ML18151A290
| ML18151A290 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/31/1990 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18151A291 | List: |
| References | |
| 90-536, NUDOCS 9009260149 | |
| Download: ML18151A290 (156) | |
Text
{{#Wiki_filter:e e VIRGINIA. ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 August 31, 1990 United States Nuclear Regulatory Commission Serial No. 90-536 Attention: Document Control Desk NO/LB:vlh Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen: VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 UPDATED FINAL SAFETY ANALYSIS REPORT REVISION 10 Pursuant to 10 CFR 50.71 (e), Virginia Electric and Power Company submits Revision 1Oto the Updated Safety Analysis Report (UFSAR) for the Surry Power Station Units 1 and 2. This represents the second UFSAR update submitted this year. Due to the considerable volume and scope of UFSAR changes for Surry Units 1 and 2, a third revision for 1990 will be submitted in December of this year. The December 1990 revision will address UFSAR changes not included in this submittal in addition to changes generated as a result of our recent UFSAR validation effort. Each revision package contains replacement instructions and Revision 1O pages, including overleafs. The Revision 1O pages replace the existing pages as described in the instructions. One signed original and ten additional copies of Revision 1O to the UFSAR are enclosed. ((?ff~ Senior Vice President - Nuclear Enclosures (Original and ten copies) r
e cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station
COMMONWEALTH OF VIRGINIA )
)
COUNTY OF HENRICO ) The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by E. W. Harrell, who is Vice President - Nuclear Operations, for W. L. Stewart, who is Senior Vice President - Nuclear, of Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief. Acknowledged before me this 31 day of ......,\"",,..,,.A_1...,,.t~....,...i<_10_t~_, "19 C,o. My Commission Expires: ,- J\t,o.,~ 31 , 193_1. __Jjje}_;_~_w_Q___ Notary Public (SEAL)
5D-Z~D
' ,,.. JJ "
't,_,.,,,....
" ) tp\:.q~*t::,---l
- )~f~.::..-*
i
'A .* , ...
f~~
- ff~l/ij t ,; /'I /(),' T~ t.)r-cJ,:_ /-e::d f~j:.-1 l?. ,,D-fdi/3 i>/9t}
~ ' ,.,,.4-1!'.;,,.~_'il': <;).~-"-.-,,.. .1.,.f "~
""'t_,,. ""Jif,{/- J:v{K);r.;,1,t(d;i'._:'
6/90 t. { ~ *~* l/ J ~ SPS UFSAR LEP-1
'~REVISION g LIST OF EFFECTIVE PAGES
~ Revision ~ ~ Revision
~
VOLtlME I 1.4-23 6/85 3 1.4-24 7/82 Original Gene~, Table of contents 1.4-25 7/82 Original i
"- 7/82 Original l.4-=-26 1.4-27 7/82 7/82 Original Original ii 7/82 Original 1.4-28 7/82 Original iii 7/82 Original 1.4-29 7/82 Original iv 7/82 Original 1.4-30 7/82 Original 1.4-31 7/82 Original Chapter 1
'\ \.
1.4-32 1.4-33 7/82 7/82 Original Original 1-i 7 / 8*~ Original 1.4-34 7/82 Original 1-ii 7/82'" Original 1.4-35 7/82 Original 1-iii 7/82 , Original 1.4-36 7/82 Original 1-iv 7/82 ~ Original 1.4-37 7/82 Original 1-v 7/82 , Original 1.4-38 7/82 Original 1-vi 7/82 '\Original 1.4-39 7/82 Original 1.1-1 7/82 Original 1.4-40 7/82 Original 1.1-2 7 /82 o*higinal 1.4-41 7/82 Original 1.1-3 7/82 ' Orrginal 1.4-42 7/82 Original 1.1-4 7 / 82 Ori<jtl.nal 1.4-43 7/82 Original 1.2-1 7/82 Origihal 1.4-44 7/82 Original
'\
- 1. 2-2 7/82 Origin~l 1.4-45 7/82 Original
'\
- 1. 2-3 7/82 Original\ 1.4-46 7/82 Original 1.2-4 7/82 Original _ 1.4-47 7/82 Original 1.2-5 7 / 82 Original \ 1 . 4-4 8 7/82 Original 1.3-1 7/82 Original 1.4-49 7/82 Original 1.3-2 7/82 Blank 1.4-50 7/82 Original 1.3-3 7/82 Original ~.4-51 7/82 Original 1.3-4 7/82 original 1'\5-1 7/82 Original 1.3-5 7 / 82 Original 1. ~2 6/86 4 1.3-6 7 / 82 Original 1. 6- 7/82 Original 1.3-7 7 / 82 Original 1. 6-2 7/82 Original 1.3-8 7 / 82 Original 1. 6-3 7/82 Original 1.3-9 7 / 82 Original 1. 6-4 7/82 Original 1.3-10 7 / 82 Original 1. 6-5 7/82 Original 1.4-1 7/82 Original 1.6-6 7/82 Original
- 1. 4-2 7 / 82 Original 1. 6-7 7/82 Original 1.4-3 7 / 82 Original 1. 6-8 7/82 Original 1.4-4 7 / 82 Original 1. 6-9 7/82 Original 1.4-5 7/82 Original 1.6-10 7/82 Original 1.4-6 7/82 Original 1.6-11 -;;_/82 Original 1.4-7 7 / 82 Original 1. 6-12 7'x.82 Original 1.4-8 7/82 Original 1.4-9 7/82 Original Chapter 2 1.4-10 7/82 Original
- 1. 4-11 7/82 Original 2-i 6/83 1 1.4-12 7/82 Original 2-ii 6/83 1
- 1. 4-13 7/82 Original 2-iii 6/83
- 1. 4-14 .
1.4-15
- 1. 4-16 1.4-17 1.4-18 8/ 89 7/82 7/82 6/ 85 7/82 Original Original 3
Original 8 2-iv 2-v 2-vi 2-vii 2-viii 6/83 6/83 6/83 6/83 6/83
\1 1 1
1.4-19 7/82 Original 2-ix 6/83 1 1.4-20 7/82 Original 2-x 6/83 1 1.4-21 7/82 Original 2.1-1 6/83 1 1.4-22 6/ 85 3 2.1-2 6/83 1
'\
I REVISION 9 6/90 SPS UFSAR LEP-2 LIST OF EFFECTIVE PAGES ~ ~ Revision ~ u.t..e. Revision VOLUME I (continued) 2.2-24 6/83 1 2.2-25 6/83 1 2.1-3 6/83 1 2.2-26 6/83 1 2.1-4 6/83 1 2.2-27 6/83 1 2.1-5 6/83 1 2.2-28 6/83 1 2.1-6 6/83 1 Figure 2.2-1 6/83 1 2 .1-7 6/83 1 Figure 2.2-2 6/83 1 2.1-8 6/83 1 Figure 2.2-3 6/83 1 2.1-9 6/83 1 Figure 2.2-4 6/83 1 2.1-10 6/83 1 Figure 2.2-5 6/83 1 2.1-11 6/83 1 Figure 2.2-6 6/83 1 2.1-12 6/83 1 Figure 2.2-7 6/83 1 2.1-13 6/83 1 Figure 2.2-8 6/83 1 2.1-14 6/83 1 Figure 2.2-9 6/83 1 2.1-15 6/83 1 Figure 2.2-10 6/83 1 2.1-16 6/83 1 Figure 2.2-11 6/83 1 2.1-17 6/83 1 Figure 2.2-12 6/83 1 2.1-18 6/83 1 Figure 2.2-13 6/83 1 2.1-19 6/83 1 Figure 2.2-14 6/83 1 2.1-20 6/83 1 Figure 2.2-15 6/83 1 Figure 2.1-1 6/83 1 Figure 2.2-16 6/83 1 Figure 2.1-2 7/82 Original Figure 2.2-17 6/83 1 Figure 2.1-3 6/83 1 Figure 2.2-18 6/83 1 Figure 2.1-4 3/88 7 Figure 2.2-19 6/83 1 Figure 2.1-5 6/83 1 Figure 2.2-20 6/83 1 Figure 2.1-6 6/83 1 Figure 2.2-21 6/83 1 Figure 2.1-7 6/83 1 2.3-1 6/83 1 Figure 2.1-8 6/83 1 2.3-2 6/83 1 Figure 2.1-9 6/83 1 2.3-3 6/83 1 Figure 2.1-10 6/83 1 2.3-4 6/83 1 Figure 2 .1-11 6/83 1 2.3-5 6/83 1 Figure 2.1-12 6/83 1 2.3-6 6/83 1 Figure 2.1-13 6/87 6 2.3-7 6/83 1 Figure 2.1-14 6/83 1 2.3-8 6/83 1 Figure 2.1-15 6/83 1 2.3-9 6/83 1 2.2-1 6/83 1 2.3-10 6/83 1 2.2-2 6/83 1 2.3-11 6/83 1 2.2-3 6/83 1 2.3-12 6/83 1 2.2-4 6/83 1 2.3-13 6/83 1 2.2-5 6/83 1 2.3-14 6/83 1 2.2-6 6/83 1 2.3-15 6/83 1 2.2-7 6/83 1 2.3-16 6/83 1 2.2-8 6/83 1 2.3-17 6/83 1 2.2-9 6/83 1 2.3-18 6/83 1 2.2-10 6/83 1 2.3-19 6/83 1
- 2. 2-11 6/83 1 2.3-20 6/83 1 2.2-12 6/83 1 2.3-21 6/83 1 2.2-13 6/83 1 2.3-22 6/83 1 2.2-14 6/83 1 2.3-23 6/83 1 2.2-15 6/83 1 Figure 2.3-1 7/82 Original 2.2-16 6/83 1 Figure 2.3-2 7/82 Original 2.2-17 6/83 1 Figure 2.3-3 7/82 Original 2.2-18 6/83 1 Figure 2.3-4 7/82 Original 2.2-19 6/83 1 Figure 2.3-5 7/82 Original.
