Figs. B2-1 through B2-4 Figs. B3-1 through B3-9 Figs. B3-10A through B3-10C Figs. B3-11A through B3-11C Figs. B3-12 through B-14 Figs. B4-1 and B4-2 THESE INSTRUCTIONS ARE TO BE FILED BEHIND THE REVISION 2 TITLE PAGE O
1 Revision 2 - September 1977 __ _ _ ___ ., ~ _. _
'O PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS i . I SHOREHAM NUCLEAR POWER STATION UNIT 1 O ! l I i l O [g[g REVISION 1
' R O M xra w r a ti;rs APRIL 1977 l
i LIST OF EFFECTIVE PAGES O Text, Table (T), Revision or Fiqure (F) Number volume 1 EP-1 through EP-7 6 i through iv 5 i y 6 vi 5 vil and viii 6 ix 3 x through xiv 5 xv 6 1-1 through 1-4 6
- T1-1 (Sheet 1 of 4) 5 T1-1 (Sheets 2 and 3 of 4) 4 T1-1 (Sheet 4 of 4) 5 F1-1 4 F1-2 6 F1-3 and F1-4 3 F1-5 and F1-6 5 F1-7 4 F1-8 5 F1-9 and F1-10 1 F1-11 6 2-1 4 2-2 5 2-3 6 2-4 through 2-7 5 4 8 6 2-9 through 2-12 5 T2-1 through T2-4 5 T2-5 (Sheets 1 and 2 of 2) 6~
F2 1 5 F2-2 3 F2-3 through F2-6 5 F2-7 6 F2-8 3 1 F2-9 4 F2-10 and.F2-11 5 F2-12 3 F2-13 4 F2-14 and'F2-15 '5 F2-16 and F2-17 3 F2-18 1 F2-19 and F2-20 5 3-1 5 3-2 6 O. 3-3 and 3-4 3-5 5 6 36 5 - EP-1 Revision 6 - December 1986
,.e , - - - - , - - - ~ ~ - , , , - , , , , , , , - - , - - ...,,~..r , , - - - - , - - - - - - - . ~ - - . - , - - - <,----e,- .. , . . - , - . . ..-, .. ,
LIST OF EFFECTIVE PACES (CONT'D) Text, Table (T), Revision or Fiqure (F) Number T3-1 through T3-4 4 T3-5 and T3-6 6 F3-1 through F3-18 4 F3-19 and F3-20 5 F3-21 and F3-22 4 4-1 6 4-2 through 4-10 5 4-11 through 4-13 6 4-14 through 4-17 5 T4-1 4 T4-2 3 T4-3 and T4-4 6 F4-1 through F4-3 4 F4-4 and F4-5 5 F4-6 1 F4-7 through F4-9 6 5-1 through 5-3 4 5-4 5 5-5 through 5-9 4 5-10 through 5-12 5 T5-1 4 T5-2 and T5-3 3 T5-4 and T5-5 4 T5-6 1 T5-7 3 T5-8 1 ! T5-9 3 l T5-10 1 l T5-11 03 l F5-1 through F5-12 4 F5-13 5 F5-14 4 F5-15 through F5-18 5 F5-19 through F5-47 4 6-1 through 6-3 4 6-4 6 1 6-5 4 T6-1 4 T6-2 through T6-6 3 F6-1 4 i F6-2 5 i F6-3 and F6-4 3 l F6-5 4 l F6-6 through F6-8 3 i i e EP-2 Revision 6 - December 1986
LIST OF EFFECTIVE PAGES (CONT'D)
, Text, Table (T), Revision v or Fiqure (F) Number 7-1 6 7-2 through 7-4 4 7-5 through 7-10 5 7-11 4 T7-1 4 F7-1 0 F7-2 through F7-5 1 8-1 through 8-3 5 9-1 through 9-7 6 9-7a and 9-7b 6 9-8 through 9-13 6 9-13a and 9-13b 6 9-14 through 9-17 6 9-17a and 9-17b 6 9-18 5 9-18a and 18b 6 9-19 6 T9-1 and T9-2 6 T9-3 and T9-4 3 i T9-5 6 :
] F9-1 through F9-11 6 F9-12 4 10-1 5 10-2 and 10-3 6 10-4 5 T10-1 and T10-2 5 F10-1 and F10-2 5 R-1 and R-2 5 R-3 6 volume 2 Appendix A Title Page 4 A-1 and A-2 5 A-3 4 A-4 through A-14 5 Appendix B Title Page 4 B-i through B-iv 4 B-1 3-B-2 through B-8 4 TB4-1 4 FB2-1 through FB2-4 4 FB2-5 5 FB3-1 through FB3-4 4 EP-3 Revision 6 - December 1986 i
-, , , , - . . -- .m- - , - - - _ ,-
LIST OF EFFECTIVE PAGES (CONT'D) Text, Table (T), Revision or Fiqure (F) Number FB4-1A through FB4-1C 4 FB4-2A through FB4-2C 4 FB6-1 through FB6-5 4 FB7-1 through FB7-4 4 Appendix C Title Page 3 C-i through C-lii 4 , C-1 and C-2 4 I C-3 6 C-4 through C-10 4 FC3-1 through FC3-15 4 Appendix D Title Page 4 D-i 4 TD-1 (Sheets 1 and 2 of 2) 4 l TD-2 (Sheets 1 through 3 of 3) 4 TD-3 (Sheet 1 of 1) 4 020-1 through 020-25 4 020-26 through 020-58 4 l 020-58a 4 020-59 and 020-60 4 1 020-61 2 020-62 through 020-64 4 020-65 3 l 020-66 through 020-68 4 020-68a through 020-68c 4 T020.68-1 through T020.68-3 4 F020.68-1 4 F020.68-2A and F020.68-2B 4 F020.68-3A and F020.68-3B 4 F020.68-4A and F020.68-4B 4 F020.68-5A and F020.68-5B 4 , F020.68-6A and F020.68-6B 4 F020.68-7A and F020.68-7B' 4 F020.68-8A and F020.68-8B 4 F020.68-9A and F020.68-9B 4 F020.68-10A and F020.68-10B 4 F020.68-11A and F020.68-11B 4 F020.68-12 and F020.68-13 4 020-69 3 020-70 4 , 020-71 3 F020.71-1 through F020.71-4 3 020-72 and 020-73 4 l 020-74 3 020-75 4 F020.75-1 3 Ol EP-4 Revision 6 - December 1986 . l 1
LIST'OF EFFECTIVE PAGES (CONT'D) { Text, Table (T), Revision ! or Fiqure (F) Number ' 130-1 through 130-6 4 , i 130-7 1 l 130-8 through 130-15 4 Appendix E Title Page 3 E-i 3 E-il and E-lii 4 E-1 5 E-2 and E-3 6 E-4 and E-5 4 E-6 6 E-7 4 E-8 through E-11 6 I E-12 through E-14 4 E-15 6 E-16 and E-17 4 TE4-1 (Sheet 1 of 1) 3 TE4-2 (Sheet 1 of 1) 6 TE5-1 (Sheets 1 and 2 of 2) 4 TE5-2 (Sheet 1 of 1) 4 FE5-1 4 FES-2 6 l FE5-3 4 FES-4 through-FES-6 6 FES-7 and FE5-8 4 Appendix F Title Page 5 F-i 6 F-1 through F-3 6 F-3a and F-3b
=
6 F-4 6 Appendix G Title Page 4 G-i 5 G-il 4 I G-lii 5 G-iv 4 G-1 through G-21 5 TG-1 and TG-2 (Sheet 1 of 1) 5 TG-3 (Sheets 1 and 2 of 2) 4 TG-4 through TG-8 (Sheet 1 of 1) 5 TG-9 (Sheet 1 of 1) 4 FG-1 through FG 4 FG-5 5 FG-6 and FG-7 4 , FG-8 through FG-10 5 f FG-11 through FG-17 4 N FG-18 through FG-21 5 FG-22 through FG-25 4 EP-5 Revision 6 - December 1986 f
LIST OF EFFECTIVE PAGES (CONT'D) Text, Table (T), . Revision or Fiqure (F) Number Appendix H Title Page 4 H-1 4 H-2 5 FH-1 through FH-6 4 Appendix I Title Page 4 Appendix J Title Page 4 J-l 6 J-2 through J-3 5 TJ-1 (Sheet 1 of 1) 5 Chart 1 4 FJ-1 5 Chart 2 4 FJ-2 5 Chart 3 4 FJ-3 5 Chart 4 4 FJ-4 5 Chart 5 4 FJ-5 5 Chart 6 4 FJ-6 5 FJ-7 5 Appendix K Title Page 4 K-i and K-li 4 K-iii 6 K-1 and K-2 5 K-3 6 K-4 5 K-5 and K-6 4 K-7 through K-10 5 K-11 4 K-12 6 TK3.1-1 (Sheet 1 of 1) 5 TK3.2-1 (Sheet 1 of 1) 6 TK3.2-2 (Sheet 1 of 1) 6 TK3.2-2 (Sheet 2 of 2) 5 FK3.1.1 through FK3.1.16 6 FK3.2.1 through FK3.2.11 6 Appendix L Title Page 5 L-i and L-il 6 L-lii 5 L-iv and L-v 6 L-1 5 L-2 and L-2a 6 L-3 and L-4 6 L-5 5 EP-6 Revision 6 - December 1986
l LIST OF EFFECTIVE PAGES (CONT'D) i Text, Table (T), Fevision l
% or Fiqure (F) Number 1
L-6 through L-8 6 , L-8a and L-8b 6 I L-9 through L-11 6 i L-11a through L-lif 6 L-12 through L-15 6 L-16 5 TL-1 through TL-3 (Sheet 1 of 1) 5 TL 4 (Sheet 1 of 3) 6 TL-4 (Sheets 2 and 3 of 3) 5 TL-5 (Sheets 1 and 2 of 2) 6 TL-6 (Sheet 1 of 1) 6 TL-7 (Sheets 1 and 2 of 2) 6 TL-8 through TL-16 (Sheet 1 of 1) 6 TL 17 (Sheets 1 and 2 of 2) 6 TL-18 (Sheets 1 through 3 of 3) 6 TL-19 (Sheets 1 through 4 of 4) 6 TL-20 (Sheets 1 and 2 of 2) 6 TL-21 and TL-22 6 TL-23 (Sheets 1 and 2 of 2) 6 TL-24 (Sheet 1 of 1) 6 FL-1 through FL-24 5 , FL-25 through FL-36 6 FL-37 and FL 38 5
\ -
FL-39 6 Appendix M Title Page 5 M-i through M-lii 6 M-1 through M 3 6 TM-1 (Sheet 1 of 1) 6 FM-1 6 f I
- O 4
EP-7 Revision 6 - December 1986
TABLE OF CONTENTS
.P_n.9.9.
SECTION 1 - INTRODUCTION i 1.1 PURPOSE 1-1 1.2 SCOPE 1-1 1.3 STATUS OF CONSTRUCTION 1-2 1.4 SUPPORTING PROGRAM 1-2 , 1.5
SUMMARY
OF DESIGN ASSESSMENT 1-2 , 1.6 GENERAL ARRANGEMENT OF SUPPRESSION POOL STRUCTURES 1-4 SECTION 2 - LOADS AND LOAD COMBINATIONS 2.1 LOAD DESCRIPTION - SUPPRESSION POOL HYDRODYNAMIC LOADS AND RELATED EFFECTS 2-1 2.1.1 Safety / Relief Valve Actuation 2 2.1.2 Steam Quenching Vibrations 2-1 2.1.3 LOCA Loads 2-2 2.1.3.1 Design Basis Accident 2-2 2.1.3.2 Intermediate Break Accident 2-3 2.1.3.3 Small Break Accident 2-4 l 2.2 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA 2-4 2.2.1 Reinforced Concrete Structures 2-5 2.2.2 Steel Structures 2-5 2.2.3 Piping and Equipment 2-6 2.2.4 Reactor Pressure Vessel and Internals 2-6 2.2.5 Combination of Dynamic Responses 2-7 2.3 OTHER DYNAMIC LOADS 2-7 2.4 APPROACH USED FOR DESIGN ASSESSMENT 2-8 2.4.1 Submerged Structure / Pool Swell Loads 2-8 2.4.2 Building Response Load 2-9 2.4.3 Related Effects 2-11 SECTION 3 - SRV LOADS
3.1 INTRODUCTION
3-1 3.2 POOL BOUNDARY LOADS 3-2 3.2.1 Design Basis Load Definition (Ramshead 3-2 Discharge Device) l 3.2.1.1 Ramshead Load Specification 3-3 0 3.2.1.2 3.2.2 Ramshead Load Summary T-quencher Load Definition 3-5 3-6 l 1 i Revision 5 - December 1981 4
. . - . , . - _ ,, ,. ~. ~ - - - , _ - _ . - . _ _ . _ , , . , - - , - . - - _ . . , . - - - _ - . --
TABLE OF CONTENTS (CONT'D) Page SECTION 4 - LOCA LOADS 4.0 GENERAL 4-1 4.1 ANALYTICAL METHODS AND DISCUSSION OF LOADS 4-1 4.1.1 Vent clearing 4-1 4.1.1.1 Submerged Structure Loads Due to Vent Clearing 4-1 4.1.1.2 Basemat Loads Due to Vent Clearing 4-2 4.1.2 LOCA Bubble Formation 4-2 4.1.2.1 Submerged Structure Loads Due to LOCA Bubble Formation 4-3 4.1.2.2 Pool Boundary Loads Due to LOCA Bubble Formation 4-3 4.1.3 Pool Swell and Fallback 4-3 4.1.3.1 Impact Loads on Small Structures From Pool Swell 4-5 4.1.3.2 Impact Loads on Large Structures From Pool Swell 4-5 4.1.3.3 Drag Loads on the Downcomer Vents Due to Pool I Swell 4-6 4.1.3.4 Drag Loads on Structures Other Than Downcomer Vents Due to Pool Swell 4-6 4.1.3.5 Loads on Grating Due to Pool Swell 4-6 4.1.3.6 Suppression Chamber Boundary Loads During Pool Swell 4-7 4.1.3.7 Drywell Floor Loads Due to Pool Swell 4-7 4.1.3.8 Fallback Loads 4-8 4.1.4 Quasi-Steady Vent Flow 4-8 4.1.4.1 Vertical Loads on the Downcomer Vents Due to l Viscous and Pressure Forces of Vent Flow 4-9 4.1.4.2 Pool Boundary Loads Due to Condensation Oscillations 4-9 4.1.4.3 Submerged Structure Loads Due to Condensation l Oscillations 4-10 4.1.5 Chugging 4-10 4.1.5.1 Lateral Loads on Downcomer Vents Due to Chugging 4-10 l 4.1.5.2 Pool Boundary Loads Due to Chugging 4-11 4.1.5.3 Submerged Structure Loads Due to Chugging 4-11 4.2 SHOREHAM PLANT SPECIFIC LOADS AND RESPONSE CONDITIONS 4-11 4.2.1 Vent clearing 4-11 l 4.2.2 LOCA Bubble Formation 4-12 4.2.3 Pool Swell and Fallback 4-12 4.2.3.1 Impact Loads on Small Structures Due to Pool Swell 4-13 4.2.3.2 Impact Loads on Large Structures Due to Pool Swell 4-13 1 4.2.3.3 Drag Loads on the Downcomer Vents Due to Pool Swell 4-13 i 4.2.3.4 Drag Loads on Structures other than Downcomer l Vents Due to Pool Swell 4-14 4.2.3.5 Loads on Grating Due to Pool Swell 4-14 11 Revision 5 - December 1981
TABLE OF CONTENTS (CONT'D) ex
\ ,] '***
4.2.3.6 Suppression Chamber Boundary Loads Due to Pool Swell 4-14 4.2.3.7 Drywell Floor Loads Due to Pool Swell 4-15 4.2.3.8 Fallback Loads 4-15 4.2.4 Quasi-Steady Vent Flow 4-15 4.2.4.1 Vertical Loads on the Downcomer Vents Due to Vent Flow 4-15 4.2.4.2 Pool Boundary Loads Due to Condensation Oscillation 4-16 l 4.2.5 Chugging 4-16 4.2.5.1 Lateral Loads on Downcomer Vents Due to Chugging 4-16 4.2.5.2 Pool Boundary Loads Due to Chugging 4-17 4.2.5.3 Submerged Structure Loads Due to Chugging 4-17 SECTION 5 - DYNAMIC RESPONSE OF PRIMARY STRUCTURES 5.1 STRUCTURAL RESPONSE TO SRV LOADS 5-1 5.1.1 Summary of Results 5-2 5.1.2 containment Structures Response to SRV Ramshead Loads 5-4 5.1.2.1 Response to All Valve Sequential Discharge 5-4 5.1.2.2 Response to ADS Discharge 5-4 (g 5.1.2.3 Response to Three Adjacent Valve (_) Out of Phase Discharge 5-4 5.1.2.4 Response to Single Valve Discharge 5-5 5.1.2.5 Response to All Valve Simultaneous Discharge 5-5 5.1.2.6 Response to Three Adjacent Valve Simultaneous Discharge 5-5 5.1.2.7 High Frequency Response Study 5-6 5.1.3 containment structures Response to SRV T-Quencher Loads 5-7 5.1.3.1 Response to All Valve Discharge 5-8 5.1.3.2 Response to ADS Discharge 5-8 5.1.3.3 Response to Tnree Adjacent Valve Discharge 5-8 5.1.3.4 Response to Single Valve Discharge 5-8 5.2 STRUCTURAL RESPONSE TO LOCA LOADS 5-8 5.2.1 Summary of Results 5-9 5.2.2 containment Structures Response to LOCA Loads 5-10 5.2.2.1 Response to Vent Clearing 5-10 5.2.2.2 Response to Condensation Oscillation Loads 5-10 5.2.2.3 Response to Chugging 5-10 5.3 STRUCTURAL' RESPONSE TO ANNULUS PRESSURIZATION LOADS 5-11 111 Revision 5 - December 1981 4
TABLE OF CONTENTS (CONT'D) SECTION 6 - PRIMARY STRUCTURES ASSESSMENT
6.1 INTRODUCTION
6-1
6.2 DESCRIPTION
OF STRUCTURES 6-1 6.3 DESIGN CRITERIA AND LOADS 6-1 6.3.1 Design Criteria 6-1 6.3.2 Loads 6-2 6.3.3 Load combinations 6-3 6.4 METHOD OF ANALYSIS 6-3 6.5
SUMMARY
, DESIGN MARGINS, AND CONCLUSIONS 6-3 6.5.1 Containment Internal Loads 6-3 6.5.2 Design Margins 6-5 6.5.3 Conclusions 6-5 SECTION 7 - CONTAINMENT LINER ASSESSMENT 7.0 GENERAL 7-1 7.1 BASEMAT LINER 7-1 7.1.1 Load Sources and Design Criteria 7-1 7.1.1.1 Load Sources 7-1 7.1.1.2 Design Criteria 7-2 7.1.2 Combining and Applying Loads 7-2 7.1.2.1 Combining Loads 7-2 7.1.2.2 Assumptions 7-3 7.1.3 Basemat Liner Analytical Methods 7-3 7.1.3.1 Strain Displacement Relations 7-3 7.1.3.2 Stress Concentration Factors 7-4 7.1.3.3 Basemat Liner Static Analysis 7-4 7.1.3.4 Basemat Liner Anchorage System 7-4 7.1.4 Basemat - Summary and Design Margin 7-4 ! 7.1.4.1 Strains and Stress from SRV Discharge 7-4 7.1.4.2 Operating Stress Range Comparison 7-5 l 7.1.4.3 Strain Evaluation per ASME III, Division 2 7-5 l 7.1.4.4 Fatigue Analysis 7-5 7.2 WALL LINER AND ANCHOR SYSTEM 7-6 7.2.1 Load Sources and Design Criteria 7-6 7.2.1.1 Load Sources 7-6 7.2.1.2 Wall Liner Design Criteria 7-7 7.2.2 Combining and Applying Loads 7-7 7.2.2.1 Combining Loads 7-7 7.2.2.2 Fatigue Load Combinations 7-8 7.2.2.3 Assumptions on Number of Events per Combination 7-8 7.2.2.4 Stress concentration Factor 7-8 7.2.3 Wall Liner Analytical Methods 7-8 l 7.2.3.1 SRV Loads 7-8 iv Revision 5 - December 1981
l l j TABLE OF CONTENTS (CONT'D) l i () 7.2.3.2 Pool Swell Loads Page 7-9 i 7.2.3.3 condensation Oscillation and Chugging Loads 7-9 i 7.2.3.4 LOCA Vent Clearing Loads 7-9 i 7.2.4 Justification of Quasi-Static Analysis 7-9 7.2.5 Wall Liner Summary and Design Margin 7-10 { 7.2.5.1 Wall Liner Strain from SRV Only 7-10 { 7.2.5.2 Wall Liner Comparisons per ASME III, Division 2 7-10 l 7.2.5.3 Wall Anchorage Summary 7-10 ) 7.2.5.4 Operating Stress Range Comparison Summary 7-10 3 7.2.5.5 Wall Liner Fatigue Analysis Summary 7-11 i 7.2.6 Piping Penetrations 7-11 l j SECTION 8 - ASSESSMENT OF THE SECONDARY CONTAINMENT AND ! OTHER STRUCTURES ! i j
8.1 INTRODUCTION
8-1 1 l 8.2 SECONDARY CONTAINMENT 8-1 : l 8.3 DRYWELL FLOOR AND SUPPORT COLUMNS 8-1 8.4 DRYWELL STRUCTURAL STEEL 8-2 8.5 DOWNCOMER BRACING 8-2 i l 8.6 PLATFORMS, LADDERS, AND WALKWAYS 8-3 8.7 CABLE TRAY AND CONDUIT SUPPORTS 8-3 l SECTION 9 - PLANT PIPING, COMPONENTS, AND EQUIPMENT 1 ASSESSMENT I 9.1 BOP PIPING AND EQUIPMENT 9-1 i 9.1.1 Piping System 9-1 1 9.1.1.1 Reevaluation Procedures 9-1 4 9.1.1.2 Analytical Techniques 9-3 ' 9.1.1.3 Results 9-5 9.1.1.4 Shoreham commitment to confirmatory Loads 9-7 l 9.1.2 Equipment 9-7 9.1.2.1 Summary 9-7 9.1.2.2 Dynamic Loads and Stress Limits 9 9.1.2.3 Evaluation Procedures and Results 9-10 l 9.1.2.4 Operability Assurance 9-12 9.1.2.5 Future Equipment Modifications 9-13 l 9.2 NSSS PIPING AND EQUIPMENT 9-13 9.2.1 Introduction 9-13 9.2.2 Evaluation Procedures 9-13 l 9.2.3 Reactor Pressure Vessel Supports and Internal O 9.2.4 Components Evaluation Floor Mounted Equipment 9-16 9-17 9.2.5 NSSS Piping and Pipe Mounted Equipment Evaluation 9-18 v Revision 6 - December 1986
TABLE OF CONTENTS (CONT'D) Page 9.2.6 NSSS Safety-Related Instrumentation Evaluation 9-18a l 9.2.7 NSSS Operability Assurance 9-19 SECTION 10 - CONDENSATION INSTABILITY DURING SRV DISCHARGE
10.1 INTRODUCTION
10-1 10.2 BULK TO LOCAL TEMPERATURE DIFFEP.ENCES DURING SRV DISCHARGE 10-1 10.3 SUPPRESSION POOL TEMPERATURE RESPONSE TO TRANSIENTS INVOLVING SRV DISCHARGE 10-2 l 10.4 SUPPRESSION POOL TEMPERATURE MONITORING SYSTEM 10-3 REFERENCES R-1 APPENDIX A LEAD PLANT ACCEPTANCE CRITERIA (NUREG-0487, APPENDIX D) POSITIONS APPENDIX B CONTAINMENT STRUCTURE DESIGN MARGIN APPENDIX C FLUID-STRUCTURE INTERACTION (FSI) APPENDIX D RESPONSE TO NRC QUESTIONS APPENDIX E FUNCTIONAL CAPABILITY CRITERIA FOR MARK II PIPING l APPENDIX F FATIGUE ANALYSIS OF DOWNCOMERS AND SRVOL'S APPENDIX G JUSTIFICATION OF MARK-II LEAD PLANT SRV DESIGN BASIS LOAD DEFINITION l APPENDIX H POOL SWELL-MODELING l APPENDIX I PROPRIETARY APPENDIX J SNPS SUPPRESSION POOL TEMPERATURE TRANSIENTS APPENDIX K SUBMERGED STRUCTURES APPENDIX L MARK II HYDRODYNAMIC LOADS CONFIRMATORY PROGRAM APPENDIX M DRYWELL FLOOR VACUUM BREAKER CYCLING DURING CHUGGING O vi Revision 5 - December 1981
LIST OF TABLES O Table Title 1-1 Summary of Loads 2-1 All Possible Hydrodynamic Load Combinations 2-2 Load Combinations and Load Factors for Reinforced Concrete Structures 2-3 Load Combinations and Stress Limits for Structural Steel 2-4 Load Combinations and Acceptance Criteria for Piping and Equipment 2-5 NSSS RPV and Internals - Reactor System Detailed Load Combinations 3-1 Summary of Maximum and Minimum Average Wall Pressures for Sequential SRV Discharge 3-2 Summary of Maximum and Minimum Wall Pressures for Automatic Depressurization Eystem Actuation 3-3 Summary of Maximum and Minimum Wall Pressures for Asymmetric SRV Discharge 3-4 Summary of Maximum and Minimum Wall Pressures for Single Valve Discharge 3-5 Loads on Quencher Body ,() 3-6 Loads on Quencher Arm 4-1 Summary of LOCA Affected Structures 4-2 Drag Coefficients of Various Shapes 4-3 Shoreham Data for DBA Transient and Pool Swell Analysis 4-4 Drywell Pressure as a Function of Time for DBA 5-1 Maximum values of Dynamic Loads in the Basemat and Superstructures from a Sequential All Valve SRV ! Discharge with Ramshead
- 5-2 Maximum Values of Dynamic Loads in the Basemat
! from a Simultaneous All Valve SRV Discharge with Ramshead 5-3 Maximum values of Dynamic Loads in the Super-structures from a Simultaneous All Valve SRV Discharge with Ramshead . 5-4 Definition of Internal Loads 5-5 Maximum Values of Dynamic Loads in the Basemat , and Superstructures from a Simultaneous Three
- Adjacent valve SRV Discharge with Ramshead 5-6 Maximum Values of Dynamic Loads in the Basemat from LOCA Vent Clearing Loads 5-7 Maximum values of Dynamic Loads in the Super-structures from LOCA Vent Clearing Loads O
vii Revision 6 - December 1986 a , - ---- ,-,,w, n-,-- , - , , ,~r - - , ,-- , , - ~ , -, ,,--rnu, .,,-,-n,r------rv vw ~w-e e -w wm----rew-',-~ ,,-r< - - *,w - -- --
l LIST OF TABLES (CONT'D) Table Title 58 Maximum values of Dynamic Loads in the Basemat from Axisymmetric 20 Hz Chugging Loads 5-9 Maximum values of Dynamic Loads in the Super-structures from Axisymmetric 20 Hz Chugging Loads 5-10 Maximum values of Dynamic Loads in the Basemat from Axisymmetric 30 Hz Chugging Loads 5-11 Maximum values of Dynamic Loads in the Super-structures from Axisymmetric 30 Hz Chugging Loads 6-1 Load Combinations and Load Factors for Reinforced Concrete Structures 6-2 Definition of Internal Loads 6-3 Basemat Design Internal Loads Just Outside Pedestal 6-4 Reactor Pedestal and Primary Containment Design Internal Loads 6-5 Minimum Design Margins for Flexure and Axial Tensile Loads 6-6 Minimum Shear Reinforcement Requirements 7-1 Stress Cyclic Data for Fatigue Analysis 9-1 Reactor Water Cleanup Piping 010 9-2 Fuel Pool Cooling and Cleanup Piping 071 9-3 Residual Heat Removal Piping 821 9-4 Functional Capability Evaluation 9-5 Submerged RHR Piping 10-1 Summary of the Shoreham Pool Temperature Results 10-2 Containment Energy Distribution for Shoreham case 3A l (SBA w/one RER HR) I 1 l l l O l viii Revision 6 - December 1986 l \ L
LIST OF FIGURES O V Ficure Title 1-1 Suppression Chamber Piping Composite - Plan El. 20'-0" to 40'-0", North j 1-2 Suppression Chamber Piping Composite - Plan El. 20'-0" to 40'-0", South 1-3 Suppression Chamber Piping Composite -
, Plan El. 40'-0" to 63'-0", North
! 1-4 Suppression Chamber Piping Composite - Plan El. 40'-0" to 63'-0", South 1-5 Suppression Chamber Piping Composite - Section 1-1 j 1-6 Suppression Chamber Piping Composite - Section 2-2 1-7 Drywell Vent Piping - Sheet 1 : l 1-8 Drywell Vent Piping - Sheet 2 l 1-9 Drywell Vent Piping - Sheet 3 ' '
- 1-10 Drywell Vent Piping - Sheet 4
. 1-11 SRV Discharge in Piping Suppression Chamber l 2-1 Event-Time Relationship for SRV Discharge Due , to Anticipated Plant Transients 2-2 Event-Time Relationship for the Design Basis i Accident i 2-3 Load Combination History Structure Affected:
- Drywell Floor
; Accident Condition: Large Line Break (DBA) 2-4 Load combination History Structure Affected:
- Downcomers
! Accident Condition: Large Line Break (DBA) ! 2-5 Load Combination History Structure Affected:
Downcomers . i Accident Condition: Intermediate Line Break i 2-6 Load Combination History Structure Affected:'
; Downcomers i' Accident Condition: Small Line Break 2-7 Load Combination History Structure Affected:
Downcomers Accident Condition: None 2-8 Load Combination History Structures Affected: Wetwell Walls Above the Water Level Accident Condition: Large Line Break (DBA) 2-9 Load Combination History Structure Affected: l Submerged Wetwell Accident Condition: Large Line Break.(DBA) 2-10 Load Combination History Structure Affected: Submerged Wetwell Accident Condition: Intermediate Line Break 2-11 Load Combination History Structure Affected: Sub- : j merged Wetwell ! Accident Condition: Small Line Break 1 2 12 Load Combination History Structure Affected: Sub-merged Wetwell . l i Accident Condition: None l ix Revision 3 - November 1978 l i ,
)
l
LIST OF FIGURES (CONT'D) Fiqure _ Title 2-13 Load Combination History Structure Affected: Small Submerged Structures, Columns, and Piping Accident Condition: Large Line_3reak (DBA) 2-14 Load Combination History Structures Affected: Small Submerged Structures, Columns, and Piping Accident Condition: . Intermediate Line Break 2-15 Load Combination History Structuras Affected: Small Submerged Structores, Columns and Piping Accident Condition: Small Line Break 2-16 Load Combination History Structure Affected: Small Submerged Structures, Columns, and Piping Accident Condition: None 2-17 Load Combination History Structure Affected: Small Structures AbovE Pool and Below Breakthrough Accident Condition: Large Line Break (DBA) 2-18 Load combination History Structures Affected: Small Structures Above Breakthrough Accident Condition: Large Line Break (DBA) 2-19 Structural Model for SRV Load Analysis 2-20 Loads Applied to Model 3-1 Phenomenon of Safety / Relief. Valve Blowdown into Suppression Pool 3-2 Cross-Section of Suppression Pool and Definition of Suppression Chamber Walls Loading Zone for Ramshead Load Definition 3-3 Orientation of SRV Line Discharge Devices (Rams-heads) for Sequential Snv Discharge 3-4 Typical All Valve Sequential SRV Discharge Forcing Function for Ramshead Device - Zone 14 3-5 Orientation of SRV Discharge Line Devices (Rams-heads) for Automatic Depressurization 3-6 Typical Automatic Depressurization System l Actuation Forcing Function for Ramshead Device - Zone 14 3-7 Orientation of SRV Line Discharge Devices (Rams-heads) for Asymmetric SRV'Dischargo 3-8 Typical Asymmetric SRV Discharge Forcing Function for Ramshead Device - Zone 14 l 3-9 Orientation of SRV Line Discharge Device (Ramshead) for Single Valve Discharge 3-10 Typical Single SRV Discharge Forcing i Function for Ramshead Device - Zone 14 i 3-11 Normalized Pressure Time History for All SRV's Discharging Simultaneously and in Phase on Ramshead Device l 3-12 Normalized Pressure Boundary Load Distribution l Around the circumferential Direction on Primary l Containment for Three Adjacent SRV's Discharging x Revision 5 - December 1981 l l l
.~ - - , . .- .-- ..-. -. - . - - - . - . -_-- _ ._ - -
l l LIST OF FIGURES (CONT'D) 'O Fiaure Title 4 Simultaneously and in Phase Based on Ramshead Device s 1 3-13 KWU T-Quencher 3-14 KKB Pressure Trace No. 35' i 3-15 KKB Pressure Trace No. 76 3-16 KKB Pressure Trace No. 82
- 3-17 Normalized Pressure Distribution on Suppression i
Pool Boundaries - Symmetric and ADS Cases j 3-18 Normalized Vertical Pressure Distribution for All Cases and for Submerged Structures i 3-19 Normalized Pressure Distribution on Suppression Pool l Boundaries - Asymmetric Case 3-20 Normalized Pressure Distribution on Suppression Pool Boundaries - Single SRV Discharge Case 3-21 Loads on Quencher (without pressure loads) 3-22 Loads on Quencher Arms (without pressure loads) 4-1 Schematic Representation of the Pool Swell Model
- 4-2 Pressure Drop Due to Flow across Grating Duration of Load
- 0.5 Sec 4-3 Assumed Pressure Distribution for Pool Boundary
() 4-4 Loads During Pool Swell Drag Pressure for Pe = 62.4.(lb./ft2) Downcomer Model for Lateral Load Assessment 4-5 4-6 Pool Boundary Chugging Loads 4 4-7 Vent Liquid C,learing Velocity Following a DBA.
- . 4-8 Containment Pressure Response During Pool Swell i Following a DBA 4-9 Pool Surface Elevation and Velocity Following a DBA
.5-1 Arplified Response Spectra of Vertical Acceleration, Top of Reactor Support Pedestal, All Valve Sequential Discharge - Ramshead 5-2 Amplified Response Spectra of Vertical o Acceleration, Primary Containment at Elevation
! of Stabilizer Truss,;All Valve Sequential Discharge Ramshead l' 5-3 Amplified Response Spectra of Horizontal (N-S) Acceleration, Top of Reactor Support Pedestal, All Valve Sequential Discharge - Ramshead 5-4 Amplified Response Spectra of Horizontal (N-S) ! Acceleration, Primary Containment-at Elevation of Stabilizer Truss,.All Valve Sequential Discharge - Ramshead 5-5 Amplified Response Spectra of Horizontal (E-W) Acceleration, Top of Reactor Support Pedestal, () All Valve Sequential Discharge - Ramshead xi Revision 5 - December 1981 J
-----re. .n-w - , , ,-.,, , , , , -. , - , , , ,,,,ww,, ,,, - - - - w- y- r- -
-+,,g,,,,,wm- wem e mw. v,.,g-ow,. -,n--w,,--- ,=e, - , - . , . - - - - . .w-- - -
I 4 l LIST OF FIGURES (CONT'D) Fiqur_e_ Title 5-6 Amplified Response Spectra of Horizontal (E-W) Acceleration, Primary Containment at Elevation of Stabilizer Truss, All Valve Sequential Discharge - ; Ramshead 5-7 Amplified Response Spectra of Vertical Acceleration, Top of Reactor Support Pedestal, 3 Adjacent Valve Out of Phase Discharge - Ramshead 5-8 Amplified Response Spectra of Vertical Acceleration, ) Primary Containment at Elevation of Stabilizer Truss, 3 Adjacent Valve Out of Phase Discharge - Ramshead 5-9 Ampliied Response Spectra of Horizontal (N-S) Acceleration, Top of Reactor Support Pedestal, 3 Adjacent Valve Out of Phase Discbarge - Ramshead 5-10 Amplified Response Spectra of Horizontal (N-S) Acceleration, Primary Containment at Elevation of ' Stabilizer Truss, 3 Adjacent Valve Out of Phase l Discharge - Ramshead 5-11 Amplified Response Spectra of Horizontal (E-W) Acceleraticn, Top of Reactor Support Pedestal, 3 Adjacent Valve Out of Phase Discharge - Ramshead 5-12 Amplified Response Spectra of Horizontal (E-W) Acceleration, Primary Containment at Elevation of Stabilizer Truss, 3 Adjacent Valve Out of Phase Discharge - Ramshead 5-13 Critical Locations 5-14 Positive Sign Convention for Internal Loads 5-15 Amplified Response Spectra of Horizontal Acceleration Primary Containment at El 21 Ft, 3 Adjacent Valve out of Phase Discharge 5-16 Amplified Response Spectra of Vertical Acceleration Primary Containment at El 83 Ft, 3 Adjacent Valve out of Phase Discharge 5-17 Amplified Response Spectra of Horizontal Acceleration Secondary Containment at El 38 Ft, 3 Adjacent Vadve out of Phase Discharge 5-18 Amplified Response Spectra of Vertical Acceleration Secondary Containment at El 38 Ft, 3 Adjacent Valve out of Phase Discharge 5-19 Amplified Response Spectra of Vertical Acceleration, Top cf Reactor Support Pedestal, T-Quencher, All Valve Discharge - Pressure Trace No. 1 5-20 Ampliied Response Spectra of Vertical Acceleration, Top of Reactor Support Pedestal, T-Quencher, All Valve Discharge - Pressure Trace No. 2 5-21 Ampliied Response Spectra of Vertical Acceleration, Top of Reactor Support Pedestal, T-Quencher, All 5-22 Valve Discharge - Pressure Trace No. 3 Amplified Response Spectra of Vertical Acceleration, lh xii Revisien 5 - December 1981 m
- _- .. - _ . - . - . - - .-- - - - - - _ - _ - .+
LIST OF FIGURES (CONT'D) () Fiqure Title Top of Reactor Support Pedestal, T-Quencher, All valve Discharge 5-23 Amplified Response Spectra of Vertical Acceleration, ' Primary Containment at Elevation of Stabilizer Truss, T-Quencher, All valve Discharge
- ' 5-24 Amplified Response Spectra of Horizontal Acceler-ation, Top of Reactor Support Pedestal, T-Quencher, All Valve Discharge '
5-25 Amplified Response Spectra of Horizontal Acceler-ation, Primary Containment at Elevation of Stabil-izer Truss, T-Quencher, All Valve Discharge , 5-26 Amplified Response Spectra of Vertical Acceleration, Top of Reactor Support Pedestal, T-Quencher, 3 , Adjacent Valve Discharge 5-27 Amplified Response Spectra of Vertical Acceleration, Primary Containment at Elevation of~ Stabilizer Truss, T-Quencher 3, Adjacent valve Discharge 6 5'-28 Amplified Response Spectra of Horizontal Acceler-ation, Top of Reactor Support Pedestal, T-Quencher, [ 3 Adjacent Valve Discharge 5-29 Amplified Response Spectra of Horizontal Acceler-ation, Primary Containment at Elevation of 4 O 5-30 Stabilizer Truss, T-Quencher, 3 Adjacent Valve Discharge LOCA Vent Clearing, Idealized Pressure Time History 5-31 LOCA Chugging Idealized Pressure Time Histories 5-32 Amplified Response Spectra of Vertical Acceleration, l Top of Reactor Support Pedestal, LOCA Vent. Clearing 5-33 Amplified' Response Spectra of Vertical Acceleration, , Primary Containment atLElevation of Stabilizer ' Truss, LOCA Vent Clearing 5-34 Amplified Response Spectra of Vertical Acceleration-- Top of Reactor Support Pedestal -: LOCA Axisymmetric Condensation' Oscillation 5-35 Amplified Response Spectra of Vertical Accele-t' ration - Primary Containment at Elevation of Stabilize Truss - LOCA Axisymmetric Condensation Oscillation 5-36 Amplified Response-Spectra of Horizontal Acceleration, Top of Reactor. Support Pedestal,-LOCA Asymmetric Condensation Oscillation 5-37 Amplified Response Spectra of Horizontal Acceleration - Primary Containment at Elevation of Stabilizer' Truss - LOCA Axisymmetric Condensation . Oscillation
- 5-38 Amplified Response Spectra of Vertical Acceleration, l Top of Reactor Support Pedestal, LOCA Axisymmetric Chugging
,O 5-39' Amplified Response Spectra of. Vertical Acceleration, Primary Containment at' Elevation of Stabilizer xiii Revision 5 December 1981
. _ _ _ _ _ . _ . . _ ~ _ _ . _ . . . _ . . . . _ _ _ . . _ . . . . . - . _ . , _ . . _ _ . - . . _ .
LIST OF FIGURES (CONT'D) Fioure Title Truss, LOCA Axisymmetric Chugging 5-40 Amplified Response Spectra of Horizontal Acceleration - Top of Reactor Support Pedestal - LOCA, Asymmetric Chugging 5-41 Amplified Response Spectra of Horizontal Acceler-ation, Primary Containment at Elevation of Sta-bilizer Truss, LOCA, Asymmetric Chugging 5-42 Amplified Response Spectra of Horizontal Acceleration - 1sactor Vessel Elevation 100, AP from Recirc Line Break 5-43 Amplified Response Spectra of Horizontal Acceleration - Shieldwall Elevation 137, AP from Recirc Line Break 5-44 Amplified Response Spectra of Horizontal Acceleration - Pedestal Elevation 90, AP from Recirc Line Break 5-45 Amplified Response Spectra of Horizontal Acceleration - Reactor Vessel Elevation 119, AP from Feedwater Line Break 5-46, Amplified Response Spectra of Horizontal Acceleration - Shield Wall Elevation 137, AP from Feedwater Line Break 5-47 Amplified Response Spectra of Horizontal Acceleration - Pedestal Elevation 90, AP from Feedwater Line Break 6-1 General Arrangement of Reactor Building 6-2 Critical Locations 6-3 Positive Sign Convention for Internal Loads 6-4 Primary Containment Wall Reinforcing Details El 8'-0" To 60'-0" 6-5 Reactor Support Wall Reinforcing Details 6-6 Containment Mat-Top Reinforcing Details 6-7 Containment Mat-Bottom Reinforcing Details 6-8 Flexure and Axial Load 7-1 Reactor Containment Liner Floor Details 7-2 Mat Liner - Details of Welds 7-3 Mat Liner - Calculation of Liner Displacements Using Mat Mid-Thickness Displacements 7-4 Liner-Anchor Model 7-5 The Amplification Factor 4 As a Function of the Frequency Ratio for Various Amounts of Viscous Damping 9-1 Reactor Water Cleanup 010 9-2 Fuel Pool Cooling 071 9-3 RHR Piping 821 9-4 ARS - Primary Containment, Upset - Horizontal xiv Revision 5 - December 1981
O LIST OF FIGURES (CONT'D) Fiqure Title SRSS (SRV + OBE) - Ramshead 9-5 ARS - Primary Containment, Upset - Vertical SRSS (SRV + OBE) - Ramshead 9-6 ARS - Primary Containment, Faulted - Horizontal SRSS (SSE + LOCA + SRV) - Ramshead 9-7 ARS Primary Containaent, Faulted - Vertical SRSS (SSE + LOCA + SRV) - Ramshead 9-8 ARS - Secondary Containment, Upset - Horizontal SRSS (SRV + OBE) - Ramshead 9-9 ARS - Secondary Containment, Upset - Vertical SRSS (OBE + SRV) - Ramshead 9-10 ARS - Secondary Containment, Faulted - Horizontal SRSS (SSE + LOCA + SRV) - Ramshead 9-11 ARS - Secondary Containment, Faulted - Vertical SRSS (SSE + LOCA + SRV) - Ramshead 9-12 Definition of Static Coefficients in USAR 10-1 Pool Temperature and Pressure vs Time With and O Without Wetwell Heatsinks - Case 2A. 10-2 Pool Temperature and Pressure vs Time With and Without Drywell and Drywell Heatsinks - Case 3A O xv Revision 6 - December 1981
SECTION 1
, () INTRODUCTION 1.1 PURPOSE I The purpose of Revision 6 of the Design Assessment Report (DAR) is to present the completed design assessment of the Shoreham l Nuclear Power Station Unit 1 (SNPS-1) for hydrodynamic loads associated with safety / relief valve (SRV) discharge and the postulated loss-of-coolant accident (LOCA) in a BWR Mark II (MK-II) containment. This revision provides the following information:
- 1. Redesign of the drywell floor. vacuum breaker due to a fast oscillation during LOCA pool swell,
- 2. Use of the LaSalle bulk-to-local pool temperature limits in conjunction with a subscale test for Shoreham-unique limits,
- 3. Evaluation results of Nuclear Steam Supply System (NSSS) piping and equipment based on confirmatory loads,
- 4. Evaluation results of 67 safety-related piping subsystems and associated 533 pipe supports in the phase II confirmatory review, and
) 5. Revision of all sections to reflect that the plant is 100% complete and all analyses have- been completed in accordance with plant as-built. conditions.
For the design basis assessment, the methods used. to define, apply, and combine the loads are in compliance with the Nuclear Regulatory Commission (NRC) Lead Plant Acceptance Criteria (***> as outlined in Appendix A. As one of the three. lead plants identified in Section I.B.1 of Reference 1, the basic supporting document for the SNPS-1 design assessment is the MK-II containment Dynamic Forcing Functions Information Report (DFFR), Revision 2,<8 *>. Additional references will be cited where the methods are different from tnose described in DFFR Revision 2 or where a particular loading condition was not addressed in DFFR Revision 2. 1.2 SCOPE Nuclear Regulatory Commission (NRC) letters of April 18 and 21, 1975 to Long Island Lighting Company (LILCO) discussed.the SRV O ^ l-1 Revision 6 - December 1986
and LOCA hydrodynamic load phenomena associated with the BWR MK II containment. Specific requests for additionhl information included with each letter formed the initial basis for the SNPS-1 design assessment. In the course of investigating the MK-II hydrodynamic phenomena, the requirements for design assessment have been refined, and are now contained for the lead plants (e.g. SNPS-1) in References 1 and 2, referred to as the Lead Plant Acceptance Criteria (LPAC). In Appendix A, the LPAC are addressed item by item to document the compliance identified in Section 1.1. In addition to meeting the LPAC requirements, the lead plants have been required to address the final generic loads accepted by the NRC. Fulfillment of this requirement is l reflected by items 3., and 4. listed in Section 1.1. The SNPS-1 DAR together with the Updated Safety Analysis Report (USAR) and reports referenced by each of these principal licensing documents, constitute the design basis for preparation of the Safety Evaluation Report (SER) and issuance of the station Operating License (OL). 1.3 STATUS OF CONSTRUCTION Overall construction of the Shoreham plant is 100 percent complete. Under the provisions of a 0.001% power license, NPF-19, the Shoreham plant loaded fuel and achieved initial criticality on February 15, 1985. Plant heatup testing was conducted under a 5% power license, NPF-36. l 1.4 SUPPORTING PROGRAM The MK-II supporting program is described in Section III of l Reference 2 and in Reference 6. The generic long term program j (LTP) is considered confirmatory for SNPS-1, 1.5
SUMMARY
OF DESIGN ASSESSMENT A design assessment has been performed on structures, equipment, and piping subjected to loads resulting directly or indirectly I from suppression pool hydrodynamic phenomena. These are as follows: 1
- 1. reactor building basemat, primary and secondary containment structures,
- 2. containment internal structures including the RPV pedestal, drywell floor and support columns, i
1 l l 1-2 Revision 6 - December 1986 l l
I 3. basemat and containment wall liner, l () 4. downcomers and bracing,
- 5. auxiliary containment structures such as platforms, ladders, and support frames,
-6. safety-related and non safety-related piping and pipe supports located within primary and secondary 1 containment, and
[ 7. safety-related equipment located within primary and secondary containment. 1 Section 2.1 provides a general description of the suppression pool hydrodynamic loading phenomena with Sections 3 and 4 i providing more detail for the SRV and LOCA loads respectively. Load combinations, acceptance criteria, and the methods of combining peak dynamic responses are given in Section 2.2. _Non- ' hydrodynamic loads with which the suppression pool hydrodynamic 1 cads must be combined are identified in Section 2.3 with the appropriate USAR reference. Section 2.4 describes the approach
- used in performing the SNPS-1 design assessment for each of the t- following three classifications of loading definitions
l 1. Submerged structure / pool swell loads - loads directly applied by suppression pool hydrodynamic. phenomena to wetwell internal structures and the drywell floor. (
- 2. Building response loads - suppression pool hydrodynamic loads applied directly to the- ' suppression pool boundaries (containment shell, basemat, and RPV 4
pedestal) and then indirectly to structures, piping, and equipment in the drywell and secondary containment due to building dynamic response. i
- 3. Related effects - other design assessment activities related to suppression pool hydrodynamic phenomena but not included .in the above classifications, such as suppression pool- temperature response, downcomer/SRV
- discharge line fatigue evaluations and consideration of drywell floor vacuum breaker cycling.
Sections 5 through 9 provide the results of the SNPS-1 design assessment. Section 5 . covers the dynamic responses ~ of the primary structures- and provides the amplified response spectra i (ARS) used in the nuclear steam supply system'(NSSS) and balance of plant (BOP) piping and equipment evaluations which results are presented in Section 9. Section 6 presents the design assessment results for the primary structures, Section 7 'the results-for the containment. liner, and r
~g Section 8 the results for the secondary structures. Section 10 (V
3 Revision 6 - December 1986 r
. . _ . , , _ . _ _. . ~ . . . - _ , . _ _ . _-_-,.,_...._-.,_,m.,. , . . .
describes the suppres.aion pool temperature response to plant transients involving 5RV discharge and describes the pool temperature monitoring system. Table 1-1 provides a summary of suppression pool hydrodynamic loads used in the SNPS-1 design assessment for each affected structure. Although SNPS-1 has installed KWU "T" Quenchers as the SRV discharge devices, the design assessment is based upon the bounding SRV-ramshead loads as requested by criterion II.2(a) of Appendix D to Reference 1 and shown in Table 1-1. The design assessment has demonstrated that sufficient design margin exists in plant structures and components to withstand the additional effects of the hydrodynamic loads. Certain design modifications were required to achieve this result, as discussed in the appropriate sections. Completion of the confirmatory program as described in Appendix L has not changed these conclusions. Plant design was reviewed for conformance with the final generic load definitions described in Appendix L. In a limited number of applications, the final confirmatory loads had been used as design loads in order to eliminate the need for further modifications. 1.6 GENERAL ARRANGEMENT OF SUPPRESSION POOL STRUCTURES Figures 1-1 through 11 are provided to show the location and size of the wetwell and internal structures. O l l l l l O 1-4 Revision 6 - December 1986 l
. - . _ _ . . ..m . _ _ . _. _. - . _ . _- . .
i i TABLE 1-1
SUMMARY
OF IAADS Load Imadea3 References ]dtts C1 ass-Description LPAC N 3 class R((g E itic& tion (*3 Qggglig i' I. IOCA RELA 7Eb , j A. $UPPRESSIOtt POOL BOUNDARIES i 1. Vent Clearing U 4 .4 .5 .1 31, II.A.1 Secondary leote 2 i
- 2. Pool Swell Air U 4.4.5.3 51, II.A.2, Secondary Note 2 4
Slug /Wetwell II.A.3 Airspace compre. 1 3. Condensation G (12) l ' Oscillation l Primary / Note 5 i L Secondary i 4 Chugging G 4.3.3 i i , b. wsTWELL COMPONENTS 3 (except downcomers) i i
- 1. Vent clearing U 4.4.5.1 III.A.1 Secondary 1
}'
- 2. Air Bubcla U (13) Secondary ,
Formation , j l } 3. Bulk Pool U 4.4.1 1.A.2 N/A Note 6
- Swell Transient Model
- 4. Bulk Pool U 4.4.4 81, II.A.2 Primary Relocated l l Swell HeightsaB vacuum Breakers i 5. Pool Swell Impact
- a. Small U 4.4.6.1 I.A.6 Primary Modiried/ Relocated i Structures Bracing 6 SRV
]' Supports
- b. Grating U 86.4.6.4 I.A.3 Primary Removea/ Redesigned [
Platforms -
.6. Pool Swell Drag U 4.4.5.2/ 111.B.1 (e) Primary Restrain Drywell Floor
, 4 . 4 .7 a j 7. - Pool Fallback U 4.4.5.4 Secondary 7 4 l i j 1 of 4 Revision 5 - December 1981 l l
T TABLE 1-1 fCOttT'D) 'I Load Loadta> References M Claps-Description Class DFFR LPACE S O CCegnents , 988 iticationt* 3
> a. Condensation 0 (13) Primary llote 7
- Oscillation
! Drag
- 9. Osugging Drag D (13) Secondary C. DRYMELL FIDOR
- 1. Bulk Pool Swell U 4.4.6.6 1.A.4 Primary Restrain Drywell Floor 4 D. DottleCOMERS
- 1. Vertical (a) Vent Flow 0 4.2.3 Secondary Drag 9
(b) Pool Swell D 4.4.8 SeconAmry
- WM
- 2. 14teral-External I
(a) Air Bubble U (13) Secondary g Pormation (b) Condensa- U (13) Secondary tion oscilla-tion drag l
, (c) Chugging U (13) Secondary
] kM ! 3. Lateral-Internal (a) Single Vent G 4.3.2.3 I.B.1 Secondary flote 8 ,i i (b) Multi Vent U 4.3.2.4 I.B.2 Secondary Note 9 1 II. SRV REIATED 1 A. SUPPRESSIOle POOL bOUN WtIES(f*), ! 1. Single Valve U 3.2.4tts> II.2 Secondary Isote 2 1
- 2. Asymmetric U 3.2.4(*** II.2 Secondary Note 2 i
] 3. ADS U 3.2.4 II.2 Primary Notes 2 6 5 i e ! 2 of 4 Revision 4 - February 1981 i
__ ___ _ _ _ _ _ ______-_-- ______ ___ . _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ __ . __. _ _ m _. _ . . , . . I w t i l. I TABLE 1-1 (CONT'D1 l 1 l Load Loadt*3 References M Clase-l Description Class DFFR LPACE**D Mg iticationt*3 QDBEente ' k
- s. All valves u 3.2.4 t a n II.2 Seea-dary l (sequential) l B. SUBagJtG3D STRUCTURES U 3.4.1.24**D III.A.2,81, Primary Note 7 II.C.1,
) III.B.2, ,! III.S.3 C. DISCHARGE DEVICE SUPPURT U 3.3.10 Secondary f .i l j tas G - Generic DFFR Imad. ! U - Generic DFFR Method, plant specific load value. ] . (O The design assessment utilises previously calculated loads which are greater than those now developed in the DAR using ,jl l the methods 01 meterance 3 of this report. This note applies caly to primary con +amt structures, not to papang and } e W pment. i j (O This 2ncludes all bulk pool swell and " froth
- during IDCA transient.
1 (O Refer to appands- 3, questica 020.26 for definition. i (** These are primary loads -for piping, secondary loads for structures. Pipe supprt modifications have been made for
! building response loads.
1 (O Pool Swell Model provides maximum airslug and wetwell airspace pressere for 1.A.2 amove and velocity for I.B.S. and l I.B.6. I tu n=ht == tion of condensation oscillation and SRV air bubble submerged structure loads is controlling for pipe support modifications required in suppression pool.
~
to A dynamic analysis will be performed in the long term as required by Imad Plant Acceptance Criteria (LPAC) I.B.1(c) - (see Appendix A). tes . A static moltivent analysis has not noen performed subsequent to bracing relocation (refer to Section 8.5). Multivent effects will be included in the dynasmic analysis identatted in Note S. The basis for the planned multivent nynamic - ammiysis is given in the appropeinte Dhm section. anos Lead Plant Acceptance Criteria (LPAC) or MURAG-0487, Supplement 1 (S1) Sectica modifying applicable DFFR key. 2 requirement (see Appendix A) . (**3 DFFR Rev. 2 includes me requirement for assessing wetwell airspace compreselon. Static analysis of suppressica cpanner hr==d=eles above marinum swell heignt ior maximan wetwell aarspace compressica has been included in the structural t l design =========t. , 2 i 3 of 4 Revision 4 - February 1981 ! I i f i
TABLE 1-1 ACOIrf *DI (**3 Condensation oscillation load not defined in DFFR, Rev 2. Potential for such a load taas identified in DFFR Mov. 2, Section 4.2.2. s e a s MCA % structure loads other .than vent clearing M and haix pool swell not identified in DFFR asv. 2 (sections 4.2, 4.3, and 4.4.5),. Beier to appropriate DAR section and Appendix K. ca*3 Ramshead load is bounding for T quencher and is applied in accordance with LPAC II.2. Aerer to Appenois A and Appendix G. l ca*D Load definition for ramahead subsequent actuation required my Imad Case 1 of LPAC II.2 (b) as moot. used in M-;---
- G includes subsequent actantion data and shoue existing ramshead load is M_r49 T-quencher data base l
(*** sasPs-1 asynenetric case conservatively employs three ad y ent valves instead of taso required by LPAC II.2 (b), Load Case 2. Ea*3 Load Case 5 of LPAC II.2(b) shown in Appendix G to be less soeure than ramshead sequential actuat.Lon (Load Case e ofl LPAC lt.2 (b) . when representative T-quencher load definition is used for simultaneous entry, in-phase oscillation case. Frequency range required by LPAC II.2(c) bonarded oy ramsheed sequential as described in Appendix G. Sequential l actuation case (Case 1 of DFFR, Rev 2, Section 3.2.4.1.2) as therefore conservative for all-valve actuatices. s a m t Jet load negligacle as described in LPAC 111.A.2 and Seq 9 1ement 1 II. Col. Air clearing load meditied by LPAC 111.8.2 and applied for T-quencher per LPAC 111.a.3 (see Appendix A) . 4 of 4 kevision 5 - December 19b1
e-* 180* S&W O' GE I ELECTRICAL PENETRATION 8P W !9 I SPARE (TYP) __
.. i
'\
t <-.6 + c. s, gt .3;. . ,...- ' g i t -- t ,
. g I SUPPRESSION POOL * =Y-LEVEL INSTR. '*;*',', ...
/ \
~.
s
/ '-
\ $
s
, -s'~, ,
t
./ ,
PLATF . O
' s EL.30'- 5" /
g , N i / \ CORE SPRAY ' SUCTION ~
/ s ,
s / '
/
l
/
s' 'ss s .
/
, x f r c. s &
SPARE .- g
~~.
'E J' '
, ':f
- Q* :. .
' $5sN(TYP)'
s \ N j i_- _ Ji + p'. 'N x . . - / 90* S SW 2f , .
, _ _ J i 'k*
270'0E RHR DISCHARGE (TESTING) !* (FIG.1 5) O S to a ..i ... . i
$CALE-FEET
HPCI TURSINE EXHAUST M BREAMER 1 H PCI TURSONE EXH AUST 15'-06-00* 8'e
], ,e RCIC SUCTION E f[ ,I s' ,/, 6 2t. PRL CONTMT .32
/
9
,,// e, HPCI SUCTION 5
o 0$ 3 i ,# RCIC TURBINE EXHAUST o I ! '
~
'x ,
RCIC TURBINE EXHAUST VACUUM l
, i /f , SREANER
'dY -Q, KEY PLAN N i. /,/ f <** RCIC VACUUM e PUMP EXHAUST x '
. \
s! , x x
. .O a-
)
( . _ . s CORE SPRAY suction
,- / s APERTURE s
g.-
,. O e' ,/
G.. "'"' CARD
- SPARE
<.-ee=*' Also Available On
/ ,
q'.. Aperture Ontd x ! /. -
/ ,'
, /'
\O / -
=uL --
""" cc"*** * *'
7
!!'i. - ' ' _,- f/ .2 10 ]2 no Sow f-MATCH LINE - ris. i.e mi-ei so aE 8 612 0 80 64 5 -Cl FIG I-l SUPPRESSION CHAMBER PIPING COMPOSITE-PLAN EL 20'-0" TO 40'-0", NORTH SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRVAND LOCA LOADS REVISt0N 4- FEBRUARY 198t - l f
t@EE
\h
., r *w '
g g' # \
'l fh
\ .-
~
\q,%,# ,
/
Sy$[1$0" ?> ,'.
' yyp, fa(y*;W;E
/s/
,s' i I - '
s I x ,
~
/
x- .
-' 1 x ,
t g4=
, /
/
/
lE (Fl4.1 5) oS65 geO' O 6 \0 , s * ,.a ' (*
=== -
I
j i#cotoo" 8 v f--MATCH LINE-rio. i-i v
,,y n v -
y 1, s . . . E _-- SUPPRESSION POOL CLEANUP / g
,. ~-{- PUMPDOWN PUMP SUCTION I
h
,- k
~
'~ -
,e Tar *W
\ j
/ \ ,
\
\ o.' , .n .
Si' s /
~ ~ ~ ~ . __ _
'x / '
\
I s
/
al. Pn con M ,_g g i, N. / N L- 's KEY PLAN SUPPRESSION POOL "T LEVEL INSTR. Tl APERTUllE "ARD Also AvahWe On Aperture Gard l 1-2 8612080645 -09_ F I G. l SUPPRESSION CHAMBER PIPING COWOSITE- ! PL AN EL. 20'-0" TO 40'-0", SOUTH SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS
-vi. ..c. . . ..: - - -
5 180*$8W G GE l+
-=
DRYWELL EQUIPMENT DRAINS , 55 3 - . 't
, .. - ~
gg ,. . . J . . . . SUP PRESS ON s _ ,.. ... h.. ,t, M d b.. . . - WFL EL6' ESS HATCH N s ,/ s TMAX Q s 4 f-ELECT PENETRA*lONS k N l i . 24*DCMNCOVERS (TYP) PLATFORM
/
EL 47'-if \ 24"VACUU , BREAKERS 1... 6'SUPP CHAMBER SPRAY NO
,-f,f
,5 lta f y'*:
DRECilON OF SP Y k CONTAINMENT - SPRAYw,e m [# e' *s's woe ,y / . l
' p W FL EL 62! 8'
...... 2L _..
n . n m ..n .... 1m_ ' a 270'0E I+ (FI G.1- 5 ) O 5 10 3 a e i i 1 a a a a a SCALE-FEET
-HYOROGEN RECOMSINER AND DRYWELL FLOOR SEAL PRESS. 159 0d 00" e 6 2t. PRI. CONTMT f2 83 4 $ [. 3* o je U g
% - b _ _ " Y U ".7-e s
<g~.....b.... e h2 e 7 KEY PLAN
@N.
- ORYWELL FLOOR ORAINS
., 4
@ o', /
\
'y '
y' ,r ' s ,a /- , . 18' PURGE AIR EXHAUST
@ 'g }
G @
- % @ r s ..n N / I
' \\ ,
} CONTAINMENT SPRAY k
\ _. SUPPRESSION CHAMBER
'\,
93 N > g AI aW ' in rh ,, A /% ,,
,% ,,2 j 2 a70.S aw APERTURE I-MATCH UNE - trio. -4) trio.i s) 80* 0 E ' 4]Q Also AvailaMe On Aperture Ontd 8 612 0 8 0 6 4 5 - op FIG I-3 SUPPRESSION CHAMBER PIPING COMPOSITE-PLAN EL 40'-0" TO 63'-O", NORTH SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 3 NOVEMBEf11978, ;_
1 1 1 1 g y -- y v- y " o.m -ty s
& $w.u - ,,y
'N
~. .$!
.= L.\=$
's-
.m h., ~ .
h .3 e e w Q e e e 'O. e . 4-
@-- e e e
r: ..
. = @
. E. ...... . - ,,
,/ ,
--.. .b l , .e ~1?
*a a a e
- 68* PURSE AHI SUPPLY 9.dL MYOROSEN .
==
1 ((FIS. I
? .t. .?
SCA LE - FEET l l
e, w
".*oYO. 50 arn t # vac e entanEns
(~ MATCH LINE- rse e-a L g u[ / ** '? lj, P V P ~ V W I
,;k i . .,.
- , ~
E'.'E ii-i ' gyg 6= # ' 4 +,/ U'Elt-
. , /
'L. muan connrr
- I g r- ,,
iM'. R ,' . 's KEY PLAN _ n n .y. .- N. _ k s \
',- b \'
.. g.. . g. -
N ..a i,..... i
- :==.e s .
D TI APERTURE CARD 1 ' Also AvdasbAs On Aportare Omrd 8612080645 FIG.1-4 SUPPRESSION CHAMBER PlPING COMPOSITE PLAN EL 40'-G" TO 63'-O" SOUTH SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 3-NOVEMBER 1978 __
M
~
, , _* y* .-
w' a. I
; GI. 2 Y, C
, s.
TM M !
' Y,' W 8ER ._
/ ruxa a ses,
. s,
/ semm-i summ umsum- <
~~ -
D 7, ,,* ~. ,.
.s m o e a y
' i. .a.-. (
,e .9, _
- h. --
M .
;r e o?. 9...
?. <
am a *e 7", 2' _._., . a ln
; *s NYCROGEN rec 00sAHfER e
l - g ,
,n wtw : i
.. . _iv , i Q NQ* 6 0
!. ! I
. i N r i i 1 8
4 I l g / l I 8 i
' l i
l i 4
, . , i .
l
.(.r----------
' =
If .
'er l . %; i 1 t;
l ; hs ,
; e 8
e r l
! l m -e i e oown consER mRAcino EL tf-f ' 1 i-8 t
yet nJf-v
?-"" __-""Ei i h"#" * "
L
- L R Mro' ..._._ _._.l ,
l r s l ,- 4 l, i -;_. 1 s
, e P---t;-
l
; e
} :
--\ -- u
! - - -- eme.'
g l l l DowiscoutRS temSTAAINE i i O ' I \ _ .!
~ r.e,n m .
/-
7v9 . . i l e / nwR a r-e, .e ., : l_ l ! ;d.t.; . /
,,,.,g,.-*.
pv v.*/,' ,, y ya # l
, q RMR PUMP SUCTION RMR PtMD SUL
& STRMNER & STREM e s==== unew
r
._ .__ " ,. I * .
CDL-6
-q y,4rj-
.** . MACTOR PEDESTAL
* ~a FLOOft SEAL 4 ,/~SMEAA flesse lTl [7 t E
ETr; 4 l I [ e l4 s: e. -
/.n a wa .
i , _ _ _ _ _ _ _ _ _ _ "q. s _ {.; tjg : r" ji.4,g a -
. ,7
'a /. . ._T_ 2, _ l l _L _
-A>j $
o, u< -!. i 1 :
'T-"-
3 ra,oa orcM I e
- Martpa4i0*1 , y_l ..) J. '
, i
.d>, - [ moi ndO' u.
-l l
- .--j [_ 1, _ _ _ _ __ __ _ _2i i amam\.. 3' c.;, ..d l f .n
,-----c ,
l w.m u i f ;r . - -- _ i
.. ..--c ,
- - 1
- i
- g__,
- i 7
.. . p, a a i L Q & l 1 ;
i__ _
/ou- _ _
TI l n s .,,RA,,A APERTURE d , CARD
~ au; ,
. . ' :-s.y {. . 5.S
.-fr
.p. :. & AvnMable On e "it]
A-m_,-- Gard y" %*?fs"'E' gr,^lT=,, =7=' E00 SECTION l-1 i 8612080645 - O S l FIG. I-5 { SUPPRESSION CHAMBER ; PIPING COMPOSITE -SECTION 1-1 ; SHOREHAM NUCLEAR POWER STATION-UNIT I l PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS I f REVISION 5-DECEMBER 1988 _ l
d & CA-16 , $ , pp
,. 9
* % =.
# , m FLOOst EL 6f-0*
, ) l, l l t t l
* , .: t" **""~
l ._ ., e,
~ . .
- f. j k 6 c--Wp q
r2, _ __ ,
.! i, j _ . _ _ _ _
l n' I Jll~ '
'l' l ,
il I i 1 R R# Di $ PRAY l l ,ie,g g y g g IN I ,1 Y~'--m. I I
'l l
I a .,& _1 w
..p '.-~(id '
t
~
d I g g QAy.7 g 'h - -- _ _ i 8 y q- - - I I J& -.
- 8 RAFF g" l l 1.l li di e IL 30'9 N "
l r
'l
'e '
g I l I FLOOR EL 4(N)*3 % i x - i, - 1
..' 'n a musu mm. ___
.l,l a r i
si l i 1
'l g l 1 g l l == C
- s x , e e
- g. .;.4 f 1
-)
i,
.w- .-w{=..
- -- . ..+ -m_4 --- . .
m t -
<R oiscumt FOR 1t31=# -~1 -l .
Ltt , , , . N T' E, ,-!' ; i s conc senAv puup suctiose Ase sinAmEn-fl It [ ' r ll -
--'N ~ -
4,' T
++ + -f.
^
g g
, jd4,$- 24* DOMsCuadE R s l l
. I I O YP) .
FLIKR ft 9 O*-
', i a %* CONC (OL e l l %,.,., ( (TyP) I , ,
t Q e s f*, . 'a' y r-O ~
- I * -; .*
. ~.s[ &* ' *. c.
AND STRAINER T-QutteCHERS {TYP) SEC' -~
- g. ' ,:,
co'-5
%pq ,
l r 1 di Tr , T F TiTJT . I n u l
. y o' "' . \ ,, .,
,. :; e. --
L' ^* ( N Y[*9T j~gg1 y
=+
-- '*^""
1-lp' L i i l l al .
,1 L
fa a a l i , . I I l
. I I l -
' $%ESf,$"s "
o
- i' , .- ,
i, l l l ~.r._ y. = we
-rn .n
,. x.
a _Lyu - -
" gy ll I
,I t,,.rf,, t av i ,
-x- ;.,; ;
,2 .; \\ ! q li D=
_;g3u '"
... -,, o- m1, N e. pg 't&==
, l
@i23 -M'_- w tnnH's+ - +W - -. .
a are N __ _ 4, .. ._ _. _ l 4 l- - N .s-- . twt n ess
, i 9'
g (u ,. e o' "'-'-
.. . ,F1 (4 o ; + l y? t APEll,pg 4'.H' s .: : w ar l
j.-l' i { e l I
- - -m..
CARD n p 4 >
,,_m -
ru Anosucr,a ssRA ntn g
.\
i) ~3 D -% Also Avahhle On
\'\ 'CA/
/ ,,,,,,,,,,uc,.
ll O'."O;L '""""'*" Aperture Card \ ' . :! av naa== . .; . ::.'./J ,
, , .. x
, r u= n **
o a, : , s. a l e. l w: 'gp 73'J?,".*tj L'"'"" "JJa"i# "'"" nu"0%$ I *~
- j.RPV spanan
'l0N 2-2 8 612 0 8 0 6 4 5 -()(p FIG. l-6 SUPPRESSION CHAMBER PIPING COMPOSITE- SECTION 2-2 SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 5-DECEMBER 1981 ,
e,# am..P Ti '. i-.i
- m. . .. A . -. .w 3 -
assar? Twassa
*ltt l WWimQ
". CawD ,
" L ou nn TABLE 6 op av ce avto
.,., (-
segmen OP Ft AAort M. ? 71 ossa oss.
& AWie b '
. # ' '- - d '4 _
mc
;wM z . t ,o cua as m0
- 43) OBJC 093F A p re Card ans g
, ,. r pp43 SFi 09Je O9he , pop e A -' 48 6S 1 093J OSSE 3ppee nses. cnnaaste .. g e '
" ( / .Caust, Sei otat 093ne ;* PtK ' '
Emv J - Pt strope it se'-s'triti .-'. w ** ido ce so ' een 3.* VACWWW SAE AAER(f,M 8
- a c g
OW g P-34
\
1 % D a D P-SF, s s %
, ... oPa ,
4 CEL4 . v h 0
- d' P2 su
,I V # OP-ir ., c,t F e' / $ r _. .
[ g 'v j DP 21 pg 7 1 R? s R ui.A**'g 40 su , /, OP-57 44 - g s I ! p.3 A T' e' . h .*,. * 'N e' j ,'4-g g j N ' ' = j,'" $ u ""
&/ ,s '<. g8 l ; , :
- . \
y. g's s' 4. ,t a. 3p g , OMJ [ -tW - -
) , - , 9 0
. I'* I T 'a NF P S - --
M 2 II
' 4
$6t) w +.
- t *, ' \ ,/ / I ,'<
t 'y
. . OP-S 3 *0P SI , **
, ' . ' pgl . -
- M # .A g\ cf s\% - j
* , f. . s g
\
Nu
,')
. IFD, UP4 no g op -,,..,
% W asa s
K . *e
.h. N, P-t eggg ,- 9 j, ting (a ..r. ., s %
ptfl CIWfl4ND6 CF M * \ , 8 U % FIF PpgG F ig e- gg % Agcig /
. APCE ,gg j (IM CfL40 U
'" EL \
SM E
,P..- .. ,h we 65 4 - .
s e - i f*f w \
, _ 4 ,. . . .o, E P-50 OP 47
,,.,, u" **
M' Scot 73 Piff VDfM(FY4 ;p . f, f
- . 41!.!..*a*.'.WP,.,.
.t . 1 <
- ',f -
icn,'s ou " **** trts m 3 . . 6 a un C:xbg7 . c~'
,~
'IV .
c nnt n.r,w ns ,
- p. .
navit tnaphs -1A00te Op .,3
' 4
- I weptuoi ca.anawa acctst
.P.,.-
OP ry sgd **' est sf t oucians H L'an,r a s w r :. : , - z..-? . .\ ~ y TT M;;I.!.$$ .M. g. >, 8.a e"m'ma.JI#ca"t u ,
.. d onew 9
.C A;B = E t r' y.
ePL crr,A.N. EL. m 8.v'...O' c
?.8 .i? .
I J .cate ager = r I
_s een o.aoean" 'k Ot/00' WY SEF) W AT'L WOTES'
- g'*"." 4 '""'**-
p{_ s]* i* Att statt iaAvt ren BAACIWG ASSEW8Litl&
'T- -
LesetDutNTS fo SE TAGLE W M *N o ( j iTre M AJpHERf To Of Cutoair ; f to io cwn stu. a _( to,c^',ma am""'"' / Pi,Ngry3*oa wate ron g'coTtra pia
! ASThe b (TaACEA518.!TY aEQ'C8
'a . j,- or, m so g* m au n m. p.
so m Tvessg' T \ a
\
M*T*86 N W - mi s or arKHec Loop Pw's V 'b . SIN 6LE-J S Aerv { gnonyg rc.ueA (tvP>p gg io, m.asTE1,o,. .,o, tan.64 .c ore
- m.n% .
,* , w
-. +
,g
,9firutT Ense6mAER. 3/ EL. 53 0' k ; .#
a{e icLE a a{$ pin -ry, ,on w 1 'A
- if .
a rism onnA eett in 4 otL =a aT fck t i a. i DMS wLo To'E*aF 7 our no ran twr c.ci n.t.=.a a t seaus
,; s i 20**a5 "
{t's h P.36 g'aggpgu g set 7ASLE t*FOR ( PIM '
, , 3,g op,,g , to ein ta ocnes enAit ,i,c g. ,,(..t tr.s e-se, orsmacswcugusta AsusII saios on n 3, DETAIL A #
De e-to 33
% CP-44 o e la la 4 f mitty' 3C>n " e. .(
et ALE - sments I, 4 g M ' -
. *, ' ' ' .1L*. ;T0ft om e' ' N" "
(; .
,, ,,, ,,3,-
f , i /
\
rM $ - ** y 4.jf Dtt 8, g e ,i e,y
, _ g
! =
'n ,,
.t' T, <
~
h~ ,.,,
.3 % g.;n.M g'_ .'.. . Y.
,, F/ a Im' . S4' EXISTING \
\ i / .' . e 'q' )
- EMBECM ENT
'[
I .* ;E OCT AI C. 3 . =* V .
., i .MTAR. J . ' "' .A WACTOR<,,,_ ]" ,
ON M8 N g L ' e
, ,,,, g}pg DI A HOLES .
*q l i ',#,) mg l'/g DI A PIN (TYP) - %
# f -DETA= F trie. e-us
't
'~ rs a
,f b' r* (l EXISTING coLLan
, p l 6' #.' ' P4' O(A A o f g .y f, eg b e, ECCWN- ,
COMER
'.,' , TO ( REACTCR
~
l '** ,.%
/ < -o p.te
, ,. t .
VIEW 1-1 4 EL ss'.o*
* - g/'
' r' W'o%f"
%'J".
m nm actow M, E n a u D.ETAll.. B ..
,,m
'C*bt*8'C"Es t scAta ss ie'en g (Tm
,,.' h - a mm p ase 3Ergumc1 Oss Pro i.e a estur j t srrr as sen sa:t oc Fan ALL troux csc j
,- g. 1,,
k,e j z rm m . cou,e o..rt. m e a..e.a.a.c ..a.wco= = ouAt .
/, , l~ W. NUCLEAR SAFETY RELATED / vo ,uhTve" L0in(4E0 F4)he tOvt .On I
QA CAT. I (tvP 90 Tee gotg) i preeseau m Wis.G'= ,,i. ...., oo a e12 o 8 o e 4 5 -07 gcagg:ne,4 accsst R LL VENT PIPING-SHEET 1
'k - *
~
SHOREHAM NUCLEAR POWER STATION-UNIT I d PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISt0N 4 -FEBRUARY 1991 __
e.n== e o sL1 w6ans Tat m C(F10.1 104 ,
^
f[ CAP & wtLD10 ameer 34' w em tp y _ _v , , -= . .r ~ ~ or.> . l / M7" [gt,33g [pg -ht h5 cRituTATeou FIE 4-7 g,g 9 9 f 9 '. scats-mcasts h l < 8888 " ss#smaa-afljua ? .gr 1] ,r---- [t n u. :- _ 5_ * 't*{ \../ > g s g;Re,7,g} -(;fe'a CA. ' y. y-oe u.r.u. ii , n- '4 's 1 2-2 * ' 1
? ? 9 9 .*
' - ~-
ANftCILSDJD5 ~,,' g,3,3, ggh.,_-[rate . iy w Nset - vasartuo A v mssLa uner, tie ron stics scoocou stRcasts rom sn m, scu:cces g. j
^*] s i es3 e R O aff5fflWCtI080s i -
T NOTES: STtf0 s#AC11st CAN DE AAR ANGEO 70 suit k 'ETAIL D K RtinfcRCiNG SARS esIIL SPACm6 6*VERTICALf. wonizonTAt unscor,aoist4=a to s, or stuo.ij. p o. u
) ocste mont (Ple. 0 * ? )'
og ALL PLATE isAtt Fom dAAcras6 A a
**
- E 8481 $8LKt "4 58d EWOEDues75 TO BE AbtfIII sA34 5 a m .n.. < -. a ,, 1 =
e
,te
. .. J . u...,<s.3"u3 am
..m
{ f .* '. ) It. westu 'wSJEKTset I
~ W.'.. co. w. n
~~p, C *. . :- u:.':
.' ^
s.7 n',,Les. P P
- p,t,4lt.h-i- 2 6,- ea f P * ;w\ -
- vs ru J.* _-'-
-. A ,.
s, e
- 3. 3-. , ,L l ,a m.r g,-9-- -. ---
n
,e , ,
eaaanmun
* . [ s, ' #
R*Pt% 044JA47 / r A
- a .
Stic apaca.noi ah
- g irr.x Au.ctncasJ'y1g -- -
<rst* oI.ao.4, I '
f,'2'Mh* k
/t \ -
, ' (y./ I '-
sa.ds,aos.ac : css .,._;
,# . -.:a.idar us a a l N b; es4straossoss.
e- -
} N. , ,,,. a a, . ,o, . w . 3
/ I N, x ,
***'^1***'"
' say sLeson. s '\ .. j. -
+
,Y sto ' '\'
e
. /
i=.e**ao,=>h 8d
; ?j- tT%- .
, s. /s %g .
q , i . 2
' p, j
,iy,aQ.
'h d [*' , - '-
. ird .. m a ,o- l Qo
\cui A.Av,7u>!runon s PARTIAL PLAN
, Eg ag'-{ ,,
- s. ....
.. .m - - '
r 4"
. /~~ "
L r"
.'lv0ky.'gRg\_ . h N
~
d - A u m___y_ . . . _ _ . o,o,,,,co . ,,,,, ev
= n *f 4, * . =%% ,
- e. u.
*=a
.s ,
.s.g h y. ..
re4 m t 7 a u ..-
. l..
4- ~% r Tw w,;:m. ! ns
*'" ~ ~ "
V ,g .:Wa%a , mn r fg n a' y y h._ a
-coeca APERTURE 4/sh],i i
w 'ne r r' * -
/
7 1 / CARD s n
/A .'e - s sy, - ,, S - -
- 9;.
'Ia .;.l. .q',
'g' a-Also Availalile 0,
. M Na uma,wt
..,,i. = ,o,,1 ete a p , fe eeuwe tsu
.m.7. .
e Wu, 7 vie. mi i-n
- . ~ . .
son: Pertare Omrd
'? '9 a,e .c.[s. r s a ,' s scett as serin
- 2. veens me aereen.w puus..uus me ,iriseos com cos .,2 ,+e s M ATL NOTE SI'T8 L** PLArg fannassa lo E Q.E A80 m apua to E CL act ALL PLatt teATERIAL asscustirs ,CR SA,f vo st asuem sas LE.4 sass om t.nro isttss oneemt secmo-Auo es to coisromu to ntouimtestnis or s.c.2 & ontrup.
4 ALL bescapL IW1 select ptst genteen Ass'vs,s.awwfs, I
. annu
- puert Ass.u.,. tic A.m As spectraeci.- - u t .w . .,,,
Ape MftetIES p.perfG
,h% N F angsgg g, . .asu.c.som.a w <,
- c. an I Suess 8.glutAt.ge ase M su i vid..t. ee.'-*p... .. . > sv.ues.t L.our,o .a.ser
.m uns m.h-,
A1 ; 6s a e6m3 7 _sese.t case,oir,,, a,tassia,,a,n,s
, ,asentvers snowuo son i
Y '
....,r,-
-.m- .x ... .,,o,. .,o,x .-
-7,
,, . ,r . u
, I
,ge,f,.,..aus sc,a r.t m ei.ett ,i,t SUPro.t (TVp) , N ,** #g[ ,, r
. . ...... <=o.. -
I (------7 it no.ses sua si.mmeson an.L pooseo.veres aseman. w as w prosseu asysrana ww I r = = arr'o-rese.e : ta Au. rato anos as e-ace unto peccranc. l -- - - - - - - l / l ,, iannaN004 oBGrED EffM Af8 (9) AAE CftflCA. m00 i m-m g,D, P- - - - -
- '> em w Ds.er as cmmao
*S.Tolfleurt onescarr enn m.(otos a,neomL, Mit omdi c'# ,g
, om .
'l'AE*r'f' 3........ as.esta e em Seelv.ta Issue su. suus.a. aESL)
- - - - - - - ...t..,-.i..,.
y k ,%
-[g,p4 . m .i .
i
... no. i.,
I -
\ M'llljga,'y o' '
L0000C004 quetM loc 4flose E';[ay ssee
$ 3,,,asspies.
i,3e$;i.l3 e 3 pie. 3.sg g NJCLEMI GAFETY REl.ATED m = *4' 2
- v. QA CAT, I S
'si! Irio.1 - 8 8612080645 -08 DRYWELL VENT PIPING SHEET 2 SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LDCA l. DADS REvistoN 5-DECEMBER 1981 _
=
nt S Fee. 0-0 6 et SAtt R isft Fil(a ior (88 A44% ( DOurNCO9 DER MPE 49 , SAFFLE It.
-% 'Q ,4-s (DeAucLE.t FCet II'Ch4 Af(.6 EOR
-1* N toL115tst tAPPLa ste nantnso "A*(rle.l-73 U
/ 'I
,f Y'f- r - -
\
a-iE DAMQLES FVft if OlA ANCHO4 saiis ron w,ica errts. = 6 crie. :-n 2 sg e -
'. OCTAL y G re -- ,_, . .pb* + -
4.',
,1 T .s ,4 .
,,, = -
sy' - f' [0
/ t [/
.=D.E,.tA,tk.N. .m ,
t.i. ei ... . . g se i.ets N -
,,y xs f
N
\
N
# s#
W F. h * '-: ;pj s U.?I.5E$ /
%s .s~. k. .... ise.p /*
r... Pl.AN
@Aw. PLATQ
? '? '?
.eata-i.cas.
I f 5 ( es %* a,
't i =
S' " EP II*gn& AbfM A-30f i 4.00LT9=Tt 4 anchoa, moi.ro sanciesi ssueg (
.v , or ' c-= 2. my] /
. imp se 'TC h I. .
,-t - a
-.a, i [ ,f ,
mass ra a#g k Q.' 7 in,arp snue
' .y w i
,' w-i
; ,*"8f" h., ----
pj 4 cm tv w 18-80 0 asau=o omw co.am
,y 88
) ,
W .sArros rea Jo (estsecucus) ite sine.ei.s a
'sart- **.'
a.
** * + um.t. as
***** **^N' **'wt?
wrecaL wt; as ymp_ g
-muses a
\ ,
N'iy cr.z.e-. _7 -- . ._ _ _ _ _ g
. . i.
N 9 9(Fie I-e) ,
? I i ? '?
S CALE-INCME S MAI= Y tattPLATL asceon toiTS 8 T '8 **
.C A L E -IIIC9 4 9 m
M
,W
TI y APERTURE CARD Ah Available 05
. saa Aperture e,-a
[, - %b g ro. c.,, ,o m ui,or. EE TAKE Cf FIO. l*to Y a% e h NY-eTll . . jossa se.
~ n T) e -
Y ,* ,.-',,, Y L*KtW,'EP,.\,ge
%'jE3 '" Asstm saas o erre sure.ar out.x DETAIL D,(ne.Y.-7)
SC ALE =lNCNCS wns. :-s a PLAN g es. (eg (narri.E a poemene nome.e rocos etAe> . massa M*T3 Kr Taett C (no.1-10) ,i
'a e 9'.=
$ C A LE,=lN CM t 3
,g, S p /g[gg=
hi h 1r y l l 3 f)
.., d- '"= Fa= = '* %K+ 'a' owmoLTs-Asm Aars4 ret i uAnnow eArrtr 'ts E SOLTE*.Astu A3ts TYPE I TYPICAL DAFFLt L'8 y ,.,
. .l QED A MTLD gm l I h3 p namz
- atrie. :-7: DETAIL P (Fle. :-11) f ffh i ~
'/
4 W , n 9CALE=ements o s ,. r -
,_ % - __m-y. - - s ,,,,.
l ;j l He.=m.w.vi-~~ ---- e,=e .s ; j Hw: me.. ..., ... ... L. P '**** * , s-gj
-s j %'
t-- NOTO j gj j g L acaut as motto gj j g 2.um.nnt woTes 4 nur.ousst. Fie. -e N ,' N Nh t d l , nooa a..re O'Yk
-;j,sg if,fg ,
NUCt. EAR SVETY REl.ATED Cm CAT. I l
) .
- C 7AI)"
?
nL-'
- t t i*
8612080645-69 s eA 6: =im C a s. pgg, j . 9 DRYWELL VENT PIPING-SHEET 3 i SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND !.OCA 1. DADS REVISION 1-APRIL i 1977
l TABLE A TABLE B oce= consta 70 oAAct arm conwetpi AssesseL7 7tPes, cousicourn Pen LotA7sosi MEeuri DoW9- Ass'? Do W80- A s s*Y DO W M- Ass 7 boWN- A s s*7 A ss*7 IdQ oF ionoF Ps Peil LoCAI CoasER TYPE CoWA 77PE CoutR TvPE ConstR tvPE T7PE AssVi CoistaC- (ANsLE e op- inatt e op- innta e DP- 7aalJE e DP- 7AeLE e thele A TIONs i 8 as 4 45 21 s7 as e 4 a 13(-0* as4* o' a i 24 sa es is as as a a 7 d-o' 44'-ad soi-as' sai-so a e as 13 47 se 69 at a e 7 6-6 44'-3o 7f-isi sai-ao 4 e as e se es 70 as 4 a 7 Gd 44*-4 77' io' sai ad s a 27 s 4s is 7e aa s a 7 d-o' 44-3o 7#is' is s'-a s e a as 10 50 is 7a as es a 7 d-c' ad-is' sf-a saaf ed 7 a as e se a7 73 a2 7 a e o'd ac*4 s' 7A ss' ias'-e s e 4 ao as sa 24 74 as e a a 7t-is' a ve"-ad 3a4*-36 e s se sa sa is 7s ma e 4 e d-d ef a# iat ad ieSd i to a sa sa 64 is 7s as to a e sf ad of ad an*ad =aet t si a as is ss to 77 aa es a s af sa et ad
- ed o' z #-D1 aa a as is ss is to as sa a 5 as-W ist-4 ied-,/ 27s'-so' is 4 -ss is 57 ao 7s at is a 4 d-o' of-ad nat-ad iso'-d 1 e4 s as is ss as eo as is a 4 se*ad sa"-so' ars5' ris*-d is e ar i7 se as et aa is sa a c'd ei-ad ind' ad iso"d j is 7 se 24 so is sa as is a 4 44-4s' as'-a# ars' ao' sis ~ is i er s as to si an es aa at a 4 Go' 44-4s' ei-so" are'-Es as 7 4o is sa as as as se a e &d ei so' pre'sd is' 'iij is e as is sa aa es aa is a e a f-is' satad its'-so' ass'-43 ac s da es 64 as es as ao a 4 af-is' e r-30 tre W W-Ed ai so es ao ss aa or aa as a 4 setas as*-ad ars' so' 32f-c f aa e 44 as as as se as an is 4 st-ad as*-so' are* ao' szAac'
~
as is a o -o' astad e7s'-ad as a e So' 4 4'-a s' es'-a6 i34ao' O as a o d-d at-ad ea*-ad sit-ed a g*p HOLE 14 3 RACE TO SE
, f LoCATEo AND DRILLED IN FIELO '
l f &, \
- FoR (PtN 10 PIN l g, thsTM SEE 7ABLE C
\
*p,t a{ 9 eets E Pim Q 7t=
',- h,p MW D*' M a,s..
,$ $ s ,1.3 -
{ ' '* ,' _ m ,se..a x,y. pmpi ( DETAIL E d .X. Z_ M 1
- 1 1 1 '* ^ ** (. ,y seasa-asen s s -- -
(FIG 4-7) a> ' < ar'" DETA.IL ar
,ii, e s e F ,, k RJ,.!,,4 t F7
+
darmie (n.. i.n + .4 +
}'i7.sT ,
[,g7,, gr,yy 4 WWs g pecyg 3 (FIG.1-7) (FIO.1-7) s.oEl M DETAL J E 4 e[DETAL H
?
seats -sacuse 8? ? ' ,8 . *l' 4 ocata -encuss (F t G- l- 7 ) - (F I S. I-7)
====--
l l w t Tl APERTURE TABLE C CARD sass Ts I smactus see 10 por neueEn 3,,, y LENisM NO RE3ER eEIR
) d inness nagR nusoEnLusETM, RGED Ospa. M (ydde On m arm eErrE 33 1o aus A I A $50000 te 4-6fA(FU F i se6-o' 22< 30' asa'-4 s' a s480eo se s' s* a t(e:S _
g reo-o' a2936 asa'-4 t' @M-fi- f-c 4 --f - _ isfe:A h soo".o* a&f-es' asa*-ai o a4 xe as i-e{ pria&m i !
";I g7 cav-o' at(is' ase'.a s' E af ts 4 a-a' EFIE be I.M. --! tt 6/-o' 7
iw-a- a v-is' asa' es' -- r' af as no s'-a- o ., iso'o' 2 3/-a s' ass'4 sof-o' 8 Cosoec e 4*H A . n Tsen4x } 6 fief A .. p 1-'[ asf-30' a rG*-36 J s'esoea 4 s-a' '. b " t. E trine .a sted' D 3 0s
-. 4 I ,(' , ,.I' ucw t nexu1 , og a -
u<.3e 4:e o J- S
. S.g o . . - r
. i.e.4x t . .4 _. ..
one . ..-
- o .
. N ,
11 9 . ocw r. o og % sur nacw a a .+ .
<-ir
. ;o,uAs,Miggi,y \t ,c oi.. g T,,,
s a4 xe e- a-.y a sa-n3-s= a oa art sa-3n-c= s on art a
- - - - eT ina4xil 4 e's f u s'4aoec so~ 49E a suo SCALE v age i. reg,,
w eye i. r..g,, DETAll V l x aig me a < >T,arien trios i r a i-s) leo'o 22s'3o 376' 3o' Sis'l e A Coastspe4 TON TOh5JE PLATE ! DETAIL 0 ren omsa op-if is arauinto ron emners 327-36 SETWEEN ANGLES 37155510 r0R u tistERS siaT PIPE SPLICEWILL SE FIELO WELDED. DTMER MEMBERS AS SPECIFIEO sofES: bM AIN ST R.u. PIPihG L SCALE A6 NOTED. k 2. CDEWL scrES 20 SE79ENCE Das F10.1-8
% s r eG.0 0'A 1 ALL G'O 004 to Pipe AfD ALL TE SECT 10N5 A9if 3 4%
GA*M S UCE to'5CM ,lb _.C.bfL -4'- MTacwMeni etArt -q j.E (W3SO4F0 OR MATL , 5. ,,, FLut0 NEAD FORGius 's WITM 5YMeh=50000 PSI) -4 4 I b EL Ese" wtgD'nG u,M!s% 08 strigV'- y+* f a mo. m *. ..- : . , m l - 7, . ggpQfar
./,*i.y d re 'p' At ,,,
b .if , . j'*
*}: e . +
ee n s
. oe: ?.}l:
(TYP)
~
""' '
- J' .{ .. , ,
, gg( g na*"$ h, ,'.Y. ' ry . , !YJ'Po4S ? d "o ? h
.. . ~ m_ .
r ' l. - ful.EM WETY RELATED k"',',",u"n ; ;.M',,^J- g T bs#3r
" ^ * *'
f (SA.333-GR_6 CR MATL I E*DIA a TMs WITM SYunie=30Doo PSI) I ** N (SA-333-G316 04 Att elTN SYMih* 30000 P5i) M0 SCALE "m"u I DETAll W l't '#.nfrial'"^* FIG. 1-1O 8 612 0 8 0 6 4 5-ld km , m . ._ . . T . .-. i (m'
***p' DRYWELL VENT PIPING - SHEET 4 SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRVAND LOCA LDADS
,141'nuttV6&'t8'faifo h Y REVISION 1- APRIL 1977 - _ _ _
)
.-- l l
E ll NE) ire-etsP-rosso 3 ex at =* j
%Eg [ ELECTIMCAL PttETRETOPA 32*R O* 6 \ be*EP 54 N Th*L - sas~
ELfCf RICAL PENETIWL?Os k I X%w'** ,\ \g, , gef / \1 Mtas% ~ I* l r a
,/ ,e '
$$gasm ,,
, , i e , , :.3,.. .y 3. ' 1.*.:. -
gg# g
. sysg s
,. y,.
a l ru== -
'm%'o**' ,-
. O* g .
, J'- ,.-
r Fw EL SYF* / , at h ~ s-g ll
'554'** * /, ,.
'/ ,,R, @ ;(.
par 29 ELM'+ s Es2P&5t **0F30P3 % [g / V N. v PSR 1 EL r-11%4
.. as. . y -aw- '
('tNd # - h - ,
/ ~
=
' -34' 147
" a* -
.[ ,, ' rd '
g i
% L's ,
148EL -11 %\ , .'. J., % POR 18e - -
} h- gs s
/ PGR 181 EL3('-S*
sx.) 1saeLe'-siv /c s
. x ,
- 1. . . - ,
f og f
.g, v .e (' \L M' a a kI .
. . e _ _
Q V \ k \ l ' t fW 5 6 y ee-
. * * *l'\
\
1 g,st' o* . %
$ 4( j "'
M. . .
,4, s' ) -
jp O P
.., a R in st. r-e A'
O @o A ~...- Pg~ b, '
- Pen 1M ELae'#
PeR 1se ste-tiV4a .
, g
%(, -
s .pg[ // ,, {
', h~- -
gogos - ,f , . n'is- -
- ~
Ot O O. , te-PMP ITS EL54' 11%
- o ,
O ' PSR 124 EL63*-4* 2 . PsR tas rLae-e- * , W PsR ice ele-si'.g. L ' s'
'",g-t g ( Q "* . E;rw' g .
s.t . , -
- i. *c.