2.2-20 6/83 1 Figure 2.3-6 7/82 Original 2.2-21 6/83 1 Figure 2.3-7 7/82 Original 2.2-22 6/83 1 Figure 2.3-8 7/82 Original 2.2-23 6/83 1 Figure 2.3-9 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-3
- t.J:S'.l: Q[ l:i((l:i!:IlY]:i ~l!i.~l:iS
* ~
VOLUME I Figure 2.3-10 Figure 2. 3-11
~
(continued) 6/87 3/88 Revision 6 7
~
2.4-47 2.4-48 Figure Figure 2.4-1 2.4-la
~
7/82 7/82 6/83 6/83 Revision Original Original 1 1 2.4-1 6/83 1 Figure 2.4-lb 6/83 1 2.4-2 6/83 1 Figure 2. 4-lc 6/83 1 2.4-3 6/83 1 Figure 2.4-2 7/82 Original 2.4-4 6/83 1 Figure 2.4-3 7/82 Original 2.4-5 6/83 1 Figure 2.4-4 7/82 Original 2.4-6 6/83 1 Figure 2.4-5 7/82 Original 2.4-6a 6/83 1 Figure 2.4-6 7/82 Original 2.4-6b 6/83 1 Figure 2.4-7 7/82 Original 2.4-6c 6/83 1 Figure 2.4-8 7/82 Original 2.4-6d 6/83 1 Figure 2.4-9 7/82 Original 2.4-6e 6/83 l Figure 2.4-10 7/82 Original 2.4-6f 6/83 1 Figure 2. 4-11 7/82 Original 2.4-7 7/82 Original Figure 2.4-12 7/82 Original 2.4-8 6/83 1 Figure 2.4-13 7/82 Original 2.4-9 7/82 Original Figure 2.4-14 7/82 Original 2.4-10 7/82 Original Figure 2.4-15 7/82 Original 2.4-11 7/82 Original Figure 2.4-16 7/82 Original 2.4-12 7/82 Original Figure 2.4-17 7/82 Original 2.4-13 6/83 1 Figure 2.4-18 7/82 Original 2.4-14 7/82 Original Figure 2.4-19 7/82 Original 2.4-15 7/82 Original 2.5-1 6/83 1 2.4-16 7/82 Original 2.5-2 6/83 1 2.4-17 7/82 Original 2.5-3 6/83 1 2.4-18 7/82 Original 2.5-4 6/83 1 2.4-19 7/82 Original 2.5-5 6/83 1 2.4-20 7/82 Original 2.5-6 6/83 1 2.4-21 7/82 Original 2.5-7 6/83 1 2.4-22 7/82 Original 2.5-8 6/83 1 2.4-23 7/82 Original 2.5-9 6/83 1 2.4-24 7/82 Original 2.5-10 6/83 1 2.4-25 7/82 Original 2.5-11 6/83 1 2.4-26 7/82 Original 2.5-12 6/83 1 2.4-27 7/82 Original 2.5-13 6/83 1 2.4-28 7/82 Original 2.5-14 6/83 1 2.4-29 7/82 Original 2.5-15 6/83 1 2.4-30 7/82 Original 2.5-16 6/83 1 2.4-31 7/82 Original 2.5-17 6/83 1 2.4-32 7/82 Original 2.5-18 6/83 1 2.4-33 6/83 1 2.5-19 6/83 1 2.4-34 6/83 1 2.5-20 6/83 1 2.4-35 6/83 1 2.5-21 6/83 1 2.4-36 6/83 1 2.5-22 6/83 1 2.4-36a 6/83 1 2.5-23 6/83 1 2.4-36b 6/83 1 2.5-24 6/83 Blank 2.4-37 7/82 Original 2.5-25 6/83 1 2.4-38 7/82 Original 2.5-26 6/83 1 2.4-39 7/82 Original 2.5-27 6/83 1 2.4-40 7/82 Original 2.5-28 6/83 1 2.4-41 7/82 Original 2.5-29 6/83 1
- 2.4-42 2.4-43 2.4-44 2.4-45
- 2. 4-4 6 7/82 7/82 7/82 7/82 7/82 Original Original Original Original Original 2.5-30 2.5-31 2.5-32 2.5-33 2.5-34 6/83 6/83 6/83 6/83 6/83 1
1 1 1 1
REVISION 9 6/90 SPS UFSAR LEP-4 LlSI QE: J:iE:E:J:i!::IlY:J:i ~AGJ:iS ~ VOLUME I 2.5-35 Figure Figure 2.5-1 2.5-la t!....a..t....e (continued) 6/83 6/83 6/83 Revision 1 1
~
Figure 3.3-5 Figure 3.3-6 Figure 3.3-7 Figure 3.3-8 tl.ali 7/82 7/82 7/82 7/82 Revision Original Original Original Original 1 Figure 3.3-9 7/82 Original Figure 2.5-2 7/82 Original 3.4-1 7/82 Original Figure 2.5-3 7/82 Original 3.4-2 7/82 Original Figure 2.5-4 7/82 Original 3.4-3 7/82 Original Figure 2.5-5 7/82 Original 3.4-4 7/82 Original 3.4-5 7/82 Original Chai;2:ter 3 3.4-6 7/82 Original 3.4-7 7/82 Original 3-i 7/82 Original 3.4-8 7/82 Original 3-ii 7/82 Original 3.4-9 7/82 Original 3-iii 7/82 Original 3.4-10 7/82 Original 3-iv 7/82 Original 3. 4-11 6/85 3 3-v 7/82 Original 3.4-lla 6/85 3 3-vi 7/82 Original 3.4-12 7/82 Original 3-vii 7/82 Original 3. 4-13 7/82 Original 3-viii 7/82 Original 3.4-14 7/82 Original 3.1-1 7/82 Original 3.4-15 7/82 Original 3.2-1 7/82 Original 3.4-16 7/82 Original 3.2-2 7/82 Original 3.4-17 3/88 7 3.2-3 7/82 Original 3.4-18 7/82 Original 3.2-3 7/82 Original 3.4-19 7/82 Original 3.2-4 6/85 3 3.4-20 7/82 Original 3.2-5 7/82 Original 3.4-20a 3/88 7 3.2-6 6/85 3 3.4-21 7/82 Original 3.2-7 7/82 Original 3.4-22 7/82 Original
- 3. 2-8 7/82 Original 3.4-23 6/83 1 3.2-9 7/82 Original Figure 3.4-1 7/82 Original 3.2-10 7/82 Original Figure 3.4-2 7/82 Original 3.3-1 7/82 Original Figure 3.4-3 7/82 Original 3.3-2 7/82 Original Figure 3.4-4 7/82 Original 3.3-3 3/88 7 Figure 3.4-5 7/82 Original 3.3-4 7/82 Original Figure 3. 4-6 7/82 Original 3.3-5 7/82 Original Figure 3.4-7 7/82 Original 3.3-6 3/88 7 Figure 3.4-8 7/82 Original 3.3-6a 3/88 7 Figure 3.4-9 7/82 Original 3.3-7 7/82 Original Figure 3.4-10 7/82 Original 3.3-8 7/82 Original Figure 3.4-11 7/82 Original 3.3-9 7/82 Original Figure 3.4-12 7/82 Original 3.3-10 7/82 Original 3.5-1 7/82 Original
- 3. 3-11 7/82 Original 3.5-2 7/82 Original 3.3-12 7/82 Original 3.5-3 7/82 Original 3.3-13 6/86 4 3.5-4 7/82 Original 3.3-14 7/82 Original 3.5-5 7/82 Original 3.3-15 7/82 Original 3.5-6 7/82 Original 3.3-16 7/82 Original 3.5-7 7/82 Original 3.3-17 6/86 4 3.5-8 7/82 Original 3.3-18 3/88 7 3.5-9 7/82 Original 3.3-19 6/83 1 3.5-10 6/86 4 3.3-20 7/82 Original 3.5-lOa 6/86 4 3.3-21 7/82 Original 3. 5-11 7/82 Original Figure 3.3-1 7/82 Original 3.5-12 7/82 Original Figure 3.3-2 7/82 Original 3.5-13 7/82 Original Figure 3.3-3 7/82 Original 3.5-14 7/82 Original Figure 3.3-4 7/82 Original 3.5-15 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-5 i.lSI QE liiEElii~nYlii ~A~liiS
- ~
VOLUME I 3.5-16 3.5-17
~
(continued) 7/82 7/82 Revision Original Original 4
~
4-iv 4-v 4-vi 4-vii 4.1-1
~
6/83 6/83 6/83 7/82 7/82 Revision 1 1 1 Original Original 3.5-18 6/86 3.5-18a 6/86 4 4.1-2 8/89 8 3.5-19 7/82 Original 4.1-3 7/82 Original 3.5-20 7/82 Original 4.1-4 7/82 Original 3.5-21 7/82 Original 4.1-5 8/89 8 3.5-22 6/86 4 4.1-6 8/89 8 3.5-23 7/82 Original 4.1-7 7/82 Original 3.5-24 6/86 3 4.1-8 7/82 Original 3.5-25 6/86 4 4.1-9 7/82 Original 3.5-26 6/86 4 4.1-10 7/82 Original 3.5-27 7/82 Original 4.1-11 7/82 Original 3.5-28 7/82 Original 4.1-12 7/82 Original 3.5-29 7/82 Original 4. 1-13 7/82 Original 3.5-30 7/82 Original 4.1-14 7/82 Original 3.5-31 7/82 Original 4.1-15 7/82 Original 3.5-32 7/82 Original 4.1-16 7/82 Original 3.5-33 7/82 Original 4.1-17 7/82 Original 3.5-34 6/86 4 4.1-18 7/82 Original 3.5-35 7/82 Original 4.1-19 7/82 Original 3.5-36 7/82 Original 4 .1-20 7/82 Blank 3.5-37 6/86 4 4.1-21 7/82 Original Original 3.5-38 6/86 Blank 4.1-22 7/82 3.5-39 6/86 4 4.1-23 7/82 Original 3.5-40 7/82 Original 4.1-24 7/82 Original 3.5-41 7/82 Original 4.1-25 7/82 Original 3.5-42 7/82 Original 4.1-26 7/82 Original Figure 3.5-1 7/82 Original 4.1-27 7/82 Original Figure 3.5-2 7/82 Original 4.1-28 7/82 Original Figure 3.5-3 7/82 Original 4.1-29 7/82 Original Figure 3.5-4 7/82 Original 4.1-30 7/82 Original Figure 3.5-5 7/82 Original 4.1-31 7/82 Original Figure 3.5-6 7/82 Original 4.1-32 7/82 Original Figure 3.5-7 7/82 Original Figure 4.1-1 7/82 Original Figure 3.5-8 7/82 Original Figure 4.1-2 7/82 Original Figure 3.5-9 7/82 Original 4.2-1 7/82 Original Figure 3.5-10 7/82 Original 4.2-2 7/82 Original Figure 3.5-11 7/82 Original 4.2-3 6/87 6 Figure 3.5-12 7/82 Original 4.2-3a 6/87 6 Figure 3. 5-13 7/82 Original 4.2-4 7/82 Original 3.6-1 7/82 Original 4.2-5 7/82 Original 3.6-2 7/82 Original 4.2-6 6/90 9 3.6-3 6/86 4 4.2-6a 6/85 3 3.6-4 6/86 4 4.2-7 7/82 Original 3.6-5 6/86 4 4.2-8 7/82 Original 3.6-6 6/86 4 4.2-9 7/82 Original 3.6-7 7/82 Original 4.2-10 7/82 Original