/ u.nu
$ * **,8 POR 130 ELS
+* PER 111 EL4 PM 123ELW PsR 113 EL57'-0" , &
M k POR PSR 114 194 EL&7'-O"% EL56'-2 2 ! \ PSR 118 EL34*-S" POR 117 ELF-11%[ Q g* g asr.nas-n -
. ?'*i i . '? ?
S C alt- P E ET
W g EL $F7
~ . G H-3 <
. ,, . g* R
- Fw EL KB' E' e*x tfnED ELh' N
SE AC104 it. . , tab-IF5LP43&&S 3M 1rx ,,,,,,IL, . ,. ("E2 i
e f am N
- g*'
Pen ina stw-av, = PsR 183 EL44'.0" par 9 84 tLF-11%4 8 1 ' l
.? , * ';, ..j- oErAit .
, ..> e . ,s sosts -ince ss EL SF
- 3/4 pgp.c y.hh3 3 wE
- YNa fast Ma
.. ue.o,4 ..e, .. oe
,. -MJa""'
. 77 .i .. m m iu.
" .,Q.h(7* 3 4 36- = 4Sb ,
' Peep 184 tL47*=g" 435 4W PER 108 ILAS' aV3
- gg qqg, PSR 104 EL34' 0" Pam 104 ELP-1IV4
- 44 444-d "O 446 Q- YO. 24 -
M , 444- 240-
,-. 442r 247-d .
E ' - r*** - Pen 141 EL57'=@* *** h #* 2lg g *-
- PM 143 EL84'.3%g *
].
{ p* Pan 148 ELaa'-o=
- a. 244 U if -2 46= I j}
{y 27treest) Pen i44 tur-117. =
---- m T.:
kgggn3 g ";.% ors - rs=> tat c.o m . ,,cr .,r.....
, 47 # I 2. M.L ELBC'eS TO GE LONS IMOlv8.
y, ,, .= x = == iune. i. ,,on oo.iram er = ==. A 10"IES ELSQg 4. IL DEBIOTSE Tal0Ktf7 LEIC.
**. g 5 0*
ima b-De .
...* ggyea3
+. * .
s % 107 EL.as' S" ") f lPSA 100 EL.84' 5" O ' 'tj';;*,a'
's ,
PSR 111 ELF-11 '/4 TI NUCLEAR SAFETY RELATED QA N. I s 4g f t.P-23&30-3 APERTURE
~
M.m - CARD
'*21/2 '
- p JAse Arallable on Aperture C, FIG. 1-11 8612080645-]
SRV DISCHARGE PIPING IN SUPPRESSION CHAMBER SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 6-DECEIASER 1986
1 SECTION 2 1 () LOADS AND LOAD COMBINATIONS 2.1 LOAD DESCRIPTION - SUPPRESSION POOL HYDRODYNAMIC LOADS AND ]. RELATED EFFECTS 4 This section briefly summarizes the source of hydrodynamic loads and general sequence of events, in addition to the loads described in the Updated Safety Analysis Report (USAR), for safety / relief valve (SRV) actuation, a design basis loss-of-coolant accident (DBA) an intermediate break accident (IBA), and a small break accident (SBA). 2.1.1 Safetv/ Relief Valve Actuation Actuation of SRV's during normal plant operating conditions may ; produce transient loadings on components and structures in the , suppression pool chamber region. Prior to actuation, the SRV discharge line contains atmospheric air and a column of water at the submerged end of the SRV line in the pool. Following SRV actuation, pressure builds up inside the piping as steam compresses the air and forces the water column out of the pipe. When the water is expelled, the air follows the water column into the pool in the form of a high pressure bubble.- Upon entering the pool the bubble expands and accelerates the surrounding pool , water since the ambient pool pressure is lower than the bubble . pressure. The momentum of the pool water then causes the bubble
- to over expand until the bubble pressure eventually- becomes negative with respect to the ambient pool pressure. This negative pressure slows down and finally reverses the motion of the water, leading to the contraction of the bubble. This sequence of bubble expansion and contraction will be repeated.
until the bubble reaches the pool surface, due to the- buoyant
- forces. During the period of the bubble oscillation, pressure l loads will be transmitted throughout the pool, resulting in
- dynamic loads on pool boundaries and submerged structures. The-analysis method used for. computation of the SRV loads of SNPS-1 is described in Section 3. The event-time relationship for SRV discharge is shown on Fig. 2-1.
2.1.2 Steam Quenchina vibrations 1
- . Steam quenching vibration phenomena occur when high pressure, high temperature steam is . continuously discharged at high mass velocity into a water pool which has a significantly high temperature. Test data demonstrate that this does not occur for
- either normal suppression pool temperature or for low steam mass
- velocity conditions. These phenomena are discussed and an analysis of the suppression pool temperature _for-the Shoreham l plant-is presented in proprietary Section 3.4 and Section 10. <
lO 2-1 Revision 4 - February 1981 l 1
----~---,.-.~,...---+--,,,---.----,,nn-~n-,vn-----,-n.,- ---,~,-,-,---,-,---~w,,n . , - - , - . - , - - - -
2.1.3 LOCA Loads f Section 2.2 of the Mark II Containment Dynamic Forcing Functions Information Report (DFFR)(3) discusses postulated accident conditions and typical time histories of the response of the Mark II (MK-2) pressure suppression containment to LOCA's. The report discusses the following accidents:
- 1. Design basis loss-of-coolant accident (DBA)
- 2. Intermediate break accident (IBA), and
- 3. Small break accident (SBA).
2.1.3.1 Desion Basis Accident When a large loss of coolant accident (LOCA) occurs, the mass of steam released into the drywell causes rapid pressurization of the drywell and expulsion of water standing in the downcomer vents. Following vent clearing, air purged from the drywell forms individual bubbles at the downcomer vent exits which rapidly expand as they are charged from the drywell. When the bubbles expand sufficiently, the upward motion of the pool becomes essentially one-dimensional and the bulk pool swell phase begins. The pool surface continues to move upward until gravity and the increasing pressure in the airspace brings the pool surface to rest. Air beneath the pool surface rises through the water slug and communicates with the wetwell air space in a relatively temperate " break through" process and pool fallback begins. At the termination of fallback, the pool is restored to its " pre-swell" condition with most of the drywell air having been transferred to the wetwell airspace. The total duration of vent clearing, bubble formation, bulk pool swell, and fallback is approximately 2 seconds for Shoreham Nuclear Power Station - Unit 1 (SNPS-1). As the purging of the drywell air continues, the vent flow becomes increasingly pure steam. The vent flow may be considered quasi-steady in that the rate of change of mass flow at any point in time is relatively small. For a containment DBA, this phase lasts approximately 25 seconds. Containment peak pressure and maximum vent flow is reached during this phase at approximately 10 seconds after the event begins. Condensation of steam vent flow during the quasi-steady flow phase is continuous, but pressure oscillations have been observed in large scale tests. This phenomena is referred to as condensation oscillations. When the vent mass flux falls below 4-6 lbm/ft2-sec, the condensation becomes intermittent. This phase is referred to as chugging. The DFFR presents the following dynamic loads for a typical MK-2 h plant DBA: 2-2 Revision 5 - December 1981
- 1. pressure load,
- 2. temperature load,
(
- 3. vent clearing loads,
- 4. pool swell loads, I 5. pool fallback loads,
- 6. pool boundary loads during condensation oscillations and chugging, and :
- 7. downcomer loads.
The analytical models used to compute these loads and the plant specific loads are presented in Section 4. The event time relationship for a DBA is shown on Fig. 2-2. No SRV actuation is mechanistically possible during the DBA due to the associated rapid depressurization of the reactor vessel. - 2.1.3.2 Intermediate Break Accident An intermediate break is an accident which results in a suf-ficiently rapid loss of reactor pressure vessel (RPV) fluid such that the high pressure emergency core cooling system (ECCS') cannot maintain reactor water level. A steam or liquid line break of approximately 0.1 ft2 is defined as an intermediate O, break (S). The drywell pressurization rate resulting from an intermediate break is significantly less than that from a DBA. Consequently, the vent-clearing load is less than that resulting from a DBA and there is no significant suppression pool swell. Condensation oscillations and chugging do occur, however. The maximum short term suppression chamber pressure is 26.6 psig, the pressure resulting from drywell air carryover. The associated short term pool temperature is less than 140 F. Since'a high drywell pressure signal scrams the reactor, the sequence of events following a scram eventually closes the main steam isolation valves (MSIV). Reactor pressure response to MSIV , closure will vary depending on the size of the intermediate break. Intermediate breaks at the small end of the spectrum. may initially result in multiple SRV actuations.at set pressure and then automatic depressurization system (ADS) actuation as reactor water level falls. Breaks in this range are.too small to produce steam condensation oscillations, but intermittent condensation. ! (chugging) occurs throughout the transient. Intermediate breaks ! large enough to produce steam condensation oscillations are too l large to result in multiple SRV actuations following MSIV l closure. Here again, ADS is eventually actuated as reactor water
- level falls. ADS actuation occurs past the point in time where 2-3 Revision 6 - December 1986 1 ,- -- . - . . - . . - . . - . - . -
steam condensation oscillations are most severe, although it is possible that some condensation oscillations may persist at that time. 2.1.3.3 Small Break Accident The small break accident is defined as a small leak in the reactor system within the drywell which does not depressurize the reactor by fluid loss. With the reactor and containment operating at normal conditions, a small break will allow discharge of reactor steam to the drywell. The drywell pressure will increase and result in a high drywell pressure signal that scrams the reactor and isolates the containment. The drywell pressure will continue to increase at a rate depending on the size of the postulated steam leak. This pressure increase will depress the water level in the vents until the water is expelled and air and steam start to flow through the vents to the suppression pool. The air flow rate is sufficiently low that pool swell is not encountered. The suppression chamber is gradually pressurized at a rate dependent on the air carryover rate. Eventually, all the air 'will be transferred to the suppression chamber and the suppression chamber pressurization rate will then be controlled by the suppression pool heatup rate. The maximum short term suppression chamber pressure is approximately 26 psig from drywell air carryover, and the associated pool temperature is less than 140oF. In the long term the pool temperature rise may be greater as discussed in Section 10.3. Vent air clearing is followed by chugging for an SBA. During SBA chugging up to 2/3 of the SRV's may actuate on pressure setpoint. Additionally, HPCI failure following a SBA could result in ADS actuation due to loss of reactor water level. 2.2 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA l The suppression pool hydrodynamic loads as well as the mechanistic relationships between SRV and LOCA loads have been described in Section 2.1. These relationships form the basis for MK-II load combinations. l All possible load combinations resulting from this basis are presented in Table 2-1. Following are several important features of these possible load combinations:
- 1. All combinations of SRV and LOCA loads could occur with or without an operating basis earthquake (OBE)/ safe shutdown earthquake (SSE), seismic event.
- 2. Without a LOCA event, any number of SRV's may actuate, as determined by pressure set points.
- 3. With an SBA or an IBA event up to approximately two-thirds of the SRV's (low and intermediate set point O
2-4 Revision 5 - December 1981
groups) could actuate on set point pressure or the automatic depressurization system (ADS) valves could be () As important actuated. as the identification of all possible load , combinations is the appropriate designation of acceptance criteria for structures and components to be evaluated. Once acceptance criteria are associated with possible load combinations it becomes evident that many simplifications to Table 2-1 can be made, while still retaining all controlling load combinations. The resulting design basis load combinations and , acceptance criteria will vary for different types of structures and components, since different design procedures and codes are applicable. These are discussed in detail in the following sections. 2.2.1 Reinforced Concrete Structurgi The load combinations and acceptance criteria for the SNPS-1 reinforced concrete structures, including the basemat, pedestal, and primary containment, are presented in Table 2-2. This is consistent with Table 5-2 of the DFFR, Rev. 2(2). As discussed there, the factored load philosophy of the strength design method is employed for the assessment of MK-II hydrodynamic loads. This is consistent with the original factored load design in accordance with ACI 318-71, " Building Code Requirements for Reinforced Concrete"(7). () Load combinations 1 and 2 (Table 2-2) for normal operation and without thermal effects are based on ACI 318-71, Paragraphs with 9.3.1 and 9.3.7. Load combinations 3 through 7a cover the same combinations of events as ASME Section III, Division 2ca), with SRV actuations also included. With the inclusion of SRV loads, load factors are adjusted to provide consistent safety margins. It is noted that substantial simplification from Table 2-1 has been achieved by two means. First, SBA and IBA effects are grouped together. This is reasonable since their effects- are-generally comparable and both can occur with the same possible SRV actuation cases. Second, the various SRV actuation cases are i not called out separately. For design assessment purposes, the l most severe SRV actuation case possible should be considered. l l For Shoreham, the maximum results from any SRV actuation case, { including an all valve discharge, are conservatively used -in ] combinations with SBA and IBA events as well as in combinations l without LOCA events. Although not mechanistically possible, the effects of a single SRV actuation are also combined with a DBA event. 2.2.2 Steel Structures ! c Load combinations and acceptance criteria for steel structures in the Shoreham plant are presented in Table 2-3 (steel structures 1
)
2-5 Revision 5 - December 1981 I s
- -- ,,,+r -, ..,,,-,..,-w_.-.~,my,v,-.-,,re.-,-r--.-~.--w-.,--r.%. . - - - . - - - - . . - , - - - - - - , , - , - . - , - , - . - - - - - -_--y-- - - , - . - - - , - . m ,.. -
are not addressed in Reference 3). Table 2-3 contains the same event combinations as Table 2-2, however, unfactored loads are used with stress allowables which reflect the probability of occurrence of each load combination. The stress allowables are in accordance with the AISC " Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings (*). This approach is identical to that used in the original design with the dynamic effects of SRV and LOCA loads now included. 2.2.3 Piping and Ecuipment Load combinations and acceptance criteria for balance of plant (BOP) and nuclear steam supply system (NSSS) piping and equipment are presented in Table 2-4. This is consistent with Table 6-1 of DFFR, Rev. 2(2) as modified by the NRC in Attachment V-A to the Lead Plant Acceptance Criteria (LPAC)(1 ) to reflect functional capability criteria. The load combination methods are described in Section 2.2.5 and are consistent with Attachment II of Reference 10. The applicable load cases and the dynamic analysis procedures used for SNPS-1 are consistent with Attachments III and V of the NRC LPAC (10) as follows: SEB-2 and MEB-4 : Use + 15 percent peak broadening of ARS SEB-3 and MEB-5 : Use Reg. Guide 1.92 to combine modal responses MEB-2 : Use OBE damping for normal / upset con-ditions Use DBE damping for emergency / faulted conditions MEB-7(a) : Annulus pressurization effects are com-bined with SSE, inclusive within Load Case No. 7 in Table 2-4 i l MEB-7(b) : "OBE plus SRV" loading condition is assessed by Load Case No. 2 in Table 2-4 MEB-8 : Criteria to assure functional capability for all essential components are in conformance with NUREG/CR-0261 and consistent with Attachment V-B of the NRC LPAC (1 ). Function capability criteria are discussed in detail in Appendix E. l 2.2.4 Reactor Pressure Vessel and Internals Load combinations and acceptance criteria for the RPV and l internals are presented in Table 2-5. Although presented in a O 2-6 Revision 5 - December 1981
h somewhat different form, these combinations cover all those presented in Table 2-4 for piping and equipment. , 2.2.5 Combination of Dynamic Responses For all the mechanical systems, components, and supports in Tables 2-4 and 5, the dynamic responses to the dynamic loads such as LOCA, SRV, and OBE/SSE are combined by using the " square root of the sum of squares (SRSS)" method. The NRC topical report evaluation (11,22) and Revision 1 of NUREG-0484 (22) accepts the SRSS method for the MK-II load combinations. For the reinforced concrete primary containment, concrete secondary structures, and for steel structures, the design basis , shown in load combination Tables 2-2 and 3 is to use the more conservative method to combine the dynamic responses due to 4 seismic and hydrodynamic loads by absolute summation. 2.3 OTHER DYNAMIC LOADS l ! The Shoreham design basis loads are defined in or are derived i from information presented in the SNPS-1 USAR(1') . The major design loads, in addition to normal operating conditions, result from LOCA and seismic events. The original design basis LOCA loads in the USAR include the j quasi-static pressure and temperature in the containment, pipe [ rupture loads, and annulus pressurization. Containment i transients including annulus pressurization are defined in USAR Section 6.2.l(24). Pipe rupture loads are described in Section 3.6 and Appendix 3C of the USAR(1'). J The .Shoreham design basis earthquake ground response spectra are defined in USAR Section 3.7(24). One special subset of seismic load not discussed in the USAR is referred to as seismic sloshing and is described below. Sloshing is a term used to describe the dynamic response of the suppression pool water due to movement of its boundary walls the primary containment, pedestal, and basemat. This pool response has been determined for both the operating basis earthquake (OBE) and the safe shutdown earthquake (SSE). A mathematical procedure has been developed which utilizes a finite element representation of the fluid and a combination of
- analytical and numerical solution techniques. The major assumptions are that the pool boundaries are rigid and that the water behaves as an incompressible, inviscid fluid having irrotational flow. The governing equations of the fluid can be written in a form which separates the spatial eigenfunctions from the equations governing the liquid motion. This results in a normal modes expansion of second-order dynamic equations under a combination of forced and parametric excitation for each mode.
2-7 Revision 5 - December 1981 f 4
,m---,- -
,, , ,,-,r_
, , - - - , , - . - - - . , , - ,,-c ,n.,n-,-----,---n-_- , - - - , - - - - - - , - ~ , .- , ,,,
The finite element method is used to find the matrix representation of the equations governing the oscillation modes. The equations are solved by standard numerical methods to obtain the eigenvalues and eigenvectors (' natural frequencies and oscillation mode shapes). Time dependent modal responses are determined by common numerical techniques and modal superposition is used to obtain the final solution in terms of free surface oscillation profiles and the pressure and velocity fields in the pool. An exact solution for the eigenvalues for liquid in a circular cylindrical tank is available. Natural frequencies obtained by the present finite element method approach exact values as more elements are used. An exact solution is not available for an annular circular tank, but calculated frequencies also converge asymptotically as more elements are used. The assumption of rigid walls has also been studied parametrically and found to be applicable for the geometry of the SNPS-1 suppression pool. The maximum free surface deflections for the SNPS-1 suppression l pool have been found to be 2.3 feet for the OBE and 4.2 feet for an SSE. Maximum dynamic wall pressures have been found to be 1.61 poi for an OBE and 2.82 psi for an SSE. These are not l considered to be significant structural loads. Other design loads include wind, flood, and missiles as described in USAR Sections 3.3, 3.4, and 3.5,(2*> respectively. 2.4 APPROACH USED FOR DESIGN ASSESSMENT The following sections describe the manner in which the SNPS-1 design assessment for suppression pool hydrodynamic loads has been carried out. The sources of suppression pool hydrodynamic loading are identified in Section 2.1 and the load combinations and acceptance criteria are provided in Section 2.2. Section 2.4.1 described the way in which loads acting directly on the suppression pool boundaries are addressed in the SNPS-1 design assessment. Indirect loading due to building response is discussed in Section 2.4.2. Section 2.4.3 covers hydrodynamic load considerations not appropriate for Sections 2.4.1 or 2.4.2. 2.4.1 Submerced Structure / Pool Swell Loads The hydrodynamic loads described in Section 2.1 act on the submerged boundaries of the suppression pool and on structures within the pool. During pool swell, hydrodynamic loads also act on the wetwell airspace boundary, and the drywell floor, as well as structures in the pool swell zone above the suppression pool. Figures 2-3 through 18 describe the time relationships of hydrodynamic loads acting on the following structures: O 2-8 Revision 6 - December 1986 e
.. . ~ _ . _. -
1 1
- 1. Drywell floor - Fig. 2-3
() 2. Downcomers - Figs. 2-4 through 7
- 3. Wetwell walls above water level - Fig. 2-8
- 4. Submerged wetwell - Figs. 2-9 through 12
- 5. Small submerged structures - Figs. 2-13 through 16
- 6. Small structures above pool, below breakthrough - Fig.
2-17
- 7. Small structures above breakthrough - Fig. 2-18 Section 3 discusses the suppression pool boundary loads due to SRV discharge. j Section 4 discusses the LOCA-related suppression pool boundary loads due to vent clearing, condensation oscillations, and chugging as well as the vertical and chugging lateral loads on the downcomers. The methods used to calculate the bulk pool swell transient are also discussed in Section 4. Appendix K describes the methods used to calculate loads on submerged structures within the suppression pool due to SRV discharge, downcomers vent clearing, condensation oscillations, and chugging as well as loads on structures in and above the pool due to pool
( swell. SRV discharge and LOCA-related suppression pool boundary
\ loads affect structures and components outside the. suppression pool by exciting primary and secondary containment motion. The , method used to calculate these building response loads are described in Section 2.4.2.
2.4.2 Buildina Response Load To determine the dynamic response of the containment structures when subjected to SRV discharge and LOCA loads, a finite element based computer program, " Dynamic Stress Analysis of Axisymmetric Structures under Arbitrary-Loading,"' developed by S. Ghosh and E. Wilson and modified by Stone & Webster (S&W) was utilized. This program is S&W code designation ST-200. The three-dimensional axisymmetric continuum is represented either as an axisymmetric thin shell, or solid of revolution, or a combination of both. The axisymmetric shell is discretized as a series of frustums of cones and. the solid of revolution as triangular or quadrilateral toroids connected at their nodal
- point circles.
The reinforced concrete containment structures which-include the mat, the primary containment, the shield wall, the reactor pedestal, and the secondary containment are modeled using axisymmetric shell elements. Two RPV shell models are developed 1 O 2-9 Revision 5 - December 1981
from the reactor vendor's simplified vertical and horizontal lumped mass representation (" beam model"). The stiffening effects of the RPV stabilizer, star truss, and inner and outer bellow seals are also included. The major dimensions of the structures and identification of their general arrangement are illustrated on Fig. 2-19. Figure 2-19 depicts the structural model used to represent the complete reactor building and supporting soil. As indicated there, solid axisymmetric elements are used to represent the soil to a radius and depth of approximately 1.5 mat diameters, with axisymmetric thin shell elements representing the structures. The external dimensions of the soil were selected to preserve free-field motions. The boundary conditions for the soil, at a radius of one and one half times the mat diameter, were tested by changing from " free" to " supported" conditions with no significant difference in building response. The depth of the soil layers was selected to preserve a uniform stress field along the radius. The shear modulus of the foundation soil is 13 ksi. Results from SRV analysis have demonstrated that, although the soil shear strains are relatively large, the dynamic behavior of the building and supporting foundation are essentially unaffected by variations in soil shear modulus. Comparisons of results by representing the soil with constant shear modulus, static or zero-strain modulus, and strain-dependent shear modulus give essentially the same results for building response. In the analysis, the soil continuum is assumed to behave essentially as a homogenous isotropic solid. Because of the large diameter / thickness ratio, the basemat is modeled as a thin circular plate. To account for the relatively large eccentricity between the mat middle surface and its intersection with the superstructures, special elements in the form of rigid links are introduced.
- A closer spacing of elements is used in the area of the suppression pool with increasing element size in areas sufficiently far from the pool. This spacing has been used because the loads are applied to the pool boundaries within this l area and a precise definition of internal loads is required.
The equations of mction are solved numerically by direct integration. The effccts of structural damping have been included in the dynamic analysis unless otherwise noted. The Rayleigh damping technique is utilized in which the damping matrix is assumed to be linearly proportional to the mass and stiffness matrix of the structure. The constants of proportionality are chosen so that the 4 percent damping for reinforced concrete structures is obtained at frequencies of 10 to 125 Hz. These limits were selected in order to conservatively 2-10 Revision 5 - December 1981 I - _
encompass the range of frequencies in which significant dynamic response occurs. O The 4. SRV and LOCA load definitions are defined in Sections 3 In general, the pressure loads on the suppression and pool boundaries due to an SRV discharge or. LOCA event vary both circumferentially and meridionally with time. At any point in time this pressure field can be represented meridionally by a l
, discretization into zones and circumferentially by a Fourier series at each of the meridional zones. Therefore, the spatial l and time-wise variation of pressure can be represented by pressure time histories at each zone for each Fourier series term.
I Figure 2-20 provides a graphical representation of the manner in which typical pressures profiles are represented and applied to the structural model in terms of line forces and moments at the respective nodal circles. f
! The equations of motion are solved numerically by direct !
integration and the acceleration time histories at selected locations are computed. Amplified response spectra (ARS) are developed from the resultant structural acceleration time-histories. The computer program "TIMHIS6", code designation ZZ-
- 126, is used to generate the ARS. This program obtains the exact analytical solution to the governing differential equations of motion for single degree of freedom elastic systems for the
- successive linear segments of excitation. These ARS are used as input to evaluate the adequacy of the piping systems and other i
mechanical equipment. 2.4.3 Related Effects The SNPS-1 design assessment for hydrodynamic loads is not
, limited to suppression pool load definition and plant response.
1 In the case of SRV steam discharge, it is necessary to avoid a
. set of conditions where the potential exists for unacceptable j suppression pool loads. The conditions of concern are
- simultaneous high mass flux and high pool temperature in the vicinity of the SRV discharge device. The local temperature / mass
, flux limit for the KWU T-quencher device is discussed in Section i 3.4. The maximum temperature difference between the mass average i (bulk) pool. temperature and that in the vicinity of the quencher i is discussed in Section 10.2 and Appendix I. Section 10.3 and l Appendix J provide a description of the suppression pool bulk - temperature transients for various events involving SRV discharge. The bulk pool temperature transients -are compared
> against the local temperature limit for the device less the i maximum bulk-to-local temperature difference. Section 10.4 describes the SNPS-1 suppression pool temperature monitoring system
, which alerts the operator to take certain actions to mitigate l the suppression pool temperature transients. These actions are consistent with the plant technical specifications. ,O 2-11 Ravision 5 - December 1981 c
Other related effects of suppression pool hydrodynamic loads include the following: the potential for fatigue damage to the downcomers and SRV discharge lines due to repetitive loading, and the potential for drywell floor vacuum breaker cycling as a result of vent depressurization during chugging, due to the location of vacuum breaker sets on six of the vents. Appendix F presents a fatigue evaluation of the subject lines, and Appendix M describes the methods employed to eliminate significant vacuum breaker cycling during chugging. O l l l l l 2-12 Revision 5 - December 1981
TABLE 2-1 ALL POSSIBLE HYDRODYNAMIC LOAD COMBINATIONS N +SRVo. ass N +0BE +SRVo.azi N +SSE +SRVo. ass N +SRVao. +SBA N +0BE +SRVao. +SBA N +SSE +SRVao. +SBA l N +SRVo.2/3 +SBA N +0BE +SRVo.2fs +SBA N +SSE +SRVo.2/3 +SBA N +SRVao. +IBA l: N +0BE +SRVan. +IBA O N +SSE +SRVao. +IBA N +SRVo.2/3 +IBA N +0BE +SRVo.2/3 +IBA N +SSE +SRVo.2/3 +IBA N +DBA
- N +0BE +DBA N +SSE +DBA Legend
N = Normal Operating Loads OBE = Operating Basis Earthquake SSE = Safe Shutdown Earthquake SRVo.aon = Actuation of up to All 11 SRV's SRVao. = Actuation of ADS Valves SRVo.a/s = Actuation of up to 8 SRV's SBA = Small Break Accident IBA = Intermediate Break Accident DBA = Design Break Accident 1 of 1 Revision 5 - December 1981
l g ; e m
$$5 E
o
. & E I a ; J ",
a i i i i i i . . s- s- a g
!o a '
<
- s s- s- s- s-5
= . . . -- . e e ea s- s s s i
a ias5 e s i i -- - i - -
,y~ 3,
. z
~
s , - 9
- ]e a$
o 8
=,e ,
gg ~
~
- s. . .
s- . i. 3e-y !e,3
,s. a' ~
a e a
!a .a , , , , , , ,
e s- s- :,y ; ,e : -
=
e
~
s- s-s e e a sa _ _ a 7 a . i -i : , i , [a: g 8 4. g :g C ~ 2- yg2 M 2 >,
..s s,, . g g gt ,1 3,!
t - as 9 g n - - , . - . . st:
- s- 9 Eg
~
o
. .. . , , - , , eage,g O e e u e - e a e e e e e u 33 5 .
; .: J a .: ; .: .: ;
nusg* s o, e-ess s s- s- e 2
- : : 3, g
. . is s
-g ~gg 1 a; p; ..~~~=s ;
s3trrrrs
- - s a : .
~
- o. c: es j
Jg Jg g! .
< 3 .g $.g g.a3 18 15 j jjjjjj
....... 2
- ., e a te 5 e n , , ,e ,o ; ;- f j u
O
as TABLE 2-3 IAAD CDMBINATIONS AND STRESS LIMITS FOR STRUCTURAL SThEL LOAD b6S
& C03hl+ E Is lo Io lo Eo 14s IB ZA I4 E *. Ef M LMIT 1.0 Normal 1.0 1.0 1.0 - - -
*e/o Temp 1.0 1.0S i
1 l 2.0 Normal 1.0 1.0 1.0 1.0 1.0 - - - - - - - 1.0 1.5s ! w/ Temp 3.0 Normal 1.0 1.0 1.0 1.0 1.0 1.0 - - - - - - 1.0 1.ss
- Severe Env.
! 4.0 Ahmormal 1.0 1.0 - - - - - 1.0 - 1.0 1.0 - 1.0 1.6s 4.a Abnormal 1.0 1.0 - - - - - - 1.0 1.0 1.0 - 1.0sa3 1.6s t ! 5.0 Abnormal 1.0 1.0 - - - 1.0 - 1.0 - 1.0 1.0 - 1.0 1.65 Severe any. 5.a Abnormal 1.0 1.0 - - - 1.0 - - 1.0 1.0 1.0 - 1.0cm3 1.6S Severe Env. 6.0 Mormal 1.0 1.0 1.0 1.0 1.0 - 1.0 - - - - - 1.0 1.bs dxt. Env. 7.0 Abnormal 1.0 1.0 - - - - 1.0 1.0 - 1.0 1.0 1.0 1.0 1.7S Ext. Env. 7.a Abnormal 1.0 1.0 - - - - 1.0 - 1.0 1.0 1.0 1.0 1.0( u 1.75 Ext. Env. DEFINITIODIS D = Dead Imads Pg = SBA or IBA Pressure and Dynorsic Loada
- L = Live Loads Pg = DBA Pressure and Dynamic Icads l i Po = Operating Pressure Imada Tg = Pipe breat Temperature Load i
To = Operating Temperature Imads R = Pipe Break Temperature Reaction Imada R'# j Ro = Operating Pipe Reactions
= Reaction and Jet Forces Associated with the Pipe Break Ea = Operating Basis Earthquake SRV = Safety / Relief Valve Loads ass = Safe aihutooun marthquake (SSE) S = Required section strength cased on the elastic design methods and allowable stress defined in Part 1 of the AldC l
" Specification for the Design, Fabrication, i
and Erection of Structural 6 teel for Buildings = i l i taBSingle valve actuation l 1 of 1 Revision 5 - Decenner 1981
TABLE 2-2 IDAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR PIPING AND EQUIPME1@ load SRV SRV Case NO) ALL ADS OBE SSE IBAta,a) Danta) Design _ Basis 1 X X - - - - - Upaet 2 X X - X - - - Opset l 3 X X - - X - - 2nergencyt** 4 1 - X - - X - haergencya.) 5 1 - X X - X - Baergencyt*3 6 X - X - X X - Daergencyt*> 7 X - - - x - X taergencyt*3 u) N - Normal load conalats of pressure, dead weight, and austained loads. u3 Ose either IBA or SBA, whichever governs. 43 SBA, IBA, and DBA shall include all event induced loads, which are applicable, such as posainle annulua pressurizatica, pool swell load, condensation oscillation load, chugging loads, etc., as defined in Section 4. (* 3 Piping functional capability la assured in accordance witn the O procedures of " Functional Capability criteria for Planta" in Appendix E. specified in this table may be used, digher stresa limit than the level MK II providing functional capability la assured. O 1 of 1 Revlaion 5 - December 1981
l TABLE 2-5 O NSSS RPV AND INTERNALS REACTOR SYSTEM DETAILED LOAD COMBINATIONS REQUIRED LOAD COMBINATIONS: Condition Loads Upset NL + (N-DELTA P) + OBE Upset NL + (U-DELTA P) + SRVcanL> Upset NL + (U-DELTA P) + OBE + SRVcann> j Emergency NL + (U-DELTA P) + CHG + SRVcaos, Y Faulted NL + (A-DELTA P) + JR + VC + SSE , Faulted NL + (A-DELTA P) + JR + AP + SSE j Faulted NL + (A-DELTA P) + CHG + SRVcans> +-SSE , Faulted NL +-(A-DELTA P) + CO2 + SRVcn.,a3 ~+ SSE Faulted NL + (A-DELTA P) + CO2 + SRVcaom> + SSE Faulted NL + (U-DELTA P)--+ AC + SSE
- Faulted NL + (A-DELTA P) + JR + SCRAM + SSE i ,
Faulted NL + (U-DELTA P) + SRVcann> + SSE Faulted NL + (I-DELTA P) + JR : + AP ' i Faulted NL + (I-DELTA P) + JR + VC i STEAM DRYER LOAD COMBINATIONS: Condition Loads Faulted NL + (A-DELTA P) + SSE Faulted NL + (I-DELTA P) O 1 of 2 Revision 6 - December 1986 4
-a w,.n, , - , , , - ,,,,ne,n----w,-w-w--, .r,.~.r,-------we-w-s ,w, ..e , - , -- -,,,_,r,--a,w,-m,mm--ame,,w,-.,-w-
-,-ve.n,~es~ -,mven-,,v-wo-,, -
l TABLE 2-5 (CONT'D) O l DEFINITIONS NL - Metal + Water Weight OBE - Operating Basis Earthquake SSE - Safe Shutdown Earthquake CHG - Chugging Loads SRV< ann, - Safety / Relief Valve Discharge Caused Loads Induced by the Actuation of All Safety / Relief Valves. Envelope of Symmetric and Asymmetric Loads. SRV<an.3 - Safety / Relief Valve Loads Associated with the Automatic Depressurization System. Envelope of Symmetric and Asymmetric Loads.
! N-DELTA P -
Normal Delte Pressure Force A-DELTA P - Accident LOCA Delta Pressure Force U-DELTA P - Upset Delta Pressure Force I-DELTA P - Interlock Delta Pressure Force COz - High Mass Flux Condensation Oscillation Loads I CO2 - Low Mass Flux Condensation Oscillation Loads SRV<ns,a3 - Actuation of Lowest Setpoint Group of Valves. Factor of Symmetric, Single SRV loads. JR - Jet Reaction AP - Annulus Pressurization Loads AC - Acoustic Pressure Loads VC - Vent Clearing Loads SCRAM - Loada Produced by the sudden Shutdown of a Nuclear Reactor as a Result of the Rapid Insertion of the Control Rods I O 2 of 2 Revision 6 - December 1986 I
O O O 4 4 i i i i z S/R VALVE WATER CLEARIMG
; 9 1 ! t-O '
2 8 /// S/R VALVE AIR CLEARING i E a S/R VALVE STE F !!! O O.25 0.35 TIME (sec) FIG.2-1
- EVENT-TIME RELATIONSHIP FOR SRV DISCHARGE i DUE TO ANTICIPATED PLANT TRANSIENTS l SHOREHAM NUCLEAR POWER STATION-UNIT I
! PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS l
REVISION 5-DECE PABER 1981
O O O DRYWEL TEMPEkkT REY RESSU F CTUATI NNNNNNNN_NN,NNN,NNNNNNNNNN,NNNNNNNNN SUPPRESS)0YCHAMBER TEMPERA URE h PRESSUREN)[C'TNTION
. < NNNNNNNNNNNNNNNsNsNNNNNsNNNNNNNNNNNNNNN\
I I I I DOWNCOMER WATER CLEARINH DOWNCOMER AIR CLEARING EkkAM FIN
. 6- NNN\\\\\\\\\\\\
l 5 q l POOL SWELL 1 1 u l POOL FALLBACK o I CONDENSATION OSCILLATION AND CHUGGING ECCS FLDODING OR SPRAY ACTIVATION lo s O.1 0.6 1.12 4 15 50100 500 TIME ( see) FIG. 2-2 EVENT-TIME RELATIONSHIP FOR THE DESIGN BASIS ACCIDENT SHOREHAM NUCLEAR POWER STATION- UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 3-NOVEMBE.'t 1978 I
O O O
=
DEADWEIGHT; SEISMIC LOADS; , LOCA PRESSURE AND TEMPERATURE TRANSIENTS l ACCUMULATED WATER ON FLOOR JET IMPINGEMENT AND
$ PIPE WHIP G
NEGATIVE LOAD 8 ' DURING REFLOOD o DOWNCOMER VERTICAL REACTION LDADS E
- o. NEGATlW LOAD
~
g . 00WNCCffER FOLLOWING SPRAY s _x ~ ACTUATION IREQU!RE*, LATERAL 6TERATOR ACTION) POOL SWELL . REACTION -
- AIR 804Ds COMPRESSION
> -SECTION 4 j
'} _ u- C.O. AND CHUGGING i . e i . ,
4 0.6 1.2 4 60 ;100 >600- ' T!WE AFTER LOCA (sec)
' - , , ,J?
.. FIG.2 -3 more.
LOAD COMBINATION H! STORY.. cows:orn ArcN is GIVEN TO nEACTION LOADS ON THE STRUCTURE'AFFECTED. DRYWELL FLOOR onywett rtoon rnou oTwen sTnuctuatzs ACC10ENT CONDITION: L ARGE LINE BREAK (DBA) SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS I ! REVISION 5-DECEMBER 1981 l i
O O O l DEADWEIGHT; SEISMIC LOADS; LOCA PRESSURE AND TEMPERATURE TRANSIENTS; HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS; I i
- , NEGATIVE NEGATIVE LOAD DURING LOAD VERTICAL REACTION LOADS REFLOOD FOLLOWING SPRAY ,
ACTUATION z S2 MERGED STRUCTURE LDADS f3 DUE TO COPOENSATION OSCILLATIONS & CHUGGING > SECTION 4 (REQUIRES OPERATOR z ACTION) 8 taTEnAt LDAD a (CHUGGING) z E o 8
.- -SWELL AND FALLBACK FRICTION LOADS a
) e B t e a ! 0. 6 0.4 15 60 100 >600 TIME AFTER LOCA (sec) FIG. 2-4 nogg. LOAD COMBIN ATION HISTJP.Y consioEnaTrow is GIVEN TO RE ACTION LOADS ON THE STRUCTURE AFFECTED: DOWNCOMERS oowwcouEn enou ornca sTeucTunEs ACCIDENT CONDITION: LARGE LINE BREAK (DBA) SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS l REVISION S-DECEMBEn 1981
- O O O i , -
DEADWEIGHT; SEISNIC LOADS; LOCA PRESSURE AND TilfERATURE TRANSIENTS;
' 'l HYDROSTATIC PRES $URE INCLUDING SEISMIC EFFECTS; l !
s-MULTIPLE SRV NEGATWE LOAD ACTUATION ON SETPOINT Up FOLLOWING SPRAY t TD f(3 TO KTf.'ATION OVEQUIRES 2/3 0F VALVES OPERATOR JCTION) j 5 > SECTION 3 C o 5 u ADS ACTUATION
$ e 3
VERTICAL REACTION LOADS i -- g SUBMERRED STRUCTURE LOADS > SECTION 4 1 fidi TD CONDENSATION OSCILLATIONS AND CHUGGING LATERAL LOADS (CHUGGING)
, , . e 1 1
l 4 T > 120 T+5 min. >600
- Tii4E AFTER LOCA (sec)
FIG 2-5 I =OTES: LOAD COMBINATION HISTORY j l T IS BREAR AREA DEPENDENT STRUCTURE AFFECTED;DOWNCOMERS auf is TypcAuf N THE ORDER OF2 TO S MINUTES. ACCIDENT CONDITION:INTERMEDI ATE 2, C088S10ERATION IS GIVEN TO LINE BREAK 4 . REACTION LO4DS ON THE SHOREHAM NUCLEAR POWER STATION-UNITI q DOWNCOMER FROM OTHER STRUCTURES. PLANT DESIGN ASSESSMENT FOR SRV AND LOCALOADS j REVislON 5-DECE MBER 1991 i ._
O O O DEADWEIGHT; SEISMIC LOADS; LOCA PRESSURE AND TEMPERATURE TRANSIENTS; f HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS: i ! MULTIPLE SRV l~ ACT W IM ON OPERATOR ACTUATION OF SRV FOR 8 1/3 T /3 REACTOR CORMN , OF WLVES ,. E C ra
> SECTION 3 ;
E z VERTICAL REACTIO*8 loa 09 i E3 4 8 l ADS ACTUATION s ! SUBMERGED STRUCTURE LOADS
- DUE TO CNUGGING
- SECTION 4
] LATERAL LOADS'DUE t 3 TO CHUGGING
- i j 4 600 1800 6 HRS DAYS TIME AFTER LOCA (seC) i i
i NOTE: FIG. 2 - 6 4 CONSIDERATION IS GIVEPI TO REACTION LOADS J ON THE DOWNCOMER FROM OTHER STRUCTURE!L LOAD COMBNATION HISTORY
; STRUCTURE AFFECTED:DOWNCOMERS !. ACCIDENT CONDITION: SMALL LINE BREAK
- SHOREHAM NUCLEAR POWER STATION-UNIT I j PLANT DECIGN ASSESSNENT FOR SRV AND LOCA LOADS 4 ) REvesl0N 5-DECEMBER 1981 1 1 1
- O O O i
DEADWEIGHT; SEISMIC LOADS; HYDR 0 STATIC PRESSURE INCLUDING SEISMIC EFFECTS i i , 0 TO ALL SRV LINES ON SETPOINT f E AIR BUB 8LE CSCILLATION PRESSURE AND DRAC
. B j [ SECTION 3 1
5 R i S 9 i 1 4 i. i
- TIME
,g7g, FIG. 2-7 ? . CON 90ERATM IS GNEN TO REACTM LOADS LOAD COMBINATION HISTORY i DN THE DowNCOMER FNOM OTHER STRUCTURES STRUCTURE AFFECTED: DOWNCOMERS 1 ACCIDENT CONDITION: NONE SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS 2 REVISION 6 - DECEMBER 1986
i O O O l DEADWEIGfT; SEISMIC LOADS: LOCA PRESSURE AND TEMPERATURE TRANSIENTS: i l P00L SWELL
- l AIR j g COMPRESSION i C 2
E > SECTION 4 5 HYDROSTATIC E LMD EE TO j WATER SWELL , a i i 4 1 ~ j o.6 1.2 - ] TIIE AFTER LOCA (sec) i i NOTE: ,' CON SIDER ATION IS GIVEN TO REAC'lON FIG.2-8
" T'* * "
fT Ndos"TRu TunEY LOAD COMBINATION HISTORY STRUCTURES AFFECTED:WETWELL WALLS ABOVE i THEWATER LEVEL ACCIDE;NT CONDITION:L ARGE LINE BRE AK (DB A) SHOREHAM NUCLEAR POWER STATION-UNIT I I PLANT DESIGN ASSESSMENT FOR SRVAND LOCA LO ADS REVISION 3-NOVEMBER 1978
. . - - . - wg 4 a. & - e-L+4- s - * - - a ~.k. .- -- 4_ -*.u 4 4 4_4.- + .-h6 --
O O O
)
j DEADWEIGHT; SEISMIC LOADS; LOCA PRESSURE AND TEMPERATURE TRANSIENTS; HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS; j DOWN-COMER VENT j CLEARINE . JET CONDENSATION OSCILLATIONS LOADS > SECTION 4
] g AND CHUGGING ! C
! .E BUDBLE 4 5 u FORMATION o 2 a o f 4 i {
) e e a _
i
! 0 0.6 1.2 4 60 i TIE AFTER LOCA (sec)
NOTE 8 l CON SIDERATION IS GtVEN TO REACTION LDAOB ON THE WETWELL WALLS FR004 ATTACHED STRUCTURES IIO E*I
; LOAD COMBINATION HISTORY ; STRUCTURE AFFECTED: SUBMERGED WETWELL l ACCIDENT CONDITION: LARGE LINE BREAK (DBA)
- SHOREN AM NUCLEAR POWER STATION- UNIT I
; PLANT DESIGN ASSESSMENT FOR SRV ANDLOC A LOADS REVISION 4 -FEBRUARY 1981 l
4 4
O O O DEADWEIGIT; SEISMIC LOADS; LOCA PRESSURE AND TEMPERATURE TRANSIENTS; HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS; s PRILTIPLE SRV ACTUATION ON SETPOINT-UP TC .- 1/3 TO 2/3 0F
@ VALVES
>SECTION 3 9
i E j j ADS ACTUATION S , CONDENSATION OSCILLATIONS AND CHUGGING SECTION 4 l l E e B 4 T >120 T + 5 MINUTES TIME AFTER LOCA (sec) NOTE:
- 1. T IS BREAM AREA DEPENDENT,80 TIS FIG. 2-10
'" " *" LOAD COMBINATION HISTORY
- r. cEsiNAreo$$s$fYEN 5RE$CT50NDS oN rHE wErwEtt WALLS FROM ATTACHED STRUCTURE AFFECTED: SUBMERGED WETWELL structures. ACCIDENT CONDITION:INTERMEDI ATE LINE BREAK SHOREHAM NUCLEAR POWER STATION-UNIT I PL ANT DESIGN ASSESSMENT FOR SRV ANDLOCA LOADS REVISION 5-DECEM8ER 1989
O O O
~
i DEADWEIGHT; SEISNIC LOADS; LOCA PRESSURE #1D TEMPERATURE TRANSIENTS; f HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS; i < ~ U ACTUATION OF 1 SRV MAY OCCUR. 1 ' SETPO!kT 1/3 TO 2/3~0F . , .@. VALVES 4 M > SECTION 3 O E W , j g ADS ACTUATION - o i CHUGGING SECTION 4 i i 1 a e a e i 3 4 1800 6 HRS Tite AFTER LOCA (sec) ,
- NOTE
- -
CONSIDER ATION IS GIVEN TO REACTION LOAOS ON THE WETWELL WALLS FROM I!O 2*II ATTACHED STRUCTURES. LOADCOMBINATION HISTORY j STRUCTURE AFFECTED: SUBMERGED WETWELL ACCIDENT CONDITION:SMALL LINE BREAK ! SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSEMENT FOR SRV AND LOCA LOADS nevismN 5-DECEM8ER 1981 i e
O O O I , DEADWEIGHT; SEISMIC LOADS;
- HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS l ACTUATION OF SRV Oil PRESSURE SETPOINT
! O TO ALL VALVES h e-Alt SUBBLE OSCILLATION PRESSURE AND DRAC l E SECTION 3 5 E s 3 5 a i a 1 k TIME i l NOTE: CONSIDER ATION IS GIVEN TO REACTION FIG . 2-12 LO AD8 0N THE WETWELL WALLS FRON f ATTACHED STRUCTURES. LOAD COMBINATION HISTORY STRUCTURE AFFECTED:SUBM.ERGED WETWELL ACCIDENT CONDITION: NONE SHOREHAM NUCLEAR POWER STATION-UNIT I j PL ANT DESIGN ASSESSMENTFOR SRV AND LOCA LOADS j REVISION 3 -NOVEMBER 1978 . l j - .
O i DEADWEIGHT; SEISMIC LOADS; LOCA PRESSURE AND TEMPERATURE TRANSIENT 3; j HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS , t m j DOWNCONER VENT CLEARING JET LOAD $ 5 , C t E POOL SWELL > SECTION 4 4 5 DRAG LOADS
- E i E FALL i
h BACK LOADS I SUBMERGED STRUCTURE LOADS DUE TO CONDENSATION OSCILLATIONS i AND CHUGGING s
- l. . i i
l
. .. .. . . . i j O.6 6 1.2 2.34 15 60 TIE AFTER LOCA (sec)
,g 7 g.
FIG.2 -13 4 consIDERATIONISGIVEN TO RE ACTION LOAD COMBINATION HISTORY
* "8 STRUCTURE AFFECTED: SM ALL SUBMERGED
, ' NE*Ew"EEEIc"EsTb* " "ES 1 STRUCTURES,COLUM NS, AND PIPING i ACCIDENT CONDITION: L ARGE LINE BREAK (DBA) 4 SHOREHAM NUCLEAR POWER STATION-UNIT I j PLANT DESIGN ASSESSMENT FOR SRVANDLOCA LOADS i I REVISION 4 -FEBRUARY 1901 4 i
1 O O O 4 DEADWEIGHT; SEISMIC LOADS; LOCA PRESSURE AND TDFERATURE TRANSIENTS; g HYDROSTATIC PRESSURE INCLUDING SEISMIC EFFECTS } I MULTIPLE SRV ACTUATION OF 1ETPOINT-UP TO . 1/3.TO 2/3 T
!. MLLVES
> SECTION 3
) B . ADS ACTUATION I g
- i .
1 , i - SUBRERGED STRUCTURE LOADS DUE TO ' CONDENSATION OSClLLATIONS AND CHUGGING f. 1 l a I . , ! 4 T>1m T + 5 min. 1 TIME AFTER LOCA (sec) WOTES: -
- 1. T is BRE4st AREA,0EPENDENT, FIG. 2 - I4 T * " LOAD COMBINATION HISTORY i S'E S Eu$ " "
y STRUCTURES AFFECTED:SMALL SUBMERGED . z.comsmEaATion is sevEn To =EacTion ! j - Lonos ou susuEheED muCTuREs FROM STRUCTURES. COLUMNS. AND PIPING
- oTHEn ATracwEn srRucmats. ACCIDENT CONDITION
- INTERMEDIATE LINE BREAK SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRVAND LOCALOADS l
1 1 t REVISI,o.N,5 -0ECEWBER 1984 i.
_ . _ . . . _ _ _. . . _ . . . . - _ . . . _ _ - _ _ . . _ _ . _ = _ _ -_ _ . _ _ _ _ . _ . _ _ . _ _ . - _ , _ . _ _ _ _ _ _ . . . _ _ . . _ . _ . o o o i {- -DEADWEIGHT; SEISMIC LOADS; LOCA PitESSURE AND TEMPERATURE TRANSIENTS; HYDil0 STATIC PRESSURE - INCLUDING SEISMIC EFFECTS MULTIPLE SRV ACTUATION OF i SRV MAY OCCUR ! ACTUATION ON ! SETPOWT I 1/2 TO 2/3 0F > SECTION 3 . MLVES l l E ADS ACTUATION CHUGGING SECTION 4 j i i E L l ' t l i
- 1 l
i i i ' i a I I I g } 4 1800 6 HRS : l TIME AFTER LOCA (SEC) ,. i l l NOTD FIG. 2-15 ! c0NseERATION IS GWEN TO REACTIONI.040S LOAD COMBINATION HISTORY ! %@,E0pg ' 5 M W OTHER STRUCTURES AFFECTED:SMALLSUBMERGED i i STRUCTURES COLUMNS AND PIPING L i ACCIDENT CdNDITION' SMALL LINE BREAK I
~SHOREHAM NUCLEAR POWER STATION-UNIT I
- PLANT DESIGN ASSESSMENT FOR SRV AND LOCA, LOADS REVISION 5-DECEMSER 1981 I
I
. .. _ _ . _ _ . . - _ . _ . _ _ ~ . .
i t. DEADWEIGNT; SEISMIC LOADS HYDR 0 STATIC PRESSURE INCLUDING SEISMIC EFFECTS 3 k 0 TO ALL VALVES ON PRESSURE SETPOINT 5 . p AIR BUS 9tE OSCILLATl0ll PRES $ tite AaD '*E i E f $ SECTION 3 . s E S 1 I 1 i TIME } FIG. 2-16 LOAD COMBINATION HISTORY l "om YiEeY E E Ea "uc N "rn"#0 o TM" ATTACH ED STRUCTUR ES. STRUCTURE AFFECTED:SMALL SUBMERGED STRUCTURES, COLUMNS, AND PIPING ACCIDENT CONDITION NONE - SHOREHAM NUCLEAR POWER STATION-UNITI I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LQADS 1 t REVISION 3-NOVEMBER 1978
- i L
. _ . .. . . . . . . . - . . . - -- . - . . .----.-- .. - - _ - - - - .-. -. . . - . ~ . . - , - - .
l O O O ' l . ! DEADWEIGHT; SEISMIC LOADS j LOCA PRESSURE AND TEMPERATURE TRANSIENTS 4 ) I li4 PACT LOADS ] a j h DRAG LOADS > SECTION 4 I 5 4 t l E i ! g FALL BACK ! g LOADS i " l i 4 3 t l . . . ; I o.s 1.2 2.3 TIME AFTER LOCA (sec) FIG. 2-17 LOAD COldBINATION HISTORY ' STRUCTURE AFFECTED:SMALL STRUCTURES ABOVE POOL AND BELOW BREAKTHROUGH j ACCIDENT CONDITION: LARGE LINE BREAK (DBA) SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSM ENT FOR SRV AND LOCA LOADS REVl810N 3. NOVEMBER 1978 l 1 l
O O O DEADWElGHT; SEISMIC LOADS .) LOCA PRESSURE AND TEMPERATURE TRANSIENTS 1
- 5 P
5 5 o E 5 a a i TIME AFTER LOCA i I FIG. 2-18 LOAD COMBIN ATION HISTORY STRUCTURES AFFECTED:SMALL STRUCTURES ABOVE BREAKTHROUGH ACCIDENTCONDITION:LARGE LINE BREAK (DBA) i
' SHOREHAM NUCLE AR POWER STATION-UNIT I ; PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS 4
REVl$10N 1- APRIL 1977
OVERSIZE DOCUMENT : PAGE PULLED SEE APERTURE CARDS NUMBER OF OVERSIZE PAGES FILMED ON APERTURE CARDS_ APERTURE CARD /HARD COPY AVAILABLE FROM RECORD SERVICES BRANCH,TIDC FTS 492-8989
=
l t e e a l e-, e-~~,+----yr- wwws-s ..wmrw,~,e--.e-ww-
SUPPORT O PEDESTAL PRIMARY CONTAIN MENT
.5 -
...e,,
'.,'.@v PRESSURE ,
ILa./ IN.a 3 SUPPRESSION POOL
/ ~
I 4 \ u o o u , o hY?? .iNI.$.k TYPICAL PRESSURE PROFILE O MOMENT (FT .K/FT.) h t oRCE(K/FT.) D
.I*UN.I 5 r '.M.d.i y4 SUPPRESSION 1 POOL /
9*7 >G h - -~ - ~ ~ ~ g3 . n n n n n . gg IDEALI2ED NODAL FORCES AND MOMENTS 1 l 1 FIG. 2-20 O LOADS APPLIED TO MODEL SHOREHAM NUCLEAR POWER STATION - UNIT I r PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION S - DECEW8ER 1881
7-~g SECTION 3 SRV LOADS
3.1 INTRODUCTION
The pressure oscillation in the BWR suppression pool resulting i from safety / relief valve (SRV) blowdown transients is a design consideration. A phenomenological representation of an SRV piping system discharging steam into the suppression pool of a BWR plant is shown on Fig. 3-1. The opening of the SRV results in the discharge of the water column from the SRV line, followed by the discharge of air, and then steam. Because of the initial inertia of the water column in the line, the air volume will be compressed, and the backpressure will increase until the watec column is expelled into the suppression pool. Upon entering the pool, the discharged air volume immediately forms air bubbles. Formation of the air bubbles provides a mechanism for the exchange of pressure energy to kinetic energy of the surrounding water medium, causing a pressure oscillation in the pool. The oscillating bubbles, in turn, impose periodic forces on the pool
; boundary and submerged structures by transmission of pressure and velocity fields through the pool.
Discharge devices are provided at the submerged ends of the discharge lines for load mitigation. A ramshead device was (
,O originally selected for the test results have shown that shoreham plant. However, various the quencher device has better performance than the ramshead: namely, the quencher device produces smaller pressure loads on the pool boundary and submerged structures; and the quencher device provides a smooth steam condensation at high pool water temperature and steam mass flux.
Therefore a quencher device was selected and installed in the Shoreham plant to replace the original ramshead device. The quencher device installed in the shoreham plant is the T-quencher designed for Mark II plants by KWU as described in Reference 15. This particular quencher device was selected because the tests conducted by KWU at Karlsteinca48 and operating plants like KKB provide a well-defined data base for improved load definition for the Shoreham plant. In Section 3.2, the pool boundary load definition is described for both the design basis ramshead and the interim T-quencher load specification. . The interim T-quencher load definition has been used in two ways:
- 1. As a means to demonstrate that the design basis ramshead load definition conservatively bounds an appropriate T quencher load definition with simultaneous bubble
- entry (see Appendix G), and 3-1 Revision 5 - December 1981
- 2. as part of the design load in a limited number of applications.
Appendix L provides a description of the hydrodynamic loads confirmatory program for Shoreham. Included is a description of the confirmatory T-quencher load and a reanalysis of each system that had previously been analyzed using the interim T-quencher load definition as the design load. Descriptions of the submerged structure loads due to SRV discharge, including the quencher device design loads, are presented in Section 3.3. 3.2 POOL BOUNDARY LOADS The formation of oscillating air bubbles in the suppression pool after the actuation of safety / relief valves produces transient loads on the wetted pool boundary. The characteristics of the transient pool boundary loads depend largely on the SRV discharge device and the plant parameters ~. This section presents the load definition for both the design basis ramshead load specification and the interim T-quencher load specification. ! 3.2.1 Desion Basis Load Definition (Ramshead Discharoe Device) In October 1978 the United States Nuclear Regulatory Commission (NRC) completed the review of the Mark II lead plants (Shoreham, Zimmer, and LaSalle) design basis LOCA and SRV loads and issued the lead evaluation report NUREG-0487(1). The report concluded that the use of the ramshead device is unacceptable for Mark II plants. However, in order to meet the lead plants' licensing schedule, a conservative load definition was required to be . established in advance of the T-quencher qualification programs. On this basis the NRC determined that the ramshead load specification, as prescribed by the analytical models and calculational procedures in DFFR Rev. 2(48 was generally acceptable as a load specification for Mark II T-quencher air discharge loads. To comply with the NRC load evaluation conclusion and to ensure safe operation of the plant, the ramshead load specification in DFFR Rev. 2 was selected as the Shoreham design basis SRV load definition. l Two exceptions were taken with the excessively conservative requirements stated in NUREG-0487. The first relates to the SRV 3-2 Revision 6 - December 1986
i i i load actuation case referred to as Load Case 5 and the second j relates to the prescribed bubble oscillation frequency range.
, Both of these issues are discussed in detail in Appendix G. It is shown there that the Shoreham design basis ramshead load ; definition, with the noted exceptions from NUREG-0487, still
- results in loads which are conservative with respect to those j resulting from the actual T-quencher device.
! 3.2.1.1 Ramshead Load Soecification !
= The ramshead load specification for Shoreham pool boundary load is in complete agreement with the analytical models and the 3 calculational procedures prescribed in Sections 3.2.2 through j 3.2.4 of DFFR Rev. 2<2). 4
! The major elements of the load specification for computing the pool boundary load due to a single SRV discharge are as follows.
l The pressure rise in the SRV line air space and the resultant
! dynamics of the water leg in the SRV line following valve l actuation are computed using the one-dimensional transient model. l j When the line clears of water, a bubble formation model is used j to compute the initial conditions of the air bubble formed by the
] complete expulsion of air from the line. A bubble dynamics model i is then used to calculate the subsequent oscillatory bubble i pressure history. Loads on the pool boundary caused by the bubble transient are computed using method of images and-l potential flow theory. $' Finally, the perturbations from a base case boundary load, as calculated by the above methodology, are presented as " influence
- coefficients" for a given change in each input parameter. These
!' influence coefficients are used to obtain the pool boundary loads with the appropriate plant parameters.
i l In accordance with DFFR Rev. 2(8), the following load cases constitute the complete basis for the plant design assessment: case 1 - All SRV discharge in sequential actuation Case 2 - ADS SRV discharge ' case 3 - Asymmetric SRV discharge case 4 - Single SRV discharge t A brief discussion of these design basis load cases is provided . as follows: I
; . Load Case 1: All SRV discharge in sequential actuation.
l The sequential actuation of all valves is discussed in
- j. paragraph A, Section 3.2.4.1.2 of Reference 3. This is a
! mechanistic discharge case and it considers the discharge of the SRVs at their setpoints for a linear pressure rise in the reactor ! pressure vessel (RPV). The plant-specific line characteristics, O i f 3-3 Revision 5 - December 1981 i i
including the line length and friction losses, are incorporated for each discharge line. Load Case 2: ADS SRV discharge The ADS SRV discharge case is described in Section 3.2.4.1.2 of Reference 3. The automatic depressurization system consists of seven valves which are symmetrically located and actuate simultaneously at the same system pressure. A conservative upper bound for the system pressure was selected. The specific discharge line characteristics were considered in the load evaluation. Load Case 3: Asymmetric SRV discharge The asymmetric loading case is conservatively constructed by considering the simultaneous blowdown of three adjacent valves having identical setpoints but different line characteristics. To yield a bounding condition for this asymmetric load case, all bubbles are assumed to enter the pool simultaneously. Load Case 4: Single SRV discharge The SRV which produces the largest pool boundary pressure is used to establish the single SRV discharge load case. To address the measured load increase under consecutive valve actuation second-pop conditions which occur when a single valve is actuated two or more times in rapid succession, Reference 1 defined the load multiplier factors of 1.6 on predicted peak negative loads and 1.4 on predicted peak positive loads to be applied on the single valve actuation conditions. DFFR Rev. 2 did not establish a load case for the consecutive [ valve actuations. For the Shoreham plant, the single valve consecutive actuation load case, with appropriate multipliers, is bounded by the other design basis load cases. Prior to the establishment of the four design basis load cases, two additional load cases had been investigated: l Case 5 - All SRV discharge, bubbles enter the pool simultaneously and oscillate in phase Case 6 - Three adjacent SRV discharge, bubbles enter the pool simultaneously and oscillate in phase It has been demonstrated in Reference 3 that these two load cases are not credible for Mark II plants. The " simultaneous and in phase" Case 5 stated above is consistent with Section III.C.2.b.1.f of Reference 1 which has been referred to as " Load Case 5" in previously published public documents. A 3-4 Revision 5 - December 1981
i 1 t 1 detailed discussion on the justification of excluding Load case 5 from design assessment is provided in Appendix G. O Since both cases 5 and 6 were originally utilized in the ! evaluation of the Shoreham primary containment structures, the
- definitions are retained here to show added evidence of design conservatism in the containment structures.
The load evaluation report <1) concluded that the use of the ramshead device is unacceptable because thermal instability i occurs at a certain threshold temperature during the steam
, quenching process. In contrast, the quencher maintains stable ; steam condensation at much higher pool temperatures. A detailed
- discussion of the temperature limit for stable SRV -discharge
! through the T-quencher device is provided in Section 3.4. 3.2.1.2 Ramshead Load Summary For the suppression pool configuration with the SRV discharge device arrangement as shown on Fig. 1-11, the pressure profile and the pressure time-history on the pool boundaries (the reactor i pedestal wall, the primary containment- wall, and the basemat) have been calculated for those six SRV load cases described in i Section 3.2.1.1. All SRV ramsheads are assumed to be located 8 4 feet from the top of the basemat cover slab. The- submerged portion of the suppression pool is divided into 27 zones for l analytical purpose. The specific pool geometry and definition of the 27 zones are shown on Fig. 3-2. For each load c a s e ', the resulting forcing functions for each zone are placed in a computer file and accessed from the file for structural analysis. It should be noted that the theoretical methodology used results in pool boundary pressures having instantaneous rises from zero to near peak amplitudes. Based on observed test results, a i pressure rise time on the order of 20 to 50 m-sec is considered i to be more realistic but has not been incorporated in the ' ) calculated pressure _ histories. The effect of this rise time on ! structural response is discussed in Section 5.1.2.7. 3 ] Case 1: All SRV discharge in sequential actuation j The specific locations of the discharge devices (ramsheads) are presented on Fig. 3-3. Table 3-1 summarizes the maximum and l minimum pressures in each of the 27. zones. Figure 3-4
- illustrates a typical time history of the circumferential average
! pressure. This specific-forcing function is the one that applies to Zone 14. The effect of including the individual discharge i line characteristics and sequential SRV discharge is that the , oscillating bubble pairs'from one discharge line to the next are ! not exactly in phase.
- l l($)
e :
- 1 j 3-5 Revision 6 - December 1986 4
l Case 2: ADS SRV discharge The automatic depressurization system. consists of seven valves O, l which are symmetrically located as shown on Fig. 3-5. The maximum and minimum pressures in each zone are summarized in Table 3-2. A typical circumferential average forcing function for the ADS actuation loading is presented on Fig. 3-6, for the forcing function applied to Zone 14. Case 3: Asymmetric SRV discharge The location of the ramsheads for the three lines used in this evaluation is shown on Fig. 3-7. The maximum and minimum pressures for each zone are summarized in Table 3-3. A typical circumferential average forcing function for the asymmetric loading is presented on Fig. 3-8, for the forcing function applied to Zone 14. Case 4: Single SRV discharge The single SRV discharge loading case has been analyzed to account for the actuation of an SRV which produces the largest pool boundary pressure. The location of the line used in this evaluation is shown on Fig. 3-9. The maximum and minimum pressures for each zone are summarized in Table 3-4. A typical circumferential average forcing function for the single SRV actuation load case is presented on Fig. 3-10 for the forcing function applied to Zone 14. Cases 5 and 6 (Non-design Basis) l l SRV discharge load cases 5 and 6 assume that the bubbles enter
- the pool simultaneously and oscillate in phase and are used for l containment structural assessment only. A normalized pressure
- time history curve on the pool boundaries is presented on Fig. 3-
- 11. For the load case with all SRV's discharging simultaneously and in phase, this pressure profile is uniformly distributed along the circumferential direction. For the load case with I three adjacent SRys discharging simultaneously and in phase, the l pressure distribution shown on Fig. 3-11 varies along the circumferential direction as shown on Fig. 3-12. For both load cases 5 and 6 a frequency range of 5 to 10 Hz was investigated.
3.2.2 T-auencher Load Definitions l See Proprietary Supplement of this report. O i 3-6 Revision 5 - December 1981
i TABLE 3-1 ,
SUMMARY
OF MAXIMUM AND MINIMUM AVERAGE WALL PRESSURES FOR SEOUENTIAL SRV DISCHARGE
\ (Sequential Discharge of.All SRVs Based on Ramshead Methodology) i Maximum Pressure fosid) Minimum Pressure (osid) 2 Zone (2) Local (2) Averace(2) Localca) Averace( )
1 1.72 0.78 -0.37 -0.21 ,
- 2 5.06 2.31 -1.07 -0.62 j 3 8.12 2.73 -f.65 -0.98 4 10.73 4.98 -2.10 -1.29 1 5 12.78 6.01 -2.43 -1.53 14.21 6 6.80 -2.67 -1.71 1 7 15.14 7.37 -2.82 -1.84
; 8 15.64 7.72 -2.91 -1.91 9 15.88 7.90 -2.96 -1.96 .
i 10 16.32 8.06 -3.01 -1.99 4 11 18.43 8.71 -3.31 -2.12 12 20.17 9.28 -3.69 -2.22 l 13 20.45 9.42 -3.69 -2.24 i 14 20.25 9.42 -3.56 -2.23 1 15 20.45 9.53 -3.55 -2.24
; 16 21.02 9.73 -3.65 -2,26 17 21.57 9.92 -3.76 -2.29 18 21.86 10.01 -3.81 -2.30 l 19 20.99 10.02 -3.84 -2.30 1 20 22.43 10.10 -4.09 -2.31 21 25.73 10.31 -4.67 -2.34 ; 22 30.33 10.51 -5.49 --2.39
! 23 31.16 10.14 -5.63 -2.42 i 24 27.23 8.59 -5.24 -2.26
- 25 26.56 6.60 -5.40 -1.93
! '26 20.36 4.00 -5.01. -1.37 27 7.32 1.33 -2.27 -0.50 1
i i i i l 1 (1) See Fig. 3-2 for. definition of zones. l (2)- Maximum or minimum ~in both' space and time.
- (2) Maximum or minimum in time of circumferential average. l
- O 1 of 1 Revision 4 February 1981
TABLE 3-2 ()
SUMMARY
OF MAXIMUM AND MINIMUM WALL PRESSURES FOR AUTOMATIC DEPRESSURIZATION SYSTEM ACTUATION (Simultaneous Actuation of Seven SRVs Based on Ramshead Methodology) Maximum Pressure (osid) Minimum Pressure (Daid) 12ng(1) Localca) Averace<2) Localca) Averace ca> 1 1.46 0.66 -0.31 -0.22 2 4.26 1.94 -0.93 -0.65 3 6.78 3.14 -1.51 -1.05 4 8.85 4.22 -2.02 -1.41 5 10.36 5.11 -2.43 -1.70 6 11.56 5.81 -2.74 -1.94 7 12.51 6.31 -2.95 -2.10 8 13.10 6.63 -3.07 2.21 9 13.40 6.79 -3.13 -2.27 10 13.68 6.93 -3.22 -2.31 11 15.56 7.53 -3.64 -2.51 12 17.41 8.04 -4.00 -2.68 13 17.49 8.15 -4.01 -2.70 14 17.00 8.13 -3.89 -2.67 15 17.14 8.22 -3.88 -2.68 O 16 17 17.78 18.39 8.41' 8.57
-3.98
-4.08
-2.72
-2.76 18 18.68 8.65 -4.13 -2.78 19 18.86 8.66 -4.16 -2.78 20 20.16 8.75 -4.37 -2.81 21 23.18 8.97 -4.87 -2.87 22 27.38 9.20 5.57 -2.92 23 29.67 8.76 -5.63 2.76 24 29.15 7.12 -5.19 -2.23 25 23.83 5.15 -5.07 -1.58 26 18.32 3.08 -4.65 -0.93 27 7.02 1.06 2.20 -0.40 (1) See Fig. 3-2 for definition of zones.
(2) Maximum or minimum in both space and time. (2) Maximum or minimum in time of circumferential average. O 1 of 1 Revision 4 - February 1981'
TABLE 3-3 x
') /
SUMMARY
OF MAXIMUM AND MINIMUM WALL PRESSURES FOR ASYMMETRIC SRV DISCHARGE (Simultaneous Discharge of Three Adjacent SRVs Based on Ramshead Methodology) Maximum Pressure fosid) Minimum Pressure (osid) Zone (1) Local: 2) Averace(2) Localca) Averace(2) 1 2.11 0.71 -0.48 -0.14 2 6.24 2.11 -1.41 -0.42 3 10.07 3.42 -2.29 -0.68 4 13.43 4.59 -3.05 -0.91 5 16.15 5.56 -3.68 -1.10 6 18.16 6.32 -4.14 -1.25 7 19.51 6.86 -4.45 -1.36 8 20.31 7.21 -4.63 -1.43 9 20.68 7.38 -4.71 1.46 10 21.19 7.54 4.84 -1.49 11 23.61 8.18 -5.46 -1.62 12 25.65 8.73 -5.99 -1.73 13 26.02 8.85 -6.02 -1.74 14 25.81 8.82 -5.88 -1.72 15 26.06 8.91 -5.88 -1.73 O 16 26.74 9.10 -6.01 1.75 'LJ 17 27.37 9.27 -6.15 -1.78 18 27.69 9.36 -6.22 -1.79 19 27.82 9.37 -6.26 -1.80 20 28.63 9.45 -6.50 -1.81 21 31.44 9.65 -7.04 -1.84 22 35.76 9.80 -7.77 -1.86 23 36.07 9.25 -7.70 -1.74 24 32.08 7.58 -7.02 -1.49 25 26.73 5.40 -6.35 -1.21 26 20.50 3.20 -5.42 -0.86 27 7.07 1.06 -2.38 -0.34 (1) See Fig. 3-2 for definition of zones.
<=> Maximum or minimum in both space and time.
(2) Maximum or minimum in time of circumferential average. O 1 of 1 Revision 4 - February 1981
TABLE 3-4
SUMMARY
OF MAXIMUM AND MINIMUM WALL PRESSURES FOR SINGLE VALVE DISCHARGE (Based on Ramshead Methodology) Maximum Pressure fosid) Minimum Pressure fosid) Isn.e<2) Local <2) Averace<2) Local (2) Averace<8) 1 1.19 0.31 -0.25 -0.05 2 3.51 0.92 -0.74 , -0.16 3 5.66 1.49 -1.16 -0.25 4 7.54 2.00 -1.51 -0.34 5 9.02 2.42 -1.79 -0.41 1 6 10.05 2.76 -2.00 -0.47 7 10.67 2.99 -2.12 -0.51 8 10.98 3.15 -2.19 -0.54 9 11.11 3.22 -2.21 -0.55 10 11.52 3.29 -2.29 -0.56 11 13.74 3.58 -2.73 -0.61-12 15.65 3.83 -3.10 -0.65 13 15.70 3.88 -3.10 -0.66 14 15.14 3.87 -2.96 -0.65 15 15.26 3.91 -2.94 0.65 16 15.93 4.00 -3.05 -0.66 O 17 18 19 16.56 16.86 17.07 4.08 4.12 4.12
-3.15
-3.20
-3.24
-0.67
-0.68
-0.68 20 18.57 4.16 -3.52 0.69 21 22.01 4.25 -4.15 -0.70 22 26.86 4.30 -5.04 -0.70 23 28.08 4.'04 -5.28 -0.66 24 23.71 3.31 -5.31 -0.57 25 25.27 2.44 -5.20 -0.48 26 19.53 1.61 -4.80 0 .~ 3 6 -
27 7.38 0.57 -2.34- -0.14 (*) See rig. 3-2 for definition of zones. (2) Maximum or minimum in both space-and time. (*) Maximum or minimum in time of circumferential average. O 1 of 1 Revision 4 February 1981
_ - . . - . - - - - .. - ._.. - ...- -..- =. . .-..-..-. -. - ---~._. - - . -.. _ -.... ... . ~ l
- /.
. / . i \ i' TABLE 3-5 . l LOADS ON OUENCHER BODY - j (Based on KWU Report - Refereact 15) [ r I l l l
'f 1
i 1, d PROPRIETARY - See Proprietary Supplement to this Report I. , , l ; 1
! 6 .
t l
/ i'
,' 4 t, :
}
\ - i i i
! 4 i t i
! i t. ,
g i. . . i j i li i 1 i
! , 4 i , -
s ,
; i 1 1
<c .
l l'. ) i j i I
; (.
i.
,r , '
J. t ; 5 a 9 I i t.
/, i
)
/
i f >
; 1 of 1 ' Revision 6 'becember 1986 [
i , 4 , [ f . J' ! (i.. !
.,l.
I
; i1 ; j,
. - - _ . - . - _ . . . _ ~ . . . _ . , _ - . . . . - - . _ . - . . . _ . . - _ . - . _ . .. - ._ . . . _ _ .
3 i. ! TABLE 3-6 1 l ! LOADS ON OUENCHER ARM , f (Based on KWU Report - Reference 15) l 1 i i i i i k 3 1 i j PROPRIETARY - See Proprietary Supplement to this Report ; i i 1 1 i I i 4 i i i l \ 1 : 1 t i i < 1 i i i i i } i I ,I . l t j l t } 1 of 1 Revision 6 -~ December 1986 i t I
; \
J i_.-_,_..,_.....,_,. __. _ _ _ _ _ _ _ _ _ _ _ . _ _ - ,
m .I SAFETY /. ,:,,, :,,,,:,;;;,,; . :,i n RE EF !!!!:::...:iij:jiii: !!- MOVING STEAM iib
*~
g MOVING AIR [\ / MAIN STEAM SYSTEM o STATION ARY AIR 4 i SHOCK WAVE WATER
/r ',
/
j --
. /~ %, :
r i ::~ ::: , O j i5~ 2 f j r
, i i
./ g WATER SURFACE l
SUPPRESSION POOL riittrit tis t ristristiriittitro
/
INITI AL RESPONSE o , w> _- r %_
- d' PRESSURE k
g% V ~~d " ,
- LOADS OSCILLATING - .
,,,, AIR BUBBLE P.
LATER RESPONSE FIG 3 - 1 O PilENOMENON OF SAFETY / RELIEF VALVE BLOWDOWN INTO SUPPRESSION POOL , l I SHOREHAM NUCLEAR POWER STATION - UN!T 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS : REVISION 4 -FEBRUARY 1981 .
\
SUPPORT PRIMARY E PEDESTAL CONTAINMENT
.,'. n in ,
y H.W.L. 27 '-0" 2'(TYP) 4 > t I v20NE NUMBER *--- SRV DISCHARGE LINE I 18' Fi : E L. 8 7 '- 0" I . BASEMAT COVER g SL A8 EL. 9'-O" BASEMAT COVER SLA8 MSMMMMMMMMM
- ii:
le-----13.00'---.
- 15.94' 1
; 18.89' =
2 21;83' - i c 24.78' : 4 27.70' : 27.72'
- 30.67' r
.: 33 61' r 36.56' :
- 39.50' --
1 PEDESTAL RADIUS -13.0 FT. 7,0NTAINMENT RADIUS -39.5 FT. NOTE POOL DEPTH-18 FT. NOT TO SCALE SUBMERGENCE DEPTH-10 FT. FIG. 3-2 CROSS-SECTION OF SUPPRESSION POOL AND DEFINITION OF SUPPRESSION CHAMBER WALLS ('] LOADING ZONE FOR RAMSHEAD LOAD DEFINITION SHOREHAM NUCLEAR POWER STATION- UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LJDCA LOADS REVISION 4 - FEBRUARY 1981
8 Fol3H - 100.7 j _ po;3L.2lt.2' 6 F015F-156*-, w p Fol3C- 257.9 8 F015J-109.3*] g o a i O O
& l 1 -
O O4 O A A O s PEDESTA _. e s __o 0d
\ F015A-265.6' O ^' '''''
& O' O v F0i s e-e n.. p O O
FOl3K-87.9' O, F015D-340.78 F01 st -e.4* ll I '
~27.7' :
b o_O ( 0.O s' : - 34.2 SECTION A-A NOTE. NOT TO SCALE. FIG. 5-5 ORIENTATION OF SRV LINE DISCHARGE DEVICES (RAMSHEADS) FOR SEQUENTIAL BRV DISCHARG E O .Noa NA Nue'sAa eo PLANT DESIGN ASSESSMENT FO SRV AND LOCA LOADS orArioN-uNir i j REvts10N 4 -FEBRUARY 1981
e-O 12 I ya - R O l -) b r 3 4 - 8 E, l
@ l 0 -
]
l d I I I I
-4 O 0.2 0.4 0.6 0.8 1.0 TIME. AFTER FIRST BUBBLE CLEARED (SEC)
FI G. 3 -4 TYPICAL ALL VALVE SEQUENTI AL SRV DISCH ARGE FORCING FUNCTION FOR RAMSHEAD DEVICE -ZONE 14 SHOREHAM NUCLEAR POWER STATION-UNIT 1 O PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4 - FEBRUARY 1981
18 0 ' N F013 L - 212.2' F013H-150 7' , O
, O. M F013 J- 109.3 O
8 F013 A -263. 6'
%*' l O
O l sc- - - d o_ O 270 s.
' '~3' "
F013 K-57. 9* O O
- V, O
i FOf 3E-60 i O' NOTE NOT TO SCALE FIG.3-5 ORIENTATION OF SRV DISCHARGE LINE DEVICES (RAMSHEADS)FOR AUTOMATIC DEPRESSURIZATION SHOREHAM NUCLEAR POWER STATION-UNIT I C PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVIS10N 4 - FEBRUARY 1981
; O O O 12 1
E8 - . S tiis % i s 1
\
l E* - 1
! !" /
E Q Uo g nu k _4 i e i e i i O O.13 0.25 Q38 0.50 0.85 0.75 088
. TIME AFTER FIRST BUBBLE CLEARED (SEC)
I FIG. 3- 6
; TYPICAL AUTOMATIC DEPRESSURIZATION SYSTEM ACTUATION FORCING FUNCTION FOR RAMSHEAD DEVICE-ZONE 14 l SHOREHAM NUCLEAR POWER STATION-UNIT 1 i PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS i
REVISION 4- FEBRUARY 1981
N
,.. h 1607* d **
136 g l CONTAINMENT ,j
\
O
% [$'UcEnr sav>
_ ,. ,0,.,go _..
~ can .#. .
(aoEEEsav> b gi ,8 ' 908 2708 l , I O NOTE NOT TO SCALE FIG. 3-7 ORIENTATION OF SRV LINE DISCHARGE DEVICES (RAMSHEADS)FOR ASYMMETRIC SRV DISCHARGE SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4 - FEBRUARY 1981
4 O O O 12.0 0
.T td z I 4
j o s.oo - 1 - 9 E i Id
$ 4.00 -
a N n. N g 0.00 -
- k8 4
- 4. o .
0.00 0.10 0.20 0.30 0.40 0.50 . 0.60 0.70 0.80 TIME AFTER FIRST BUBBLE CLEARED (SEC) !
' FIG. 3- 8 .
l TYPICAL ASYMMETRIC SRV DISCHARGE FORCING ! FUNCTION FOR RAMSHEAD DEVICE -ZONE 14 , SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENTFOR SRV AND LOCA LOADS i REVISION 4 - FEBRUARY 1981_
N b isoa I I CONTAINMENT Fol3F-135' O O , h _ 34,2 N! I so.3* PEDESTAL i l O' NOTE NOT TO SCALE FIG. 3-9 ORIENTATION OF SRV LINE DISCHARGE DEVICE (RAMSHEAD) FOR SINGLE VALVE DISCHARGE O SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENTFOR SRV AND LOCA LOADS REVISION 4 - FEBRUARY 1981
O O O 4 3 - ' E n 2 ,L E e b W E 2 \ E1 - W W k 0 - l j O.0 0.1 Q2 0,5 0.4 03 OA O.7 0.8 TIME AFTER FIRST BUBBLE CLEARED (SEC) I i FIG, 3-10 TYPICAL SINGLE SRV DISCHARGE FORCING I FUNCTION FOR RAMSHEAD DEVICE-ZONE 14
.SHOREHAM NUCLEAR POWER STATION-UNIT 1 i PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS l
REVISION 4 - FEBRUARY 1981
l i p SUFFORT k .g F PEDE5TAL a-PRIMARY f6 CONTAINMENT {d ... p-1 l *-
."* SUPPRESSION POOL .',,
l 8
'.- 26 PSI *. '
i.o _ - w .- l 13 PSI ,!;T ' i o : 11 PSI
- . l . . , . / >
. . . . !c , . . *. . : [ l l
PRESSURE DISTRIBUTION o.a - o.s -- E x o.4 - O Q. o w t! s o.x - 4 2 x o - z O -
-o.x -
g4 I I I I I I O O.1 0.2 0.3 0.4 a5 0.6 TIME (SEC) FIG. 3 - 11 , NORMALIZED PRESSURE TIME HISTORY FOR O ALL SRV'S DISCHARGING SIMULTANEOUSLY AND IN PHASE BASED ON RAMSHEAD DEVICE SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVisifMi 4 -FFassaamY lest
2 -' t to - id
- g as -
E Os
- a. -
O - N 3 0.4 - a: E a2 - o 1 1 1 I I I I I I 1 i -11 0 70 -50 -30 -10 0 0 10 30 50 70 90 11 0 l CIRCUMFERENTI AL COORDINATE (DEG) i j i i J FIG. 3- 12 i- NORMALIZED PRESSURE BOUNDARY LOAD l DISTRIBUTION AROUND THE CIRCUMFERENTIAL l DIRECTION ON PRIMARY CONTAINMENT FOR THREE ADJACENT SRVh DISCHARGING SIMULTANEOUSLY AND IN PHASE BASED ON RAMSHEAD DEVICE SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DEClGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4-FEBRUAR1r 1981
i ! O O O h i i 1 i THIS FIGURE CONTAINS PROPRIETARY INFORMATION 1 1 4 j i FIG. 3-13 i KWU T-QUENCHER i SHOREHAM NUCLEAR POWER STATION-UNIT 1 j PLANTDESIGN ASSESSMENT FOR SRV AND LOCA LOADS I - j REVISION 4 - FEBRUARY 1981 1
O O . O THIS FIGURE CONTAINS PROPRIETARY INFORMATION i I 1 4 j j FIG. 3-l'4 KKB PRESSURE TRACE NO.35 SHOREHAM NUCLEAR POWER STATION-UNIT 1
- l. PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS i
REVISION 4- FEBRUARY 1981
- l 1 .
. O THIS FIGURE CONTAINS PROPRIETARY INFORMATION t
I - FIG. 5715 KKB PRESSURE TRACE NO.76 2 SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND J OCA LOADS
., s '-
REVISION 4 ' FEBRUARY 1981 e
4 1 i i l t l THIS FIGURE CONTAINS PROPRIETARY INFORMATION 4 i 4 i i i 1 l I i FIG. 3-16 KKB PRESSURE TRACE NO.82 SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4 - FEBRUARY 198l e i
O O O i I I i 'I l THIS FIGURE CONTAINS PROPRIETARY INFORMATION l l l i ! FIG.3-17
- ' NORMALIZED PRESSURE DISTRIBUTION ON SUPPRESSION POOL BOUNDARIES-
- SYMMETRIC AND ADS CASES
- SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4- FEBRUARY 1981 i
o i 1 l O THIS FIGURE CONTAINS PROPRIETARY INFORM ATION l l l l i l l FIG. 3-18 NORMALIZED VERTICAL PRESSURE DISTRIBUTION Os FOR ALL CASES AND FOR SUBMERGED STRUCTURES SHOREHAM NUCLE AR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REV1SiON 4 - FEBRUARY 1981 l i
O i l f i l l l l i I O THIS FIGURE CONTAINS PROPRIETARY INFORMATION FIG.'3-19 NORMAll2ED PRESSURE DISTRIBUTION , O ON SUPPRESSION POOL BOUNDARIES-ASYMMETRIC CASE SHOREHAM NUCLEAR POWER STATION-UNIT t PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REV1810N 5-DECEM8ER 1988
f O O V THIS FIGURE CONTAINS PROPRIETARY INFORMATION 4 F10. 3-20 NORMALIZED PRESSURE DISTRIBUTION O oa sueeassaios roo' =ouaoaaiss-SINGLE SRV DISCHARGE CASE SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 5-DECEMSEA 1981 __ _ . . . _ _ _ _ _ _ _ _ . , _ _ . . _ _ , . . _ ~ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ _ _ . _ _ . . _ . _ _ _ _ _ . , , . _.. _ _ _ _ _
O . THIS FIGURE CONTAINS PROPRIETARY INFORMATION FIG. 3-21 LOADS ON QUENCHER O' WITHOUT PRESSURE LOADS SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4- FEBRUARY 1931
- -. _ W , -
O l I i O THIS FIGU RE CONTAINS PROPRIETARY INFORMATION i FIG. 3-22 LOADS ON QUENCHER ARMS WITHOUT PRESSURE LOADS O SHOREHAM NUCLE AR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4- FEBRUARY 194[
~
\ _ ___ _
SECTION 4 LOCA LOADS 4.0 GENERAL A general description of the loss-of-coolant accident (LOCA) is given in Section 2.1.2. The sequence of events described in Section 2.1.2 causes direct dynamic loading of the structures and components comprising the vapor suppression portion of the Mark II (MK-II) containment system (suppression chamber boundaries,
, downcomer vents, and other piping and structures within the suppression chamber). The transient nature of these forces is generally termed the dynamic forcing function for the loading condition in question. The bases for specifying forcing ,
functions for given loading conditions are presented in Section 4.1. Plant-unique input and calculations of plant-unique loads for Shoreham are presented in Section 4.2. Specifications of other LOCA related loads including the pressure and temperature loads on the containment pressure boundary; downward and upward differential pressure loads on the drywell' floor resulting from maximum vent flow and condensation of steam in the drywell; pressurization of the annulus formed by the reactor vessel and : the biological shield wall and seismic loads due to the assumed design basis earthquake concurrent with the design basis accident (DBA) are presented in the USAR. 4.1 ANALYTICAL METHODS AND DISCUSSION OF LOADS As outlined in Section 2.1.2, there are five. periods of interest during a LOCA transient related to direct dynamic loading of
~
the vapor suppression system. These are: vent clearing, air bubble formation, bulk pool swell and fallback, quasi-steady vent flow, and chugging. Table 4-1 describes the structures directly affected during each phase and identifies the . specific section where the load is defined for Shoreham. 4.1.1 Vent clearina Vent clearing loads result from the accelerating water _ jet appearing at the vent exit during the vent clearing process.. The jet loads may act on submerged structures located near the downward projection of the downcomer vent and on the basemat. 4.1.1.1 Submerced structure Loads Due to Vent'Clearina l The vent clearing jet acts on structures in the jet path either by drag or by direct momentum transfer. The generic approach' for calculation of vent clearing jet loads is provided in Reference l 17 and is based on the jet model described in Reference 16.- l Criterion III.A.1 of . Appendix D to Reference 1 raises concern about the method described in Reference 17 in two major areas: ( consideration of acceleration drag and the calculation of submerged structure loads outside the jet but near the jet 4-1 Revision 6 - December 1986 l
boundary. The latter concern arises from the realization that some motion may be imparted to the pool mass by the vent clearing jet which is not presently included in the jet model. To l overcome this modeling limitation, work has been completed on an improved jet model which includes a treatment of the significant kinetic energy imparted to the pool mass by vortex ring formation at the vent exit. High-speed movies of the vent clearing process in sub-scale, single cell dye tests have shown that energy dissipated from the jet in the formation of this vortex ring greatly reduces the degree of penetration into the pool when compared to the predictions of Reference 16. Althcugh the lead plants support the generic ring vortex as part of the MK-II long term program, expediency dictates adoption of the position outlined in Reference 1, Criterion III.A.1, parts (b) and (c) for consideration of induced flow loads. Acceleration drag will be modified as described in Criterion III.A.1, part (a). Further discussion is provided in Appendix K. 4.1.1.2 Basemat Loads Due to Vent Clearina The vent clearing a jet acts on the basemat by direct momentum transfer. For a maximum jet velocity of 60 fps at the source, the maximum impingement pressure with the basemat 10 ft below the vent exit is shown to be 33 psid using highly conservative methods as outlined in Section 4.4.5.1 of Reference 3. Section III.B.2 of Reference 1 accepted the 33 psi overpressure, but inferred that the MK-II generic position has been to apply the overpressure uniformly below the vent exit and then linearly attenuate the overpressure to zero at the pool surface. This is not a correct interpretation of the generic approach which calls for application of the overpressure to the basemat only. This issue has been resolved in Section II.A.1 of Reference 2 which states that a 24 psi overpressure is acceptable based on a review of 4T test data. The overpressure is applied uniformly to the basemat and to the pool boundaries up to the vent exit elevation, then linearly attenuated to zero at the pool surface. The 24 psi overpressure is acceptable as long as (mhL)/(AP/(AV) Vow ) 1 55 Btu /ft2-soc where: i m = Vent mass flow - lbm/sec l h = vent enthalpy - stu/lbm L = vent submergence -ft Vow = drywell volume - ft$ AP/AV = pool area / vent area ratio
- For plants exceeding the limit, the 29 psi overpressure is increased by
- 0.27 [ (mhL)/(AP/AV) Vow-55) psi.
4.1.2 LOCA Bubble Formation i once the water in the downcomer vents is expelled, drywell air j vent flow begins to form bubbles at the vent exits. The bubble j 4-2 Revision 5 - December 1981 1
. - ____ . _ _ . __ __ __ _ ~_ ___._. _ _ . _ - __ ___
1 center is assumed to be stationary on the vent centerline, one O vent radius below the vent exit. the drywell, it dimensional acceleration and velocity fields in continues to As the bubble is charged expand establishing the pool. from three-The motion of the pool water results in drag loads on small submerged structures. At the.same time, the air bubble pressure adds to i the local hydrostatic pressure on the pool boundaries. The end 1 of the LOCA bubble formation phase is referred to as the " switch j time" as described in Reference 17. The " switch time" is the time the spherical bubbles first touch each other or the pool boundary. 4.1.2.1 Submerced Structure Loads Due to LOCA Bubble Formation The generic methodology for the calculation of submerged structure loads due to LOCA bubble formation is given in References 17 and 18. The bubble dynamics and charging relationships may be coupled or uncouple'd at the option of the
- user. Criterion III.B.1 of Appendix D to Reference 1 calls for
, additional margin to be applied to the generic methodology in the i areas of bubble asymmetry, steady versus unsteady drag
- coefficients, superficial versus maximum local velocities and l accelerations and interference effects. The lead plant generic position on each of these items is given in Reference 19 as modified by Section II.C.2 of Reference 2. Further discussion is provided in Appendix K.
() 4.1.2.2 Pool Boundary Loads Due to LOCA Bubble Formation
- The bubble formation phase of pool swell is considered part of l- bulk pool swell for purposes of pool. boundary load assessment.
Pool boundary loads due to pool swell are discussed in Section 4.1.3.6. 4.1.3 Pool Swell and Fallback i In general, structures and components within the suppression chamber including piping, valves, piping supports, platforms, and the downcomer vent bracing system may be subjected to impact, drag, and fallback loads resulting from bulk pool swell. Bulk
- -pool swell is the term describing the upward movement of the
- suppression pool water above the exit plane of-the vents due to
- the injection of drywell air beneath the pool surface. The pool swell phenomenon occurs immediately after vent clearing and the j formation of air bubbles at the vent exits following a large
- LOCA. As air flow continues from the drywell, the bubbles expand i
and coalesce. At the time the bubbles contact one another or the . pool boundaries, the pool motion becomes essentially one-l dimensionally upward as described in Reference 17. The continued ! expansion of the air forces the slug of water above- it to accelerate and rise upward causing the pool swell. The velocity. of the pool surface associated with this phenomenon causes impact-and drag forces to be exerted on structures which are within the ,\ swell zone. Pool swell is eventually terminated due to the compression of air in the suppression chamber freespace and the 4-3 Revision 5 - December 1981 '
negative acceleration due to gravity. Once the swell is terminated, communication is established between the air bubble and the air compressed in the suppression chamber freespace during a relatively temperate process (as compared to the breakthrough characteristics of the Mark III containment system) without generation of any significant froth. A pool swell analytical model (PSAM) used to predict bounding pool swell velocity and acceleration transients for MK-II containments is detailed in Section 4 of Reference 20 and shown schematically on Fig. 4-1. A general description of the computer code used to implement the model is included in Appendix H along with the results of benchmark problems for comparison with the results of the three classes of MK-II containments provided in Section 4.4.4 of Reference 3. These problems are used to verify the proper operation of the code. Qualification of the PSAM itself is provided by comparison with test data in Section 6 of Reference 20 and in Reference 21. Criterion I.A.2 of Appendix D to Reference 1 calls for the application of a 10 percent margin to pool swell velocities calculated using the approach outlined in Section 6.7 of Reference 20. The 10 percent margin has been applied. As noted in Section 1.2 of Reference 22 and Section 5 of Reference 20, the maximum pool swell height observed in the 4T Test Program for values of initial vent submergence and drywell charging rates characteristic of MK-II containment systems is less than 1.5 times the initial vent submergence, However, Run 31, a minimum vent area, minimum submergence, maximum blowdown case where saturated liquid rather than saturated steam was initially discharged due to a water slug left inadvertently in the blowdown line, shows a swell height on tne order of 1.6 times the initial vent submergence. Because of this particular run and because a very minimal amount of splashing was observed several feet above the bulk pool swell height in the Run 29 conductivity probe data submitted in response to NRC Question 020.46, Criterion I.A.1 of Reference 1 calls for the use of the PSAM l described above to specify swell height (with a lower limit of l 1.5 vent submergence). The use of the PSAM to specify loading criteria for events occurring near the end of the transient represents excessive conservatism due primarily to the assumption of constant slug mass during the bulk swell process. A second alternative method proposed by the MK-II Owners Group for specifying maximum swell i height has been accepted by the NRC as outlined in Section III.A.2 of Reference 2. This alternative method retains l 1.5 times vent submergence as the lower limit on maximum swell height, but bases the maximum swell height specification on a conservatively high estimate of wetwell air compression. As noted above, froth loading is expected to be insignificant in a Mark II containment, and no load is specified. However, it is prudent to avoid locating structures potentially sensitive to 4-4 Revision 5 - December 1981
j y, - ) . froth loading in the region. immediately above te maximum specified swell height to ensure negligible effects from froth 4 Os loading.