- 4. 2-11 7/82 Original VOLUME I I 4.2-12 7/82 Original 4.2-13 7/82 Original Ch,:1r;i:ter 4 4.2-14 6/83 1 4.2-15 6/83 1 4-i 7/82 Original 4.2-16 7/82 Original 4-ii 6/83 1 4.2-17 6/83 1 4-iii 6/83 1 4.2-18 6/83 1
REVISION 9 6/90 SPS UFSAR LEP-6 LlS:i: Q[ J:i((J:i!::l:lYJ:i ~A~J:iS ~ 4.2-19 4.2-20 4.2-20a
~
VOLUME I I (continued) 6/83 3/88 3/88 Revision 1 7 7
~
4.3-20 4.3-21 4.3-22 4.3-23 4.3-24
~
6/83 6/83 7/82 6/83 6/83 Revision 1 1 Blank 1 1 4.2-21 8/89 8 Figure 4.3-1 7/82 Original 4.2-22 6/83 1 Figure 4.3-2 7/82 Original 4.2-23 6/83 1 Figure 4.3-3 7/82 Original 4.2-24 6/83 1 Figure 4.3-4 7/82 Original 4.2-25 6/83 1 Figure 4.3-5 7/82 Original 4.2-26 6/83 1 4.4-1 7/82 Original 4.2-27 8/89 8 4.4-2 7/82 Original 4.2-28 6/83 1 4.4-3 7/82 Original 4.2-29 6/83 1 4.4-4 7/82 Original 4.2-30 8/89 8 4.4-5 7/82 Original 4.2-31 6/83 1 4.4-6 7/82 Original 4.2-32 6/83 1 4.4-7 7/82 Original 4.2-33 6/83 1 4.4-8 6/83 1 4.2-34 6/83 1 4.4-9 7/82 Original 4.2-35 6/83 1 4.4-10 7/82 Original 4.2-36 8/89 8 4. 4-11 7/82 Original 4.2-37 6/83 1 4.4-12 7/82 Original 4.2-38 6/83 1 4.2-39 6/84 2 Chapter 5 4.2-40 6/83 1 4.2-41 6/83 1 5-i 6/84 2 4.2-42 6/83 1 5-ii 7/82 Original Figure 4.2-1 6/87 6 5-iii 7/82 Original Figure 4.2-2 6/87 6 5.1-1 7/82 Original Figure 4.2-3 7/82 Original 5.1-2 7/82 Original Figure 4.2-4 7/82 Original 5.2-1 7/82 Original Figure 4.2-5 7/82 Original 5.2-2 6/84 2 Figure 4.2-6 7/82 Original 5.2-3 6/84 2 Figure 4.2-7 7/82 Original 5.2-3a 6/84 2 Figure 4.2-8 7/82 Original 5.2-4 7/82 Original Figure 4.2-9 7/82 Original 5.2-5 6/83 1 Figure 4.2-10 7/82 Original 5.2-6 6/83 1 Figure 4. 2-11 7/82 Original 5.2-6a 6/83 1 Figure 4.2-12 7/82 Original 5.2-7 7/82 Original 4.3-1 7/82 Original 5.2-8 6/83 1 4.3-2 7/82 Original 5.2-9 6/85 3 4.3-3 7/82 Original 5.2-10 7/82 Original 4.3-4 7/82 Original 5. 2-11 7/82 Original 4.3-5 8/89 8 5.2-12 7/82 Blank 4.3-6 7/82 Original 5.2-13 7/82 Original 4.3-7 8/89 8 5.2-14 7/82 Blank 4.3-8 7/82 Original 5.2-15 7/82 Original 4.3-9 7/82 Original 5.2-16 7/82 Blank 4.3-10 7/82 Original 5.2-17 6/83 1
- 4. 3-11 7/82 Original 5.2-18 7/82 Blank 4.3-12 7/82 Original 5.2-19 6/84 2
- 4. 3-13 7/82 Original 5.2-20 7/82 Blank 4.3-14 7/82 Original 5.2-21 7/82 Original 4.3-15 7/82 Original 5.2-22 6/84 2 4.3-16 6/83 1 5.2-23 6/84 2 4.3-17 6/83 1 5.2-24 6/84 2 4.3-18 8/89 8 Figure 5.2-1 7/82 Original 4.3-19 6/83 1 5.3-1 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-7 LIST OF EFFECTIVE PAGES
~ ~ Revision ~ ~ Revision VOLUME II (continued) 6.2-16 6/85 3 6.2-17 6/85 3 5.3-2 7/82 Original 6.2-18 7/82 Original 5.3-3 7/82 Original 6.2-19 7/82 Original 5.3-4 7/82 Original 6.2-20 6/85 3 5.3-5 7/82 Original 6.2-21 6/85 3 5.3-6 6/85 3 6.2-22 6/85 3 5.3-7 3/88 7 6.2-23 8/89 8 5.3-8 6/85 3 6.2-24 6/85 3 5.3-9 6/85 3 6.2-25 6/85 3 5.3-10 6/85 3 6.2-26 7/82 Original 5.3-11 7/82 Original 6.2-27 7/82 Original 5.3-12 7/82 Original 6.2-28 7/82 Original 5.3-13 6/84 2 6.2-29 6/83 1 5.3-14 6/90 9 6.2-30 7/82 Original 5.3-15 7/82 Original 6.2-31 7/82 Original 5.3-16 7/82 Original 6.2-32 6/85 3 5.3-17 7/82 Original 6.2-33 7/82 Original 5.3-18 7/82 Original 6.2-34 7/82 Original 5.3-19 7/82 Original 6.2-35 7/82 Original 5.3-20 7/82 Original 6.2-36 7/82 Original Figure 5.3-1 7/82 Original 6.2-37 7/82 Original Figure 5.3-2 7/82 Original 6.2-38 7/82 Original Figure 5.3-3 7/82 Original 6.2-39 7/82 Original
- 5. 4-1 6/83 1 6.2-40 7/82 Original 5.4-2 7/82 Original 6.2-41 7/82 Original 5.5-1 7/82 Original 6.2-42 7/82 Original 5.5-2 6/85 3 6.2-43 7/82 Original 5.5-3 6/85 3 6.2-44 7/82 Original 6.2-45 7/82 Original Chapter 6 6.2-46 7/82 Original 6.2-47 7/82 Original 6-i 6/85 3 6.2-48 7/82 Original 6-ii 7/82 Original 6.2-49 6/85 3 6-iii 6/83 1 6.2-50 7/82 Original 6-iv 7/82 Original 6.2-51 7/82 Original 6-v 7/82 Original 6.2-52 7/82 Original 6.1-1 7/82 Original 6.2-53 6/85 3 6.1-2 7/82 Original 6.2-54 6/85 3 6.1-3 7/82 Original 6.2-55 7/82 Original 6.1-4 7/82 Original 6.2-56 6/85 3 6 .1-5 7/82 Original 6.2-57 6/85 3 Figure 6.1-1 6/85 3 6.2-58 7/82 Original 6.2-1 7/82 Original 6.2-59 7/82 Original 6.2-2 6/85 3 6.2-60 7/82 Original 6.2-3 6/85 3 6.2-61 6/85 3 6.2-4 7/82 Original 6.2-62 7/82 Original 6.2-5 7/82 Original 6.2-63 7/82 Original 6.2-6 6/85 3 6.2-64 7/82 Original 6.2-8 7/82 Original 6.2-65 6/85 3 6.2-9 6/83 ,1 6.2-66 6/85 3 6.2-10 6/83 1 6.2-67 7/82 Original 6.2-lOa 6/83 1 6.2-68 6/85 3 6.2-11 7/82 Original 6.2-69 7/82 Original 6.2-12 6/85 3 6.2-70 7/82 Original 6.2-13 7/82 Original Figure 6.2-1 6/87 6 6.2-14 6/85 3 Figure 6.2-2 6/87 6 6.2-15 6/85 3 Figure 6.2-3 6/87 6
REVISION 9 6/90 SPS UFSAR LEP-8 i.lS'.l: Q[ :r::rr:r::c:ny:r:: ~A~J::S ~ VOLUME II Figure 6.2-4 Figure 6.2-5
~
(continued) 6/87 7/82 7/82 Revision 6 Original Original
~
7.2-8 7.2-9 7.2-10
- 7. 2-11
~
7/82 7 /82. 7/82 7/82 Revision Original Original Original Original Figure 6.2-6 7.2-12 7/82 Original 6.3-1 6/83 1 7.2-13 7/82 Original 6.3-2 6/86 4 7.2-14 7/82 Original 6.3-3 6-83 1 7.2-15 7/82 Original 6.3-4 6/83 1 7.2-16 7/82 Original 6.3-5 6/83 1 7.2-17 7/82 Original 6.3-6 6/83 1 7.2-18 7/82 Original 6.3-7 6/83 1 7.2-19 7/82 Original 6.3-8 6/83 1 7.2-20 7/82 Original 6.3-9 6/83 1 7.2-21 6/85 3 6.3-10 6/83 1 7.2-21a 6/85 3