- Loads resulting from pool' swell include impact on structures above the initial elevation of the pool. Wut below the maxidum swell height, drag on structures and grating', and drag loads 'due
, to fallback between the maximum swell height and the elevation of ! the vent exit plane. The impact force on,a.s hody occurs over a time period, t, and typically the force :versus time profile during this period is such that the force increases to a maximum value during the first half of the time peri.od and then decreases to the value of the drag force duringothe second half of the period assuming flow continues around the structure.' A ' typical force profile measured during PSTF tests is shown on fig. 4-33 of Reference 4. The duration, t, of the: force varies from about 7 msec for small structures to i,about 100 msec for la'rge structures. The drag forces apply during the tima the slug ..'is translating past the elevetion.of thelobstruction. } e
, , ,\ , ; 4.1.3.1 Imoact Loads on Small Structures from Pool: Swell 4 i c i i A small structure subjected to poolu well impact,.caus'es %he water to flow around it, thereby imparting drag forces in addition to the impact forces. I-beams, and othsr similar j structures having any one horizontal dimension 1cjs,thanor equal
- to 20 in. are considered small structures.
The-impact loads on small structures are calculated in becordance ' with criterion I.A.6 of Appendix D to Reference 1. ' Criterion I.A.3 of Appendix D to Rhferetice li spec'ifiesN a ! multiplier to account for the dynamic naturefol pc,ol swell s(loads , on grating. This is discussed in Section 441.3.5. Further discussion is provided in Appendix K. , y 4.1.3.2 Imoact Loads on Laroe Structurer., from Pqgl Swell q y The pool surface impact . on a large structure leads to<' higher impact loads than those which occur onismallerstargets. This is because the larger structures cause de.:elerat. ion of ths ' entire water slug whereas, in the smaller structuras, the water slpg is almost entirely diverted around the Jbodyeexcept for the 4 limited l amount of water which is decel'erated in the immediate vic'ijtity; of the impact area.
,/ , , , w' A structure is considered' large if it has<a dimension in;'a.
- horizontal plane greater than 20 in.' The f.n0mber of' large structures in MK-II containment i blimited and each'is treated on a '
case-by-case basis. I s t
- i
,;>,l p
- j. e t
; g / ;/< s i'
'!l
'. ? t'
', V, .
, i N. 1 4-5 Revision 5 - Lecarber 1981
* * " " " * " " * " ' ' ' ' - * * ' - * ' ' ' ' " ' ' * ' ' ' ~ ~ '
- 4.1.3.3 Draq Loads on the Downcomer Vents Due to Pool Swell A vertical load is imposed on the.downcomer vents due to the O upward movement of the slug of water initially above the vent exit during pool swell. In Equation (4-32) of Reference 3, the shear stress for flow along the axis of the downcomer vents is given as:
T = Cc # v2 (1) 2 g c Where C is given as 0.0023 for a geometry, viscosity char-acteristic length and velocity typical of MK-II containment. The total area to which the shear stress is applied is equal to: Ac = r Dr L (2) where D is the outside diameter of the vent and L is equal to the thickness of the water slug which is taken to be equal to the initial vent submergence. Calculation of the drag load is based on the maximum velocity for both the swell and the fallback phase and is calculated in Section 4.2.3. 4.1.3.4 Draq Loads on Structures Other Than Downcomer Vents Due to Pool Swell Reference 17 describes the method used to-calculate drag loads on structures within the swell zone with a projection onto the plane of the pool surface. Some of the concerns raised by Criterion III.B.1 of Appendix D to Reference 1 with respect to the methodology of Reference 17 apply here as well as in Section 4.1.2.1. Resolution is provided in References 2 and 19. One concern not applicable to Section 4.1.2.1 and not covered in Reference 19 is blockage effects due to dovncomer bracing (Criterion III.B.l(e)). The MK-II Owners Group finds this criterion acceptable. 4.1.3.5 Loads on Grating Due to Pool Swell Based on Section 4.4.6.4 of Reference 3, the drag loads on grating are calculated using the expression: Fo = Pa Aa (3) where: Po = pressure differential across the grating given on Fig. 4-2, and Ao = solid area of the grating (for grating with open area less than 60 percent) or Ao = total area of the grating (for grating with open area equal to or greater than 60 percent) lh 4-6 Revision 5 - December 1981
The results,given on Fig. 4-2 are based on a velocity cf 40 fps. The duration of the load is taken to be 0.5 sec, beginning () with the time the pool surface reaches the elevation of the grating. . To acco'unt for the dynamic nature of the initial load application the F is increased by the f actor 1 + V1 + (0.0064 Wf)2 for Wf i <2,000 in./sec where W is the width of the grating bars in inches and f is the natural frequency of the lowest mode of bar vibration. Application of this factor brings the load specification for grating into compliance with Criterion I.A.3 of Appendix D to Reference 1. i 4.1.3.6 Succression Chamber Boundary Loads Durina Pool Swell The pressure developed in the air bubble (air vented into the suppression pool) will cause additional loading on .the suppression chamber walls. At the same time, the suppression' chamber boundaries above the instantaneous pool surface are loaded by the increase in suppression chamber airspace pressure. The i maximum value of these pressures can be obtained from the pool swell analysis and are applied statically-as uniform increases in the suppression pool. hydrostatic pressure. In Section 5.2 of 4 Reference 22, this is shown to be conservative. The loading . condition is shown conceptually on Fig. 4-3. I B Criterion I.A.S of Appendix D to Reference 1 requests -an f asymmetric bubble pressure loading case. This criterion, ,as modified by- Section II.A.3 of Reference 2, requires that an d-O asymmetric pool boundary load equal to 20 percent of the maximum-t air bubble pressure be statically applied. The MK-II Owners Group finds this acceptable. Section III.B.3.a.1 of Reference 1~ establishes that the wetwel1~
- pressure transient may be taken ' from the PSAM. .Section II.A.2 of -
2 Reference 2 further establishes that ,the wetwell pressure may be limited to that consistent with .the- drywell floor uplift . specification; 4.1.3.7 Drvwell Floor L'oads Due to Pool Swell , The drywell floor is;the- structure that separates the drywell'.and the wetwell (suppression chamber). It is referred to as J the diaphragm floor in Reference' 3. At the end of pocl swell the potential exists for an uplift -differential pressure. to,'act on j the drywell floor due. .to the increase in; suppression -chamber
- airspace pressure during the final stages of bu?k pool swell..
> :)
- The net. upward' load on the drywell
~
floor was shown in the 4T Test Program to be less than 2.5 paid. Table 2.1 'of, Reference 22 calls for the . application 'of a 2.5 paid upliftt pressureias a1 bounding condition. No net downward load is specified 'as a result of pool ' dynamic considerations.- j() Criterion I.A24 modification to- the above of Appendix ~~D t
~'
to Reference specification whereby 1 calls for the. uplift:5\- a. . 4-7 Revision S - December 1981: I 4
) ,l r >
- - os..o - m- -
differential pressure is equal to 8.2 - 44.(F) psi if F 10.13 and 2.5 psi if F >0.13 where F= (break area) (pool free surface area) (suppression chamber airspace volume)/(drywell volume) (vent area)2 The MK-II owners Group has found this alternate specification acceptable. Note that this criterion reveals that the worst uplift differential pressure may not occur at the largest break size; i.e., as F decreases the uplift differential pressure increases. The Mark II owner's Group has proposed a maximum uplift differential pressure independent of break size equal to 5.5 psid which the NRC staff has accepted in Reference 31. As noted previously, there is no significant froth generated in the MK-II containment system as a result of pool swell. Reference 22 does not specify a froth impingement load on the drywell floor. 4.1.3.8 Fallback Loads There is no pressure increase in the suppression pool boundary during the pool fallback. The structures within the suppression pool will experience drag forces as determined by Fig. 4-4 and Table 4-2. The fluid density is taken to be that of water although it is now a two-phase mixture of air and water. The velocity during fallback is computed by assuming acceleration by gravity as follows: Vra = 9.83s/H2 (4) which is taken from Section 4.4.5.4 of Reference 3 where H is the initial vent submergence in feet and V is the terminal fallback velocity in feet per second. Equation (4) is based on a swell height of 1.5 vent submergence. For a swell height of any multiple of vent submergence: i Vra = 8.03 s/REo (5) l l where R is equal to swell height / vent submergence. The terminal l fallback velocity is used to compute all drag forces on structures within the swell zone. The duration of the fallback phase is based on the average fallback velocity. 4.1.4 Ouasi-Steadv Vent Flow After the end of pool swell for a large LOCA or during the initial phase of an intermediate LOCA, there is a period of quasi-steady vent flow. This regime is characterized by drag forces on the downcomer vents acting downward and condensation oscillation loads on the pool boundaries. As described in Section 6.0 and Table 2.1 of Reference 22, no ! significant lateral loads on the downcomer vents due to high-and medium-mass flow condensation have been observed in the 4T Test Program and none have been specified in Reference 3. 4-8 Revision 5 - December 1981 l
-- . _= . .
4.1.4.1 Vertical Loads on the Downcomer Vents Due to Viscous and Pressure Forces of Vent Flow i Section 4.2.3 of Reference 3 describes the method for calculating
, the vertical load on the downcomer vent due to viscous and ! pressure forces resulting from the two phase, two component vent flow. Equation (4-2) of Reference 3 is as follows:
Fror = Cro Io LGA (6) 2pr Dgo i where: i Fror = total force on the vent under consideration Cro - friction factor for all liquid flow Io = two - phase multiplier from Fig. 4-10 of Refer-ence 3 (given as a function of pressure and quality) L = length of the downcomer vent j G = mass flux in the vent
/~ A = flow area of the vent V} #c = density of the liquid phase in the vent 4
D = diameter of the vent pipe go = gravitational constant ! The above equation neglects the load due to unbalanced pressure forces on the pipe which- is shown in Reference 3 to be negligible. Fror itself is small when compared to the lateral loads. 4.1.4.2 Pool Boundary Loads Due to Condensation Oscillations As a result of the 4T Test Program, high and medium mass flux + condensation oscillation loads have been defined. In Section 6.1 of Reference 22, the specified pool boundary load for high mass flux condensation is 4.4 psi peak-to-peak amplitude (PPA) with a frequency content of 2 to 7 Hz for a 24 in. downcomer vent. A sinusoidal wave form should be assumed over the entire submerged pool boundary. As noted in Section 6.0 of Reference 22, condensation loads may begin as early as t = 4.0 see following the DBA. 1 4-9 Revision 5 - December 1981
When the mass flux decreases to approximately 11 to 12 lbm/ft2-sec as discussed in Section 6.2 of Reference 22, the condensation oscillations change amplitude and frequency content. The pool boundary load specified in ~the above reference for medium mass flux condensation is 7.5 psi PPA with a frequency content of 2 to 7 Hz. This load is applied as above, with a sinusoidal wave form over the entire submerged pool boundary. 4.1.4.3 Submerged Structure Loads Due to Condensation Oscillations There is no generic methodology for calculating submerged structure loads due to condensation oscillations. The method developed by Stone & Webster for Shoreham is described in Appendix K. 4.1.5 Chuqqing Chugging is the term applied to the intermittent condensation of steam at the vent exit which occurs after the air has been essentially purged from the drywell and the vent steam mass flux has fallen below a threshold value of 4 to 6 lbm/ft2-sec (section 5.5.2, Reference 23). Chugging is characterized by the collapse of the steam bubble at the vent exit which results in the generation of loads on the downcomer vent and on the pool boundary. Pressure fields generated in the pool may also result in loads applied to submerged structures. 4.1.5.1 Lateral Loads on Downcomer Vents Due to Chugging As described in Reference 22 and further discusr,ed in Reference 14, a review of the 4T Test Program results has indicated that the equivalent static load on any single downcomer vent will be l less than 3,000 lb. Section 4.3.2.3 of Reference 4 and Table 2.1 i of Reference 22 call for the application of an 8,800 lb equivalent static load for the assessment of lateral loads on the vents due to chugging. In analyzing a particular structural element, the effects of be considered. multiple vent chugging may have to Once the number of vents affecting the loading of a particular element is defined, the appropriate load from Fig. 4-10a of Reference 4 can be applied to each downcomer vent in the direction which maximizes the combined load on that element. Figure 4-10a of Reference 4 is based on a probability of 10 4 that the loading condition specified would be exceeded in 265 chugs. Criterion I.B.1(b) of Appendix D to Reference 1 calls for increasing the equivalent static load of 8,800 lb by the ratio of downcomers natural frequency to 7 Hz for 24 in. vents with natural frequencies between 7 and 14 Hz and bracing (if any) located at least 8 feet above the vent exit. This requirement will be met by the MK-II containment lead plants. 4-10 Revision 5 - December 1981
. - _.m__.-. _._ ~ _ _ - - _- -._ .__ __ _ _=__ _ - __.___ _
1 A dynamic analysis of downcomer response to chugging is provided >
- in Appendix L as required by Criterion I.B.1(c) of Appendix D to i , Reference 1. The load definition included in criterion I.B.l(c)
! is not the load definition established by the long term program. The MK-II Program design basis for dynamic vent lateral loads due 1 to chugging is that given in Reference 24 (see Fig. 4-5). For the static load definition, Criterion I.B.2 of Appendix D to Reference 1 calls for application of a multiplier of 1.26 to multivent loads taken from Fig. 4-10a of Reference 4 in addition ! to the frequency ratio multiplier mentioned above. This l requirement is acceptable to the MK-II containment lead plants. i For the dynamic load definition, the NRC staff requires an
- increase of the maximum amplitude of the single vent forcing i function from 30 to 65 Kips (81). For the multivent dynamic j load (21) states that the multivent reduction factor proposed by j the MK-II Owner's Group is acceptable and may be based on the
- original 30 Kip maximum amplitude. The increase in the single !
vent lateral load is acceptable to the MK-II Owner's Group. 4 4.1.5.2 Pool Boundary Loads Due to Chuoaino As suggested in Reference 25 and further discussed in Reference 14, the pool boundaries are subjected to a uniform loading with a ,
; pressure intensity of +4.8/-4.0 psi and an asymmetric loading condition with a maximum pressure of +20/-14 psi. The asymmetric l circumferential variation in pressure is shown on Fig. 4-6 which
! is taken from Reference 25. , The. chugging data presented in Reference 25 are for 20 and 30 Hz, ' respectively. The load is assumed to be applied uniformly below
- the elevation of the vent exit plane and then decrease linearly
- to zero at the pool surface.
i 4.1.5.3 Submerced Structure Loads Due to Chuaaina
- There is no generic methodology for calculating submerged .
structure loads due to chugging. The method developed by Stone &' j j Webster for shoreham is described in Appendix K. 4.2 SHOREHAM PLANT SPECIFIC LOADS AND RESPONSE CONDITIONS The Shoreham specific loads resulting from a LOCA are discussed ! in this section. Reference is _made to the lead plant generic
- load specification methods outlined in Section 4.1.
9 y 4.2.1 Vent Clearina l The generic approach to calculating submerged structure and basemat loads due to vent clearing is given in Sections 4.1.1.1 i and 4.1.1.2, respectively. Shoreham will follow the lead plant j generic approach. Shoreham has completed the assessment- of
- basemat loading during vent clearing using the water jet- clearing i velocity presented on Fig. 4-7. The maximum velocity is about 60 fps. Using this value, the maximum pressure at the basemat would l l
4-11 Revision 6 - December 1986
,,e-ee e,-a-+vm --.--,,ar-- ,~~n-,------,e -w,, ,---.-,-vwm,w---,- -,,-+-w-a~,w,,--ee- ~~n,--,-wpg m-, e nm v-
be < 33 psi as described in Section 4.1.1.2. For added conservatism, however, an impingement pressure calculated using the unattenuated vent exit velocity was applied to the Shoreham basemat as shown on Fig. 5-30. This forcing function was applied to the basemat only. The Shoreham design has also been assessed for the statically applied 24 psi overpressure applied to the basemat and containment walls below the vent exit (attenuated linearly to zero at the pool surface) as described in Section 4.1.1.2. The value of (mhL)/(AP/AV) Von for Shoreham is less than 55.0 Btu /ft -sec. 4.2.2 LOCA Bubble Formation Shoreham has completed the assessment of submerged structure loads due to LOCA bubble formation using the " coupled" option of Reference 17. Shoreham has followed the lead plant generic resolution of issues raised by Criterion III.B.1 of Appendix D to Reference 1 as described in Section 4.1.2. 4.2.3 Pool Swell and Fallback Pool swell is analyzed as described in Section 4.1.3. The plant parameters used in the analysis are presented in Table 4-3. Table 4-4 presents the short term pressure transient of the containment for a DBA calculated using the LOCTVS computer code (26) . This short term pressure transient includes the effect of inventory in the broken recirculation suction line. For comparison, the drywell pressure predicted by NEDM-10320 with subcooled Moody slip flow is included. This comparison demonstrates the conservatism of the LOCTVS model. Figure 4-8 shows the variation of bubble pressure and suppression chamber pressure with time up to maximum swell height. After maximum swell height has been reached, the suppression chamber pressure is taken to be that from the containment analysis transient presented on USAR Fig. 6.2.1-6. [ Figure 4-9 gives pool surface elevation and velocity as a I function of time following the DBA. A margin of 10 percent has been applied to the velocity when used for design assessment. Vent clearing occurs at t = 0.53 and 0.49 sec and maximum swell height occurs at t = 1.24 and 1.20 sec for high and low water level respectively, in the suppression pool. All results presented on Figs. 4-8 and 9 are for vent submergence of 9 and 8 feet for high and low water level in the suppression pool. However, for structures near the pool surface, the minimum submergence case (8 feet) produces the highest velocity and is used for design assessment. l Structures within the suppression chamber may be divided into two groups, those above the initial pool surface (plant el 26-0 to 27-0) and those below the initial pool surface. Structures above the initial pool surface may be subjected to impact as well as drag loads, while those below the initial pool surface are ~ subject only to drag. In general, all loads discussed in this section occur within the pool swell zone, defined as the region above the vent exit plane (plant el 18-0) and below the maximum 4-12 Revision 6 - December 1986
v swell height. Shoreham has no structures potentially sensitive to froth loading in the region immediately above maximum O height. swell The maximum swell height for Shoreham is less than 2.2 vent submer- I gence (plant el 47-0). This value was obtained by the alternate I method described in Section 4.1.3 (wetwell airspace compression approach). The plant specific calculation for Shoreham is presented in the plant unique response to NRC Question 020.68 (Appendix D). 4.2.3.1 Impact Loads on Small structures Due to Pool Swell The impact loads on small structures located above tre initial pool surface have been calculated as described in Section 4.1.3.1. In Section 4.1.3.1 a small structure is defined as a pipe, an I-beam, or other similar structure having one horizontal dimension less than or equal to 20 in. For wedge-like flow deflectors, the methods of Reference 32 have been used to confirm the impact load calculations performed using Reference 33. 4.2.3.2 Impact Loads on Large Structures Due to Pool Swell A structure is considered large if it has a dimension in a horizontal plane greater than 20 in. There are no large structures within the Shoreham swell zone. 4.2.3.3 Drao Loads on the Downcomer Vents Due to Pool Swell As described in Section 4.1.3.3, the shear stress on each downcomer vent during pool swell is calculated using Equation (1) and the maximum pool swell velocity of 39.4 ft/sec (see Fig. 4-9). Then, with a 10 percent margin applied to velocity: l Te = (0.0023) (62.4) (39.4)2(1.1)2 = 4.19 lbf (2) (32.2) fta The area over which the force acts is: Ac = r De L = r (2.0) (9.0) = 56.5 ft2 The upward drag force per downcomer vent during pool swell is: Fe = Ac Te = 244 lb = 0.24 kips To calculate the drag load during fallback, it is necessary to know the fallback terminal velocity. From Equation (5), the terminal fallback velocity is: Vr. = 8.03f(2.2) (9.0) = 35.7 ft/sec O 4-13 nevision 6 - December 1986-
For fallback then: Tc = (0.0023) (62.4) (35.7)2 = 2.85 lbf (2)(32.2) ft2 For the same A , the downward force acting on each downcomer vent during fa'iloack is then: Fr = Ar Te = 160.8 lbf = 0.16 kips For reference, the static weight of each downcomer vent is approximately 4.3 kips. vertical drag on the downcomer vents due to pool swell and fallback is insignificant compared to other loads. 4.2.3.4 Draq Loads on Structures other than Downcomer Vents Due to Pool Swell Shoreham has completed the assessment of drag loads due to pool swell using the approach outlined in Section 4.1.3.4 and Reference 17. Shoreham has followed the lead plant generic resolution of issues raised by Criterion III.B.1 of Appendix D to Reference 1. Appendix K describes in detail the manner in which the Shoreham assessment of pool swell drag loads on structures other than the downcomer vents has been conducted. In particular, compliance with criterion III.B.l(e) of Appendix D to Reference 1 has been , demonstrated. This criterion, which was not assessed in l Reference 19, covers the effects of lockage on the drag coefficient of the downcomer vent bracing system. 4.2.3.5 Loads on Gratino Due to Pool Swell l There is no grating in the pool swell zone. 4.2.3.6 Sunoression Chamber Boundary Loads Due to Pool Swell As discussed in Section 4.1.3.6, the lead plant position is to apply the pressure of the air bubble and the wetwell airspace statically as a uniform increase in the suppression pool hydrostatic pressure from the basemat to the vent exit and above the maximum swell height, respectively. For Shoreham, the air bubble and wetwell pressures are obtained from Fig. 4-1. Note that the maximum wetwell pressure is taken to be the instantaneous drywell pressure plus the maximum uplift differential pressure of 2.5 psid from Section 4.2.3.7. Pool swell boundary loads are not applied to the pedestal since there are downcomer vents within the pedestal and the ratio of pool area to vent area for the region outside the pedestal is not suffficiently different from that inside the pedestal to indicate that a significant difference in bubble pressure would exist during pool swell. 4-14 Revision 5 - December 1981
The asymmetric case called for by Criterion I.A.5 of Appendix D (~% to Reference 1 and Section II.A.3 of Reference 2 has been ( ,) included in the Shoreham design assessment as well as the uniform case described above. 4.2.3.7 Drvwell Floor Loads Due to Pool Swell As discussed in Section 4.1.3.7, a 5.5 psid uplift pressure is l applied to the drywell floor during pool swell for purposes of design assessment. Since the DBA "F-factor" for Shoreham (that corresponding to the largest break) is greater than 0.13, a 2.5 psid uplift differential pressure meets the requirements of Criterion I.A.4 of Appendix D to Reference 1 and NRC 020.69 (Appendix D). The DBA "F-factor" and the 2.5 psid Question uplift pressure appropriate for use in calculating maximum wetwell air space compression and swell height since both of these design parameters are maximized by large breaks. The long term depressurization of the drywell and consequent upward load on the drywell floor is mitigated by the vacuum breakers as discussed in the USAR. The drywell floor is also designed to react to the impact and drag loads or the vent bracing during pool swell. 4.2.3.8 Fallback Loads Fallback loads are calculated as described in Section 4.1.3.8. In Section 4.2.3.3, the terminal fallback velocity for Shoreham was shown to be a maximum of 35.7 ft/sec. The duration of the fallback phase of pool swell using the average velocity f~'-) x- ft per sec and the pool swell height of 20.0 ft is 1.12 sec. of 17.9 Post pool swell wave loads associated with fallback are considered to be less severe than the seisnic sloshing condition described in Section 2.3. During seismic slosh, the pool moves as a coherent mass of high density and is, therefore, capable of generating larger loads than the highly aerated pool present immediately after pool swell. There is no generic load specification for post pool swell waves. 4.2.4 Ouasi-Steadv Vent Flow For drag on the downcomer vents and condensation oscillation loads on the pool boundaries during quasi-steady vent flow, Shoreham has used the lead plant generic methodology described in Section 4.1.4. Submerged structure loads due to condensation oscillations are described in Appendix K. Confirmation of the design basis loads for MK-II program long-term loads accepted by Reference 31 is provided in Appendix L. 4.2.4.1 Vertical !!oads on the Downcomer vents Due to Vent Flow , As described in Section 4.1.4.1, the vertical load on the
'w downcomer vent due to vent flow is calculated using Equation (6),
The following parameters are used in this analysis:
'~
Ceo - 0.02 1 4-15 Revision 5 - December 1981 l l
I co = 150 (based on a quality of 33 percent) L = 45 ft h G = 133.0 lbm/ft2.sec (maximum) A = 2.95 ft2 Pe = 62.4 lbm/ft8 D = 1.94 ft go = 32.2 ft/sec a The resultant force is 0.90 kips on each downcomer vent. This load is small compared to the weight of the downcomer and will not be discussed further. 4.2.4.2 Pool Boundary Loads Due to Condensation Oscillation The condensation oscillation forcing functions for design assessment of the pool boundary are specified in Section 4.1.4.2. High mass flux condensation is assumed to begin at t - 4.0 sec following the DBA. At t = 20.0 sec following the DBA, the steam mass flux approaches 11.0 lbm/ft2-sec (threshold value for transition to medium mass flux condensation) and medium mass flux condensation is assumed to begin. At t = 25.0 sec following the DBA, the steam mass flux approacnes 6.0 lbm/fta-sec and chugging is assumed to begin as described in Sections 4.1.5 and 4.2.5. 4.2.5 Chuquino Design basis chugging loads for Shoreham are defined below. Confirmation of the design basis loads for the MK-II program long-term loads accepted by Reference 31 is provided in Appendix L. 4.2.5.1 Lateral Loads on Downcomer Vents Due to Chuagina The assessment of the Shoreham downcomer vents and vent bracing system has employed the generic methods described in Section 4.1.5.1. In the summer of 1979, the Shoreham vent bracing was lowered from plant el 33-0 to el 27-9. Prior to this design change, a satisfactory static analysis had been completed for both the single and multivent cases. Since the design change the static analysis has been reevaluated for the single vent case, and a dynamic single vent analysis has also been performed. Because of the complexity of multivent analysis, it was decided that only a dynamic multivent analysis would be performed. This work has been completed. The analysis is in compliance with the requirements of Reference 31. O 4-16 Revision 5 - December 1981
4.2.5.2 Pool Boundary Loads Due to Chuacino Section 4.1.5.2 specifies the chugging forcing function to be used for assessment of the pool boundary.
)
4.2.5.3 Submerced Structures For submerged structure loads due to chugging, a plant unique approach has been employed as described in Appendix K and the response to NRC Question 020.75 (Appendix D). O i i i 4 l e 4-17 Revision 5 - December 1981
-.___, . . . , _ - , _ , , , _ . _ . ~ _ - . , , . _ , , m- ,. -c-.e._.,w,---,.,, _ , , , . . , _ __..,-3._,
.._,_,.__.y_ . . . _ _ , _ . . _ , _ . , , _, , , , . . - _ _ , - _ . , . . . . . , . _ _ ,
- _ . .m___ ._ _ _ _ ! TABI.E 4-1
SUMMARY
OF IDCA AFFE.CTED STRUCTURES 1 , Type of Loadina Conditions . Structures Drag Impact Suppress 1ast j 2xperiencing Steam Steam Water Air Bubble from Due to Cnamber Air LOCA Loads Condensation _ Flow Jet Pressure Pool Spell Pool Swell rallcact Compressicq i. Structures below 4.2.4 4.2.1 4.2.2 4.- asol surf ace 4.2.5 4.2.3.4 , Structures above Pool surface 4.2.3.4 4.2.3.1 4.2.3.s 4 .2 . 3.2 Downcomer Vents 4.2.5.1 4.2.4.1 4.2.3.3 4.2.3.3 Drywell Floor 4.2.3.7 4.2.3.7 ! Containment Wall 4.2.4.2 4.2.1 4.2.3.6 4.2.3.6 , 4.2.5.2 l Pedestal 4.2.4.2
! 4.2.5.2 1
, Basemat 4.2.4.2 4.2.1 4.2.3.6
) 4.2.5.2 l i
.i M E!!: 4 14 sabers refer to sections of this report. i I 1 of 1 Revleiost 4 - February 1981 i
TABLE 4-2 DRAG COEFFICIENTS OF VARIOUS SHAPES i Body Shade V Co Reynolds Number Circular Tube : Q 1.2 10* to 1.5x105 Lenoth/ Width Elliptical Tube 2:1 0. 6 . 4xina Q 0.46 105 4:1 0.32 2.5x10* to 105
~
8:1 0.29 2.5x104 O 0.20 2x105 Square Tube - 2.0 3.5x104 1.6 10' to 105 4 Anale of Imoact Triangular Tubes 120a 2.0 10*
? 120a 1.72 10*
r 90a 2.15 10* r 90a 1.60 104 , _ r 60= 2.20 e 104 r 60a 1.39 104 e 30= 1.8 105
- . ( 30= 1.0 105 Semicircular' Tube e
} 2.3 4x104
=
( 1.12 4x10* O 1 of 1 Revision 3 - November 1978
TABLE 4-3 SHOREHAM DATA FOR DBA TRANSIENT AND POOL SWELL ANALYSIS (J DRYWELL
- 1. Free Air Volume 192,500 ft$
- 2. Temperature (initial) 110 F
- 3. Pressure (initial) 2.0 psig
- 4. Relative Humidity 20%
SUPPRESSION CHAMBER
- 1. Free Air Volume 138,500 ft3
, 2. Water Volume 81,385 ft3 l
- 3. Temperature (initial) 90 F
- 4. Pressure (initial) 2.0 psig l
- 5. Relative Humidity 100%
BREAK ARBA
- 1. DBA - Recirculation Line Break (with pipe 4.34 fta i inventory considered)
VENT SYSTEM
- 1. Submergence 9ft f-sg 2. Diameter (Inside) 23.25 in.
(_j 3. Number of Vents. 82(1) l ADDITIONAL DATA FOR POOL SWELL ANALYSIS
- 1. Drywell Pressure Transient Table 4-4
- 2. Suppression Pool Surface Area 4,250 fta
- 3. Vent Area 241.8 fta l
- 4. Vent Loss Coefficient (excluding exit loss) 1.0 NOTES:
(1) Six vents out of a total of 88 have been deactivated as described in Appendix M. O 1 of 1 Revision 6 - December 1986
a TABLE 4-4 ) . () DRYWELL PRESSURE AS A FUNCTION OF TIME FOR DBA (EFFECTS OF PIPE-INVENTORY INCLUDED) l NEDM-10320 Subcooled Moody Drywell Pressure j Time after DBA(sec) Slio Flow (osia) fosia) LOCTVS { 0.0 15.45 16.70 0.53 37.12 39.59 i . 0.70 39.34 41.59 0.80 - 42.67 t 0.90 41.66 43.69 l 1.01 42.80 44.75 i ? 1.20 44,58 46.45 ! I i i o i i i i { 4 i I 4 j.O i 1 of 1 Revision 6 - December 1986 1 J. _ . . , . . , _ . _ . ~ . . . _ _ _ . _ _ _ _ _ _ . _ . . - . . _ _ - , . _ _ _ _ _ _ . _ . . _ . _ . _ _ __ _ _ . , - - , , _ . _ . _ , . , _
b- sL e a- - w- ma-,--
- O PRESSURE OF SUPPRESSION AIR '
SPACE, Ps% \ \\\\\\\\\(\\\\ DOWNCOMER VENT g
\ POOL SURFACE AT
\ ANY TIME, t
~, , ~ ~ ,~~~ g ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9
\
! \ Vw t:0, INITI AL POOL
- - = -_
9 SURFACE g , O g N g - WATER SLUG ye e (s _ - - - - - - f g : AIR SLUG
-_-=z__;-__,__ m- _-=
(PRESSURE Ps)
\ /
=
\ SUPPRESSION
\ POOL NNNANNxxNNNxxxxs i
FIG 4-1 SCHEMATIC REPRESENTATION OF THE POOL SWELL MODEL O SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REYlSION 4- FEBRUARY 198I I
i .25 4 1
- 20 ;
"c i~
E _ E j is -
! i'
- z i w -
, = i w i I 5 10 - 4 W
=
3 1 m - m ' w 4 = E 5 - l .O O ' ' ' ' ' ' ' ' ' O5 0.6 0,7 0.8 0.9 f.O , OPEN AREA FR ACTION
- I 4 FIG. 4-2 l PRESSURE DROP DUE TO FLOW ACROSS GRATING l DURATION OF LOAD: 0.5 SEC l
! SHOREHAM NUCLEAR POWER STATION - UNIT 1 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS i l REVISION 4-FEBRUARY 1901 r .- .--- . 1
'O t
00WNCOMER VENT PRESSURE OF SUPPIESSION 1 1 CHAUSER AIR SPACE,Ps \\\\\\\\\\\Y\\' 8
\ MAXIMUM F POOL SURFACE N ELEVATION
}
/ \
/ \ c WATER SLUG
/ \
f \= ::::_ ::::::_-::. x- - - _v: - l r AIR SLUG (PRESSURE Ps) ! -> Pe [r i
\ , y_ _ _- : _ _- _-_-_- - -_ v_-
_ : : - -x :=
\ rSurPRESSION POOL D
\
4 __ \ l N N NN N NN N NN NNN NNT l I
~ ~ ~ - - ACTU A L ASSUMED 1
1 i { FIG. 4-3 ASSUMED PRESSURE DISTRIBUTION FOR POOL
' BOUNDARY LOADS DURING POOL SWELL l SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4-FEBRUARY 1941
_ . - _ . ~ . . , . . . . _ . _ . _ - . . _ _ . _ _ . _ . _ . _ _ . _ . . . . _ _ . , _ _ _ . - . . . _ . _ _ _ . _ _ , _ . _ . _ . _ _ - . . . . - _ . .
f ( O - so - 26 -
^
= .
c - 2 c 20 - n ,. ki o m hi ' :s ti i i M oO l N o. to - O : m - s9 o 10 - 1 Ob 6 i
~
O .1 0
' ' f i -
0 to 20 3o " 40 so I VELOCITY, y (fps) i i FIG. 4-4 DRAG PRESSURE FOR p, = 62.4(Ib m /U O SHOREH AM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVltl0N S-DECEh00ER 1981
)
///// //#
4W N K i I
- : F
! STATIC: F = 8.8 KlP qs NATURAL FREQUENCY OF DOWNCOMER OYNAMIC: CASE A. F(t) = A(t) SIN 1r t/ r FOR 04t*r Airl = (50-20 r/3ms) Ma lbf FOR 3mser 8ms Ms SPRING CONSTANT OF BRACING SYSTEM CASE S. F(t) = SSKIP a SIN w t /r re 3ms 1 4 , FIG. 4 - 5 DOWNCOMER MODEL FOR LATERAL LOAD ASSESSMENT i SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS Revision 5-DECEMBER ISSI i
- O O O l
i,
- 1 1: + 20 i
5
- + 15
] N i
l j + 1o i n.
- - A w +5 - - - - -- -- -- --- -
m 3 4 m 0 t 1 E j o. ~""~ ~ ~ - -
~~'~ ~ ~ ~ ~ ~ -
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - - -~~~
-5
~'
l
-15 V
3 08 90* 180* 270* 3E0* 3 ANGLE OF APPLICATION (DEGREES) i i l
- NOTES
i A. UNIFORM LOADING CONDITION j + 4.8 /-4.0 PSI, O' TO 3608 1 B. ASYMMETRIC LDADING CONDITION l MAXIMUM +20/-14 PSI CIRCUMFERENTI ALLY 3 ATTENU ATED AS SHOWN. ) POOL BOUNDARY CHUGGING l. DADS i SHOREHAM NUCLEAR POWER STATION-UNIT 1 i PLANT DESIGN ASSESSMENTFOR SRV AND LOCA LGADS 1 REVISION 1- APRIL 1977 i
s .f L 6 8 9 S 1 D A R O E L B M 4 A E I C C 6 TO I E 0 N L D UD - 4 Y N 8 N A 4
- T N I
C O I R V O O T A S I S L TR I V E SO E V R F R G ETN
- N W E 5
I R O P M
' S 0 AARS EBA E S LD C
EL S D ACUNA I GNG
' UN I S
QI M E 7U-WA O HD 4TL ETRN
.NLOA 0EOHL 4 1 0 FVFSP 4
b
)
S D N O C
' 3 E a
0S ( E + M I e T b = 2 0 w B.
' O 4
- - ~ - - -
O o 0 0a o M s m M 2 a, 5eC > SW. ESo s a> - e m i , ! , ' 1tI ! ,
O O O 1
; 80 i
i. l l DRYWELL PRESSURE i so _ l +2.5 PSIO DRYWELL i
$M -
_E N i l a m BUBBLE E 30 - l A-I SUPPRESSIO;J CHAMBER 20 - i ? i j jQ R f I f 8 I I f i O Q2 0.4 0.6 0.8 LO 1.2 1.4 1.6
! TIME AFTER LOCA(SECONDS) l FIG. 4 -8 CONTAINMENT PRESSURE RESPONSE DUdlNG POOL SWELL FOLLOWING DBA SHOREHAM NUCLEAR POWER STATION-UNIT i l PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS i
REVISION 6 - DECEMBER 1986 1
O O O i Ak E wU
$N
$ 40- g POOL SWELLING VELOCITY
"- /
_J >- \ 8 n. t- / \ NIGH WATER LEVEL OPERATION _j g9
/ ---- LOW WATER LEVEL g
Eo OPERATION W6
>g
, 8mn /
\
POOL SWELLING HEIGHT l 4 .J / ! 20 / /
$2 g / /
I w & j / d /
/
1 o . . . . . , ,
=
.2 .4 .6 .8 10 L2 1.4 t
j TIME AFTER LOCA (SECONDS) FIGURE.4-9 4 POOL SURFACE ELEVATION AND VELOCITY I FOLLOWING A DBA SHOREHAM NUCLEAR POWER STATION PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS I
~
REVISION S.DECEMER 1906
. . . ~ _ _ - - _ _ _ ~ _ _ - - .- . - _ . _ _ _ . - . - - . . _ - - .-- -
l l SECTION 5 () DYNAMIC RESPONSE OF PRIMARY STRUCTURES 5.1 STRUCTURAL RESPONSE TO SRV LOADS l The safety / relief valve (SRV) discharge loads have been defined in Section 3. As stated in Section 3, the ramshead discharge ( device loads consitute the Shoreham design assessment basis and bound the loads associated with the actual T-quencher (TQ) device installed in the plant. In a limited number of applications where the ramshead load definition was found to be too , conservative, building response data based on the TQ load definition was used. Therefore, loads associated with both the 3 ramshead and TQ discharge devices are. considered here. The i piping systems to which the TQ tesponse data were applied are
- discussed in Section 9.
I The following SRV load cases are considered for the ramshead j discharge device:
- 1. All valve sequential discharge All valves fire sequentially according to setpoint pressures. Individual line characteristics are accounted f for in the phasing of the bubbles. Figure 3-4 shows a l j typical pressure time-history. The maximum average j() basemat pressure is 10.0 psi.
I 2. Automatic depressurization system (ADS) discharge l ! The ADS valves fire simultaneously and the bubbles enter the pool according to line characteristics. Figure' 3-6 l
- shows a typical pressure time-history. The maximum
- ' average basemat pressure is 8.65 psi.
- 3. Three adjacent valve out of phase discharge l The three adjacent valves fire such that the bubbles i
enter the pool simultaneously but based on identical setpoints and consideration of the different line characteristics, rigure 3-8 shows a typical pressure l
- time' history. The maximum average basemat pressure' is
- 9.36 pai. ,
- 4. Single valve discharge i
A typical pressure time-history is shown on Fig. 3-10. The maximum average basemat pressure is 4.12 psi. 1
- 5. All valve simultaneous discharge a
- The bubbles enter the pool simultaneously and in phase.
; The normalized time-history and corresponding pressure l profile were shown on Fig. 3-11. Maximum basemat l 5-1 Revision 4 - February 1981
}
pressure is 26 psi. This most conservative load case has been used for the design assessment of the primary structures.
- 6. Three adjacent valve simultaneous discharge The bubbles enter the pool simultaneously and in phase.
The time-history and meridional pressure profile are the same as those of Case 5. A 60 degree circumferential variation of load was considered to be representative of three adjacent valves and was depicted on Fig. 3-12. SRV load Cases 5 and 6 described above were originally utilized in the design assessment. Reference 3 does not recognize these as credible cases, but rather specifies load Cases 1 through 4 above for the design assessment. Load Cases 1 through 4 are used to evaluate the adequacy of piping components and equipment. Load Cases 5 and 6 produce a more severe state of stress for the reinforced concrete, they are retained for the structural assessment. The actual discharge device installed in Shoreham is the T0 The following four TQ load cases are considered here:
- 1. All valve discharge
- 2. Automatic depressurization system (ADS) discharge
- 3. Three adjacent valve discharge
- 4. Single valve discharge r
As described in Section 3.3, for the interim To load definition the entry of air bubbles into the suppression pool is assumed to be simultaneous and in phase for all load cases considered. Three pressure traces from Brunsbuttel, as discussed in Section 3, are selected for plant assessment. To ensure a conservative l load definition, a pressure amplitude multiplier of 1.1 is used. The frequency content of the time histories is varied by stretching or compressing the time history traces into a longer or shorter duration by a factor varying from 1.8 to 0.9. 5.1.1 Summary of Results The containment structures were subjected to a dynamic analysis under SRV discharge loadings. The time-histories and time wise maximum values of internal loads have been conservatively determined. Acceleration time-histories and ARS were also generated. The major conclusions from the analysis are:
- 1. The dynamic response of the containment structures is oscillatory in nature and the maximum values of internal i
5-2 Revision 4 - February 1981 l l
,7-t *
y
,\
s . ' d-loads are generally comparable to tho'ee resu) ting from a small or intermediate break accident. 3; 3
- 2. The primary structural responses of< the con'tainment structures to SRV loads are transverse shears, j longitudinal bending moments, and axial forces at the junctures of the reactor pedestal and prisgry containment '
with the foundation mat. y
- 3. The maximum integrated vertical load -o' the basemat occurs in ramshead load Case 5, the all valve simultaneous discharge. ,,i 1 ,
- 4. Axisymmetric (all valves) SRV ramshead discharge loads, Case 5, result in the largest internal loads on the containment structures in the area of the suppression l i i pool.
- 5. A comparison of ARS from the different SRV ramshead discharge events indicates the following overall behavior:
, a. Vertical response to an all valve sequential l discharge (case 1) 'i s comparable to a three. adjacent valve out of phase discharge (Case 3).
- b. Horizontal response; 'to an ,all valve sequential lO discharge (Case 1); is comparable to a adjacent valve out of phase discharge (Case 3).
three
- 6. TQ loads result in less severe structural internal loads '
than those.from the design case of all valves discharging' simultaneously with a ramshead device., -
- 7. ARS from the TQ loads are lors severe in the frequency range of importance for. piping and components"rhan those from ramshead-loads.
t' ..
..i' s
- 8. The following observations are made regarding'< HRS resulting from the threle TQ. design pressure traces (refer to Section 3 for descriptions): g j . ..
- a. ARS from . pressure trace 42 very nearly envelope,
~
results from- curves #1s-ani a#3 in a majority of:
-i locations.' -
~ .
- b. ARS from curve #1 are comparable to curve #2 at N high frequencies, but less severe at lower,,
frequencies. %- j}
- c. ARS from curve #3 are,compara'ble to curve #2 at low frequencies, but leen severe at higher frequencies.
O 9. Q
*t For all three~ TQ pressure troyea the important high frequency response tends to decrease'as the time scale is expanded, i.e., as the predominant 5 frequency is lowered.
5 .? Revision 4 - February 1981
+i .
The conclusions summarized here, in part, are utilized in Section 6 for evaluation of the design adequacy of the containment structures. 5.1.2 Containment Structures Response to SRV Ramshead Load _s Both structural forces and ARS were computed for all six load cases. Note that ARS from Cases 1 through 4 are used for design assessment of piping and equipment while structural forces from the more conservative cases 5 and 6 are used in Section 6 fot structural evaluation. 5.1.2.1 Response to All Valve Secuential Discharuc The dynamic response of the containment structures was determined for the SRV discharge event of all valve sequential firing (load Case 1 of Section 5.1). Results in terms of internal load time histories and ARS are presented here. The circumferential variation of pressure for this event was represented in the dynamic analysis by the use of five Fourier harmonic terms. Time-wise maximum values of internal loads at critical locations in the suppression pool region are shown in Table 5-1. A comparison with Tables 5-2 and 3 shows that sequential firing of all valves results in less severe internal loads at critical locations than does simultaneous firing of all valves. ARS of overall vertical and horizontal accelerations were generated and representative results are presented here. Figures 5-1 through 6 contain vertical, horizontal (N-S), and horizontal (E-W) ARS at the top of the reactor support pedestal and the primary containment at the elevation of the stabilizer truss. . 5.1.2.2 Response to ADS Discharae A comparison of applied load time-histories for the two events of all valve sequential firing and ADS discharge show general similarities. On this basis, it is considered that the structural response due to an ADS discharge event can be represented by applying a factor to the results obtained from the detailed sequential firing analysis. The factor based on the ratio of pool boundary pressures is 0.86. 5.1.2.3 ResDonse to Three Adiacent Valve out of Phase Discharae A detailed dynamic analysis of the containment structures when subjected to the pressure loads of a three adjacent valve out of phase discharge was performed. ARS developed from the acceleration time-histories were determined with results presented here. 5-4 Revision 5 - December 1981
/
i i l Three Fourier. harmonic terms were used to represent the circumferential variationaof the pressure loads.' Figures 5-7 and 8 show ARS of overall vertical acceleration at the top of the reactor- support pedestal and the primary containment at the elevation of the stabilizer truss. Overall horizontal ARS in the principal direction of loading are shown on Figs. 5 9 and 10. Figures 5-11 and 12 show overall horizontal l ARS in the perpendicular direction. This response is associated l with the fact that the bubbles are out of phase. 4 5.1.2.4 Resoonse to Sinole Valve Discharoe l As in the case of an ADS discharge event, it is considered here that the results of a single valve discharge can be obtained by multiplying the results of the three adjacent valve out of phase discharge by a factor. The factor based on the ratio of pool boundary pressures is 0.44. 5.1.2.5 Response to All Valve Simultaneous Discharoe , The dynamic internal load time-historiesi for the reacdor l pedestal, primary containment, and foundation basemat resulting from an all valve simultaneous discharge loading (load case 5 of Section 5.1) are defined in? this section. Figure 5-13 identifies the critical structural locations. l Table 5-4 provides a definition of internal loads and Fig. 5-14 l
, indicates the corresponding positive sign convention for the j basemat and superstructures.
One Fourier harmonic term was used to define this axisymmetric i event. Structural damping is omitted when determining the time history response of the . internal loads for the containment structures. This is done in order to provide an upper bound for the containment stresses calculated in Section 6.- It should .tne noted, however, that with the inclusion of structural < damping .and-a realization that the applied hydrodynamic pressure is negligible after 0.60 sec, the ensuing dynamiciresponse will decrease correspondingly. Table 5-2 shows the miximum values of bending' moment, transverse shear, 1 and axial load at l various radial positions along the basemat. 'For the reactor pedestal. and primary containment in the region of the suppression pool, Table 5-3 provides a definition of the maximum values of internal ! loads. 5.1.2.6 Response to Three Adiacent Valve Simultaneous Discharas-l The dynamic response of the containment structures to an asymmetric SRV discharge load (load case 6 of Section 5.1) was ! determined. A 60 degree circumferential variation of load was considered to be representative of three adjacent SRV's discharging as described in Section 3. O 5-5 Revision 4 - February 1981
~ - ,--._-,,-e-, - , , , , -. -.-----ev,--a, . - - . - _ - - - - , - - , - - - - - - - - - --n- ,aw-n- , - - - - - -
Ten Fourier harmonic terms were used to represent the circumferential variation of the load in the dynamic ana lys,i s . The use of 10 Fourier terms provided not only a detailed solution to the dynamic problem, but also insight regarding the significance of the higher harmonic terms. As in determining internal loads due to an all valve simultaneous discharge, structural damping was omitted. Again, this provides an upper bound for the containment structures internal loads. Table 5-5 shows the maximum values of bending motent, transverse shear, and axial load at selected locations along the basemat, primary containment, and reactor pedestal. A comparison of Tables 5-2 and 3 and Table 5-5 for the all valve and three adjacent valve simultaneous discharge cases indicates that an all valve SRV discharge loading generally results in higher internal loads for the containment structures in the suppression pool. 5.1.2.7 Hich Frecuenev Response Study As described in Section 3.2.1.2, the wall pressures calculated for the ramshead loadr used the conservative assumptions of instantaneous rises from zero to near peak amplitudes, while pressure rise times on the order of 20 to 50 m sec are actually considered to be more realistic. The effect of this rise time on structural response has been studied and found to be significant on predicted high frequency (above 100 Hz) response. As discussed below, it has been determined that the incorporation of a realistic short pressure rise time more realistically , represents and greatly reduces the high frequency component of structural response currently predicted based on pressure time histories with instantaneous pressure rises. The SRV ramshead discharge case used in this study is the sequential actuation of three adjacent valves. This case was selected because it generally results in the largest calculated high frequency responses (although the effect is similar for all ramshead load cases). For this case a very short pressure rise time of 10 m-sec was incorporated at the beginning of the pressure transient. No other changes to the pressure time histories were made. This revised time history was then applied to the structural model with no changes in the model or method of analysis. Building response spectra were generated from the resulting acceleration time histories and compared to the original results. Results are presented on Figs. 5-15 through 18 for the four curves which had the largest high frequency responses horizontally and vertically in the primary and secondary containments. The greatest reduction occurred where the response had been the highest, horizontally in the primary containment. g The original curve had high emplification from 50 Hz to well over 100 Hz. The revised curve has no significant amplification above W 5-6 Revision 4 - February 1981
i 100 Hz. In all cases the response spectra were unaffected at the lower frequencies. i Although the study described above has shown that the building response resulting from the ramshead loads contains i unrealistically high response in the high frequency range, due to the instantaneous pressure rise hypothesized, the curves continue i to be used as the conservative design basis. In Section 9.1.2, ;
- however, this effect is taken into account in the reevaluation of '
! equipment. ! 5.1.3 Containment Structures Response to SRV T-Ouencher Loads ' The dynamic response of containment structures has been determined for the SRV discharge loads specified for the TQ device. The analysis has considered four load cases - all i valves, ADS, three adjacent, and single valve discharge. For - 1 each load case the structural response was determined for all three TQ pressure traces with five time scale factors applied to each trace to encompass a broad range of predominant frequencies. The time scale factors utilized were 0.8* (providing the highesc frequency), 1.0, 1.2, 1.4, and 1.8 (providing the lowest frequency). The individual ARS resulting from the three pressure j curves each with five time scale factors are all enveleped to 4 produce conservative ARS for the assessment of piping and mechanical systems. i The magnitudes of the three TQ pressure traces are to be increased by a constant multiplier of 1.1 to obtain the design i time histories. However, no multiplier has been included in the dynamic structural analysis for TQ loads nor does it appear in i the results presented here. All results must be increased by the multiplier before use in the design assessment. In order to view the differences in response due to the three pressure traces and- also to see the effect of altering the frequency content by time scaling the curves a set of comparison plots of ARS have been developed. The vertical response at the top of the pedestal due to an all valve discharge is used as
- representative of the behavior. Figures 5-19 through 21 contain l ARS plots developed for the three pressure _ traces, with the plot i for each curve identifying the general effect of time scaling (altering tho frequency). Two conclusions can be drawn from these results. First, in the frequency range above approximately-10 Hz, the response tends to be most severe for the higher '
frequency cases of each curve (for the smaller time scale factors). Also, it is noted that the response to curve No. 2 nearly envelopes the response to curves No. I and 3 at all frequencies. Curve No. 2 can be characterized as having the most regular shape- of the three, generally resembling a decaying sinusoidal curve. ( *This is even slightly beyond the range specified in Section 4 3.2.2.1. J 5-7 Revision 4 - February 1981
5.1.3.1 Response to All Valve Discharoe Structural response to the all valve discharge has been developed h for all three curves each with five time scale factors. Enveloped ARS have been developed of the 15 individual ARS. In order to present a more realistic description of TQ ARS, the effects of the time scale factor of 0.8 are not included in the representative figures presented here, since this factor is outside the range of the TQ load definition. Vertical response is predominant from this axisymmetric event. Representative l vertical ARS are presented on Figs. 5-22 and 23 at the top of the pedestal and at the primary containment at the elevation of the stabilizer truss. Horizontal response to this event is exclusively in shell redial " breathing" modes. Horizontal response at the same structural locations is presented on Figs. 5-24 and 25. Again, note that the multiplier discussed in Section 3 has not been included in the results. 5.1.3.2 ResDonse to ADS Discharoe As in the case of SRV discharge with the ramshead device, ADS results can be obtained by applying a factor to the results of the all valve discharge. Since ADS is very nearly axisymmetric this factor is obtained by taking the ratio of integrated pressure on the circumferential pressure distributions for the two load cases. 5.1.3.3 Response to Three Adiacent Valve Discharoe Structural response to this asymmetric event has also been completely determined. Enveloped ARS of both vertical and horizontal response are presented on Figs. 5-26 through 29 for the top of the pedestal and the primary containment at the elevation of the stabilizer truss. These results do not include the load multiplier of 1.1. 5.1.3.4 Response to Sinole Valve Discharoe Single valve discharge response can be obtained by factoring the j results of the three adjacent valve discharge. Factors for the vertical and horizontal response are obtained by comparing the ratios for Fourier series coefficients of the circumferential l pressure distribution of these two load cases. 5.2 STRUCTURAL RESPONSE TO LOCA LOADS This section describes the behavior of the containment structures when subjected to dynamic LOCA loads. The LOCA transient events considered in this section are vent clearing, air bubble pressure, condensation oscillation, and chugging. These events are described in detail in Section 4. I Vent clearing jet and air bubble pressure loads occur consecutively and are treated together here. These loads are characterized by an initial impact on the basemat followed by 5-8 Revision 4 - February 1981
transient suppression chamber air bubble pressures on the pool boundaries. This event is axisymmetric with a maximum uniform b,) pressure on the basemat of 44.7 psi at a time of 0.57 seconds \# after the LOCA begins as shown on Fig. The value of 44.7 5-30. l psi is the pressure at the main vent exit and has not been attenuated in order to provide an upper bound to structural response. Sections 4.2.4.2 and 4.2.5.2 describe the condensation oscillations and chugging pool boundary loads, respectively. Figure 5-31 depicts tb? idealized chugging load time-history used for the dynamic analysis. Internal loads are computed for the basemat, reactor support pedestal, and primary containment. l Amplified response spectra were generated for these dynamic events and are used for the assessment of piping systems and equipment. l Results of this section are used in the assessment of primary structures (refer to Section 6), as required by the load combinations discussed there. 5.2.1 Summary of Results l The dynamic response of the containment structures to transient LOCA loads was determined. Structural internal load time-histories, acceleration time-histories, and ARS have been developed. The following is a summary of significant results:
- 1. Containment structure internal loads in the suppression pool region due to LOCA vent clearing are less severe than those due to the long term static pressures and temperatures associated with a LOCA. Therefore, the assessment of primary structures (Section 6) is based on the long term effects of LOCA as defined in the USAR.
- 2. Chugging and condensation oscillation loads are not a source of significant containment structure internal loads.
- 3. Vertical ARS due to vent clearing and axisymmetric chugging are generally comparable, with vent clearing response the greater at frequencies lower than 15 to 20 IIz and chugging response greater at the higher frequencies.
- 4. Vertical ARS due to axisymmetric chugging are generally comparable to, but somewhat less than, vertical ARS due to all SRV sequential firing with ramsheads.
- 5. Horizontal ARS due to asymmetric chugging are less
[]s A severe than those due to three adjacent SRV. discharge with ramsheads (either simultaneous or out of phase). 5-9 Revision 4 - February 1981
,1 . .- .
5.2.2 containment structures Response to LOCA Loads 5.2.2.1 Response to Vent Clearino The transient pressure loads described in Section 5.2 were applied to the structural model and the dynamic response deternined. One Fourier series term was used to represent this axisymmetric event. Time-wise maximum values of internal loads have been computed and are presented in Tables 5-6 and 7 for primary structures in the region of the suppression pool. Selected ARS of overall vertical accelerations are presented on Figs. 5-32 and 33. 5.2.2.2 Response to Condensation Oscillation Loads The dynamic response of the structure from the axisymmetric condensation oscillation load was determined. The condensation oscillation load as defined in Section 4.2.4.2 is a sinusoidal varying load having a peak-to-peak pressure amplitude of 7.5 psi for the controlling medium mass flux regime with a single frequency ranging from 2 to 7 Hz. As stated in Section 5.2.1 containment structures internal loads due to condensation oscillation loads are small and Appendix B shows that the containment structures have ample design margin. ARS curves of vertical acceleration which envelop the 2 to 7 Hz condensation oscillation load are shown on Figs. 5-34 and 35. They have higher peak amplitudes than the axisymmetric chugging ARS. However, since the peak occurs at a frequency of 7 Hz (period = 0.14 sec) and there is no significant power in the range above 7 Hz, they will not significantly affect the piping system and equipment response since most of their natural frequencies lie above 7 Hz. The ARS curves of horizontal ARS are shown on Figs. 5-36 and 37. They have a smaller peak acceleration than the asymmetric chugging ARS. 5.2.2.3 Response to Chuacino The dynamic response of the containment structures to the oscillatory chugging loads was determined. Both axisymmetric and asymmetric chugging events were considered with characteristic frequencies of 20 and 30 Hz. The axisymmetric chugging loads were represented by one Fourier series term while two terms were used to represent the overall behavior of the asymmetric event. Internal load time histories were developed, as well as acceleration time histories and ARS. Tables 5-8 and 9 present time-wise maximum values of containment structure internal loads 5-10 Revision 5 - December 1981
in the suppression pool region for the axisymmetric 20 Hz chugging O for 30 Hz event. Tables 5-10 and 11 present the same information axisymmetric chugging. Results due to chugging are of comparable magnitude. asymmetric ARS curves which envelop the response from the 20 and 30 Hz cases 1 were generated. Figures 5-38 and 39 present ARS of overall 4 vertical acceleration due to axisymmetric chugging while Figs. 5-40 and 41 present overall horizontal ARS due to asymmetric chugging. 5.3 STRUCTURAL RESPONSE TO ANNULUS PRESSURIZATION LOADS Annulus pressurization (AP) refers to the dynamic asymmetric pressurization of the annular space between the reactor pressure vessel (RPV) and the shield wall following a double-ended rupture (DER) of a high energy line at the safe-end weld to the RPV nozzle. As stated in Section 2.3, AP loads are described and discussed in Section 6.2.1 of the Shoreham USAR(14). This section contains a description of the method used to calculate the dynamic structural response to AP loads. Representative ARS of the resulting building accelerations are also presented for the two postulated pipe rupture events that result in significant AP loads. These are DEd's of the feedwater or recirculation suction lines at the RPV nozzle. A similar () rupture.of a main steam line does not result in AP loads since. it is attached to the RPV above the top of the shield wall. The dynamic structural analysis for AP loads utilized the mathematical structural model previously developed for seismic
- analysis and described in Section 3.7 of the Shoreham USAR.(14)
It is a " lumped mass" beam type model with a total of-47 mass points included to represent the RPV, shield wall, pedestal, primary containment, secondary containment, and basemat.. This includes the addition of three mass points on the shield wall to obtain more detail in the area of direct load application. The forcing functions applied to the structural model consist of three components. They include the annulus pressure transient, the direct nozzle blowdown force, and the pipe rupture restraint reaction force, all of which occur concurrently. The annulus pressure has a time varying spatial distribution. Therefore, at each time step of the calculated pressure transient, the pressures are integrated over the surface areas of action to obtain the unbalanced forces acting on the RPV and shield wall, respectively. These integrated pressure loads are then applied to the appropriate nodal points on the mathematical structural model simultaneously with the direct nozzle blowdown and rupture restraint reaction forces. l The dynamic solution is obtained by time history modal analysis using the computer progran STRUDL (S&W Computer Code designation [} 1 5-11 Revision 5 - December 1981
ST-15 as described in the USAR). Since AP is a faulted condition, the structural damping values used are those & prescribed for a safe shutdown earthquake. The dynamic solution consists of time histories of structural displacements, forces, W and accelerations. ARS of buildino accelerations are generated using an exact analytical solution as described in Section 2.4.2 for the hydrodynamic loads. Results of the dynamic analysis indicate significant horizontal accelerations in the immediate vicinity of the applied loads, i.e., on the RPV and shield wall. The response is very localized, and substantially attenuated at the base of the RPV and shield wall (top of the pedestal). Response of the primary containment is not significant while the secondary containment is virtually unaffected. No significant vertical response occurs. ARS of horizontal accelerations at points of maximum response in the RPV, shield wall and pedestal are presented on Figs. 5-42 through 44, respectively, for AP loads resulting from a recirculation suction line break. In this case, the break location coincides with a principal axis of the model and therefore, there is no perpendicular component of horizontal response. Note that the peak structural acceleration (zero period acceleration) is about lg while the peak ARS value is about 4g. Figures 5-45 through 5-47 present horizontal ARS at points of maximum response for a feedwater break. In this case, the break is located 45 degrees from the principal axis of the model. Therefore, two perpendicular components of equal amplitute result. For this event, the peak structural acceleration component is again about lg while the peak ARS valve is about Sg. Since these curves represent two equal perpendicular components, the peak resultants are actually 1.414 times greater. l A complete set of ARS curves including vertical responses were generated at several points throughout the reactor building (including primary and secondary containments) for both the l recirculation suction and feedwater line breaks. These results are utilized in the evaluation of all plant piping and equipment components. i l l 1 l l 5-12 Revision 5 - December 1981 l
4 TABLE 5-1 MAXIMUM VALUES OF DYNAMIC LOADS IN THE BASEMAT AND SUPERSTRUCTURES FROM A SEQUENTIAL ALL VALVE SRV DISCHARGE WITH RAMSHEAD(1) BASEMAT Radius M. Me, Q. N.. New (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 11.10- -49 -50 5 12 12 e 11.10+<2) 144 -63 -31 21 12 26.40 -79 -62 11 17 12 41.70-(2) 145 -28 23 16 12 41.70+ -125 -39 12 -11 10 1 PEDESTAL Elevation M. Mw, Q. N.. Nww (ft) (ft-k/ft) (ft-k/ft) (k/ft) (%/ft) (k/ft) iO 8.00(*> -33 -6 10 27 -8 12.00 6 1 3 24 -9 PRIMARY CONTAINMENT Elevation M. Me, Q. N.. Nw, (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) I 8.00<*> 107 18 -21
~
13 -4 12.00 29 4 -20 13 5 (2) For design calculations all values can be considered positive or negative. (2) Radius to pedestal wall centerline is 11.10 ft.
-($> - Radius to primary containment wall centerline is 41.70 ft.
(*) Base of pedestal and primary containment is at El 8-0.
^
1 of 1 Revision 4 - February 1981
.= _
TABLE 5-2 (.-
'~' )
MAXIMUM VALUES OF DYNAMIC LOADS IN THE BASEMAT FROM A SIMULTANEOUS ALL VALVE SRV DISCHARG WITH RAMSHEAD(1) l Radius M.. Mrr Q. Nas Nrr (ft1 (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 6.00 -249 -250 -21 -35 -35 11.10-c2> -244 -247 .-11 -35 -35 11.10+ -530 -291 58 -19 -32 16.20 -266 -286 40
-16 -26 21.30 198 -239 28 -17 -23 26.40 138 -7'2 -19 -17 -22 31.50 92 169 25 -17 -21 36.60 84 140 31 -17 -20 41.70-(2> 179 119 27 -17 -19 41.70+ 234 153 -21 14 -17 48.02 123 136 -18 12 -14 54.35 -76 113 -14 -11 -14 60.68 -87 88 -12 -10 -13 67.00-(4) 134 61 -11 -9 -13 67.00+ 123 65 -7 -7 -12 72.00 -87 58 -6 -7 -12 77.00 -55 53 6 6 -11 C1 82.00 -23 50 5 4 -10 \~ / 87.00 0 49 4 -3 -10 (1) For design calculations all values can be considered positive or negative.
ca) Radius to pedestal wall centerline is 11.10 ft. ( > Radius to primary containment wall centerline is~41.70 ft. (*> Radius to secondary containment centerline is 67.00 ft. 1 of 1 Revision 3 - November 1978
TABLE 5-3 MAXIMUM VALUES OF DYNAMIC LOADS IN THE SUPERSTRUCTURES FROM A SIMULTANEOUS ALL VALVE SRV DISCHARGE WTTH RAMSHEAD5 1) PEDESTAL Elevation M. Mer Q. N. New (ft) (ft-k/ft) (ft-k/ft) (K/ft) (k/ft) (k/ft) 8.00<a> 114 18 -34 -56 -49 l 12.00 16 2 -15 -56 .-46 16.00 -13 -3 -5 -55 -18 21.00 -10 -2 2 -55 -13 26.00(2) -3 -1 1 -55 -5 l PRIMARY CONTAINMENT Elevation M. Mer Q. N. Nrr (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(2) -126 -21 16 25 29 l 12.00 -68 -11 -14 25 43
\- 16.00 31 5 9 24 42 21.00 -29 -5 5 24 32 26.00(2) -20 -3 4 24 -22 l (1) For design calculations all values can be considered positive or negative.
(2) Junction of basemat with superstructure. ($) Top of suppression pool. 4 1 0 l V- ; l l 1 of 1 Revision 3 - November 1978 l j l i i
TABLE 5-4 l DEFINITION OF INTERNAL LOADS M.. - Radial (longitudinal) bending moment (ft-k/ft) positive
- (+) when causing tension on top of mat and inside surface of superstructure N.. -
Axial (longitudinal) force (k/ft) positive (+) when causing tension in mat and superstructure Q. - Radial (transverse) shear (k/ft) positive (+) when
- acting upward on outer face of mat and radially outward on superstructure ,
Mer - Tangential (Hoop) bending moment (ft-k/ft) positive (+) when causing tension on top of mat and inside surfaceuof superstructure Nrr - Tangential (hoop) force (k/ft) positive (+) when causing i' tension in mat and superstructure Q, - Tangential (inoop) shear (k/ft) 4 N., - In plane membrane shear (k/ft) 4 (1) Refer also to Fig. 5-13 r O 1 of 1 Revision 4 February 1981
- v. e- ,-- , , --..,-n,. ,w. ,,,e,-,o,-em.,,,--en, e w, en ,,~,,.,..w-r,., , N ., N, 4..,,, ,e ,r --r, ,,,-e,a.A,,,,-.
-, ,y,--
TABLE 5-5 _MANIMUM VALUES OF DYNAMIC LOADS IN THE BASEMAT AND FUPERSTRUCTURES FROM A SIMULTANEOUS THREE ADJACENT VALVE SRV DISCHARGE WITH RAMSHEAD(1) BASEMAT ~ Radius M.. Mrr Q. N. New (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 11.10-(2) 52 42 -11 -8 -8 11.10+ 152 52 -48 15 -8 26.40 -125 -81 -5 12 -6 41.70-(2) 136 25 27 -7 -7 , 41.70+ 60 36 6 -8 -6 ? i 1 PEDESTAL Elevation M.. Mer Q. N.. New (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(*) -45 -10 14 33 -11. d 12.00 4 -2 9 27 -9 PRIMARY CONTAINMENT Elevation M. Mrx Q. N. Nrr (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(4) 67 11 -14 22 -5 12.00 17 -3 -14 22 10
~
(1) For design calculations all values can be considered positive or negative. (2) Radius to pedestal wall centerline is 11.10 ft. .. (28 Radius-to primary containment wall centerline is 41.70 ft. (*) Base of pedestal and primary containment is at El 8-0. j 1 of 1 Revision 4 - February 1981
- a. _ _ _ _ _ _ . _ _ . _ . _ . . . , _ _ - ,_ ._ __...--._-.-._.: _ _ _ _ . _ _ _ . - . _ . . _ -
t _ TABLE 5-6 -
, )
N/ MAXIMUM VALUES OF DYNAMIC LOADS IN THE BASEMAT
>FROM LOCA VENT CLEARING LOADS (1)
Radius Mas Mer Qs Nas Nrr (ft) (ft-k/ft) (ft-k/ft) Ik/ft) (k/ft) (k/ft) 6.00 -432 -424 22 56 56 11.10-<2) -386 -407 12 56 56 11.10+ -802 -474 46 16 50 16.20 -650 -533 36 21 39 21.30 -571 -531 33 25 35 26.40 -459 -501 40 28 33 31.50 -303 -451 52 30 32 36.60 -97 -377 65 31 32 41.70-<'> 175 -277 57 31 32 41.70+ -529 -369 47 -27 23 48.02 -266 -315 41 -21 17 54.35 -111 -262 27 -15 14 ' 60.68 108 -216 18 -11 12 67.00-<4) 135 -178 12 -73 10 67.00+ -117 -188 10 -11 10 72,00 -70 -169 10 -10 9 77.00 50 -154 7 -7 8 82.00 23 -144 5 -5 8 0 87.00 0 -136 3 -3 7 (' ) For design calculations all values can be considered +. (2) Radius to pedestal wall centerline is 11.10-ft. ca> Radius to primary containment wall centerline is 41.70 ft. (4) Radius-to secondary containment wall centerline is 67.00 ft. f i 1 of 1 Revision 1 - April 1977 _. -._ _. _ . _ _ __ . ~ . _ - -- -- , . _
4 TABLE 5-7 v MAXIMUM VALUES OF DYNAMIC LOADS IN THE SUPERSTRUCTURES FROM LOCA VENT CLEARING LOADS (1) PEDESTAL Elevation Man Mer Os Nas Nrr (ft) Ift-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00c2> 170 27 -49 -51 -68 l 12.00 29 4 -51 -70 -22 16.00 -17 -4 -4 -50 -31 21.00 -14 -3 2 -49 -2 26.00(2) -3 0 1 -48 3 l PRIMARY CONTAINMENT Elevation Ms. Mer Qs No. Nrr O' (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(2) 361 60 -53 -17 -73 l 12.00 178 29 -47 -17 -108 16.00 72 12 -30 -16 -102 21.00 -57 -10 -15 -16 -72 l 26.00(3) -59 -10 -2 -16 -39 (2) For design calculations all values can be considered' positive or negative. (23 Junction of basemat with superstructure.
<2) Top of suppression pool.
1 of 1 Revision 3 - November 1978
TABLE 5-8 MAXIMUM VALUES OF DYNAMIC LOADS IN THE BASEMAT FROM AXISYMMETRIC 20 Hz CHUGGING LOADS (2) Radius Meu Mrr Q. No. Nrr (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 6.00 -23 -24 -4 4 4 . 11.10-(2) -16 -20 -2 4 4 11.10+ 47 -21 -11 9 5 16.20 -39 -27 -8 8 6 21.30 -37 -29 -4 7 6 26.40 -25 -28 5 7 6 31.50 -13 -23 6 6 6 36.60 -18 -15 8 6 6 1 41.70-(3) 50 12 7 6 6 41.70+ -35 -10 2 -3 5
- . 48.02 -21 -9 2 -2 4 54.35 -13 -8 1 -2 4 60.68 11 -7 1 2 3 67.00-(*> 11 -6 1 2 3 67.00+ 10 -7 1 1 3 72.00 7 -6 1 1 3 77.00 4 -5 -1 1 3 82.00 2 -5 0 1 -2
-5 87.00 0 0 0 2 (1) For design calculations all values can'be considered +.
(2) Radius to pedestal wall centerline is 11.10 ft. (2) Radius to primary containment wall centerline is-41.70 ft. (*) Radius to secondary containment wall centerline is 67.00 ft. 1 i i i 4-
!O i
1 of 1 Revision 1 - April 1977
_ , _ - ._ - ~. -. . . _ . - -- - i TABLE 5-9 I i MAXIMUM VALUES OF DYNAMIC LOADS IN THE SUPERSTRUCTURES FROM AXISYMMETRIC 20 Hz CHUGGING LOADS (1) l l PEDESTAL Elevation M.. M,,. Q. N.. Nr, (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(2) -14 -2 4 -12 4 l 12.00 2 0 4 -12 -4 16.00 6 1 -2 -13 -5 21.00 3 0 -1 -13 -5 26.00(3) -1 0 -1 -13 -2 i l PRIMARY CONTAINMENT Elevation M.. Mer Q. No. Nrw (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00c2) 21 4 -5 6 4 l 12.00 6 1 -5 6 6 16.00 -11 -2 -3 6 9 Os 21.00 -10 -2 2 6 9 26.00(3> -4 -1 1 6 7 A l l l (1) For design calculations, all values can be considered positive or. negative. , (2) Junction of basemat with superstructure. (3) Top of suppression pool. T 1 of 1 Revision 3 . November 1978
TABLE 5-10 . r MAXIMUM VALUES OF DYNAMIC LOADS IN THE BASEMAT
- FROM AXISYMMETRIC 30 Hz CHUGGING LOADS (1) ,
t Radius -M.. Mer Q. No. N2r (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 6.00 -17 -18 -5 -8 -8
- 11.10-ca) -16 -17 -3 -8 -8
! 11.10+ 46 -19 -9 -10 -8 I l 16.20 24 -20 -6 -9 -9 . I- 21.30 -25 -19 -3 -9 -9
- - 26.40 -23 -16 2 -8 -8 '
- 31.50 -13 -14 5 7 -8
! 36.60 22 -10 7 7 -8 50 16 ! 41.70-(2) 6 7 7
- 41.70+ -25 -9 -2 -5 -7 4
48.02 -16 -7 -2 -5 -6 54.35 -12 -5 -1 -5 -6 60.68 8 -5 1 -5 4 67.00-(*) 10- -4 1 -4 5 4 67.00+ 10 4 -1 3 5' 72.00 8 4 -1 3 5
- i. T7.00 -4 4- -1 2 5
+
1 82.00 -2 3 - l' 2 4 {, 87.00 0 0 0 4 i ? i ' (2) For design calculations all values can be' considered +. (2) Radius to pedestal wall centerline.is 11.10 ft.
<2) Radius to primary containment wall. centerline is 41.70 ft.
l (4) Radius to secondary containment wall centerline is 67.00 ft. l l i i i i i O t ! 1 of ll Revision 1 - April 1977 1 f _ . .; .--_._._,_.--,____.__._._u...__..,_....___...,_..._,_-, - - _ , . _ . . - _ - - . - _ , _-
4
,- TABLE 5-11 s- l MAXIMUM VALUES OF DYNAMIC LOADS-IN THE SUPERSTRUCTURES FROM AXISYMMETRIC 30 Hz CHUGGING LOADS (1)
PEDESTAL Elevation Mas Mrr Qs Nas Nrr (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(23 14 2 4 9 3 l 12.00 -2 0 4 9 -3 , 16.00 6 1 -2 9 -5 21.00 3 0 -2 10 -5 26.00(2) -1 0 0 10 -2 PRIMARY CONTAINMENT ! Elevation Mas Mer Os Nas Nrr (ft) (ft-k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 8.00(2) 26 4 -6 4 4 12.00 -6 -1 -6 4 7 fN g 16.00 -14 -2 -4 4 12 ( ,f 21.00 -15 -2 -2 4 14 26.00(2) -9 -1 1 4 13 (1) For design calculations all values can be considered positive or negative. (2) Junction of basemat with superstructure. (2) Top of suppression pool. (s / 6 1 of 1 Revision 3 - November 1978
3.00 -
.r- -
i t 2.50 - -- -. e 2.00' - - - b ---- z ;'. . o . i a . 6 4 l.50 - - - - - - -i-- - ' -- - - - - - e g i
. i a ; . .. ,. . . . . _ .
g . o i. - o 4 ' - - - - - - - - ~ - - - - - - - - I.00-- -- - - -- - / - - --- - t : . .
,^
- l '., ....---.,
0.5o --l- , ' --- -'.. -- - -- *. . .p :. . . . . 9 . e~ --- - .... - . . . .. : --- ' i , o.co i t i i i O.00 0.02 0.04 6.'D 6 'O 08. 0.i d' PERIC 4 g FIG. 5-1 ppgrUBE AMPLIFIE CARD TOP OF F ALL VAL \ Also Av M SHOREHAh Aperture p i 1
.- I
....;_.....-- . . _ . - . . . . .. . .. .__. . . . . . . . . . . . =.
. -. , ' 1 r
,- 7 .._-....__.r"r~-'--'
t . . H . E rl'r . EIE Elll'lE Il1]E lIEEIbEI . . , DI-DIT -% in _ I_ n pr!j .. !j
- ~J
- i1 L f.43-L < "---
J-h ,
..LJEA' .
;.p: ngip .a Ljupl '
rui. u r 2 :rrn ;rigu - '
._.t om 1, mu a- . . i-.
... fb 1i I 'il_ ._2 p. 4_ - I a.a ! M.L-.
- l4 -
g- . . i L; s L.L u..! .L .. - r.
~
l
+.
f ; I . . _: , i ._.._.___..._.2 . . . . . . . -.. .... . .. ... .. . . _ . . _ .
.l i;.
l j j i - 4 i
. . t_ . .. .._ .#_
Fl . i . i l . i , I . ; 1 . .,. , i
; l
..,.'.g.
. ,, ...,,m.._.... . . . ..... , . .... . . . . . . ..
;> i H. ... }..!. i I I 1 i I i . !
l 0.12 0.14 0.16 0.18 0.20 0.22 0.24 l lD (SECOND) I I i l l lD RESPONSE SPECTRA OF VERTICAL ACCELERATION iEACTOR SUPPORT PEDESTAL 7E SEQUENTIAL DISCHARGE-RAMSHEAD 1 l NUCLEAR POWER STATION-UNIT 1 ,L 8612080645-[3 IGN ASSESSMENT FOR SRV AND LOCA LOADS l REVISION 4-FEBRUARY 1981 l ;
5.00 > j j
--+-
l l
',l t l
; l
. . . . L i.. . . j_...._ 4 . '!. . .
9.. . .. . t 2.50- - -- - - -- - -- a
- . ~
o 2.00 - -- - -
, ; p Z , !
O_.
.j ..
- --4 t r -- t, -
y .
! i '
4 i ! - x l.5 o -
-- a - --- -- g - i - -- --l ,
w :
,i i
J , y ; O - i !' . o 'l . l< 1 l' ., . , 4 l 00 ,0 .. _ ....L. - al...hu L. -..... . . ~ ~ . . . . - . . - f.
- l. l s.h f^ , .
a j y. i, . t. c ., ; ,. ., o.so --- -- y ., e....... .s.p... .... l
.e I i 1
o.oo ! o.co o.o2 o.o4 o.'os o.os o.'io PERI rpt FIG. 5-2 AptmTURI AM PLIFC CARD PRIMARi ALL VAL Also Avallable On SHOREHAl Aperture Card PLANT DE . l
r . - - l l 1 . . . . . . . . . _ . . . .. . . . . r . l 4 ,.,
.l'.
1 y'. _ l_ }_. g m
~~
r="" [ . ..
.. . .] . .
a ! J_im iin ' - - - i'. _,
;.1g3;nI.. in 7 opr tr iy mi.1- pmg ,
F ' ME
- rTq rgp} ggrrmy
.; 4 1 P a U_._.;g rnu m uw m . an
.. . ._ _. .. _..__.__.. i , p.v
. 1 t
. p. . ....;_ -
;4,. ..
. 3 !
I i .! '! . f
.. . _d. .'
i ;- ,
~
! i ; .
- I 4
8 j
, t l .
i t
.-...-_.u.... .
. . .. !..._ ... ... . . _ . . . . . . . . 1 . ....
t
! i
. . _ . . . . . . . . .. _ .u . .. :. :. . . . , ,
. p .. 4 . . . . '
%. _. _. ,. m -
i
- . , ' pi i
, i i
- : ,. i e-1 0.12 0.14 0.16 0.18 0.20 0.'22 0.24 OD (SECOND)
ED RESPONSE SPECTRA 0F VERTICAL ACCELERATION ,
' CONTAINMENT AT ELEVATION OF STABILIZER TRUSS VE SEQUENTI AL DISCH ARGE-RAMSHEAD -
A NUCLE AR POWER STATlON-UNIT 1 SIGN ASSESSMENT FOR SRV AND LOCA LOADS 8612080645-lY : REVISION 4-FEBRUARY 1981 I
-_...,____._,_)
n-
- 3. 0 0 - - - - -- -
--- - . - - i .
r - - r.
- - r. . - - - .
.t r .
. . ~ . . . . . .
. . i 2.50 - -- - --
n 2.00 - - - - - - - - - o ihms' . z . . . . ...... ._ _. . .. o l: ; :-! ;
. , ,i 1.50- - -- - - - - - -
[l: E to .-
.J ,1 g . _ . . . .._. _ _ . . .
o :9
.o : '
4 1.00 ~ - - - - --- - - - - --- - - i \
- - ~ ~ - - - ' - - ' - ----
l : j:* 3 . . _ . ._ I e p
* ;1 0.50 -^ '
-- - - - -- - - l --\ - - - - - ' - ~ ~ - -
,i .
' 's ,
.- .2,
.- ;s = . p.~ ..- ,. . ' g.... s .t........p.. .w.....,.
; .i 1
- y. , 6. i W - l i [.! . I q!' ,
** I 1 g
0.00 O.OO O.02 0.04 0 06 0.08 0.10 PERK l TI FIG. 5-3 APERT M AMPLIFIE CARD TOP OF R ALL VAL; Also Ara 11able On W ar. Card SHOREHAM PLANT DES l
.a . ,, .
g .. .. . 4
~. . .. .._.-
m _
,,_ ,m. m . _ ,....
3 ,
--- L nr0 wnn pO' EEILo Ld nkina n.am,ihm -Wia rnki hurter t. :
t.- ,, . . . . a uu-m ,.. _.u . .-- .---- n_oxi. o- ,,1=
' _. i u ,
I '!
.1 I I - I '
g igg .
.Ll l.._ .' ' . . .
9
.ilr.~.
e . l . r j ' .. . . r _ ' .._ ... _ ._. ._........ _. ._... .-- B - g 9 e
..._.._....z...._. .. . . .
l : l i e i!i h ... k r l'
..m_... _ ..: . ..
. _ .L .
! i i *.. t
! 0.12 0.14 0.16 0.18 0.20 hD (SECOND) l 1 lD RESPONSE SPECTRA OF HORIZONTAL (N-S) ACCELERATION EACTOR SUPPORT PEDESTAL E SEQUENTIAL DISCHARGE-RAMSHEAD , NUCLE AR POWER STATION-UNIT 1 IGN ASSESSMENT FOR SRV AND LOCA LOADS 8 612 0 8 0 6 4 5-t 5 l REVISION 4-FEBRUARY 1984.
e.- . , I
- 5. 0 0 - - - - -- - - - - - -
i ~- t1 4
. , . . . . . ....a . . - . . . . . . . . . . .. ...
2.s 0 - -- - - - - -- -- - 2.00 - a w z ' ' ' ' i _O. .e . s ::
$ i.50 - - - - - -
l,c-
-- r - - - '
W i ! i i _ ie =
, iI y _ . _ _ .
;j i . _ . . . . ..
O . 9 - O i ', 4 l,00 _ - . _ . . . . - . .
\ .-
I
..- 4 .- . i a .
j ' ---- / *g 0.50 - - - - l- -
~
- . ~.
,. i_. _. . . .. ._ ... ...:..... ..; . .
i .... i 1--.. i . 0.00 i I ' 8 O.00 0.02 0.04 0.06 0.08 0 10 l J l PL i i TI APERTURE FIG. 5-4 CARD AMPLIFIE PRIMARY I Al" Aperture ^'"M ALL VALV SHOREHAM PLANT DES
! ; I , f
,7 ; g l ;
7;
--*- , ,j l .
c .,.. - .
o i -. o.
s , i l 2 4 1.50 - - _ - ~ . . - _ _ - _..._..______1. g -- l , W I . w O i O ; 't l .00 - . - . . . . . . . . . . . - . . . .. - . ... ....__.....i..._ , y . ,S I ,d . _ . ;.. . . 0.50 - . - - - . . . - - - - - . . . 1. . . . . . . . . . . . . . . . . . . j.'?',..... l. i P - m -- ---T-----' 0.00 - i O.OO O.02 0.04 0.06 0.08 0.10 PER i ( FIG. 5-5 l l TI AMPLIFIED RE APERTURE TOP OF REACl C A" ALL VALVE SE SHOREHAM NUCL & AvrJ!able On PLANT DESIGN A: Aperture Card 1 . i ' i . . . . . t. ? . , p.- , s ,4 : . ; _i... -- - -a ! ,. m--- - mmu m_ _ c. . ..g".: . . . . ., - = --a1 . sr Irn4anauswno ,; - Z~T- .y J1 . IE 3 grit t 3 ng lygtlejl; , . 3.-3, , ,,, ,, y ,. y,y,,, m ,, ; - - !. f .' . . !i ., l .1_ .1, i ! i i-J l . r V. i . s I. ;a .. 3 * *t . . e ..1. t . .' - i i o - - - =; i: : ... . . .. .. .;. . .- . ~. . , ;;i L ._ . i. n_.. .:.._.. . u,,. 3, l ,,4 [i! - . -). i - - ----- . i. ;!,- . . .. , 1 O.12 0.14 0.16 0.18 0.20 .O.22 0.24 IOD (SECON D) k SPONSE SPECTRA 0F HORIZONTAL (E-W) ACCELERATION : 'OR SUPPORT PEDESTAL i ;QUENTIAL DISCH ARGE-RAMSHEAD t .E AR POWER STATION-- UNIT 1 8612080645 --l7 BSESSMENT FOR SRV AND LOCA LOADS I REVISION 4-FEBRUARY 19411 3.00 - -- 'i 3 , , ' l i - : , 2.50 - - - 2.00 - e w I Z -- O . . t l.50 -- - ---~--'--------i -- -- --< E : W ' i , .J W l l 0 I o ; c - 4 g 00 - -. . h . . . . .. . .. .. . ,9 : . v .y, . i H .' ' O.50 ; . . . . . . . - ,-*,g n_ . 3 e - , g 1 0.00 O.OO O.02 0.04 0.06 0.08 0.10 PEI i TI FIG. 5-6
- APERTURE AMPLIFIED RE!
CARD PRIMARY CONT / i3 , ALL VALVE SE l Aperture Gud SHOREHAM NUCLI , PLANT DESIGN AS j i I 1 ..f. .m+ ..__ .....'. L-- ,m.... u ,. r a 4 ... r,.m i
- 1.
J2EYOU .F - - - - 17 - J LG. 15 ; E' Ud N i* ; ; '~ _t. a na . - www .11 senn . . . g .,. . .. ... ..... 13 . y%a ru - - a le rin jI35j " LJH ,
- p. _...__.____
< ~ i l 1 l , 4 i -i
- i r : :)- ti 9 ..___._.<.. . . _ . .
- - - ; .. 9.,._ _ , i f . .b , . ; . .. :i i i t . i - ' 6 . . t l . a . _ . . . . . . . . .j . . . .. - .. ', . .a . : '. _. . .I , m ..' i , i :. ;- i,. . l ....... 1. , --- .. , - , ,. i : , p .; ., ; l 0.12 O.I4 0 16 O. I 8 0.20 0.22 0.24 ) LOD (SECON D) l kPONSE SPECTR A 0F HORIZONTAL (E-W) ACCELERATION INMENT AT ELEVATION OF STABILIZER TRUSS aUENTI AL DISCH ARGE- RAMSHE AD AR POWER STATION-UNIT 1 %
- ESSMENT FOR SRV AND LOC A LOADS 8 6120 80 64 5 -(E REVISION 4- FEBRUARY 1981-l
l 3.00 - -- -- i . . I i . ?. i. . . _ _ . .I . - . . . ~ .. . - ._ J. . . ; _. . . . i .. . .i. 2.50 - - - - ~ e 2.00 - -- - - w - . z - O. . . .i,+ . 6- .-
- i . . . . i.
------,-+ <c l.50 . , .. . ., . . (E . . . . W j i * ' ' _.I ._. .
- Y s
..+. ... W . ! l l o ; ;. , Q = ! i a <t l .00 -- -- - l- --+--,- ' -----i------- , , i i. . I I .
- g. . ,mg. .._ . ..._ '
. .. . i. . ., l .. ; ..._..,,[' - , ; . r 1 . : < . 0.5 0 '---'.---,.--r---t'- - ./ ~ ~ i.1 - . ' '---.~-- ....... t ,d's r ' 'A . - _. ___.2.... . . .* . l __. . u . _ __.. 3'. 1 L ~ i l1 i ;. ' ' l l't;. . .. ;.t i . , ; ;. . a 0.00 O.OO O.02 0.04 0.06 0.08 0.10 PEF FIG. 5-7 TI AMPLIF APERTURE TOP OF CARD 3 ADJA' Also Anthble On SHOREHA Apertu Card PL ANT D . . . . . . s. . . i
- i . ;
i _y.... I i ~ l' 4r }-. . . , _ ._ irr ., o ry i- y - l in our - rima i ; , .v... ....- .. g .g I y ira ' I I la rime 11 IlWPr!ill,; r f 11 .1177 ITE Trn~lITifil 11 RND fldil l . .= - - - ~i .I JIIblJ ~ 133 ;Lh_i_ju rimr i - -' T _L_... rien ...._.__..L._ __ 1,- _. .1. . . . t .p...... . .. I ! 5 __ _ . _ . . . . . - _ . . . . . . . _ .u. .._ . . . :s m. . _... . .L..~. . l i , t ! + . . . . . . = l , g 'O** L. ., .. . . J . l . . 's ' ; l- . _ - - . . . . ~. .. . . . . . . . .-. . . . . . l * .t ..-.
- p. _ _ _.. , . . %,7 _ . . . . _
; . . . . p. .- . g. ., . . . ...-,.._p_r. { , i r . .. : !>iii: i: - ; i ' i ! ,6, is I i l I ,l '
- I .3 0.12 0. l4 0.16 0.18 0.20 0.22 0.24 IOD (SECOND) 1 l
OED RESPONSE SPECTRA OF VERTICAL ACCELERATION . REACTOR SUPPORT PEDESTAL $ENT VALVE OUT OF PHASE DISCHARGE-R AMSHEAD l 'M NUCLEAR POWER STATION-UNIT 1 NSIGN ASSESSMENT FOR SRV AND LOCA LOADS 8G12080645 -{9 REVISION .4- FEBRUARY 198f .. _ n --- 1 3.00 - . I l J 2.50 7 - - 2.00 - o w z .. _ . . . : ._ . . . t _ . . . i . o_ e : :* : . . 4 l.50 _ . . . . . . - . . . . - _ - . - . - . . , . . j QC : , ; W , 1 I . W -
- i O '
O 4 \.00 -. - - - - -. .i : . O.50' -- ,--- '.~,'. - - .. . ~ . - ~ . . . .-~ .. 7..'. ., , , q .._.,., _ , ,,.; ,,, ._7_. ; i . o.00 i > i i- i i 0.00 0.02 0.04 0.06 008 0.10 PEF l FIG. 5-8 TI ! AM PLIFl APERTURE PRIMAR'
- CARD l 3 ADJA(
i Ah Anthble on SHOREHA Apedes Gard PL ANT Di m.
- g ; t
. . , , ; i 4 . . merner 1: ter- - .. i irrmr,u t - - H - - " ;).a.pi 3 . ' as:LIA Ollt IIup uit . 4 ;3.pp3 ;;38:KLyl 0lp It urpW l ! j as . o.Ln i m .1- pu ~ ~ ' - t- z ps - . i ; .l l , f - l j i iI [ ! I
- i !
I l ! . . . . . . _ _ . _ . .. . _ t . i 4 . ! . . -- s a . ; i , . q._ . ; l , e
- i L. .
. . I_ .. . 6 t .' k i l .3 , Q , - - . { . . . . . ; h' . . _ . . . L . .....u... ..._W.. ,! i t. , r- -..____ __ ~
- .- . .! i;
{ l l ! [ ! E a l 0.12 0.14 0.I6 0.18 0.20 0.22 0.24 IOD (SECOND) lED RESPONSE SPECTRA OF VERTICAL ACCELERATION t 7 CONTAINMENT AT ELEVATION OF STABILIZER TRUSS )ENT VALVE OUT OF PHASE DISCHARGE-RAMSHEAD M NUCLEAR POWER STATION-UNIT 1
- SIGN ASSESSMENT FOR SRV AND LOCA LOADS 8612080645 4 D i REVISION 4- FEBRUARY.1981 --
l ( 3.00 - ; L' .
- s. .s.. . . . ,. .!....
2.50 -- - . n 2.00 - - - <o S ~" I O. -- --; ]-l-;l t - 4 150 g . ti - -- - - - = ------, --- w ' J ! ! 4 y : . . . . ;_ .. o . . ; ,. . i o / t. [- .... 'j. .\ ;. .. . . . 4 l,00 - ... .,', . - . 2 . . . . . . . . . {_ i P g q
- . ' r i . ,. I o.so u-- -
. *. . ! / . t* . .., , . 2. . . g.4 . . j ._ . .. t ' 4 i , : j : : (7 1 i t . ; !y
- '
- i i i 0.00 - - I O.00 0.02 0.04 0.06 0.08 0.10 Pi TI FIG. 5-9 APERTURE AMPLIFIED RE CARD TOP OF REAC'
, 3 ADJACENT ' Aperture O'Td SHOREHAM NUC i PLANT DEdlGN / '7W - . .. . . .. .. . . .. ... . . . .. .. p o. . . . . . . . . .j & i .. . _. _ ._ _ _: . in _ - n i in u-r i - si== .. ... .. m . mg uz . .u ; r i_la rimri >i _uu _. * ~~ ^ ire .it s:ri e IEE' Il.I: R Ilmr i _._. hl EIR _
- m. min. El l'E Ilmr _r NINE
...,. min <m-> lY _l'ilhl'm immi i _ _. e I ..j *' .I f * ...<I_ ... i., i l jf r .. .. l . ~ . . . - - -. 7- . - . _ _ ' .... - . !.. i. f 8 t . J . .. . ._ . e
- s . .. . . . ... .. .. . . . . . . . . . . . .. .. ..
j . 1 , _.....y.. p .. . . . . . C o........ . 3 . ..~~~~- ... O.12 0.14 0.16 0.18 0.20 0.22 0.24 RIOD (SECOND) SPONSE SPECTR A OF HORIZONTAL (N-S) ACCELERATION OR SUPPORT PEDESTAL /ALVE OUT OF PHASE DISCHARGE-RAMSHEAD LEAR POWER STATION- UNIT 1 LSSESSMENT FOR SRV AND LOCA LOADS 8612080645 c2l REVISION 4'- FEBRUARY 1981 - 1 3.00 - - - + - - - i , , , ,. l ,- .
- a. 2 . .
2.50 - - - 2.00 - - --- e l z O F- - 4 150 -a---- I r2 = .. , , .f.' w :t s . . w ,., , o , o . 4 g.Oo . . t.. .. .. L ....;.....;..-.._-. ', . . . . ... c l ' ' '., j'., 0.50 - - , i , j : ;. t , .- j .._ . ._ : ,,,,,,,..p.. .p ._. . .q .7,_,,, i \ I, .' i j 000 i i i i l l 0.00 0.02 0.04 0.06 0.08 0 10 PE T1 FIG. 5-10 ! APERTUBE AMPLIFIED RE l CARD PRIMARY CONT
- 3 ADJACENT V
* *" SHOREHAM NUCL Aperture Our PLANT DESIGN Al i 1 i l l i - . !! i ! i . p .. . ,, 1 .. .i . . r -- .j . .,.. , ;EEw . ..... p p m n i nmritmernor 1; [ . _7_ '._.,Z ~ LIT7 is r.m 7= ll 1 un IILer -_ gg,' c . ....... . .. .... ym_. g ., y_( ,, ,,, _ F ........__.j....._. .
- g. .
. u... I ? i 7 9' .
- ._ _ ' . ... j; '
_-_ 4 . I ...I . . _ . i i ,.....m.. . .l. 4 ' I 6
- 2. ... .. . . . . . . . . . . . . .. .. .. . _ ... ,.