- 6. 3-11 7/82 Original 7.2-22 7/82 Original 6.3-12 7/82 Original 7.2-23 7/82 Original 6.3-13 7/82 Original 7.2-24 7/82 Original 6.3-14 7/82 Original 7.2-25 7/82 Original 6.3-15 7/82 Original 7.2-26 7/82 Original 6.3-16 7/82 Original 7.2-27 7/82 Original 6.3-17 6/83 1 7.2-28 7/82 Original 6.3-18 7/82 Original 7.2-29 7/82 Original 6.3-19 8/89 8 7.2-30 7/82 Original 6.3-20 7/82 Original 7.2-31 7/82 Original 6.3-21 7/82 Original 7.2-32 7/82 Original 6.3-22 7/82 Original 7.2-33 8/89 8 6.3-23 6/83 1 7.2-34 6/86 4 6.3-24 7/82 Original 7.2-34a 6/86 4 6.3-25 7/82 Original 7.2-35 7/82 Original 6.3-26 7/82 Original 7.2-36 7/82 Original 6.3-27 7/82 Original 7.2-37 7/82 Original Figure 6.3-1 6/83 1 7.2-38 7/82 Original Figure 6.3-2 6/83 1 7.2-39 7/82 Original Figure 6.3-3 7/82 Original 7.2-40 6/86 4 Figure 6.3-4 7/82 Original 7.2-41 7/82 Original Figure 6.3-5 7/82 Original 7.2-42 7/82 Original Figure 6.3-6 7/82 Original 7.2-43 7/82 Original 7.2-44 7/82 Original Chs:n.2:t~r 7 7.2-45 7/82 Original 7.2-46 7/82 Original 7-i 7/82 Original 7.2-47 7/82 Original 7-ii 8/89 8 7.2-48 7/82 Original 7-iii 7/82 Original 7.2-49 7/82 Original 7-iv 7/82 Original 7.2-50 7/82 Original 7-v 6/83 1 Figure 7.2-1 7/82 Original 7-vi 6/83 1 Figure 7.2-2 7/82 Original 7-vii 7/82 Original Figure 7.2-3 7/82 Original 7-viii 7/82 Original Figure 7.2-4 7/82 Original 7-ix 7/82 Original Figure 7.2-5 7/82 Original 7.1-1 7/82 Original Figure 7.2-6 6/85 3 7.2-1 7/82 Original Figure 7.2-7 7/82 Original 7.2-2 7/82 Original Figure 7.2-8 7/82 Original 7.2-3 7/82 Original Figure 7.2-9 7/82 Original 7.2-4 7/82 Original Figure 7.2-10 7/82 Original 7.2-5 7/82 Original Figure 7. 2-11 7/82 Original 7.2-6 7/82 Original Figure 7.2-12 7/82 Original 7.2-7 7/82 Original Figure 7.3-13 7/82 Original
'REVISION 9 6/90 SPS UFSAR LEP-9 LIST OF EFFECTIVE PAGES
- ~ ~ Revision ~ ~ Revision VOLUME II (continued) 7.5-5 7/82 Original 7.5-6 7/82 Original 7.3-1 7/82 Original 7.5-7 7/82 Original 7.3-2 7/82 Original 7.5-8 7/82 Original 7.3-3 7/82 Original 7.5-9 7/82 Original 7.3-4 7/82 Original 7.5-10 7/82 Original 7.3-5 7/82 Original 7. 5-11 7/82 Original 7.3-6 7/82 Original 7.5-12 7/82 Original 7.3-7 7/82 Original 7.5-13 6/90 9 7.3-8 7/82 Original 7.5-13a 6/90 9 7.3-9 7/82 Original 7.5-14 6/83 1 7.3-10 7/82 Original 7.5-15 6/83 1
- 7. 3-11 7/82 Original 7.5-16 6/83 1 7.3-12 7/82 Original 7.5-17 6/83 1 7.3-13 7/82 Original 7.5-18 6/83 1 7.3-14 7/82 Original 7.5-19 6/83 1 7.3-15 7/82 Original 7.5-20 6/83 1 7.3-16 7/82 Original 7.5-21 3/88 7 Figure 7.3-1 7/82 Original 7.5-2la 6/84 2 Figure 7.3-2 7/82 Original 7.5-22 6/84 2 Figure 7.3-3 7/82 Original 7 5-22a 6/83 1 7.4-1 7/82 Original 7.5-23 7/82 Original 7.4-2 7/82 Original 7.5-24 7/82 Original 7.4-3 7/82 Original 7.5-25 6/83 1 7.4-4 7/82 Original 7.5-26 6/85 3 7.4-5 7/82 Original 7.5-27 6/85 3 7.4-6 7/82 Original 7.5-28 7/82 Original 7.4-7 7/82 Original 7.5-29 6/83 1 7.4-8 7/82 Original 7.5-30 6/83 1 7.4-9 7/82 Original 7.5-31 7/82 Original 7.4-10 7/82 Original 7.5-32 7/82 Original
- 7. 4-11 7/82 Original 7.5-33 7/82 Original 7.4-12 7/82 Original 7.4-34 7/82 Original 7.4-13 7/82 Original 7.5-35 7/82 Original 7.4-14 7/82 Original 7.5-36 7/82 Original 7.4-15 7/82 Original Figure 7.5-1 7/82 Original 7.4-16 7/82 Original Figure 7.5-2 7/82 Original 7.4-17 7/82 Original Figure 7.5-3 7/82 Original 7.4-18 7/82 Original Figure 7.5-4 7/82 Original 7.4-19 7/82 Original 7.6-1 7/82 Original 7.4-20 7/82 Original 7.6-2 7/82 Original 7.4-21 7/82 Original 7.6-3 7/82 Original
- 7. 4-22 7/82 Original 7.6-4 7/82 Original 7.4-23 7/82 Original Figure 7.6-1 7/82 Original 7.4-24 7/82 Original Figure 7.6-2 7/82 Original 7.4-25 7/82 Original 7.7-1 3/88 7 7.4-26 7/82 Original 7.7-2 6/83 1 7.4-27 7/82 Original 7.7-3 6/83 1 7.4-28 7/82 Original 7.7-3 6/83 1 7.4-29 7/82 Original 7.7-4 6/83 1 7.4-30 7/82 Original 7.7-5 3/88 7 Figure 7.4-1 7/82 Original 7.7-6 6/83 1 Figure 7.4-2 7/82 Original 7.7-7 6/86 4 Figure 7.4-3 7/82 Original 7.7-8 3/88 7 7.5-1 7/82 Original 7.7-8a 3/88 7 7.5-2 7/82 Original 7 7-9 3/88 7 7.5-3 8/89 8 7.7-9a 3/88 7 7.5-4 7/82 Original 7 7-10 6/86 4
i REVISION 9 6/90 SPS UFSAR LEP-10 i.;i;s~ Q[ l::((l::!::~J::ll:: ~.!.Gl::S
~ ~ Revision ~ ~ Revision.
VOLUME II (continued) 8.5-6 6/83 1 8.5-7 6/90 9 Figure 7.7-1 6/87 6 8.5-7a 6/90 9 7.8-1 8/89 8 8.5-8 6/83 1 7.8-2 7/82 Original 8.5-9 6/83 1 7.8-3 7/82 Original 8.5-9a 12/86 5 7.8-4 7/82 Original 8.5-10 6/83 1 Figure 7.8-1 7/82 Original 8. 5-11 6/83 1 Figure 7.8-2 7/82 Original 8.5-12 6/83 1 7.9-1 7/82 Original 8.5-13 6/83 1 7.9-2 7/82 Original 8.5-14 6/83 Blank 7.9-3 7/82 Original 8.5-15 6/83 1 7.10-1 7/82 Original 8.6-1 7/82 Original 7.10-2 7/82 Original 8.6-2 7/82 Original 7.10-3 7/82 Original 8.6-3 7/82 Original 7.10-4 7/82 Original 7.10-5 7/82 Original Cha12ter Sl 7.10-6 7/82 Original 7.10-7 6/86 4 9-i 7/82 Original 7.10-8 7/82 Original 9-ii 7/82 Original 7.10-9 7/82 Original 9-iii 7/82 Original 9-iv 7/82 Original VOLUME III 9-v 6/83 1 9-vi 6/83 1 ChaI2ter 8 9-vii 6/83 1 9-viii 8/89 8 8-i 6/83 1 9-ix 7/82 Original 8-ii 6/83 1 9-x 6/83 1 8-iii 6/83 1 9-xi 7/82 Original 8.1-1 6/83 1 9-xii 7/82 Original
- 8. 1-2 7/82 Original 9-xiii 6/83 1 Figure 8 .1-1 7/82 Original 9-xiv 6/83 l 8.2-1 7/82 Original 9-xv 6/84 2 8.2-2 6/84 2 9.1-1 7/82 Original 8.2-3 6/90 9 9.1-2 7/82 Original 8.2-4 7/82 Original 9.1-3 6/85 3 8.3-1 6/87 6 9.1-4 6/85 3 8.3-2 6/86 4 9.1-5 7/82 Original 8.3-3 6/86 4 9.1-6 7/82 Original 8.3-4 6/86 4 9.1-7 7/82 Original Figure 8.3-1 6/86 4 9.1-8 6/85 3 Figure 8.3-2 7/82 Original 9.1-9 12/86 5 8.4-1 6/86 4 9.l-9a 12/86 5 8.4-2 6/86 4 9.1-10 7/82 Original 8.4-2a 3/88 7 9 .1-11 7/82 Original 8.4-3 6/86 4 9.1-12 7/82 Original 8.4-4 6/86 4 9.1-13 7/82 Original 8.4-4a 6/86 4 9.1-14 7/82 Original 8.4-5 6/84 2 9.1-15 7/82 Original 8.4-6 6/86 4 9.1-16 7/82 Original 8.4-7 6/86 4 9.1-17 6/85 3 Figure 8. 4-1 6/86 4 9.1-18 7/82 Original Figure 8. 4-2 6/87 6 9.1-19 7/82 Original 8.5-1 7/82 Original 9.1-20 7/82 Original 8.5-2 7/82 Original 9.1-21 7/82 Original 8.5-3 7/82 Original 9.1-22 7/82 Original 8.5-4 7/82 Original 9.1-23 7/82 Original 8.5-5 6/90 9 9.1-24 7/82 Original J