... - ..,s + e . 4, , _ . . _ . . . i .; .,-- .; g , . i. . . ; . , . . . . t 4 . - . : i ! + . r, m - i , , 5 1 0.12 0 14 0.16 0.18 0.20 0.22 0.Z4 i fRIOD (SECOND) 1 SPONSE SPECTRA OF HORIZONTAL (N-S) ACCELERATION AINMENT AT ELEVATION OF STABILIZER TRUSS 4LVE OUT OF PH ASE DISCHARGE-R AMSHEAD EAR POWER STATION-UNIT 1 ISESSMENT FOR SRV AND LOCA LO ADS 8 612 0 8 0 6 4 5 - 9 Q. REVISION 4- FEBRUARY 1981 r l 1 l 1 0.60 . ; i [ ..t . .. . . . .a . . l O.50 - - l l l O.40 -. . . . . .. <D., l z __ i o - + H il ' i . 4 0 30 __...-.-L..- _M .. --.i .1. ..- m i i g! . w I . ! ! I.{ w . . o . . o , .- 4 0.20 - . - _ . . . - ._ . ..._..- . . _ j;g .. . .. . _. .-- . 1 ,l 1 , } . .. . . .~ O.I O ., -- a- f .-{p,. - , . . .- . . . .. - . . - - . n ,
- n '.
f-[,, ...+.... a, , ,. .i - .. .y ' . !,i _j__ o.00 ' ' l ' ~ i~ 0.00 0.02 0.04 0.06 0.08 0 10 Pl FIG. 5-11 U AMPLIFIED RE i APERTURE TOP OF REAC' CARD 3 ADJACENT N Ab Available On SHOREHAM NUCL Apertare Card PLANT DESIGN A l ~- j f ; ; , ,l i - .* inr...-... - r la rime i simia , , , , . . .. . .... ... la .ms'y m - r;.1.is rair iy _ _. t sulp ,J -4 .@P"I JE pit,a rize 17 nr y d '* .-23 :- - -- .hpl..J1 . LlWLLUM la - Lyt . I ~ + ~k {' f i ' ,I i t . ..t g .l......L......._...i. . . . . . . ... . . l .; _ . i . . i f , ; . I . 1 . . . l . .. ?. _ .. . . . . . . i t .. - . . . .. . . . . ...! r [ i -. h .. ..._........ . . . . . . . . . .. . . . .e , i ....... .. . . !._ .. . - .. _ . . . . ; . . .. . .. . . . . . . . . . . .-..._..i. - + l * , ! { l l t . I o.iz o.i4 o.is o.is o.zo o.za o.24 l ERIOD (SECON D ) [ l t a h SPONSE SPECTRA OF HORIZONTAL (E-W) ACCELERATION 'dR SUPPORT PEDESTAL ! 'ALVE OUT OF PHASE DISCH ARGE-RAMSHEAD . EAR POWER STATION-UNIT 1 l SSESSMENT FOR SRV AND LOCA LOADS 8612080645 cQ3 i REVISloN 4-FEBRUARY 1981 - t l t 1 l W& 0.6 0 - L , -..i._.. . . - . . . . . . I O.50 -- - -
- O.40 - -
O w Z .. _ . 9 - . I r - 4 o,30 L_ _ _ _ _ . _ _ _ _ _ . _ _ _ _ :___._. _ _.___ ._ --c m W ! 1 I ' g . . O : . I/,' _ O ! l' 4 0.20 ---- +- b - - - - - - - - i;-' I { ' i 1 =- , ! ;- '+4, r, , . *., 0.10 - - , - ' 'y".*.* . - -- - - - -- - :l - ' I 0 00 O.00 0.02 0.04 0.06 0.08 0,li l FIG. 5-12 'n APERTURE AMPLIFIED R CARD PRIMARY CON 3 ADJACENT Aliv Avd.lable 0" SHOREHAM NUC Agn. 6d PLANT DESIGN J 0 .i . l y e l 4 .___ . ..._,.._. .... ... a surrep, i , y .1. , . - . 4 .. ., . '~~;;;m - :E -).UtT :58:015 rat.E urIkli . fi-:.uC.i. Y'_ JnUE1: JI5"Jf.LM$ $ W5p3l } ' ..p wppa. .op masum .. a sprtu j !, .. . . a .. . .: . .I_ . 1. . . . t s i.I , 1 . g I I i l , l ! . l : r .. g i i , 0.12 0 14 0.16 0.18 0 20 0.22 0.24 gERIOD (SECOND) i I (SPONSE SPECTRA OF HORIZONTAL (E-W) ACCELERATION lTAINMENT AT ELEVATION OF STABILIZER TRUSS
- VALVE OUT OF PHASE DISCHARGE-RAMSHEAD l LEAR POWER STATION-UNIT 1 hSSESSMENT FOR SRV AND LOCA LOADS 8612080645 c2h 1
i REVISION 4-FEBRUARY 1981- - l l ry l f
- 41.70 FT.
i 4--11.10 FT.---> / d . :,$- s .; . A{ n. '. A - O. > *: i. if'. - . A, , , SUPPRESS 10N P00L J TOP 0F P00L E L. 27'-0" <- -_ _ _ _ _--- _-- - _ .=-._ - ._ -_-_=_- -_ _ = _ = :_::__ _- k, 2,. . s: A % ,> . i.: BASE OF */ . d' COVER MAT L ;-F, - EL. 9'-CT, SLAB OUTSIDE . .l-SUPPORT BASE OF PRIMARY MAT OUTSIDE CONTAINNENTg '8 PED ESTAL ,. ( 3 - CONTAINMENT \ . / PEDESTAL \ . /. .' . , . ' a . * ~s '. ? . *;': * : *i ' * . . L '.X - E L 8,_0,, ,',5., .. e ,- Z> .
- 17 s
FIG. 5- 13 CRITICAL LOCATIONS SHOREHAM NUCLEAR POWER STATION-UNITI PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS ;tEVISION 5-DECEMBDt 1991 a.,,a ., -.m a a - =an--~ ~ A.- 'O Nss a Qs _ \ N s, , OUTSIDe -.S,-,-OF N}. l :/'G 1 w"M .s SS y ( Nss TOP OF MAT f O Tg J M as M YT 6 N ss N/g Nst NT FIG. 5 - 14 POSITIVE SIGN CONVENTION FOR INTERNAL LOADS
- O eNO eNAN ,uc<eA Fe erAr ON-uNir .
PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4 - FEBRUARY 1981 - ..---...-,---...-,---.-,...,...-m. .,__-_.--.,---,-,,-__....c.~. - - - --,._,-,.. ,.-,--- --..- , - , - -.. ,--.- - - ---,. O O O ~ .! I 6.oo 4 s.co - 1 o 4.00 - l 2 i 2 ! 4 soo - - } E d I l l N L fg l\ i \l j i.oo i i \/g\ ~' o.oo ' ' ' * - * - ' - " - ' - o.oo o.oz o.o4 o.os o.'oe o.io o.it o.I4 0.16 o.is 0.2o o.22 o.z4 PERIOD IN SECONDS LEGEND: l DESIGN B ASIS CURVE ~ ~ ~ *" "' " " "' '" ncoaP RATED A PLIFIED RESPONSE SPECTRA OF HORIZONTAL ACCELERATION 0.Olo DAMPING VALUES l PRIMARYCONTAINMENTAT EL 21 FT 3 ADJACENT VALVE OUTOF PHASE OfSCHARGE t SHOREHAM NUCLEAR POWER STATION- UNIT I l PLANT DESIGN ASSESSMENT FOR SRV & LOCA LOADS REVISloN 5-DECEWeER 1981 1 . - - - _ - - - - . - . -. . .- ~ ... . . . . _ _ . - .- _. - . . . - - - - . . O O O i l j 3.00 l 2.50 - t I ! h f , 2.00 - f II l 2 II i 9 f 11 f.50 - 1 J l f \ El I \ I I\ ! E i.00 - I ,4 /\ I I f 't I I \ I ,\ i I \ I \ \ 0.50 . f l \l V \ I} g \ r - ~~~, - ~ l g i i i i i r ~ ~ T* ~ ~ r - - - J _ _ l 0.00 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 0.18 0.20 0.22 0.24 l PERIOD IN SECONDS i i ! LEGEND: I DESIGN BASIS CURVE l ---- SHORT PRESSURE RISE TIME FIG. 5-16 i INCORPORATED AMPLIFIED RESPONSE SPECTRA 0F i VERTICAL ACCELERATION 0.oio oAurius vAtuts PRIMARY CONTAINMENT AT EL 83 FT 3 ADJACENT VALVE OUT OF PHASE DISCHARGE ~SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV E LOCA LOADS l' REVISloN 5-DECEMSER f981 i ~ O O O l.20 f 1.00 - [ lII I O l.80 - l z l1 9 g1 $ I.60 - il d Il 1.40 - i i 1 /\ l I \r \ 1.20 - p ,hf Y sJ \_ _s" % 0.00 ' ' ' ' ' ' ' ~ '- ~~ --- 0.00 0.02 0.04 0.05 0.08 0.10 0.I 2 0.14 0.16 0.18 0.20 0.22 0.24 PERIOD IN SECONDS LEGEND: DESIGN BASIS CURVE g , 7 --- - S H " "' nC0RP AE AMPLIFIED RESPONSE SPECTRA OF HORIZONTAL ACCELERATION O.010 DAMPING VALU ES SECONDARY CONTAINMENT AT EL 38 FT 3 ADJACENTVALVE OUT OF PHASE DISCHARGE 'SHOREHAM NUCLE.".R POWER STATION- UNIT I PLANT DESIGN ASSESSMENT FOR SRV C LOCA LOADS REVISION 5 DECEMBER 1981 O O O i o.70 \ k o.so - [g l 1 t o.so - II l\ . I\ z j l I) i o o.4o - l l I g l 4 I[ l ! ! $ l} 1 l I i da - I \l \ i 8 0 \I \ J 11 li l1 ) i o.zo - II II I - , i V \ l
- d I \
\ o.lo e \ ,p 1 N 1 N % ~~ o.oo I I I I I I ~ * ~ ~ ~ , , o.oo o.o2 o.04 o.os o.oa o.io 0.12 o.14 o.'ls o.is o.20 c.22 0.24 i PERIOD IN SECONDS I LEGEND: FIG. 5-18 , l AMPLIFIED RESPONSE SPECTRA 0F j DESIGN BASIS CURVE VERTICAL ACCELERATION , ---- SHORT PRESSURE RISE TIME '" "" """ SECONDARY CONTAINMENT AT EL 58 FT I' 5ADJACENTVALVE OUT OF PHASE DISCHARGE o.olo DAMPING WALUES 'SHOREHAM NUCLEAR POWER STATION - UNIT I PLANT DESIGN ASSESSMENT FOR SRV E. LOCA LOADS j
- nEvlStoN S-oECEM8ER 1981
( 3.00 m 1.0 PERCENT OSCILLATOR DAMPlNG 2.50 - 5 l ! j 0 =TI ME g 2.00 - ,.- / 1. 2 = TIME w - l #. 1.4 :TIMG 2 ,; _'l'i*l L8: TIME , J.l __j O F [..Jd. . . i 1.50 /. . - ....a._..i...E ! ~ 0: L u.) . I _ ..t- t
- 1. J o . . . . - . ..l i...-
O -- l-[ J < 1.00 - - . .+4-i . .. ' I *. _ 4* 1. s i [ . . ./ W... ! - O.50 i 7.ii -.i , I I1 -- i -I,.; 33-rij$,.$,p/....-.l'.? - 3 ._yg.gTgr, ;. i - ; j,4. . i . . . i. aw;p ym-mww . .....u. . . . . . .._ l t' l o.00 ~ i i ' ' ' I l o.00 0.02 0.04 0.06 0.0 a o.io PE i l NOTE: THE PRESSURE MULTIPLIER USED l HERE IS EQUAL TO 1.0 FIG
- 5- 1 i <n AMPLil APERTUM TOP Of CAUD T-QUEL SHOREH Alro Available O PLANTI l
Aperture Car l -- ~~ l 1 I 7 ;.ill
- .t I!'.'.l._ - .. . i
! ACALE'1%CroN j SCALE FACTdR SCALE FACTOR . SCALE FA GTOR - +t L ,,. ; r, ; . . .. g . ,. .. . .. . l ....q_ _ i . 4 a i * * ~ 1 i '*~~~~~~~~~ . ...g u _ - , , r T'F D llj ' ..., ]. .{-. qg j._. _
- .- ! i . . . <
7 9l ., .. . . _i i i ! j. i ;; i - , ,. q"y . . .%.. ,,5y. nes _. i , . , og i . e 0.12 0.14 0.16 0.18 0.20 0.22 0.24 00 LOD (SECOND) 9 [ LED RESPONSE SPECTRA 0F VERTICAL ACCELERATION REACTOR SUPPORT PEDESTAL 4CHER ALL VALVE DISCHARGE-PRESSURE TRACE NO.1 AM NUCLEAR POWER STATION-UNIT 1 DESIGN ASSESSMENT FOR SRV AND LOCA LOADS 8612080645 oD REVISION 4-FEBRUARY 198t 3.00 - - l.0 PERCENT OSCILLATOR DAMPING 2.50 - I.. t i ' [. ik, l.0 kU5 y 2.00 - ;, . f r-- 1.2 = !Tih + u+ - 1.4 : Tin z , 1.8 : Tih o . _ _ ._p;- - - ~ ~~~' q F ; ; .l- . 4 1.50 -- j~ 1i.1 (r li . i 1. y L_ _y. .. o : .. .. . I g i u _ _ _ . W : i . U - < 1.00 N~ '~ _ ;. r --b_..'[',., p a 2 .I . c.p . .i ., .._f ___ , ,,? i . . ,_ s . i,
- . , i ...
I y ._. -- . H _. . ( _ , .
- ..g g.4 y. f_.y 0.50 g, . ,
.. . ,, . . l -. .
- p. 7: - ,
]i,
- 0.00 ' -
O.00 0.02 0.04 0.06 0.08 0.1C P NOTE: THE PRESSURE MULTIPLIER USED FIG. 5-HERE IS EQUAL TO l.0 AMPL TI TOP C APERTURE T-QUI ( C \RD SHORE PLANT Also Available On Aperture Card I .s ).' . l: ! !j ,; i lE 3dAIt!..E FACit . , . . .Nt' fE SCA'LE FACTOR ; IE SCALE FACTOR IE SCALE FACTOR , [- ;. , j L- . ji,1 . ; I'll .I'j ; t 4, . '.I.. 2 [. i i , i" t 2!;! . i i.i l 'l ,i ! ,i..i., ,i I} l !. ._ H., . i , , i e ! . . I * , ? . g
- l. _.. . , .
*f.;;. _"., 1 i .i; . 4 i l l l 3 . 't f .j- , .- - . f ! l . f. . . . . c . , ! i
- i. -(, . ,
i , - ,i . . p ., . ..L_. .a Lt. i . .., .. ; ,. . 1.cs , .:w.4 - . -?,- .r-. p-i , .. s - . *' g=:-. t . t.. . -
- s. -
a j..- r, . . . , _ . m i.... % _ .. A LC,..- _. il ! 'l'., -d. , i - . . . .. ,'!' :'M lli .. ,.I+ D#' ..^** l ll'! i r '}..gjd.j i l >: - I t 31 ~ iii, . l *: m - .4 h. :lt ! , i . .l . li i i O.12 0.14 0.16 0.18 0.20 0.22 0.24 3RIOD (SECOND) l L20 FIED RESPONSE SPECTRA 0F VERTICAL ACCELERATION )F REACTO.R SUPPORT PEDESTAL lNCHER ALL VALVE DISCHARGE-PRESSURE TRACE NO.2 %AM NUCLEAR POWER STATION-UNIT 1 DESIGN ASSESSMENT FOR SRV AND LOCA LOADS 8612080645d[p l l ! REVISION 4-FEBRUARY 1981 l n -~ 3.00 - 1.0 PERCENT OSC1LLATOR DAMPlNG 2.50 - I ! I I i ._...i... . ' 1.5 -. . . . p l . - 2.00 - l' . ,-j/ o . l ,, 1.'2' = TIME ;, ij j_. f 1.4 = TIME Z , 1.3 : TIME .o ..L_'-- ~!.} .. , I s -. j < 1.so - ' .. x - ,. -u s .l ,. !. W i I i i .I y : o I,lc I o ;jp > ;- ci 4 1.00 9 i ) i , ,, _ ,, I . , {! I, _;__ I _. a 1 .Iji g 41 : .i 3 J .. %+ r Q ,. 4.
- c. . ... . .. i,. >
g' J. . . . . . .i , 'y, l \. a , L.! . 1 . .! ' . ,ej 0.50 . - - : v -, ; y,7!. 'c y., 4 . , .. pE ;. q t . 4 ';"b.7,4 I - ~ i." f +htL 3u .I+ l- l * ,J - A .. 4 ' ~ ,@sh,,4 "1 +- ,,Mtin)2 y,u , z._ .- J. 0.00 1 r- ' ' ' i O.00 0.02 0.04 0.06 0.08 0.10 PEI NOTE: THE PRESSURE MULTIPLIER USED FIG. 5 - 2 HERE IS EQUAL TO l.0 11 AMPLIF ! APEllTURE TOP OF c utD T-QUEh SHOREHJ Also Available Os PL ANT D Aperture Card l .- ~ ^ - , - l l 4 I w i i.i; . \ -- ' i \ SCALE! IA'CYOR ' SCALE FACTOR SCALE FACTOR SCALE FACTOR ,
- s r
; ii ll ; j : J -- i - i -. i ., ;, }' ._ .1. , s l'; ; j . \ . t.,. , t, ) ! 4 , ; . 1 .; - I . i i, li .. , . Ac . ... . ii . . . . . . . . _ ...l .,lI , ~ r _ . . . . . . . _. . . . . . ,e . ; , . , _ . . . .4 . .p l...,41 I . . . . ;. . . ; ; l e , , . _, , ,. a* ? 9 w .. .; . . j . c' # .. . .9, : . . , _... ,; i. r- - -
- a.
I C ----- ' (( / ' ' }i!!'; Arrli.l -- ,__lp , f ,' - k..;.,.I-P-7"~,r-'i!E-[,- S ._ . . . i T t l,'
- i.;
I .2 4- ' N.,s.i.! V . .:../ ( i ,. ! , _ s n,,." ! - , A , ; .L ,- l i 'i i w;. .n., - - .o ,' ,L m, "*-=H. . .7. .. i l i .i ii a. !j i--- l- ;{ti o.12 0.14 0.16 0.18 0.20 0.22 0.24 ilOD (SECOND) . , , ~ 1 'IED RESPONSE SPECTRA 0F VERTICAL ACCELERATION REACTOR SUPPORT PEDESTAL ICHER ALL VALVE DlSCHARGE- PRESSURE TRACE NO.3 ' (M NUCLEAR POWER STATION-UNIT 1 ESIGN ASSESSMENT FOR SRV ANDLOCA LOADS 861208064547 REVISIONA-FEBRUARY 1981 3.00 - -- t r- - - - j 250 - - - - l - 2.00 --- - .* .o. , z . .o I-4 1.50 -; :-- -- -- ---- - -- . a: W i J i - - y . , O ; U i .
- 1. -
4 1.00 -- I -l-'.. . j.o.,:. ,'.....'..q l'. 0.50 -l. - - -- - - -- - ~- ~ ' *r * . . . - - . - I l : l i !. l 0.00 O.OO O.02 0.04 0.06 0.08 0.10 PERI NOTE: THE PRESSURE MULTIPLIER USED FIG. 5- 22 HERE IS EQUALTO l.0 AMPLIFC U TOP OF APERTURE T- QUENt CARD SHOREHAl PLANT DE Also Available Om .- - Aperture Card l f l i t ,i ;. . l , ! i . . . . _ . .. ;. . . . .- .; . . j ...:....... 4 ; , , , .nma in -. - .. .. 1 , , i . . 1 n ... , , , W. . '. . .r. n. . .. _ ; . . .:. . -- . i . .. i ; l a f l ! 8 8 i i
- .w , . . .._
_n..-. _ , i . - - - ..-- n.-. - .a . .._ . . .. ; . * ~ ~ ~ * "' ~~ ' . "e' * ; n.-in - _u n a mu- .- -_r-7 . I
- g. .
i, , ' I - I i . t i i j f . ? ..-.,...m.. ..A . . . . .f. . ...J., , i i } . . I l t i - ' i l .. l , . ! i . . . . .I . .. . . .y. . ,,* *,,v.'% . . . j . . . . -. . . . ', .. i i **,4 %g .e....* ..= ........ ,1 .,,.? * . , , .,, .* ? g,. , l l . - I l l t t ' l I i l i 1 b 1 0.12 0.14 0.lG 0.18 0.20 0.22 0.24 , 3D (SECOND) ~ / \ \ D RESPONSE SPECTRA OF VERTICAL ACCELERATION REACTOR SUPPORT PEDESTAL ALL VALVE DISCHARGE fHER NUCLEAR POWER STATION-UNIT 1 8612080645 if 'IGN ASSESSMENT FOR SRV AND LOCA LOADS ~ REVISION 4-FEBRUARY 1981 1 t 3.0 0 -- - -- -- - - > -- --- , , ! . , I i t 2.50 2.00 -- -- -- - - - .o, z ' .O F- - 4 1.50 -- e i w i ! J ; ..a., g i. O O h ! . i 4 1.00 -- ,kk, - -- - ,l ~ .'<, - --V- -- ,\ ' ' . , f l . ) .1 Y p , '.'.... l . \r. '!... ' s',.-- 0.50 - . 2 - -
- ...~~ .s~~~ ..'. -
., 7; j . . !, t, . 0.00 ' ' ' ' ' O.00 0.02 0.04 0.06 0.08 0.10 PERI NOTE: THE PRESSURE MULTIPLIER USED FIG. 5- 2E HERE IS EQUAL TO l.O AMPLIF[ i T1 APERTURE PRIMAR} T- QUEN C AD SHOREHA PL ANT Dd Also Available On ) Aperture Card { _ i l - _ _ _ _ _ _ . _ \ l i t i . . . ~ . . . . . . . . . . . . . . . . . _ . . . . . . .....7_._._.__-..,... , . ._ y i e i l-- - .' l' . , L . . l i ! u' s I l l
- 1. . .. .. - . t. . ., . . ..._:
? i i I 1 i t i g ,, . g.. , . . ..i. . .. .1 i - - i . r --,n .r. . i ! i _ , . . _ . _ . _ . .. ....... m m - 1 i- man rur . . , r .i-- i na fgi akan n. mygyg .i e 6 i .. 6 o.- - , j i i , .. g..; .,t-- mi i r-J- . r > - - i i i l , ; , F Wm - = = = == & i - - = - ': , i . . f E i t . . . . _ . . _ _ _ . . . . .. . . . _ l . . . . . . . . . . . . .. . . . _ . . . ' 7 i t, , p . . - --- - - ~ . . . _ - - - - ....2..---._... . I , 6 ' t . i. i i , , , , , , . . . . .. - ;' ,.. .j .. . -3 , . 4 - - = - - - .. ~ . - , , g .. ;.*'...... i . ........ 9* . . . .. .. - . ,-- .....1 1 ~ m j.- ..... ' .. 0- ... ... m.. . .,......... , l' { , t ( , t 0.12 0.14 0.16 0.18 0.20 0.22 0.24 r TD (SECOND) : ED RESPONSE SPECTRA OF VERTICAL ACCELERATION CONTAINMENT AT ELEVATION OF STABILIZER TRUSS HER ALL VALVE DISCHARGE 4 NUCLEAR POWER STATION-UNIT 1 8612080645 -2 ' IGN ASSESSMENT FOR SRV AND LOCA LOADS S REVISION 4-FEBRUARY 1981 . - L ._________ .___ _ ______.__ _. -- - -.-_-- , t 1 0.60 ~ , . i i I l ? l 1 . l ) i O.50 -- -- - -- -l- ! I
- i. .
! . i , I . i e Q,fQ -. - .
- l. . .c,.
o I 2 ) . ( .. ... _ .. : _. . . . . . . O. .. I- . 4 0,5Q - [ ...,...........__L._.......;........._..._............ , E
- t i i W l t i
J : . g . g._. 7. . .. - .i . . . . .._; o :[ i t o : :. ; i i 4 0.20 - -- 1,' ' -, l--- ----- - - -, {-- - - - -- i- - - - . - - --- ! -- --- ,~ - -- . - - E
- n. ,
i . ~~ i n l . .; . I 't f., , .! - l- ; -
- i. .,
j' p,.s , . .1 , . . . i r ,1 .. . ' ( - - *,. t 0.10 - -, ?*-- .- - - - ^ " - .' N .. .. k- .j. %g- , . . . . ! + .. 4. ._..,. ._ , 0.00 i ' i ~ . l , ...': i Phuam O.00 0.02 0.O? 0.06 0.08 0.10 PEF NOTE: THE PRESSURE MULTIPLIER USED HERE IS EQUAL TO 1.0. FIG. 5- 2 AMPLIF 'O TOP OF APERT M T- QUEN CARD SHOREHA PLANT D <W Available On A neture Card u l l _ . . _ . . _ . - . . _ _ _ m _ . __ _ _ . . . I i i l l l 1 l L. . . . . . , . . _ , - - - . . . . . . . ~ . . . . ...y...~.._. . ... .. .. - . ... ... . .. ( .l.l , !, i I i . , . , ... . . a. . . ... ... .. : i O q l ......t..._._____.,_._ .. . . . . I r I f 9 1 L 11 _.:.' R!E 'II I M I rm = . . . sr T M11I _ e , i i , r ,- , - - rr r i- r- r i .~ - . [ , 4 ><me<m,< moves.< G 3 IlF lbE @NNkfl M"* l' ' l FE ~ l ~l_ ~ T l # 1 I t I ~FI . ! F ' 4 ~ l l l. 2 - ymymv..e i . . T U i"" o_,._ .i i 7 mi i - 4" ? WM .- - - r -.=i i {
- l. ....i........ -;
i ;y ;_ gg L -. ,-- r. ~ .j. . . . . . i r l . t I : . . .., . .. . . ...i, . i . .. _. ..... . . . l . t t ' t . t i p......__.,.-._.,._~.2... i . 4 : - , , .. . . . .. p _ _ .. ... 4 . [.... i .. . i : - i ' ? ; .; 1 .._-__.J ._ .1... . . . . . p .. . . .. . . . . .. . g . j . _. . : l l. ;- l f - <
- s !
3 p _ .. .. = I ___ _. . . . . _ . . . . . . . . . . . .:........,... _ .. i........_. ; I' I ! , i j l t * - 5.._ .._ _ _ _ _ . __ _ . . . . . . . . ... . . _ _ . . _ . . . I . h . . . . _ . . . . , _ _ . _ _ -. . . I. . - ...._. .. . ., . l i ! i b=MN-m ~ l O.12 0.14 0.16 0.18 O.20 0.22 0.24 i i \ t . LOD (SECOND) ; i ? 1 ED RESPONSE SPECTRA OF HORIZONTAL ACCELERATION l REACTOR SUPPORT PEDESTAL CHER ALL VALVE DISCHARGE 8 612 0 8 0 6 4 5 -3C ! M NUCLEAR POWER STATION-UNIT 1 $ SIGN ASSESSMENT FOR SRV AND LOCA LOADS t REVISION 4-FEBRUARY 1981 l _ _ _ _ 3.00 - i
- i ,
t 2.50 -- - - -- -- - - - , 2,00 - . . . . o w Z . O 4 1.50 - - - -- -- - - - - -~ - - E uJ ' _J .. . i O ; ! U 4 1.00 --.-:.....'..-.......: i .. - . . . . ~ ..: (.. 6 i t ......;..__..I,..... _ ,._._..i. . . . . . ! .. . 1 t , O.50 ~ ~ ~ + - -; - - + --- - -- } - , I '! i e t * ! : I i i ; . , ; . .i . . . . . .. . 0.00 ' ' ' O.OO O.02 0.04 0.06 0.08 0.10 1 . PER( f NOTE: THE PRESSURE MULTIPLIER USED FIG. S- 2 HERE IS EQUAL TO l.0 AM PLIFO' T1 PRIMAR' r APERTURE T- QUENo CARD SHOREHA[ PL ANT D@ Alm A;411able On , %m Aperture Gard l j l I b....... , l , t t }- ; . . . i
- I ,
,. ; .. . . _ . . . . .J..... , ' ~ II . rind M R
- Il l El rnr. ni anr r ula:
.vk.h k .a 11.nl il RR MI I e 11 Inu r mui enerma r-i '- -~ , ;j J. _:_ . . . f JmKJ , , UDydI I M IN ~ "' '"F I EA I f , .. - - a 1 - DaUD J [. . - II I u rein unpreg **.-~~- - -**'~~ _ ,, , , y ,t m .i rim 2 i a na i-lu anyam2: e' ~ ~ - ' ' - ' - l j i t t t ; .....s... . 2 . . , . I l . I , l : . I i t , . . . ._.2.... . . . . . t . l t l . i I ! : l , i I . . ; I. I ,. I i ' 1 0 t I i 1 j o.12 o.14 o.1s o.is o.20 o.22 o.24 3D (SECOND) f I l D RESPONSE SPECTRA OF HORIZONTAL ACCELERATION CONTAINMENT AT ELEVATICN OF STABILIZER TRUSS )HER ALL VALVE DISCHARGE NUCLEAR POWER STATION-UNIT 1 8612080645.-3' -IGN ASSESSMENT FOR SRV AND LOCA LOADS l REVISION 4-FEBRUARY 1981 . I i 3.0 0 -- - - - - ! i i ; I . 1 2.50 -- - - '- - .. 2.00 - . . to w z .
- o. .
4 1.50 -- - - - -- er , w . J , w .. i- r o . t o 4 1.00 ---- -- r. i ...j. _ : ; . . . . .. I .... . _ ,_, ,I , o.50 ~ 3.... , - - . - . . . ... . _. '[- - - t ~~ !: ..~. ,'..': _ _ ___ j~ - - - - , i l i f , 0.00 0.02 0.04 0.06 0.08 0.10 PER NOTE: THE PRESSURE MULTIPLIER USED FIG. 5-26 HERE IS EQUALTO l.0 p TI TOP OF APERTURE ' T-QUEN cAno SHOREHA PL ANT DE Ah Avanable On Aperture Card
- i i .
! i ! i e i l , . w.rn e i nn - m .o* a .....m; i j ==.- = w. l. n.n i + i,& w' s ,. o . i ,. .: .' e ,,;* . -' ;. - + . dt m ,*,! . o.s z,u . ,o. o . 1.2 .; D ' . o .... . ! .,< , , , :6 ,o. m . aur . pg ; u.on , i on. a i i .i.o. u enr,mu. .- I e r 8 ..;...l . i . . i
- -; . ;.1 !
- i !
- . i.
1... . l . l . ! : : - i , .i . I i. .l..... g ........y...... ...... - mg -. p ............ ..........--. .. n l ' l i i I h , ll l 0.12 0.14 0.16 0.18 0.20 0.22 0.24 POD (SECONL'.) I l l ED. RESPONSE SPECTRA OF VERTICAL ACCELERATION ! REACTOR SUPPORT PEDESTAL $HER 3 ADJACENT VALVE DISCHARGE $ NUCLEAR POWER STATION-UNIT 1 8612080645-30 lSIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4-FEBRUARY.1981 3.00 -- - r- ! i I i 2.50 -- - - ~- 2.00 o w 2 - - O V 4 1.50 -- - -- 0: W - J W o ' o 4 1.00 -- - i ~ j' ?1 0.50 - -- .^. . .s ... ' ' .l . .___. O.00 O.00 0.02 0.04 0.06 0.08 0.10 PERIO i l NOTE: THE PRESSURE MULTIPLIER USED FIG. 5- 27 HERE IS EQUAL TO l.0 AMPLIFIE l U PRIMARY APEitTURE T-QUENCI ct3D SHOREHAM pug PLANT DES Aperture Owd m ~, - - - . a _ y ...q.._....... , i
- I i,
= i : i
- i. .
r a 4 . .. .. y . . - -. 1 .. n ' 43 j . Jial.IU b. . ,.Ug 2 I I I M E Hw- H 4MFI Nll i . .- g..sa --F1 rpr
- .t.w 1 un ii i u i nw : n anr a iuli.
~ - * .ri' hp r nn top E.ichilui u. . .__ Ti i a i) ! ---wer. e _ bin ;..uumiu rn= " anr s =r ! I ~ ~ ~~~ ***F'* ~ ' ,u un nh ' .l a, im .riiu inn nienr a me .i 6 l .. . 92 , t *t l i, 4l ; .....;.. o n ..
- i. .
i i i e - ! !, I.; . l . i ;. ..' . . . . .'..-e. s . - - * -l . *. . y = , . = l i '
- : j !
j . :. . . . . .; - - y . .- .. . : i. e ! ...; _ . ._.I........;......-... . . . . . . . '. ... *.. - - . . s ; . , , , I * . . . :. u . '- . ; ;j , ~ .,_-_- _. w.... .c......,. .........-.....g .,.. i !
- l l '
4 O.12 0.14 0.16 0.18 0.20 0.22 0.24 D (SECOND) 4 D RESPONSE SPECTRA OF VERTICAL ACCELERATION CONTAINMENT AT ELEVATION OF STABILIZER TRUSS iER 3 ADJACENT VALVE DISCHARGE NUCLEAR POWER STATION -UNIT 1 IGN ASSESSMENT FOR SRV AND LOCA LOADS 8612090645-33 REVISION 4-FEBRUARY 1981 .- 1 I 3.00 - - -- - '- l 2.50 -- i - 2.00 -t-e '~* . z 9 l N 1.50 - l 1 oc \ w - W - o ! o - 4 1.00 - l 0.50 - l $.; a ; i . .- !: ' 'i !, - i-, 'ti i' t. ' ' ~ 0.00 ' ' ' O.00 0.02 0.04 0.06 0.08 0.10 peril NOTE: THE PRESSURE MULTIPLIER USED FIG. 5-28 HrRE IS EDUAL TO l.O. TL AMPLIFil APERTURE TOP OF 5, AnD T-QUENC SHOREHAN Ak Avalable On PLANT DE! Aperture Card . - - = .j_- - U-"'at*mi yt !(,LAfspp.3ligeIggj ,., Mi_rhr r 7 t_ e_a_ _ r_ t a r a_ m&s __ m_. - , r.m - n- .. .l!!- 523LLA OR;a tatig; a 1 -pra.e n ,- 0$$.2 $$ 2LLA ial harINe! - l - ,- ,. Etapilussi;I HJam-ainrzus! { 4 I l L l i l. l l . . * ' ' , , -. .- - - - - . . . . . . . . . . , ' ' ) ~ 1 I I I i 1 i i 0.12 0.14 0.16 0.18 0.20 0.22 0.24 l (D (SECOND) , I [D RESPONSE SPECTRA OF HORIZONTAL ACCELERATION [ aEACTOR SUPPORT PEDESTAL i HER 3 ADJACENT VALVE DISCHARGE !- l NUCLEAR POWER STATION -UNIT 1 8612080645-3 . l-l flGN ASSESSMENT FOR SRV AND LOCA LOADS i REVISION 4-FEBRUARY 1981 ; 1 4 - - , i l l 3.00 -~ l . l l 2.50 - - 2.00 -- to 2 ' 0 E 4 1.50 ' 1 E l w \ .J l uJ : C u 4 1.00 - - ' i l' i ., ,* ) ' ,i . l'., . , l 0.50 - ~ ,a .. . . I ~~~.....__, ~- .nu 0.00 ' 0.00 0.02 0.04 0.06 0.08 0.10 PERI l l NOTE
- THE PRESSURE MULTIPLIER USED FIG. 5-29 '
HERE IS EQUAL TO l.O AMPLIF T1 PRIMARj APElyrunE T-QUEN4 (ARD SHOREHAh PLANT DE Also Avsnable 03 - m. APcrture Card l n -enis: n .nm - ne, es':rd hv c.m.- ;r-vai riinEn. a, ~ p . I k M .-..e.r7,r. - opau m: rot =;a ; n=> I ml ' mee, 'nyp ny:Ig=inF = ==15* . . 1 l . l l 1 l g :::::: g 'h=== ---- ':::::... _ 1 l 0.12 0.14 0.16 0,18 0.20 0.22 0.24 fD (SECOND) t D RESPONSE SPECTRA OF HORIZONTAL ACCELERATION CONTAINMENT AT ELEVATION OF STABILIZER TRUSS HER 3 ADJACENT VALVE DISCHARGE NUCLEAR POWER STATION-UNIT 1 861208064 5- . lGN ASSESSMENT FOR SRV AND LOCA LOADS 35 i l ~~~~- - I REVISION 4-FEBRUARY 1981 SUPPORT PRIMARY PEDESTAL CONTAINMENT { O 3.s 46 .s e i s- ., . __ - _ ,_ _ _- ._ _ _.__ ._ _-_ s _ g .: w .a. U iP . ** 1r 1P 1P it .= . * ;/. i. v, . s,..-4 * , , ' * . . s. .. . % g . .. . , . , . .* > . . a. r . ..'s . * * .s. s ... .. . .;.*, . .... .. .. . .., . e. * .s :. . '. s . ..,.o. s:. t. . PRESSURE DISTRIBUTION 50 - O , ON BASEMAT E S 3g _ ON CONTAINMENT w - E 3 m 20 - - m W x 10 - 0 I I I I i O.0 0.5 1.0 1.5 2.0 TIME AFTER LOCA (SFCONDS) FIG 5-30 LOCA VENT CLEARING IDEALIZED PRESSURE TIME HISTORY O SHOREHAM NUCLEAR POWER STATION -UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4-FESRUAltY 1991 k % W __=, = _ = _ = _ _ = = = _ =_. p.* SUPPORT .* ...- PRIMARY PEDESTALN, cf . s, / CONTAINMENT i l 4 1 e .o* ,,t. I ' t, 1 r 1r 1r 1r 1r ,. .,*,.',',,f.'.
- k ,' .' * *
[.'.,' ; .- =.
- .*...,o. .
g l s;s. .
- s. g,a. ;
3 . s, ., .: * . . .. a . s. . ...s - s' PRESSURE DISTRIBUTION m 20 HZ $ b bA" _ i W - ) V V E \ L f I I f I O.0 0.1 0.2 0.3 0.4 0.5 TIME (SECOND) 2 30 HZ = t k AA A . i m g W vvy- ~ t Q. I i I t i I i CLO O.1 0.2 0.3 0.4 0. 5 TIME (SECOND) FIG. 5-31 LOCA CHUGGING IDEALIZED PRESSURE TIME HISTORIES , SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS Revision 4-FURUARY 1981 : . ,--..,,,,w-y.-,,,-e, -,, y,y,, - - +
- 3. 0 0 .=.4 i.. - - - . - . - - .
5 l I i . l i 2.50 - - 2.00 -- - . .. . - O , w i . r t .~ . . . - . .. .. i ; i r h< 1.50 - .t.---..... . g g. w ; J - , w - .: l, o . . . 3 o i l l ,' 4 1.00 -- ..-- .. - . .,=, ... .-. -.-{- - -A.,, -.u,.' . ,i , l. , . ' * , . .- .....~~.... . . . . - - i ,,, . . ....... ' 3..:...... i. . .. .- . . . - . . . 0.50 -~- -A- - . . . . . . . . . . . - . ,. . . . . , _ . . . _., , _ ... . _ . 4, _ . _ .. 2.t . _ 1, ' t i .i 0.00 - 0.00 0.02 0.04 0.06 0.08 0.10 PEl FIG. 5-TI AMPL1 APERTURE TOP Of CARD LOCA SHOREH 4 Also Availahta On Aperture Card PLANT C NN i t. S l ._ J i'. . . . . . . . .:._ . . . . . . .: l . . > .j._._ ..
- 7. . ..;
'~ i !* ! l > I i l 1 I _ _ . I i
- Em um ii- ul Eimrimm Kimm ! *!
i * ~ - ' r -[ l El IIn r1Er rime *i .~.w.= w. ..gi ium : ..=<= f+3.7 1]Il] _
- 4. <
y , , li 1 In. w i 'Iin : ri nr-WMr r1 RIB Y r t-t-i- - i i ;i , i i - - ^ ~ ~ - * ~ ,pg ; =.-<.- -* n.i- _ i-- si il ms uir - rim. L ] l.;IH H + -5 . i . . ---..+...a.. ......y , , 'i.. I i s . i : t
- g .
w + . ~~ .... ,,*, l . . .. s ., : ....,,.g ------- ,,- * } q-.. p g,.... ... 7-- ...,,,..... . ..,,, .. . . .. ~ ; . i ! t ! e ; . ...4 .. i.i , . s t , . ,' 4 i , .l,. i 4 ; I . I' i - 8 . l ; ;, t > 6 -i 1 1 I i . . l O.12 0.14 0.16 0.18 0.20 0.22 0.24 hlOD (SECOND) l l 2 hED RESPONSE SPECTRA OF VERTICAL ACCELERATION ! REACTOR SUPPORT PEDESTAL ENT CLEARING 8612080645 -39 M NUCLEAR POWER STATION-UNIT 1 (ESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4-FEBRUARY 1981 y- i 1.20 .~- - - - - - - - - - - - - - - -- 1.00 - - i - 0.80 - C B 0 z .. - l o , i l-0.60 4 ' ...----:--'---+- % 1l i Laj , J I o ,:.... ., o . , , ; a 0.40 - - -- - . 4- , I f 0.20 - ---- - - - - -- - - - - - . . . . . . . . . . . . ....J....-. ' . - . , . *. . J. . s 0.00 ! 1 t 0.00 0.02 0.04 0.06 0.08 0.10 T1 APERTURE FIG 5 . CARD AMPLil ne Avanable o. PRIMAL Aperture Ca,a LOCA' SHOREF PLANT g g . . __ = v_-n - , , ,-- -. ~ .. a 2pwrdEm:x20rrm:m oat. aern
- _t 7 LihnlIE1 -. hill [$.i TM IE@!M.
..wur- m7 -... . _p 2.au - a.ma %, oo ,, q . .. . - t , . _ _ _ . _ . i . . ~ i ' . . i m 'w - _ - __ _ . . . . . . . . - . . - - - - - - - i ' , 1 . i. .. h { !r i .t .} * ; j ; t _ . .. . __ _ . -- l.- - . : . ;. ... - t . . . . . . . . . . . . .;., ., ' ., ~ i > , i %....... . . . r '[. . 2. . a i""" ~ t" -.... '.'.'.,, ,,,..... ., . r . _ . . _ . - - - ..t,.._........ --- . .
- ; 1 . .......
; i ; i , __ \ i 3,, ..__g - . --'-~? * * * , .; . . . . -- =- . i t : ; 1: {\ ' , i ' i I , - i i O O 14 0.16 0.18 0.20 0.22 W ERIOD ( SECOND) 53 71ED RESPONSE SPECTRA OF VERTICAL ACCELERATION RY CONTAINMENT AT ELEVATION OF STABILIZER TRUSS /ENT CLEARING AM NUCLEAR POWER STATION-UNIT 1 8612080645 -37 DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4.-FEBRUARY 1981 O O O 0.08 I c.oS - b o c.04 - l l z 9 I
- 4 l g o.os -
l d o - 8 'l ^
- o.or - i ,
s ,, % t ,{ s' .), 0.01 - v\\, _ _ _,- ', - 0.00 -. r____,./~~~~~~. i f f = a i I i O.00 0.,10 0.10 0.30 0.40 0.50 0.60 0.70 0.00 0.90 1.00 1.10 1.20 l PERIOD IN SECONDS LEGEND: FIG. 5-37 0.005 OSC1LLATOR DAMPlNG VALVES AMPLIFIED RESPONSE SPECTRA OF ~ ------ o.olo oSCILL ATOR DAMPING VALVES CONTAINMENT AT ELEVATION OF o.omo oSCiLLATon cAurlNs VALVES STABILIZER TRUSS - LOCA ------- o.oso OSCILLATOR D AMPING VALVES MISYMMETRIC CONDENSATION OSCILLATION SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND 1.DCA LOADS REYlSION 4-FEBRUARY 1981 O O O 1 1s.00 15.0o 8 12.00 - 2 9 - l >- ! j EM - U W U l's N LM " f\ i \ ) 'g r 3.00 - 4 g j ,-s , y ,. -- - sk *Ns -m. . -----.- x i i i , , , 0.00 0.10 0.20 0.30 o.40 0.50 0.6o 0.7o o.80 0.90 1.00 1.iO l.2o l PENiOD IN SECONDS ! LEGEND: j o.cos OSCILLATOR DAMPINs VALVES pgg 3,34
- ------o.oso osciLLAron o4MPius valves AMPLIFIED RESPONSE SPECTRA OF
------ o'.oso osc!LL To DAM N6 VALVES VERTICAL ACCELERATION - TOP OF REACTOR SUPPORT PEDESTAL- LOCA l AXISYMMETRIC CONDENSATION OSCILLATION J SHOREHAM NUCLEAR POWER STATION-UNIT l l PLANTDESIGN ASSESSMENTFOR SRVAND LOCA LOADS l REVISION 4-FEBRUARY 1981 l ! O O O 1 10.00 e.00 - ! o 8 S.00 - M . ,r a: ks i d o 4.00 - N f'. \ i 2.00 - \ l %,,
- ,, Nj, --- - .g O.00 ~~~~"~ ~~~"~ 8 I I I 1 8 O.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00 1.10 1.20 '
I PERIOD IN SECONDS LEGEND: 0.005 OSCILLATOR DAMPING VALVES ---- - 0.010 OSClLL ATOR DAMPlNG VALVES I 0.020 OSCILLATOR DAMPING VALVES FIG. 5- 55 ----- 0.0s0 OSCILLATOR DAMPING VALVES AMPLIFIED RESPONSE SPECTRA 0F VERTICAL ACCELERATION- PRIMARY CONTAINMENT AT ELEVATION OF STABILIZER TRUSS-LOCA AXISYMMETRIC CONDENSATION OSCILLATION SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENTFOR SRVAND LOCA LOADS REVISION 4-FESRUARY 1981 4 O O O o.ao - i o.So - I
- e 0.40 -
E D g 0.30 - 1 d 8 o.20 - ,g \ t 0 so - ' \ , . . l -# o,no d - 4 . _ _ . _ , - - q - - - ~ ~.,.- _3 , , , , , l l 0.00 0.to 0.20 0.30 0.40 0.50 0.so o.7o o.so c.s o iso tio i.go PERIOD IN SECONDS LEGEND:
- o.005 OSCILLATOR DAMPING VALVES FIG. 5- 36 i ------
o.olo osctLLATon DAMPING VALVES AMPLIFIED RESPONSE SPECTRA OF l o.ogo oscillator cAMPiNo valves HORIZONTAL ACCELERATION-TOP OF ------ o.oso OSCILLATOR DAMPING VALVES REACTOR SUPPORT PEDESTAL - LOCA AXISYMMETRIC CONDENSATION OSClLLATION SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENTFOR SRV AND LOCA LOADS REVISION 4 FEBRUARY 1981 J i 3.00 . - 2.50 - ^ 2.00 - e w z , .O 4 1.50 - a: W : a I: ..: ' w : : o : :' o ! 4 1.00 - 0.50 < - -- * ~.
- c ....... .. ... _,' .~....
i , , ' ' o.co 000 0.02 o.04 o.og g,os o.io PE FIG. 5-3 U AMPLIF APERTURE TOP OF C'ARD LOCA / SHOREH/ h o A u & hle on PLANT C Apertars Gud 6: I, *i. '!l
- g,' '
..i. n' nhlr me 'i . .l! '.irl (gehe' haish.g! ' i -h*- k- b 'O.01 ~ 3 0 61 ILLATORi(M MNNb - D'.Oth. OSI:!LLATOR'OlW ING: I --*-r--- 0.05 3- 06-;JLLMrcR cimF1C i i ! .a i . . I s i 4 F l ! l ' 4 , , ------- ,-___........ 3........__...r---- ----- ,- - - - ---1 j o.12 0.14 o.is . o.1 e o.2o o.22 0.24 l 1100 (SECOND) )ED RESPONSE SPECTRA OF. VERTICAL ACCELERATION l REACTOR SUPPORT PEDESTAL IXISYMMETRIC CHUGGING 8 612 0 8 0 6 4 5 -3$ 1M NUCLEAR POWER STATION-UNIT 1 , @ SIGN ASSESSMENT FOR SRV AND LOCA LOADS i l REVISION 4- FEBRUARY 198t - l _ i l 3.00 - 2.50 - - 2.00 - w e z [ O- .s kx 1.50 -
- / i.
W re i a ! : w : : o : o : ' <t 1.00- - l 0.50 - ,r .. . ' )' ' . ,~~ _______ O.00 O.OO O.02 0.04 0.06 0.08 0.10 PE FIG. 5-39 11 AMPLIF[ APERqmdE PRIMAR' CARD LOCA A) Aleo Avdlable On SHOREHAl Aperture Cara PLANT DE O.005 OSCLILLATOR DAMPING - - - ---- 0.010 OSCLILLATOR DAMPlNG O.020 OSCLILLATOR DAMPING - - - -- - - 0.050 OSCLILLATOR DAMPING i r -- , , . i O.12 0.14 0.16 0.18 0.20 0.22 0.24 RIOD ( SECOND) ED RESPONSE SPECTRA 0F VERTICAL ACCELERATION f CONTAINMENT AT ELEVATION OF STABILIZER TRUSS (ISYMMETRIC CHUGGING A NUCLEAR POWER STATION-UNIT 1 8612080645 -39
- SIGN ASSESSMENT FOR SRV AND LOCA LO/.DS
' REVISION 4-FEBRUARY 198i l 1.20 - -- - 1 1.00 - 1 - 0.e0- -- o c. 2 _. ; .I
- 'n, ,
_O. , ' i s : < o.eO - - - - e LU i a g . . _ _ o I
- ~.
. l' '.' o : . < o.40- -- /- i', , i . \. i
- ~~. .
O.20 -. t '.s ' / * .g. . _ . . . . - - - 1 ~ ~ i ' ' ' ' 0.00 O.O O O.02 0.04 0.06 0.08 0.10 PEq g FIG. 5- 40 AMPLIFIED MERTURE TOP OF R CMto LOCA ASY Also Avanable On SHOREHAM i APerture Cud PLANT DESIG t I , . , r. - g . . . k - 2 _Id$.[ LL- ,.M ?rb. E Ih M8' = == a ikk'+ - g[fHPF 1Dl.851y EDS : J'- t- - 8
- ' ~-
. , tr.orr os: I$j$$t2L*... rran urrn :- * *Fm-'WU,;050' : US#1LLapimt .i .or tmu. . . >. ,,,, .J - >> ,
- t. . .i.e.-
E ump . 9 \ i l . 7 - - - - - , - _ _ . . . . . . , . _ , __ ______ , l O.12 0.14 0.16 0.18 0.20 0.22 0.24 IOD (SECOND) RESPONSE SPECTRA OF HORIZONTAL ACCELERATION ACTOR SUPPORT PEDESTAL 1 METRIC CHUGGING 8 6l'2 0 8 0 6 4 5 -Yd UCLEAR POWER STATION-UNIT 1 N ASSESSMENT FOR SRV AND LOCA LOADS REVISION 4-FEBRUARY 1981 1.20 - . ..l 1.00 - ~ 0.80 - 1 o ~ I
- f. l 2
o [ .I. .. .t 1 F ' g 0.60 - -- - 1, ,~ ~ - , ... - + W ' ,,,g ' w f. O ' ! '. O ! 4 0.40 -. t
- t. . ,
t' '.\ , 0.20- - I. . . ., a / .,. - . . - . - - - - . ~ . . l ..- . '. ,s +.;,!,, , i i 0.00 ! I "i O.00 0.02 0.04 0.06 0.08 0.10 -P[ g FIG. 5-41 APERTURE AMPL!FIE CARD PRIMARY LOCA AS' & Avdable On SHOREHAK Apertar. Card PLANT DE8 , . l j , *9 e 1 e I i .l l , ;,. s , , . )) .i - - ima. II I in ,im, i-' r--- ; f i tn .IT, i;- .'i.gA. 1 .... ii .ii-o, I.. , ,, , ,. ,,.. _ ,-= 2 ,xu:pnwiann r - r-- ....... . . . . , , __, . z , , , . ,,.. . _ .y [ . : j p 2. r q.ggg , .-r ; -pt; . .. . . . t.._. .. i , , i . . i t '
- i. _ _.
. - - - . . . i. ; - i j ....: - . . . . . .
- b. . . . . . .
!.. i . . i . \ l: : ; : L'i a ai j ; l. I 1 . . _ ____I o.12 o.14 o.ts o.is o.ao o.22 c.24 lRIOD (SECOND) 1 ) RESPONSE SPECTRA OF HORIZONTAL ACCELERATION . l CONTAINMENT AT ELEVATION OF STABILIZER TRUSS MMETRIC CHUGGING NUCLEAR POWER STATION-UNIT 1 8 612 0 80 6 45 -Y[ lIGN ASSESSMENT FOR SRV AND LOCA 1 LOADS REVISION 4-FEBRUARY 1981- --- . - - - , , . ~ . - - - - - - , - - , O O O i ! 5.00 i 4 l l 4.00 - I l 5 s.00 - q a 0ll p 3 /* n ==
- rs i o
\ i I 1.co - 0.00 I ' ' I ' ' ' ' ' " 0.00 0.02 0.04 0.06 048 0.10 0.12 0.14 0.16 0.18 0.20 0.22 0.24 l PERIOD IN SECONDS 4 LEGEND; i 0.010 OSCILLATOR DAMPINS VALVES FIG. S- 42 ------- 0.0 0 OS ClLLATOR DAMPING VALV ES p p p 0.030 OSCILLATOR DAMPINS VALVES HORIZONTAL ACCELERATION REACTOR VESSEL ELEV 10 0 ! AP FROM RECIRC LINE BREAK 4 SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANTDESIGN ASSESSMENT FOR SRV AND 1.DCA LOADS REVISION 4-FEBRUARY 1981 i O O O 1 l J - i , s.00 I l t t j 4.00 - i i
- b
) $ 3.00 - i ! P i E [) >.
- 2.00 -
l p I O
- l'kg I.00 -
l o.co ' ' ' ' ' ' - - - > >
- ODO 0.02 0.04 0.06 0.00 0.10 0.12 OJ4 0.16 0.18 0.20 0.21 0.24 PERIOD IN SECONDS l LEGEND
- ,
O.010 OSCILLATOR DAMPING VALVES l ---- 0.020 OSCILLATOR DAMPING VALVE 8 FIG. 5-43 l 0.os0 OSCILLATOR DAMPING VALVES AMPLIFIED RESPONSE SPECTRA OF i HORIZONTAL ACCELERATION
- SHIELDWALL ELEV 137 l AP FROM REClRC LINE BREAK '
SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESKIN ASSESSMENTFOR SRV AND LOCA LOADS l REVISION 4-FEBRUARY 1981 i l O O O 1 l 3.00 - - - I i 1.50 - i 2.00 @ s I.
- s.
x o I1 l li I.s0 - \ E / ' i d / ' $ i.00 - l i
- 0.50 1 1 1 t 1 I i t i I i O.00 0.02 0.04 0.04 0.08 0.10 0.12 0.14 0.15 0.18 0.20 0.11 0.24
- PERIOD IN SECONDS ,
LEGEND: FIG. 5-44 o.o iO oscittaron naurius valves MPLIFE MPME SPmp OF ----- o.o20 osciLLaTon OAurine VALVES HORIZONTAL ACCELERATION ! 0.030 osCILLAT0n DAMPlNG VALVES PEDESTAL ELEV 90 i AP FROM RECIRC LINE BREAK
- SHOREHAM NUCLEAR POWER STATION- UNIT I i PLANT DESIGN ASSESSMENT FOR SRV AND 1.0CA LOADS REVISION 4-FEBRUARY 1981 l
O O O i i 5.00 i 4.00 - l s
- 2
- o 3.00 -
l E /g = , ! U \ s w .00 - g . i ! l.00 - [ E I 0.00 0.02 0.04 0.04 0.08 0.10 0.12 0.I4 0.16 0.18 0.20 0.22 0.24 PERIOD IN SECONDS LEGEND: l
- 0.010 OSCILLATOR DAMPINS VALVES f ----- 0.020 OSclLLATOR, DAMPING VALVES 0.050 OSCILLATOR DAMPING VALVES AMPLIFIED RESPONSE SPECTRA 0F
- HORIZONTAL ACCELERATION
- REACTOR VESSEL ELEV 119 i AP FROM FEEDWATER LINE BREAK l SHOREHAM NUCLEAR POWER STATION-UNIT 1 i PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS j REVIS10N 4-FEBRUARY 1981
O O O 4 6.00 d 1 s.oo - A l! \. e 4.00 - ,i ./ 8 y\ k S.00 - it; I .i m i , u l l Q 2.00 i < I i 1.o0 J O2 I ' ' I I I 1 I : : o.os O.co 0.02 0.04 o.oa o.lo a.12 0.14 0.16 o.18 0.20 0.22 0.24 PERIOD IN SECONDS I ! LEGEND FIG. 5-46 o.oio OSCILLATOR DAMPING VALVES AMPLIFIED RESPONSE SPECTRA OF 4 ------- o.oto OSCILLATOR oAMP1NG VALVES HORIZONTAL ACCELERATION - o.oso osciLLATom oAurius VALVES SHIELD WALL ELEV. 137 ! AP FROM FEEDWATER LINE BREAK 1 SHOREHAM NUCLEAR POWER STATION- UNIT l ! FLANT DESIGN ASSESSMENT FOR SRV AND 1.DCA LDADS > REVISION 4-FEBRUARY 1981 O- O . O l 2.50-i t 2.00 - l l.5o - i p ! 4 E W
- d 1.00 -
l 0 ? l 0.50 - ..r%N I on f I I I I ' O.oo 0.02 0.04 0.os one o.io o.'12 0.14 o.ls o.is o.to o.22 o.24 PERIOD IN SECONDS 1 1 LEGEND: o.olo OSCILLATOR DAMPiHS VALVES . ------ o.oto OSCILLATOR DAMPING VALV ES pyg*$_47 ! o.oso OSCILLATOR DAMPsNG VALVES i HORIZONTAL ACCELERATION i PEDESTAL ELEV - 90 l AP FROM FEEDWATER LINE BREAK
- SHOREHAM NUCLEAR POWER STATION-UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA t.OADS l
, REVisloN 4-FEBRUARY 1981 I i ' . _- . . . . - - - ~. - _ l 1 SECTION 6 4 () PRIMARY STRUCTURES ASSESSMENT
6.1 INTRODUCTION
This section presents an evaluation of the design adequacy of the primary structures for loading conditions which include the effects of a safety / relief valve (SRV) discharge and loss-of-
, coolant accident (LOCA) events. The primary structures defined ;' herein consist of the foundation basemat, reactor pedestal, and primary containment. The general arrangement of the reactor building is shown on Fig. 6-1. These structures are evaluated in
, terms of their capacity to resist the factored load combinations j discussed in Section 2 and presented again in Table 6-1. Section 5 provides a qualitative and quantitative description of the dynamic behavior of the containment structures when subjected i to the effects of SRV and LOCA loads. As discussed in Section 5, l the structural assessment for SRV loads is based on the ramshead i load definition with simultaneous air- bubble entry into the i' suppression' pool. This load definition bounds that associated with the T-quencher discharge device which is.actually installed l at SNPS-1. Therefore, margins of safety for the primary structures will actually be greater than those calculated and presented here. The containment structure design margin is farther discussed in Appendix B. l
6.2 DESCRIPTION
OF STRUCTURES 4 The primary containment is a reinforced concrete structure 4.5 ft thick with a~79 ft inside diameter at its base and 142 ft ~high. The reactor pressure vessel (RPV) and primary shield wall are
. supported on a 3 ft-10 in, thick reinforced concrete reactor support pedestal with an approximate inside diameter of 18 ft.
The reactor pedestal and primary containment are supported on a common-10 ft thick reinforced concrete foundation aat- with a l diameter of'174 ft.
! 6.3 DESIGN CRITERIA AND LOADS 6.3.1 Desion Criteria The containment structures are designed to accommodate .the effects of environmental conditions associated with normal
! operation, seismic events, and ' postulated accidents. The structures have also shown to have sufficient capacity to sustain l the effects of SRV discharge and LOCA loads when selectively . combined with the normal and ' abnormal load events as delineated in Table 6-1 and in the USAR. l ' ((() ' 6-1 Revision 4 - February 1981 l __ _ ~ _ _ _ _ _ _ . . . . - . _ _ _ _ _ . _ . _ _ . . _ _ . . . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ , . _ _ , , ,
To assure the structural integrity of the structures under the postulated events, the strength capacity of the members under , flexure, and compression includes a capacity reduction l tension, factor "9" in accordance with the building code requirements of ACI-318-71(7) . The capacity reduction factor "#" shall be determined from the following table: ; Description of Load Reduction Factor &
- l. Tension and flexure 0.90
- 2. Diagonal tension 0.85 For independent, short duration, vibratory loads, such as seismic and SRV discharge, the maximum dynamic responses resulting from the individual loads are combined by the square root of the sum of the squares (SRSS) method. However, a positive margin of safety is maintained even when loads are combined by the absolute sum method as described in Appendix B.
The following is a summary of material and strength properties of the primary structures and supporting soil. Reinforcing steel:
. Yield stress = 50 ksi (#14, #18 bars)
= 40 ksi (for #11 and smaller bars)
. Modulus of elasticity = 29,000,000 psi concrete:
. Ultimate compressive strength = 4,000 psi (pedestal) 3,000 psi (containment and mat)
. Ultimate compression strain = .003 in./in.
. Modulus of elasticity = 3,000,000 psi
. Poissons Ratio = 0.167 l Soil:
i . Shear Modulus = 13,000 psi 1
. Poissons Ratio = 0.30 6.3.2 Loads All defined loads on the containment structures have been considered in the design assessment. Load events that have not been previously identified in the USAR include the small and O
6-2 Revision 4 - February 1981
i l l intermediate break accident (SBA and IBA), SRV discharge loads, l () and the transient LOCA loads. ' l i For a SBA and IBA, the suppression pool temperature and pressure i are defined in Section 2.1. l l I 2
. The simultaneous SRV discharge loading (obtained from Section 3), l normalized pressure time-history, and corresponding pressure profile are shown on Fig. 3-11. l l
i In addition to the quasi-static LOCA pressures and temperatures defined in the USAR, the transient LOCA loads defined in Section 4 have also been considered. , 1 6.3.3 Load combinations a As discussed in Reference 1, Section 2.2 of this report provides a set of loading conditions and factored load combinations used for evaluating the structural integrity of the containment structures. These load combinations and corresponding load descriptions are shown in Table 6-1. 6.4 METHOD OF ANALYSIS l The method of analysis of the-containment structures under the j effects of the suppression pool loads has been described in i Section 2.4. The results of the analysis in terms of primary
- structure response are described in Section 5, including maximum values of structural internal loads (moments, shears, axial
,j loads). Resulting concrete and steel stresses and strains are evaluated as described below. The S&W computer program referred to as NEWSECT is utilized to i determine the detailed steel reinforcement stresses and concrete i compression stresses resulting from the containment structure internal loads. The program performs a nonlinear stress analysis l of reinforced concrete sections. The stress-strain curve for the reinforcing steel is represented as elastic-perfectly plastic i while that of the concrete is represented' as nonlinear in i compression. i 6.5
SUMMARY
, DESIGN MARGINS, AND CONCLUSIONS i
- 6.5.1 containment Internal Loads An evaluation of the structural capacity of the containment structures to sustain the additional effects of SRV discharge and transient LOCA loads has been completed. The foundation basemat, reactor pedestal, and primary containment were evaluated for the factored load combinations discussed in Section 2.2.
]. The predominant structural internal loads are longitudinal l bending moment, axial load, and transverse shear at the junctures 4 O 6-3 Revision 4 - February 1981
of the reactor pedestal and primary containment with the foundation mat. Figure 6-2 identifies those locations on the containment structures that are most heavily stressed. l Tables 5-2 and 3 show the maximum values of bending moment, transverse shear, and axial load at selected positions within the suppression pool for an axisymmetric (all valve) discharge. Table 5-5 provides the maximum results for an asymmetric (60a) discharge load. These results omit the effects of structural damping. A comparison of the tables indicates a substantial decrease in the magnitude of the significant design parameters for the asymmetric (60a) SRV discharge loading. The results contained in Tables 5-2, 5-3, and 5-5 for SRV load cases 5 and 6 of Section 5.1 are the SRV results utilized in the assessment as described below. The internal loads due to the all valve simultaneous discharge are used in load combination Equations (1), (2), (3), and (6) of Table 6-1. Since this case produces more severe effects than the sequential firing of all valves, and since positive design margins are shown, combined results are not recomputed for this case. Load equations (4), (5), and (7) of Table 6-1 include the com-bination of an SBA or IBA event and SRV ADS effects. The long term quasi-static pressures associated with an SBA or IBA event result in greater structural internal loads than do the hydrodynamic phenomena and are therefore used in these load combinations. Also the effects of SRV ADS are conservatively accounted for by utilizing the results from the all valve simultaneous discharge case. Load equations (4a), (Sa), and (7a) of Table 6-1 include the com-bination of a DBA event and single SRV discharge effects. Again, the long term pressures associated with a DBA are more severe than the hydrodynamic effects and are used in these load combinations. The effects of a single SRV discharge are conservatively accounted for by utilizing the results from the simultaneous discharge of three adjacent valves. It was found in Section 5.2 that the long term pressures and temperatures associated with a LOCA are more severe than the transient results of vent clearing. Therefore, the quasi-static results are used in Equations (4a), (5a), and (7a). It is also considered that the quasi-static effects of a small or intermediate pipe break (SBA or IBA) are greater than those due to the vent clearing associated with those accidents. For a description of nomenclature and positive sign convention refer to Table 6-2 and Fig. 6-3. The internal loads resulting from the SRV and LOCA events have been combined with the corresponding internal loads from the O 6-4 Revision 6 - December 1986
l
, other individual events in accordance with Table 6-1. The summed values for internal loads, Equations (1) through (7a), are shown , in Table 6-3 for the basemat and in Table 6-4 for the pedestal and primary containment. In each of these load equations the directions of the maximum SRV load and other transient internal ; loads are assumed in such a manner as to produce the most ; critical conditions for stress calculations. As previously mentioned, the SRV loads are combined with other transients in accordance with the SRSS method.
6.5.2 Desian Marains The critical locations for each containment structure, in terms to be at the base of of design margin, have been found the reactor pedestal, at the base of the primary containment, and i just outside the pedestal and primary containment for the foundation basemat. Figures 6-4 through 7 show the reinforcement i details for the primary containment, reactor pedestal, and
, foundation basemat.
The critical load combinations have been found to be
- equations (6), (7), and (7a) of Tables 6-3 and 4. Minimum design t
margins and shear reinforcement requirements obtained from the stress analysis of the critical areas are shown in Tables 6-5 and
- 6. rigure 6-8 provides the definition of design margins as applied to the reinforced concrete wall and mat sections under
; flexure and axial load.
From Table 6-5 the minimum design margin is 1.15 at the base of 2 the primary containment. The critical load equation is (7a), i' abnormal / extreme environmental, and results in a maximum reinforcement stress of 43.1 ksi and a maximum concrete stress of i 2.18 ksi. Table 6-6 shows the requirement for shear reinforcement to be most critical at the base of the pedestal; however, as indicated in the table, adequate shear reinforcement l . is provided. 6.5.3 Conclusions ! An evaluation of the structural capacity of the foundation basemat, reactor support pedestal, and primary containment to sustain the load combinations, including SRV discharge and LOCA transient loads, indicates an adequate design margin. Furthermore, as described in detail in Appendix B, a positive design margin is maintained even if the hydrodynamic SRV and LOCA . loads were arbitrarily increased by more than 100 percent. I 1 6-5 Revision 4 - February 1981
i O - 4 a i b b1 C'
. . . 5: 3 ;
s*1 - -
$9k k- 9 9
k- 3 3,2 ,
~
8 li 3 9 9 g y ij,g g
= ,
4 i i i i i i i i - - . a Eo W
- 92 1- 3 +
9 9 9 9 i 9 9 :g ,j- gg y, y i
$ a= y. y. E' **
99 9 9 9 9 gj ig 4 e i i - - - - . - - gyg g ,. g g v5 i-
.8
=
i i i i - . - i 9
- 5a .- .Ias 4 al g, u
, i 1
e - i
- i i 9
- i ..e x =
a l
'gl
- k. . i e i i i O s u lh,,3 . .
S
-ii
- . . i a
e$*33 g
. z
- 3,8n g I-l 9 9 9 *:
g 3 3 ,,. a, . . - i i - i i
- a:
$l 4 i
; ;ii i i ; . i t &R 88
' 4
.: . i i .
; i i alrg:rg, * *
.i .
a = =
- II ****
9 9 9 a 9 9 9 8 ,1 & & & & 4 a - - - - - - - - -- 8, ! j
, . .* . . .* . .* . . l
'. =' m'=
l cu .
- JJ 4 J J o .'.t a. s y
' o, > e -e . $ -o $ $
-s a :
4
-3 i d Ic 3 g:e H np - - -
3ai51$ aas I la as.: 35 1 I. : e . m o e $ a em ? w O 1 l l
. _ _ - . - . - . - - - . . - - . - - .. - _ _ = . , . -
I l TABLE 6-2 O DEFINITION OF INTERNAL LOADS (18 l I I M. - Radial (longitudinal) bending moment (ft-k/ft) positive (+) when causing tension on top of mat and inside surface of superstructure N.. - Axial (longitudinal) force (k/ft) positive (+) when causing tension in mat and superstructure Q. - Radial (transverse) shear (k/ft) positive (+) when acting upward on outer face of mat and radially outward on super-structure Mer - Tangential (Hoop) bending moment (ft-k/ft) positive (+) when causing tension on top of mat and inside surface of superstructure Nrr - Tangential (hoop) force (k/ft) positive (+) when causing tension in mat and superstructure or - Tangential (hoop) shear (k/ft) Nx - In plane membrane shear (k/ft) O (1) Refer also to Fig. 6-3 O 1 of 1 Revision 3 - November 1978
TABLE 6-3 BASEMAT DESIGN INTERNAL LOADS JUST OUTSIDE PEDESTAL Equation Ms. N.. Q. Me, New Q, N., No. fft-k/ft) (k/ft) (k/ft) (ft-k/ft) (k/ft) (k/ft) (k/ft) 1 -3059 29 293 -1827 48 0 0 2 -2307 25 222 -1311 42 0 0 3 -2667 40 264 -1402 40 72 7 4 -3275 38 299 -1891 106 0 0 4a -2840 62 249 -1667 109 6 4 5 -3420 49 319 -1868 94 63 6 Sa -3298 74 288 -1798 99 63 7 6 -2881 49 288 -1441 33 102 10 7 -3674 62 332 -1916 87 102 10 7a -3688 ja 335 -1913 93 109 15 JUST OUTSIDE PRIMARY CONTAINMENT 1 -2945 21 200 -1585 26 0 0 i O 2 3
-2157
-2956 18 45 147 176
-1167
-1401 22 26 0
27 0 5 4 -1858 -4 134 -1195 54 0 0 4a -1630 -32 113 -1030 66 1 3
- 5 -2580 19 160 -1379 54 24 4 Sa -2551 1 153 -1333 68 24 5 6 -3370 ja 194 -1537 28 38 7 7 -3155 41 185 -1563 56 38 7 7a -3163 27 184 -1546 72 38 8 I
4 HQIE1
- 1. Refer to Table 6-1 for load combination equations.
- 2. Refer to Figure 6-3 for internal load definitions.
- 3. The individual load events have been superimposed in such a way as to produce the most critical conditions for design calculations.
Oi 4. The SRSS method has been used when combining dynamic loads. 1 of 1 Revision 3 - November 1978
4 TABLE 6-4 ( REACTOR PEDESTAL AND PRIMARY CONTAINMENT DESIGN INTERNAL LOADS BASE OF PEDESTAL Equation Ms. N. Q. Mww New Qw Nx No. (ft-k/ft) (k/ft) ( k /f t )_ (ft-k/ft) (k/ft) (k/ft) (k/ft) 1 560 -160 -177 27 74 0 0
- 2 426 -101 -134 23 64 0 0 3 477 -22 -146 28 80 10 3 4 631 -159 -194 43 -22 0 0 4a 552 -141 -168 32 -74 1 4 5 637 -89 -192 42 -2 9 3 Sa 609 -51 -181 37 -26 9 5 6 504 24 -151 29 87 15 5
- 7 671 -20 -199 45 21 15 5 J 7a 121 .11 -197 44 12 16 9 BASE OF PRIMARY CONTAINMENT l
1 1077 -193 -114 32 44 0 0 . 2 798 -132 -85 27 38 0 0 i 3 937 -67 -103 38 80 6 8 4 1141 -20 -137 61 41 0 0 da 1310 37 -173 72 35 1 3 5 1216 23 -146 65 77 5 7 Sa 1415 78 -181 80 88 5 8
- 6 1019 -32 -114 44 105 8 10 l 7 1311 60 -159 73 112 8 10 7a ligt 113. -192 91 127 8 10 4
i i NOTES
- 1. Refer to Table 6-1 for load combination equations.
- 2. Refer to Figure 6-3 for internal load definitions.
- 3. The- individual load events have been superimposed in such a way as to produce the most critical conditions for design calculations.
- 4. The SRSS method has been used when combining dynamic loads..
I O . 1 of 1 Revision 3 - November 1978 4
C TALIE 6-5 l F1EIMOM DESIGN FAEGIllS FOR FLEKtTRE AND AXIAL TENSILE LOADh8 s 3 4 . I calculated Allowable Max Max. 4 Location Mode Critical Loade Imadssas Coccrete Steel DesignE*3 of of Imad Mss Nss Ma P: Stress (ksi) Stress (ksi) Margin , Structure Failure Cambinatice trt-k/tti Outti tit-k/ft) ik/ft) estrair.1 f6traint WH3 ) Base of Coebined Eq. 7a i Reactor Eending and Abnormal 673 29 933 40 1.91 34.9 1.39' . Pedestal Axial Tension Eat. Inv. (.0005) (.0012)
- Base of Combined Eq. 7a Primary . Sending and Abomi 1509 300ta) 1729 344 2.18 43.1 1.15 Containment Axial Tomeica Ext. Env. 4 0008) ( 00M)
! Basemat- Combined Eq. 7a Just out- monding Abnormal -360s 89 -4790 116 1.64 38.4 1.30 side and F.xt . Inv. (.0095) (.0013)
- Pedestal Axial Teemion ;
Basemat- Combined Eq. 6 q Just Out- Bending Abnormal -3370 60 -4289 76 1.56 39.3 1.27 i side and Ext. Env. (.0005) (.0013) 1 Pri. Cont. Axial Tension i i i I
<s 3 combined flexure and axial coscreasic 2 is not critical
- <*3 Ms . yet , ps . WP , W = .90 (capacity reduction factor) 3 ca3 Includes effect of vertical thermal stress i (*) Design Margin = Allowable load l Calculated load (See Figure 6-8) i i
1 1 7 i t i i 2 i 1 ot - 1 Revision 3 - November 1978 i
l 1 TAELE 6-6 ELW1FUH ENEAR REINFORCEMENT REDUIRD'ElstS TRAhSVERSE SPEAR Ehear Calculated Carried Minimun
- Incation Mode Critical loads by Pequired Steel Area l of of Imad Q5 WSS(*3 Concrete steel Area Provided
] Structure Failure combination fir /ft) ik/ft) Ow ik/ft) tima fttsca3 gim*/ft) Coseent
- Base of Diagonal Eq. 7a -197 29 57 1.18 1.46s u Adequate ahear
- menctor tension Abnorsal reir. forces.ent
. Pedestal Ext. Enw. 1 Base of Diagonal Eq. 7a -192 300 4 1.63 7.814 0 Mequate shear Prieary Tension Abnormal reinforcement Containment Ext. Env. I
- Basemat- Diagonal Eq. 7 352 62 136 0.94 1.124 0 Mequate shear i
Just outside Tension Abnormal reinforcement Pedestal Ext. Dtv . Basemat- Diagonal Eq. 6 194 60 136 0.43 1.128 u Adequate shear Just outside Tension Normal reinforcement i Pri. Cont. Ext. Env. i t o Tensile axial load j cas shear steel area per foot or circumference sus Eased on 18-4= vertical spacine t o Based ca 28-0* vertical spacing 4 su Based on 2*-3" radial spacing j <*a Based on 38-0* radial spacino i l 4 b 1 oi 1 Revision 3 - November 1978
4 / } i 1 0 i [_ REACTOR e7*- 2 " R e7'- 2" 4 T.O.S. (LOW F0lNT) 70P 0F PARAPET T.0.8. (H.P.) E L 22 9'-9" EL 2 2 S'- 4" UC- 4 5
- UC-SB / E L. 220*4*
El- . D I
\ l t BOT OF STEEL T l EL. Its'- 4 /e "
NN ! *---- 525/30 TON POLAR CR ANC , TOP CP CONC. -
- S*d (TYP)
EL 203'-I* YOP OF St AIL EL. 20S*-9* 5 7", 5 7 IT 31- CR M -00 2 _ 1
*I
, PLATFORM EL.192*rl0" O- ..
A l ' .
/. '
ggg.,, :/ o IFIS-CRsI-006 Is0VASLE NORtIAt r Wi-PLATE EL. lT4 - 9 ' REMOVABLE 3 ** 4 Li NER DRYWELL
" ' L' "
T 7 E L .17 5'-S" :4.y.s SHlELD PLUS$7 ,_.4. 4f __ .- d { { h (- 6EL .178'-9*
.s,8'#
p yr LINER el / Yhel w Is*.; Ig / oRYER STOR POOL ssEP i " F ?TEL.167,-0%* RWCU FLTR DMNRLZR MI ##FMs W VI b. y <ni $ t g WORit TABLE j'" REMOVE / e [DRYWELL ft l ] - .
$;. myg M 3 EL .te4" ItfLET _
,,[' IG33 DE 006 TO RLOS cogy*
,, L b f# l t. .:,w; .
R00 (,,,,, N -h ,J BE n N [LJ y .*24 l(LAYDfM0)[ if'
. STE Aa4 DRYER: '
j ; E L. iS$*
.e
* "; g .1C -
TOP OF PLUSI
'b Gy ,. y +
g?.,s)g .-
,g '., l p';.. s d.
+
L 3 -0,, ' m
'; g - --
'J, Piet sv4Lvs AR tA / riell TE-107A
, , Il . ,
blA!N STRI(TYP) A / PHASE SEPARATOR E L.137'- 0* N* *,.' .
~ ~
RWCU HEAT EXCNANGERS g ' Y . . .*,9. i
- *. S'- 0~ sutsP f Rpy emaT:0es -._
? C
/ W
.; ELEC itTS .e- I i
pes y g,py , x, -> . . P. E L .128'S 1R 244t M CC-Il23 58%CE ' '**^* *' ' Q IR 24 sitsCC-lil3
~
g..d. SPACE
- 3, eg.c g ga '
win " UC-f ?b _" Rwle33-P OlsA CLEAss uP Puher f',
~
3 " VENT . DUCT *
.. f T PUMP SUC E u it'-s- .a _ __
- ; ; w.,g, &-u--- ~~_~ .
Et . 2 ..-
$/' ' -
'((( I lI k NYOR06EN g f a T.O. S.
" PLATFORIf R C -00 0
*ITCg 8 EL.estn%
l[ e
. l J l
L - _ _ __ ~ - . E L. se'. 0 (T Y P)
_-___.-----.-,y --.v- m - - r- v -- ~~ '-- '- ' "' 7 - . R
- ~
h } , PLATPost) E L. 9 5'- 4{ ~ ~ ~ - - ~ ' ' - - 4 j I l IL o
*==
b *
- a mus EoM PuTPDR, M,1W g 5 KAUC MW I
- C ED *I 8
ll l C4'A teYDRAULiC -, f, si gg sua 8 u2BTs (TYP)
' ,y s
J<i s..-4 T 8 E Mm m ei n'-?" t , V ; ' fr*.e.a ' 4wL 3 2 f I ' 4 ELtc PULL BOX # $__.- -i EL.6/ S, , _ r_ ' , - 7
, I i
$ pgyg g
- @~
- T **C' pget a mEcInc. e =
- lf wmLWE g EL Ss*-o,,
i LWE y PW88P L AREA e
- Et es'-o= M A EA
( g ar see war d G- eh ' = M y'4b N 1!ri1" er h { ._ s* ,,,,,,,,,,,eg,,,,, f* P
,{,
iMTar
- iG j -
i bM : *
.6 rj 'i oRYwELL seal FIS. sal-14 l'h{W...d I I IS Il I I I"
- 4
-i "
e , N,..?. l. . , P L}*j g ', p ' . s ca. i - P-DRYwELL SE AL 3$RY,iCL t
~
h' .5' . " , ,
, , si d -
T PLATFORM EL. 54 *7 p,n- .
'~In-~ L# '-nem stocu WALL rL. 4o'- o" 2 +. . -
g, E ' *** - **
,<3 G .
- s.foown. EE 65 ann 6-
~
l E b ,' i coman Pens -
! nunsnaarruenauets 4
inii.s ncAT es Excnas.ste-l]- l
- mm Irii*osae
' 4
- p
' a -d tst as'.uf.f,ust ,m., -in mte,r, d-
=='
H ig ,6. waT :n Levtt rLaf
. ,,- wt].p '-
e
/ s'.o*ung.
t space
==om m FOR P- = 8UNOLE P'a ==>
hM416 ' L w. Ltv. I.. .
/
gg4 -
;""P'o. .
1
]' ,, i
sm.P'.i j[4 , ; ".lis'ar. edR,1Gwf. Eta 25To ,,o
- .oc wi n we m ,i 'a *.
l Pu.P n - E8Q j - o . 4*** et.v.o g a f,
) " *'[
Je
- -- [M 3!
4 d Lij;I riio= ===P
<TYP
- Places M m *E E-** e if '
S - - - - -- - '" D' b ' L ' g2%igg T?-
*).Y* ' estywaLL coouse usarJ
'[j. '
,h.
i g rLoom onasu st.se'.o*Pum'
- - suut Puur eggy agecu (TYPD rLoon onnes sens-P oset
,s,u,ese Puers P.osec 3 . .
m 00 4 z z-i O 4 b% ,
~
w i- i > m >3 j 4 =4 o o > O t'd , e r-iE - CD 4* > ;;c -
- , 2"*C A
- z en
$ yY y
W ~ e 1 h S \ Y -
- < = m O
l $ ' 1 4 i .+-.-,-g . .m.____,._,, ,._ ,, .-m.-_
l f
- 41.70FT. ;
+--I I.lO FT.--*
/
l' Y
.f
- s ,- s ,9 f.
.': . * ?R'
.. Lf,c SUPPRESSION POOL f,,*'
-., T O P O F P M, m ::- ---:--:. - : : ::::::L-w ::::::_____:
.y '
a:
' .6 - l\E BASE OF
. MAT ,i- BASE OF ,
SUPPORT 'b ' . If' COYER, SLAB- OUTSIDE ' T.
. PRIMARY PEDESTAL ;. ' ,.
EL 9-O CONTAINMENT ;f' CONTAINMENT
'.3 %
/ PEDESTAL MAT OUTSIDE \L b8
/ h ,mR M . ; :. . . . .~ . . '. < : .' . fr.\ 6/
-st.S'-v l'.W:'t ) ( :.;.:i.y. .
f, >:-
*7.
CRITICAL LOCATIONS FIG. 6-2 O CRITICAL LOCATIONS sHORENAM NUCLEAR POWER STATION -UNIT I PLANT DE0left ASSESSMENT FOR SRV AND LOCA' LOADS Revision 5 - DECEMBER 198f
1 O Nss a Os _ 1f NsT \ OUTSIDE 'OF SUPERSTRUCTURE or
/
\/
J)M,g m, a if N ss ( TOP OF MAT O 4 o, ' P ,,, s l h N/
- M ir vj o, Ns s
Nar Ny - l i FIO. 6-3 POSITIVE SIGN CONVENTION FOR INTERNAL LOADS O SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS 1 REvis10N s -IfOVsIISER 1978
. . ~ _ _ . _ - _.._._-._...,_.,___._.,_-____.,..,._.-_.._.._2. _
a= se m (Isf Cu E) 9# 8 - vetar tsip # "** ** I4 l 370* G E) s- - _________.__._______._1 __m_ _____ _ _ _ _ _ _ _ _ _ _ _ ________________ _ _ , _ _ _ . _7 n.so ___.
.[ M *
$ "g * 'H I'L IS ' g 40 fL gy le w i xn . ..eu. si i
x
'Iff r'
4 w w .. s e t e
." } -
'kWh g
T i em c-
~
j l\ '
' L e 2.',s ., i J s # at 84 . . .sa l $ t ie t,E. P.
sm,- it 3-
* . R 19 E1 2* g 9 EL %4 r=+ .....,,, _ _ . t .....m .,.., ~ m,
, , = . = 'c . . .., . - - r jg.j;.
. ,, 4,, ,
b,# b . a ,,: fdllll'b
"^
a.~, m .. .. .
- 'r 'n ="7x t
~.a -.a
. . . .ma.t..
.m 4 i% . ,, ,,, .<c . ..u... . > .r- .. ,
\ ,,,-a. --
~
g-c .,, . .
...w...\> ,,e
[.... w m w, ..g
. . .- .. a - .
...... /,
y a w.g. ..[. ,s ,,.. . . .. r. ,/ ~ W j , _,,_fg L _.. .
. r.. w y n< m#
L. g LIT t or, t U'D p, . PR IM A R Y CONT AI N M E NT 'WAL L (PART E) L._ . . .dtwm=. _ _lll_ -. - - - lll\ 1- . __- - . _ _ - ...m.--n-- __ gg--- .._ m_--..... %r~-- ~-_- _ K_ 1.- "U- - . - - - -
\ f.4 SI*C kk b i M aer*, N ,,,' .f y [-8%,], EL57-4* 8 h 'W 'MY' ggg*
, .$ f, f e. x 1 "k., / Y .b v s ma m a,.
i < seu L_
._g.<
cut or .e,s'.*o-
,n ,- s T v4r ISS OtA4086ALS 84f. At e.' .
ge-0aMe.4. -esee p.44s a.a _ l \ ~. IM , gage .* t.. tr At tvPunts. R( OE NN cur or s a ==..\ ( -e.es.:v* sva nas mess.4 s.,.
- at n.st'#ist now /***, ; . ,,,. x ,,,
g l < / =7 L.H n.e f n as-s*
.j J .r.ft M' d-t o m. ...
,,k s gg j . .
s s'+ rd i nyv4 -sP y*','j bs g 8.s..3-p... uff "' Tp 37 .%. u
,Egg ( me.9.[' f .,
~. '#v44h. ...
.. ., ).s .? e. g1/2.
.....,1-
.g. . .f
' ,,,.e (fe..g-c.
_. . . - . .. .r . a,. e
, n F. s A ,,, g r eso eges es.or a eseeso ,
sL tr. IstTr".N.ap *.wele... 1
/ ~~ -
- m. u. =, ;-..w. .icu
~
1~s mesa em asseau seems [
,1 -
3 MiuMARY CONTAINMENT WM L (PARTE)
t i . t Wi3 %"" W .-,s r
;32,*2"M
ouo ' ate r'3
= >< a- = .st
. -e -}-- --
M k. w
<.d
- wmff"h"M. 2.c.a ,
s Q. eu, n..ast.
,.e t 2-....,
,______________________A. p~ . .
~ "
%'j ) ,. ,,, s.e Ig,g 1. c., l " -
. , . ,; a., .... .
t .. , y $ i g
/
4
, 3 u .a.s L;} l. uim i ,.
n? - r-f
.g i -/
.1..,'rt."L .
, . . .i , s k , . ;r. .c -r, i
1 , 4 u ; - 7a.ce. 4 11
' M.
, .. .o rio. gN o ner. 5:.. => q.7 T <.*-_)
s . jlf,",'apa.fl. " h t . " '" I [/ MI eau, i!: , N - m m2- . _ _m m. c.::
- nte- .'
l - % %.s,. ... . . . .m ..
......, ~ . . . "Y
,,. y ,
5 p<.'. --
.1 -
p "Y.' *'. %;n'. g2'" 'hI 6 - -
~
\
N, d,.,,,u.ii m ,.....,s. '. . n.M. m-i q f,p. - - Ii
. , c ,,
+,
,, u ,' m. 1. .w x.^au N .a 2- ear e .; , ' n. t.
- ie s a.bmc 1. s.,s . 4.ivM. s. , g
- a. 3 - .:.
s ...r., suub. i . - -. ' ,".g. g","u.;.';;l ya m yy ,
- p. me e'as n.see seu n .t. y1 i_ ,+=gg g, ,,*
.g _
-) _
f J l su5__6, . c c i e o g r,**
" *S g ag; 7 1, 3
' .3yaar..,
i
't g 24.<
_?~ = '"J".J.C
- P-I* f l .. -
,'4s6.4# u o> Y~ i l
V , ENLARGE 0 OETAIL OF SECT l-l ELasV w i F
-. ^ F8 '85
{ l sit-.D
.':t'a';#
h ,
! c t
s 1 M ! .-,...,,u..... m.w
,,,,,,,,,,c_--.........
/> in%L '
l hu.u ,e mee.n= rar>a. . - P f > 3.l, . .
, f a ,,e e .e c" ,
4 , .0, 't - S b A m( ,_. t TI i '# _
/ @'F '
e r**" . APERTURE $paa < .... . t-' CARD ACC 419 Ni w L
~ pa g.g=
R $* 1 t" 1 'M b did% g BiM" ,o, . .- I!P"*1.,L.w AP *
- Cud ni e -
& " Q'.,R " .=..~. p ' 8 612 0 8 0 6 4 5 _ y_3
=**'a****= "8 g t W' < FIG. 6-4
\' PRIMARY CONTAINMENT
, WALL REINFORCING DETAILS
! ,e.. . .. . h EL. 8'- O" TO 6 0'-O"
\ _ y SHOREHAM NUCLE AR POWER STATION-UNIT l C 1 e x / PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS N,
~
REVISION 3-NOVEM8ER 1978
l EEca8th
.,, e m ,.
- m .._.
esa .i Q
- ~ ~ *t +
g, ,g ,ntoor agon mnes utstaAsur p to0P mmwG ( ep 45' rw ae'os sjs' I TE,%YM u ;.h "_i. 5
'T n .r, l .-
i saq'mt
. ,@ k\ m- f *# "'. '.3*~
e a';5 05: 9 I v_ I I __
.,L J J i_ i n es ar L f .
i g , ,_, n .-s e
'n W----+-+h+* 4 -- * +1t*:-- - - +-' + -' \ - / h 88~ -4
.-b - _ .-- _. 3,,g, *. . ? -
pg..- L.h*y : .h-. D --3',
- s. phi
*- .*l '.t
., -hht. :--%,,,,, ! . 7, m-. . . . .
%c, m ..+-+. A, . . 4,- 1,@
G- W 4 . m, # +i w
. %<,t,qfm,
~
eage q #; S ing" u.. a a "- 1 4, r :- I f n .r+</.,i ut
,; a . % .U, 1 . [ t m ' 1S
% ! #m.,4 Io I!
pt,C - r m .u,%6w4 c m a 6 v.cr s m, r% er t , i . cme sv = /_ ~ . m w y.., i -%.*oet p. . ,., _L _ _c- ,_ge a- ,7r_rt A.2L=p,%:>V . wsn nvgj ~ h,' 'a ' j
~
$ 2 ,,E L i-s .7h.N[.7 7 dm-
,v.
MQ -u ,, j n... -s y n,- %e n i / q _ , ix , N. , , . _
- n. . . .r - 3 \ .., . n , , 3 o, .ok 'h asen. ((
'n \ y?.f1-Q-..._Q-Q . p' - - - ( "",f, . El1&$$5$,?&n n .. x , ,
-gr - gw.px.m7 g .m 7
._ x ~- e.! + -V ' e - - , .
\M- si.um ~ @
_ __. .'[ ' 3 ~.su:n-- . - ~
)-- J ,y[dyrrgygmrr Q ---
x --- r d I y} i ,a. b . m. ! a
- a. ..
l *d s r, 1 /(.J 1 %. ,e s
** C
, , .. J j %c. i
,-- , ,-i.st. f .t.2' '
a *l7 4 J--
.g .. -. , -
- u. ,,s _
, #"*j- . tasw-e.-d d '.b - ep-@'"
. ... .. J- . .- . ~ . . . -
2- ES- -EES ---EEEl-- C.. gyg7._
- - --EE -
C -C 3 -- -EEE3- - -E25- J
>,..,r._i 1
t
/ -
l l.
, ..>,.,S . ...
'/ _-- t
\ s !
t i
,.c 4. m i i
- s. 4'.og 3 .
'i -- . f 4s' # \ aear m e a # l *
(neeYo"s.I.'U.*v 4 ELEY. 'REACTOR SUPPORT vnLL (outsiot FAca) l
v
' r- "s ;t.,t.h. **A. .W. .., "9-
, ~c .
F ,,c gc==! . ~ i . . u. n , . . . . l - i= r ,el I ta wd. 4 ._
, Y M !==
ir -'Jg,,";,'/
..8 f = u " ?
4 .,
-i y .;( 7yg.
-- U 3, ya L h;- , p.x, 4, '- ,
7,..a d 3etL ,939g
=m j 9,,f ),,,
um-
-gl1_,. ,,pOg;l -, - .- . . . . . . . .
g, pf" [- f1. ,,Tyyg.c j _ ,,,, ,,,,_,,g
",,,_,.. - {9 lpg gwpw L t ---
e-A .c e. w .
- 4. a L'a"w" ni i
-q e ., . . .
.v tse g% g
. aomq -m y
.2 - --
...o, o h
ar . ., o m e '6u G=. - G
- - -M.,,s o,, , ,f.
jg 4' ,_ , 4 ;gg DET.Y Lenac ) 7, (
$y 7Qep I- %. F
, 4 ggcgig,,gt:g g ,.
4 .;t 1-E- -* t g_.;, _a, e -- #. aig
],9; i \N ['
J
'T.
w- .Jt g M.y DET A 4[${# eat tr. 8,'Eg f. pt, DET S 2_ _ '
,]
<-st sof ,,,,,_
g ,se. s e.s.e N (j l.
*istf- e, -p .** 8' ass 'g".','y ,g a x i gr. ,, o-
*smaa ar.o' Ls.
( 1 4 l ,-u a r.<.-
= -
j A '3'J11')
-s les 74~2 agog /
I
* 'I' ,, _., **
,. . DETAIL X .
(4 REQUIRED) egim , I .t f bsessa
,; ,< c , s -
...u.:.4+ sa,w l
p'.c . n .D}\ 2t I n< + - - . . A gg : g i pg,g'I ;e 3 e E a ,n.. t g, .e Tiff"-'g
. ' " W ! @ r ar .a r s.e 4/L ..
TI 4 %
%- k APERTURE CARD
; 4 ?. ._
p* men.9r l F, # 2 m! .u - Df
!* ~
_e. Also Available On
,ra.r E .u .u ... i= N =a) .
l "' oa Aperture Card En'%mp . a.i i , i l v , "L*;e's au ;-a ", _ . . - - 8612080645 L \( k ,,,g-y 5 <
'" pramen.nwe.
FIS. 6-5 o REACTOR SUPPORT WALL l __J R2lNFORCING DETAILS A.~A SHOREH AM NUCLEAR POWER STATION -UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS l - - ( REVISION 4 - FEBRU ARY 1981 1
Pa
=n, =-
_] Q\,T - - l k\\N 5'.-
, ~ . . . . .
\\\\*~:~s
/
\\\\b ~g .
Wul tw. c 6. 3 ; , 4 DET."A"
~~. .
===
==,..
- S a g,==" "- ,, ,
7 . 4'" pa, 0 's / ; ' _- %m;y m
. _..._ _a -__ m =_ .u
- _ a_ ,
- - = -
u m _.___ ~
" 5%
~
l
'! )f'f! !! .
*= =
A fig ~ ~ '
'llIll HllFc fllll i
l l
,r. .~; . .. -- a.
4
- ,.a . , ~ . -
l I I Iga l $$,4 9f M '),,,g
~' '
rn-- rs V:::=::: e y,,,....... ,.
...ard , ! ,. t,,. *l,8'*' 0 4 y.
l , PLAN. MAT TOP AEINFORCING
) ; '""""",'"f1'". ,,] ,
= O"' =7.J" *""
22 (T hP)
9 l m y . i t . -- _
+%..w .
=
t _ d t'Hi d4%i - '
+ 'M
+.k.
+ e \
am,. rou.A.c. . ,, enou ce=ren eaur = _ es W I THIS AnEA p gag ) i g w ' l}}