, REVISION 9 6/90 SPS UFSAR LEP-11 LIST OF EFFECTIVE PAGES P..sul..e. ~ Revision ~ ~ Revision VOLUME I I I (continued) Figure 9.2-2 3/88 7 Figure 9.2-3 6/87 6 9.1-25 6/85 3 Figure 9.2-4 6/87 6 9 .1-26 6/85 3 9.3-1 7/82 Original 9.1-27 6/85 3 9.3-2 7/82 Original 9.1-28 7/82 Original 9.3-3 12/86 5 9.1-29 7/82 Original 9.3-4 12/86 5 9.1-30 6/83 1 9.3-4a 6/86 4 9.1-31 6/83 1 9.3-5 6/85 3 9.1-32 7/82 Original 9.3-6 7/82 Original 9.1-33 7/82 Original 9.3-7 7/82 Original 9.1-34 7/82 Original 9.3-8 7/82 Original 9.1-35 6/85 3 9.3-9 7/82 Original 9.1-36 7/82 Blank 9.3-10 7/82 Original 9.1-37 7/82 Original 9. 3-11 7/82 Original 9.1-38 7/82 Blank 9.3-12 6/84 2 9.1-39 6/85 3 9.3-13 7/82 Original 9.1-40 7/82 Original 9.3-14 7/82 Original 9.1-41 7/82 Original Figure 9.3-1 6/87 6 9.1-42 7/82 Original 9.4-1 7/82 Original 9.1-43 6/85 3 9.4-2 7/82 Original 9.1-44 7/82 Original 9.4-3 7/82 Original 9.1-45 7/82 Original 9.4-4 6/86 4 9.1-46 7/82 Original 9.4-4a 6/86 4 9.1-47 7/82 Original 9.4-5 7/82 Original 9.1-48 7/82 Original 9.4-6 7/82 Original 9.1-49 7/82 Original 9.4-7 12/86 5 9.1-50 7/82 Original 9.4-8 12/86 5 9.1-51 6/85 3 9.4-8a 6/86 4 9.1-52 7/82 Original 9.4-9 7/82 Original Figure 9.1-1 6/87 6 9.4-10 7/82 Original Figure 9.1-2 6/87 6 9.4-11 7/82 Original Figure 9.1-3 6/87 6 9.4-12 7/82 Original Figure 9.1-4 6/87 6 9.4-13 7/82 Original Figure 9.1-5 3/88 7 9.4-14 3/88 7 Figure 9.1-6 6/87 6 9.4-14a 6/85 3 9.2-1 7/82 Original 9.4-15 7/82 Original 9.2-2 6/85 3 9.4-16 7/82 Original 9.2-3 3/88 7 9.4-18 7/82 Original 9.2-4 6/83 1 9.4-19 7/82 Original 9.2-5 6/83 1 9.4-20 7/82 Original 9.2-6 7/82 Blank 9.4-21 7/82 Original 9.2-7 7/82 Original 9.4-22 7/82 Original 9.2-8 7/82 Original 9.4-23 7/82 Original 9.2-9 7/82 Original 9.4-24 7/82 Original 9.2-10 7/82 Original 9.4-25 7/82 Original
- 9. 2-11 7/82 Original 9.4-26 7/82 Original 9.2-12 7/82 Original 9.4-27 7/82 Original
- 9. 2-13 7/82 Original 9.4-28 7/82 Original 9.2-14 7/82 Original 9.4-29 7/82 Original 9.2-15 7/82 Original 9.4-30 7/82 Original 9.2-16 7/82 Original 9.4-31 7/82 Original 9.2-17 7/82 Original 9.4-32 6/86 4 9.2-18 7/82 Original 9.4-33 8/89 8 9.2-19 7/82 Original 9.4-34 7/82 Original 9.2-20 3/88 7 9.4-35 7/82 Original 9.2-21 7/82 Original 9.4-36 7/82 Original Figure 9.2-1 6/87 6 Figure 9.4-1 6/87 6
l REVISION 9 6/90 SPS UFSAR LEP-f2 i.lSI Q[ l:i((l:H:~UY:l:: ~AGl:iS ~ ~ Revision ~ ~ Revision. VOLUME III (continued) 9.7-10 6/83 1
- 9. 7-11 6/83 1 Figure 9.4-2 6/87 6 9.7-12 6/83 1 Figure 9.4-3 6/87 6 Figure 9.7-1 6/87 6 Figure 9.4-4 6/87 6 Figure 9.7-2 6/87 6 Figure 9.4-5 6/87 6 Figure 9.7-3 6/87 6 Figure 9.4-6 6/87 6 Figure 9.7-4 6/87 6 Figure 9.4-7 6/87 6 9.8-1 6/83 1 Figure 9.4-8 6/87 6 9.8-2 6/85 3 Figure 9.4-9 6/87 6 9.8-3 6/83 1 9.5-1 7/82 Original 9.8-4 6/83 1 9.5-2 7/82 Original 9.8-5 6/84 2 9.5-3 7/82 Original 9.8-6 7/82 Blank 9.5-4 7/82 Original 9.8-7 6/86 4 9.5-5 7/82 Original 9.8-8 7/82 Original 9.5-6 7/82 Original 9.8-9 7/82 Original 9.5-7 7/82 Original 9.8-10 7/82 Original 9.5-8 7/82 Original Figure 9.8-1 3/88 7 9.5-9 7/82 Original Figure 9.8-2 6/87 6 9.5-10 7/82 Original 9.9-1 8/89 8
- 9. 5-11 7/82 Original 9.9-2 8/89 8 9.5-12 7/82 Original 9.9-3 8/89 8 Figure 9.5-1 6/87 6 9.9-4 6/90 9 9.6-1 6/84 2 9.9-5 8/89 8 9.6-2 6/84 2 9.9-6 6/90 9 9.6-3 6/84 2 9.9-7 8/89 8 9.6-4 6/83 1 9.9-8 6/90 9 9.6-5 6/84 2 9.9-9 6/90 9 9.6-6 6/84 2 9.9-10 8/89 8 9.6-7 6/84 2 Figure 9.9-1 6/90 9 9.6-8 6/84 2 9.10-1 7/82 Original 9.6-10 6/84 2 9.10-2 6/85 3 9.6-11 6/84 2 9.10-3 7/82 Original 9.6-12 6/83 1 9.10-4 6/85 3 9.6-13 6/84 2 9.10-5 7/82 Original 9.6-14 6/83 1 9.10-6 7/82 Original 9.6-15 7/82 Original 9.10-7 7/82 Original 9.6-16 7/82 Original 9.10-8 7/82 Original 9.6-17 6/84 2 9.10-9 3/88 7 9.6-18 6/84 2 9.10-9a 3/88 7 9.6-19 7/82 Original 9.10-9b 3/88 7 9.6-20 6/84 2 9.10-10 6/83 1 9.6-21 6/84 2 9.10-11 6/86 4 Figure 9.6-1 6/87 6 9.10-12 6/87 6 Figure 9.6-2 6/87 6 9.10-12a 6/87 6 Figure 9.6-3 6/87 6 9 .10-13 6/84 4 Figure 9.6-3a 6/87 6 9.10-14 12/86 5 Figure 9.6-4 3/88 7 9.10-14a 12/86 5 Figure 9.6-4a 6/87 6 9.10-15 6/83 1 9.7-1 7/82 Original 9.10-16 6/83 1 9.7-2 7/82 Original 9.10-17 12/86 5 9.7-3 6/83 1 9.10-17a 12/86 5 9.7-4 6/87 6 9.10-18 6/83 1 9.7-5 6/83 1 9.10-19 6/83 1 9.7-6 6/83 1 9.10-20 6/86 4 9.7-7 6/83 1 9.10-21 8/89 8 9.7-8 6/83 1 9.10-22 8/89 8 9.7-9 6/83 1 9.10-22a 8/89 8
. ' REVISION 9 6/90 SPS UFSAR LEP-13 LIST OF EFFECTIVE PAGES
~ ~ Revision ~ ~ Revision VOLUME III (continued) 9.12-2 7/82 Original 9.12-3 7/82 Original 9.10-23 6/84 2 9.12-4 6/86 4 9.10-24 6/86 4 9.12-4a 6/86 4 9.10-24a 6/86 4 9.12-5 7/82 Original 9.10-25 7/82 Original 9.12-6 7/82 Original 9.10-26 7/82 Original 9.12-7 7/82 Original 9.10-27 6/87 6 9.12-8 6/87 6 9.10-27a 6/87 6 9.12-9 7/82 Original 9.10-28 7/82 Original 9.12-10 6/83 1 9.10-29 7/82 Original 9. 12-11 6/83 1 9.10-30 7/82 Original 9.12-12 3/88 7 9.10-31 7/82 Original 9.12-13 3/88 7 9.10-32 7/82 Original 9.12-13a 3/88 7 9.10-33 7/82 Original 9.12-14 6/83 1 9.10-34 7/82 Original 9.12-15 6/87 6 9.10-35 7/82 Original 9.12-16 6/83 1 9.10-36 3/88 7 9.12-17 3/88 7 9.10-36a 3/88 7 9.12-18 6/83 1 9.10-37 7/82 Original 9.12-19 7/82 Original 9.10-38 7/82 Original Figure 9.12-1 7/82 Original 9.10-39 7/82 Original 9.13-1 6/90 9 9.10-40 6/90 9 9.13-2 6/90 9 9.10-41 7/82 Original 9.13-3 6/90 9 9.10-42 7/82 Original 9 .13-4 6/90 9 9.10-43 7/82 Original 9.13-5 6/90 9 9.10-44 7/82 Original 9 .13-6 6/90 9 9.10-45 6/83 1 9.13-7 6/90 9 9.10-46 6/83 1 9.13-8 6/90 9 9.10-47 6/85 3 9.13-9 6/90 9 9.10-48 7/82 Original 9.13-10 6/90 9 9.10-49 3/88 7 9.13-11 6/90 9 Figure 9.10-1 3/88 7 9.13-12 6/90 9 9.11-1 6/84 2 9.13-13 6/90 9 9.11-2 6/84 2 9.13-14 6/90 9 9.ll-2a 6/84 2 9.13-15 6/90 9 9.11-3 7/82 Original 9.13-16 6/90 9 9.11-4 7/82 Original 9.13-17 6/90 9
- 9. 11-5 6/84 2 Figure 9.13-1 6/87 6 9.11-6 6/84 2 Figure 9 .13-2 3/88 7 9.11-7 6/84 2 Figure 9.13-3 6/87 6 9.11-8 6/84 2 Figure 9.13-4 3/88 7 9.11-9 6/84 2 9.14-1 7/82 Original 9.11-10 6/84 2 9.14-2 7/82 Original 9 .11-11 6/84 2 9 14-3 7/82 Original 9.11-12 6/84 2 9.14-4 7/82 Blank
- 9. 11-13 6/84 2 9.14-5 7/82 Original 9.11-14 6/84 2 Figure 9.14-1 3/88 7 9.11-15 6/84 2 9.15-1 7/82 Original Figure 9.11-1 6/87 6 9.15-2 7/82 Original Figure 9.11-2 6/87 6 9.15-3 7/82 Original Figure 9.ll-2A 3/88 7 9.15-4 7/82 Original Figure 9.ll-2B 6/87 6 9 15-5 7/82 Original 9.15-6 7/82 Original Figure 9.ll-2C 6/87 6 Figure 9.ll-2D 6/87 6 9.15-7 7/82 Original Figure 9. ll-2E 6/87 6 9A-i 7/82 Original Figure 9.11-2F 6/87 6 9A-ii 7/82 Blank 9.12-1 7/82 Original 9A-iii 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-14 *.. i.is:i: Q:1£ J::1£1::J::CIJ;YJ:: PAGJ::S ~ ~ Revision ~ Revision. VOLUME III (continued) VOLUME IV 9A-iv 7/82 Original Ch!;!I:it~r 10 9A-v 7/82 Original 9A-l 7/82 Blank 10-i 8/89 8 9A-2 7 82 Original 10-ii 6/83 1 9A-3 7/82 Original 10-iii 7/82 Original 9A-4 7/82 Original 10-iv 7/82 Original 9A-5 7/82 Original 10-v 7/82 Original 9A-6 7/82 Blank 10.1-1 7/82 Original 9A-7 7/82 Original 10.1-2 7/82 Original 9A-8 7/82 Original 10.2-1 7/82 Original 9A-9 7/82 Original Figure 10.2-1 6/87 6 9A-10 7/82 Original Figure 10.2-2 6/87 6 9A-ll 7/82 Original Figure 10.2-3 6/87 6 9A-12 7/82 Original Figure 10.2-4 6/87 6 9A-13 7/82 Original Figure 10.2-5 6/87 6 9A-14 7/82 Original Figure 10.2-6 6/87 6 9A-15 7/82 Original Figure 10.2-7 6/87 6 9A-16 7/82 Original Figure 10.2-8 6/87 6 9A-17 7/82 Original Figure 10.2-9 6/87 6 9A-18 7/82 Original Figure 10.2-10 6/87 6 9A-19 7/82 Original 10.3-1 7/82 Original 9A-20 7/82 Original 10.3-2 8/89 8 9A-21 7/82 Original 10.3-3 6/86 4 9A-22 7/82 Original 10.3-4 7/82 Original 9A-23 7/82 Original 10.3-5 7/82 Original 9A-24 7/82 Original 10.3-6 7/82 Original Figure 9A-l 7/82 Original 10.3-7 6/83 1 Figure 9A-2 7/82 Original 10.3-8 6/83 1 9B-i 7/82 Original 10.3-9 6/86 4 9B-ii 7/82 Blank 10.3-10 6/86 4 9B-iii 6/83 1 10.3-lOa 6/86 4 9B-l 6/87 6 10.3-11 6/86 4 9B-2 7/82 Original 10.3-12 8/89 8 9B-3 7/82 Original 10.3-12a 8/89 8 9B-4 6/83 1 10.3-13 7/82 Original 9B-5 6/83 1 10.3-14 7/82 Original 9B-6 6/83 1 10.3-15 7/82 Original 9B-7 6/83 1 10.3-16 7/82 Original 9B-8 6/83 Blank 10.3-17 6/83 1 9B-9 6/83 1 10.3-18 7/82 Original 9B-10 6/83 1 10.3-19 7/82 Original 9B-ll 6/83 1 10.3-20 7/82 Original 9B-12 6/83 1 10.3-21 7/82 Original 9B-13 6/83 1 10.3-22 7/82 Original 9B-14 6/83 1 10.3-23 7/82 Original 9C-i 7/82 Original 10.3-24 7/82 Original 9C-ii 7/82 Blank 10.3-25 7/82 Original 9C-iii 7/82 Original 10.3-26 6/90 9 9C-l 7/82 Original 10.3-27 6/90 9 9C-2 7/82 Blank 10.3-27a 6/90 9 9C-3 7/82 Original 10.3-28 6/90 9 9C-4 7/82 Original 10.3-29 6/90 9 9C-5 3/88 7 10.3-29a 6/90 9 9C-6 7/82 Original 10.3-30 6/90 9 10.3-31 7/82 Original 10.3-32 6/90 9
l REVISION 9 6/90 SPS UFSAR LEP-15 LIST OF EFFECTIVE PAGES
~ ~ Revision ~ ~ Revision VOLUME IV (continued) 11-iii 6/83 1 11-iv 7/82 Original 10.3-33 6/83 1 11-v 7/82 Original 10.3-34 6/83 1 11-vi 7/82 Original 10.3-34a 6/83 1 11-vii 7/82 Original 10.3-35 6/83 1 11.1-1 7/82 Original 10.3-36 7/82 Original 11.1-2 7/82 Original 10.3-37 8/89 8 11. 2-1 7/82 Original 10.3-38 6/83 1 11. 2-2 7/82 Original 10.3-39 6/90 9 11. 2-3 8/89 8 10.3-,40 6/90 9 11. 2-4 6/86 4 10.3-41 6/83 1 11.2-5 6/86 4 10.3-42 6/83 1 11.2-6 6/90 9 10.3-43 8/89 8 11.2-7 6/83 1 10.3-43a 8/89 8 li.2-8 6/83 1 10.3-44 6/83 1 11. 2-8a 6/83 1 10.3-45 6/83 1 11.2-9 7/82 Original 10.3-46 6/83 1 11. 2-10 7/82 Original 10.3-47 6/83 1 11. 2-11 7/82 Original 10.3-48 6/83 1 11. 2-12 7/82 . Original 10.3-49 6/83 1 11. 2-13 7/82 Original 10.3-50 6/83 1 11. 2-14 7/82 Original 10.3-51 6/83 1 11. 2-15 7/82 Original 10.3-52 6/83 1 11. 2-16 7/82 Original 10.3-53 6/83 1 11.2-17 7/82 Original 10.3-54 6/83 1 11. 2-18 6/83 1 10.3-55 7/82 Original 11. 2-19 6/83 1 10.3-56 7/82 Original 11.2-20 6/83 1 Figure 10.3-1 3/88 7 11.2-21 7/82 Original Figure 10.3-2 6/86 4 11. 2-22 7/82 Original Figure 10.3-3 7/82 Original 11. 2-23 7/82 Original Figure 10.3-4 7/82 Original 11.2-24 7/82 Original Figure 10.3-5 7/82 Original 11.2-25 7/82 Original Figure 10.3-6 7/82 Original 11. 2-26 7/82 Original Figure 10.3-7 3/88 7 11. 2-27 7/82 Original Figure 10.3-8 6/87 6 11.2-28 7/82 Original Figure 10.3-9 7/82 Original 11.2-29 6/83 1 Figure 10.3-10 7/82 Original 11.2-30 6/83 1 Figure 10.3-11 7/82 Original 11. 2-31 6/83 1 Figure 10.3-12 7/82 Original 11. 2-32 6/83 1 Figure 10.3-13 7/82 Original 11. 2-33 6/83 1 Figure 10.3-14 7/82 Original 11. 2-34 7/82 Blank Figure 10.3-15 7/82 Original 11. 2-35 7/82 Original Figure 10.3-16 7/82 Original 11. 2-36 7/82 Original Figure 10.3-17 7/82 Original 11.2-37 7/82 Original Figure 10.3-18 7/82 Original 11.2-38 7/82 Original Figure 10.3-19 7/82 Original 11.2-39 6/90 9 Figure 10.3-20 6/90 9 11.2-40 7/82 Original Figure 10.3-21 3/88 7 11.2-41 7/82 Original Figure 10.3-22 3/88 7 11.2-42 7/82 Original Figure 10.3-23 3/88 7 11.2-43 7/82 Original Figure 10.3-24 3/88 7 11.2-44 7/82 Original Figure 10.3-25 3/88 7 11.2-45 7/82 Original 11.2-46 7/82 Original Chai;2ter 11 11. 2-4 7 7/82 Original 11.2-48 7/82 Original 11-i 6/83 1 11.2-49 7/82 Original 11-ii 6/83 1 11.2-50 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-16 1.lS:I: Q[ !:i[J::!:i!:'.U:il:!:i fAG!:iS ~ ~ Revision ~ ~ Revision. VOLUME IV (continued) 11. 3-38 7/82 Original
- 11. 3-39 6/86 4
- 11. 2-51 7/82 Original 11.3-40 6/86 4 11.2-52 7/82 Original 11.3-41 6/84 2 11.2-53 7/82 Original Figure 11. 3-1 6/86 4
- 11. 2-54 7/82 Original Figure 11. 3-2 7/82 Original
- 11. 2-55 7/82 Original llA-i 7/82 Original 11.2-56 7/82 Original llA-ii 7/82 Blank 11.2-57 7/82 Original llA-iii 7/82 Original
- 11. 2-58 7/82 Original llA-iv 6/83 1
- 11. 2-59 7/82 Original llA-v 6/83 1
- 11. 2-60 7/82 Original llA-1 7/82 Original Figure 11. 2-1 3/88 7 llA-2 7/82 Blank Figure 11. 2-2 3/88 7 llA-3 6/83 1 Figure 11. 2-3 6/87 6 llA-4 6/83 1 Figure 11.2-4 7/82 Original llA-5 6/83 1 Figure 11.2-5 7/82 Original llA-6 6/83 Blank Figure 11.2-6 7/82 Original llA-7 6/83 1 Figure 11.2-7 6/87 6 llA-8 7/82 Blank Figure 11. 2-8 3/88 7 llA-9 6/83 1
- 11. 3-1 7/82 Original llA-10 6/83 1 11.3-2 7/82 Original llA-11 6/83 1
- 11. 3-3 7/82 Original llA-12 6/83 1 11.3-4 7/82 Original llA-13 7/82 Original
- 11. 3-5 7/82 Original llA-14 7/82 Original
- 11. 3-6 7/82 Original llA-15 6/83 1
- 11. 3-7 7/82 Original llA-16 6/83 1
- 11. 3-8 7/82 Original llA-1 7 6/83 1
- 11. 3-9 7/82 Original llA-18 6/83 1 11.3-10 7/82 Original llA-19 6/83 1
- 11. 3-11 6/84 2 llA-20 6/83 1
- 11. 3-12 7/82 Original llA-21 6/83 1 11.3-13 7/82 Original llA-22 6/83 1 11.3-14 7/82 Original llA-23 6/83 1
- 11. 3-15 8/89 8 llA-24 6/83 1 11.3-16 8/89 8 llA-25 6/83 1
- 11. 3-17 6/83 1 llA-26 6/83 1 11.3-18 6/85 3 llA-27 6/83 1 11.3-19 6/83 1 llA-28 6/83 1 11.3-20 6/83 1 llA-29 6/83 1
- 11. 3-21 6/83 1 llA-30 6/83 1 11.3-22 6/84 2 llA-31 6/83 1 ll.3-22a 6/85 3 llA-32 6/83 1
- 11. 3-23 7/82 Original 11.3-24 6/90 9 Chi;!i:;lter 12
- 11. 3-25 7/82 Original 11.3-26 7/82 Original 12-i 3/88 7
- 11. 3-27 7/82 Original 12-ii 3/88 7 11.3-28 6/90 9 12-iii 3/88 7 11.3-29 6/85 3 12-iv 3/88 7
- 11. 3-30 7/82 Original 12.1-1 3/88 7
- 11. 3-31 6/85 3 12.2-1 3/88 7
- 11. 3-32 7/82 Original 12.2-2 3/88 7 11.3-33 6/85 3 12.2-3 3/88 7
- 11. 3-34 7/82 Original 12.2-4 3/88 7 11.3-35 7/82 Original 12.2-5 3/88 7 11.3-36 7/82 Original 12.2-6 3/88 7 11.3-37 7/82 Original 12.2-7 3/88 7
REVISION 9 6/90 SPS UFSAR LEP-17 lilSI Q[ l:i((l:i~Il:!ll:i ~.Mzl:iS
- ~
12.2-8 12.2-9
~
VOLUME IV (continued) 3/88 3/88 Revision 7 7
~
13.2-5 13.3-1 13.3-2 13.3-3
~
7/82 7/82 7/82 7/82 Revision Original Original Original Original 12.2-10 3/88 7 13.3-4 7/82 Original
- 12. 2-11 3/88 7 13.3-5 7/82 Original 12.2-12 3/88 7 13.3-6 7/82 Original 12.2-13 3/88 7 13.3-7 7/82 Original 12.2-14 3/88 7 13.3-8 7/82 Blank 12.2-15 3/88 7 13. 3-9 7/82 Original 12.2-16 3/88 7 13.3-10 7/82 Original 12.2-17 3/88 7 13. 3-11 7/82 Original 12.2-18 3/88 7 13.3-12 7/82 Original Figure 12.2-la 3/88 7 13. 3-13 7/82 Original Figure 12.2-lb 3/88 7 13.3-14 7/82 Original Figure 12.2-2 3/88 7 13.4-1 7/82 Original 12.3-1 3/88 7 13. 4-2 7/82 Original 12.3-2 3/88 7 13.4-3 7/82 Original 12.3-3 3/88 7 13.4-4 7/82 Original 12.3-4 3/88 7 13.5-1 7/82 Original 12.3-4a 6/85 3 13.5-2 7/82 Original 12.3-4b 6/85 3 13.5-3 7/82 Original 12.3-5 6/84 2 13. 5-4 7/82 Original 12.4-1 6/83 1 13.5-5 7/82 Original 12.4-2 6/83 1 13.5-6 7/82 Original 12.5-1 6/84 2 13.5-7 7/82 Original 12.5-2 7/82 Original 13. 5-8 7/82 Original 12.5-3 6/83 1 13.5-9 7/82 Original 12.6-1 6/83 1 13.5-10 7/82 Original 12.6-2 6/83 1 13. 5-11 7/82 Original 12.7-1 7/82 Original 13.5-12 7/82 Original 12.8-1 6/83 1 13.5-13 7/82 Original 12.9-1 3/88 7 13.5-14 7/82 Original 12.9-2 3/88 7 13.5-15 7/82 Original 13.5-16 7/82 Original Chai;its;:r 13 13.5-17 7/82 Original 13.5-18 7/82 Original 13.1-i 7/82 Original 13.5-19 7/82 Original 13.1-ii 7/82 Original 13.5-20 7/82 Original 13.1-iii 7/82 Original 13. 5-21 7/82 Original 13.1-1 7/82 Original 13.5-22 7/82 Original 13.1-2 7/82 Original 13. 5-23 7/82 Original 13.1-3 7/82 Original 13.5-24 7/82 Original 13.1-4 7/82 Original 13. 5-25 7/82 Original 13.1-5 7/82 Original 13. 5-26 7/82 Original 13.1-6 7/82 Original 13.1-7 7/82 Original VOLUME V 13 .1-8 7/82 Original 13 .1-9 7/82 Original Chi;li;iter 14 13.1-10 7/82 Original 13.1-11 7/82 Original 14-i 8/89 8 13.1-12 7/82 Original 14-ii 7/82 Original 13.1-13 7/82 Original 14-iii 7/82 Original 13.1-14 7/82 Original 14-iv 7/82 Original 13.2-1 7/82 Original 14-v 6/83 1 13.2-2 7/82 Original 14-vi 7/82 Original 13.2-3 7/82 Original 14-vii 7/82 Original 13.2-4 7/82 Original 14-viii 7/82 Original 11
l REVISION 9 6/90 SPS UFSAR LEP-18
;t.IS? QI:: J::1::1::J::C?I:Y:J:: ~a.GJ::S
~ ~ Revision ~ ~ Revision. VOLUME V (continued) 14.2-43 7/82 Original 14.2-44 7/82 Original 14-ix 7/82 Original 14.2-45 7/82 Original 14-x 7/82 Original 14.2-46 7/82 Original 14-xi 7/82 Original 14.2-47 7/82 Original 14-xii 7/82 Original 14.2-48 7/82 Original 14-xiii 7/82 Original 14.2-49 6/90 9 14-xiv 7/82 Original 14.2-50 6/90 9 14-xv 7/82 Original 14.2-51 6/90 9 14-xvi 7/82 Original 14.2-52 7/82 Original 14-xvii 7/82 Original 14.2-53 7/82 Original 14-xviii 6/85 3 14.2-54 7/82 Original 14.1-1 7/82 Original 14.2-55 7/82 Original 14.1-2 7/82 Original 14.2-56 6/90 9 14.1-3 7/82 Original 14.2-56a 6/90 9 14.2-1 7/82 Original 14.2-57 6/90 9 14.2-2 7/82 Original 14.2-58 7/82 Original 14.2-3 8/89 8 14.2-59 7/82 Original 14.2-3a 8/89 8 14.2-60 7/82 Original 14.2-4 8/89 8 14.2-61 7/82 Original 14.2-5 8/89 8 14.2-62 7/82 Original 14.2-6 8/89 8 14.2-63 7/82 Original 14.2-7 7/82 Original 14.2-64 7/82 Original 14.2-8 7/82 Original 14.2-65 7/82 Original 14.2-9 7/82 Original 14.2-66 7/82 Original 14.2-10 7/82 Original 14.2-67 7/82 Original. 14.2-11 7/82 Original 14.2-68 8/89 8 14.2-12 7/82 Original 14.2-69 7/82 Original 14.2-13 7/82 Original 14.2-70 7/82 Original 14.2-14 7/82 Original 14.2-71 7/82 Original 14.2-15 7/82 Original Figure 14.2-1 8/89 8 14.2-16 7/82 Original Figure 14.2-2 7/82 Original 14.2-17 7/82 Original Figure 14.2-3 8/89 8 14.2-18 7/82 Original Figure 14.2-4 8/89 8 14.2-19 7/82 Original Figure 14.2-5 7/82 Original 14.2-20 7/82 Original Figure 14. 2-6 7/82 Original 14.2-21 7/82 Original Figure 14.2-7 7/82 Original 14.2-22 7/82 Original Figure 14.2-8 7/82 Original 14.2-23 7/82 Original Figure 14.2-9 7/82 Original 14.2-24 7/82 Original Figure 14.2-10 7/82 Original 14.2-25 7/82 Original Figure 14.2-11 7/82 Original 14.2-26 7/82 Original Figure 14.2-12 7/82 Original 14.2-27 7/82 Original Figure 14.2-13 7/82 Original 14.2-28 7/82 Original Figure 14.2-14 7/82 Original 14.2-29 7/82 Original Figure 14.2-15 7/82 Original 14.2-30 7/82 Original Figure 14.2-16 7/82 Original 14.2-31 7/82 Original Figure 14.2-17 7/82 Original 14.2-32 7/82 Original Figure 14.2-18 7/82 Original 14.2-33 7/82 Original Figure 14.2-19 7/82 Original 14.2-34 7/82 Original Figure 14.2-20 7/82 Original 14.2-35 7/82 Original Figure 14.2-21 7/82 Original 14.2-36 7/82 Original Figure 14.2-22 7/82 Original 14.2-37 7/82 Original Figure 14.2-23 7/82 Original 14.2-38 8/89 8 Figure 14.2-24 7/82 Origina. 14.2-39 7/82 Original Figure 14.2-25 7/82 Origina 14.2-40 7/82 Original Figure 14.2-26 7/82 Original 14.2-41 7/82 Original Figure 14.2-27 7/82 Original 14.2-42 7/82 Original Figure 14.2-28 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-1-9 l.IS:I: QJ:: iJ::J::iC:l:UZi PAGES
- ~
Figure Figure Figure 14.2-29 14.2-30 14.2-31 12..a.t..e. VOLUME V (continued) 7/82 7/82 7/82 Revision Original Original Original
~
Figure 14.2-85 Figure 14.2-86 Figure 14.2-87 Figure 14.2-88 Figure 14.2-89 12..a.t..e. 7/82 7/82 7/82 7/82 7/82 Revision Original Original Original Original Original Figure 14.2-32 7/82 Original Figure 14.2-90 7/82 Original Figure 14.2-33 7/82 Original Figure 14.2-91 7/82 Original Figure 14.2-34 7/82 Original Figure 14.2-92 7/82 Original Figure 14.2-35 7/82 Original Figure 14.2-93 7/82 Original Figure 14.2-36 7/82 Original Figure 14.2-94 7/82 Original Figure 14.2-37 7/82 Original Figure 14.2-95 7/82 Original Figure 14.2-38 7/82 Original Figure 14.2-96 7/82 Original Figure 14.2-39 7/82 Original Figure 14.2-97 7/82 Original Figure 14.2-40 7/82 Original Figure 14.2-98 7/82 Original Figure 14.2-41 7/82 Original Figure 14.2-99 3/88 7 Figure 14.2-42 7/82 Original Figure 14.2-100 3/88 7 Figure 14.2-43 7/82 Original Figure 14.2-101 3/88 7 Figure 14.2-44 7/82 Original Figure 14.2-102 3/88 7 Figure 14.2-45 7/82 Original Figure 14.2-103 7/82 Original Figure 14.2-46 7/82 Original Figure 14.2-104 7/82 Original Figure 14.2-47 7/82 Original 14.3-1 7/82 Original Figure 14.2-48 7/82 Original 14.3-2 7/82 Original Figure 14.2-49 7/82 Original 14.3-3 7/82 Original Figure 14.2-50 7/82 Original 14.3-4 7/82 Original Figure 14.2-51 7/82 Original 14.3-5 6/85 3 Figure 14.2-52 7/82 Original 14.3-6 7/82 Original
- Figure 14.2-53 7/82 Original 14.3-7 7/82 Original Figure 14.2-54 7/82 Original 14.3-8 7/82 Original Figure 14.2-55 7/82 Original 14.3-9 6/85 3 Figure 14.2-56 7/82 Original 14.3-10 6/85 3 Figure 14.2-57 7/82 Original 14. 3-11 6/85 3 Figure 14.2-58 7/82 Original 14.3-12 6/85 3 Figure 14.2-59 7/82 Original 14.3-13 6/85 3 Figure 14.2-60 7/82 Original 14.3-14 6/85 3 Figure 14.2-61 7/82 Original 14.3-15 6/85 3 Figure 14.2-62 7/82 Original 14.3-16 6/85 3 Figure 14.2-63 7/82 Original 14.3-17 6/85 3 Figure 14.2-64 7/82 Original 14.3-18 6/85 3 Figure 14.2-65 7/82 Original 14.3-19 6/85 3 Figure 14.2-66 7/82 Original 14.3-20 6/85 3 Figure 14.2-67 7/82 Original 14.3-21 7/82 Original Figure 14.2-68 7/82 Original 14.3-22 7/82 Original Figure 14.2-69 7/82 Original 14.3-23 6/85 3 Figure 14.2-70 7/82 Original 14.3-24 6/85 3 Figure 14.2-71 7/82 Original 14.3-25 7/82 Original Figure 14.2-72 7/82 Original 14.3-26 7/82 Original Figure 14.2-73 7/82 Original 14.3-27 6/85 3 Figure 14.2-74 7/82 Original 14.3-28 6/85 3 Figure 14.2-75 7/82 Original 14.3-29 7/82 Original Figure 14.2-76 7/82 Original 14.3-30 7/82 Original Figure 14.2-77 7/82 Original 14.3-31 7/82 Original Figure 14.2-78 7/82 Original 14.3-32 7/82 Original Figure 14.2-79 7/82 Original 14.3-33 7/82 Original Figure 14.2-80 7/82 Original 14.3-34 7/82 Original Figure 14.2-81 7/82 Original 14.3-35 6/85 3 Figure 14.2-82 7/82 Original 14.3-36 6/85 3 Figu:te 14.2-83 7/82 Original 14.3-37 7/82 Original Figure 14.2-84 7/82 Original 14.3-38 6/85 3
REVISION 9 6/90 SPS UFSAR LEP-20 LIST QF EFFECTIVE PAGES ~ ~ Revision ~ ~ Revision VOLUME V (continued) 14.5-7 7/82 Original 14.5-8 7/82 Original 14.3-39 6/85 3 14.5-9 7/82 Original 14.3-40 7/82 Original 14.5-10 8/89 8 14.3-41 7/82 Original 14. 5-11 7/82 Original 14.3-42 7/82 Original 14.5-12 7/82 Original Figure 14.3-1 7/82 Original 14.5-13 7/82 Original Figure 14.3-2 7/82 Original 14.5-14 7/82 Original Figure 14.3-3 7/82 Original 14.5-15 7/82 Original Figure 14.3-4 7/82 Original 14.5-16 7/82 Original Figure 14.3-5 7/82 Original 14.5-17 7/82 Original Figure 14.3-6 6/85 3 14.5-18 7/82 Original Figure 14.3-7 6/85 3 14.5-19 7/82 Original Figure 14.3-8 6/85 3 14.5-20 7/82 Original Figure 14.3-9 6/85 3 14.5-21 7/82 Original Figure 14.3-10 6/85 3 14.5-22 7/82 Original Figure 14. 3-11 6/85 3 14.5-23 7/82 Original Figure 14. 3-lla 6/85 3 14.5-24 7/82 Original Figure 14.3-llb 6/85 3 14.5-25 7/82 Original Figure 14.3-12 7/82 Original 14.5-26 7/82 Original Figure 14.3-13 7/82 Original 14.5-27 7/82 Original Figure 14.3-14 7/82 Original 14.5-28 7/82 Original Figure 14.3-15 7/82 Original 14.5-29 7/82 Original Figure 14.3-16 7/82 Original 14.5-30 7/82 Original Figure 14.3-17 7/82 Original 14.5-31 6/85 3 Figure 14.3-18 7/82 Original 14.5-32 6/85 3 Figure 14.3-19 7/82 Original 14.5-33 6/85 3 Figure 14.3-20 7/82 Original 14.5-34 7/82 Original Figure 14.3-21 7/82 Original 14.5-35 7/82 Original Figure 14.3-22 7/82 Original 14.5-36 7/82 Original 14.4-1 7/82 Original 14.5-37 6/85 3 14.4-2 7/82 Original 14.5-38 7/82 Blank 14.4-3 7/82 Original 14.5-39 7/82 Original 14.4-4 7/82 Original 14.5-40 7/82 Original 14.4-5 7/82 Original 14.5-41 7/82 Original 14.4-6 6/83 1 14.5-42 6/85 3 14.4-6a 6/83 1 14.5-43 7/82 Original 14.4-7 7/82 Original 14.5-44 7/82 Original 14.4-8 7/82 Original 14.5-45 6/85 3 14.4-9 6/85 3 14.5-46 7/82 Original 14.4-10 7/82 Original 14.5-47 7/82 Original
- 14. 4-11 6/85 3 14.5-48 6/85 3 14.4-12 6/85 3 Figure 14.5-1 7/82 Original 14.4-13 7/82 Original Figure 14.5-2 7/82 Original 14.4-14 7/82 Original Figure 14.5-3 7/82 Original 14.4-15 6/85 3 Figure 14.5-4 7/82 Original 14.4-16 6/85 3 Figure 14.5-5 7/82 Original 14.4-17 6/83 1 Figure 14.5-6 7/82 Original 14.4-18 7/82 Blank Figure 14.5-7 7/82 Original 14.4-19 6/85 3 Figure 14.5-8 7/82 Original 14.4-20 7/82 Original Figure 14.5-9 7/82 Original 14.5-1 8/89 8 Figure 14.5-10 7/82 Original 14.5-2 8/89 8 Figure 14.5-11 7/82 Original
- 14. 5-2a 8/89 8 Figure 14.5-12 7/82 Origina.
14.5-3 7/82 Original Figure 14.5-13 7/82 Origina 14.5-4 7/82 Original Figure 14.5-14 7/82 Original 14.5-5 7/82 Original Figure 14.5-15 7/82 Original 14.5-6 7/82 Original Figure 14.5-16 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-21 LIST OF EFFECTIVE PAGES
~ ll..a.!....e. Revision ~ ll..a.!....e. Revision VOLUME V (continued) 14B-27 7/82 Original 14B-28 7/82 Original Figure 14.5-17 7/82 Original 14B-29 7/82 Original Figure 14.5-18 7/82 Original 14B-30 7/82 OriginaL_
Figure 14.5-19 7/82 Original 14B-31 7/82 Original Figure 14.5-20 7/82 Original 14B-32 7/82 Original Figure 14.5-21 6/85 3 14B-33 7/82 Original Figure 14.5-22 6/85 3 14B-34 7/82 Original Figure 14.5-23 6/85 3 14B-35 7/82 Original Figure 14.5-24 6/85 3 14B-36 7/82 Original Figure 14.5-25 6/85 3 14B-37 7/82 Original Figure 14.2-26 6/85 3 14B-38 7/82 Original 14A-i 7/82 Original 14B-39 7/82 Original 14A-ii 7/82 Blank 14B-40 7/82 Original 14A-iii 7/82 Original 14B-41 7/82 Original 14A-iv 7/82 Original 14B-42 7/82 Original 14A-l 7/82 Original 14B-43 7/82 Original 14A-2 7/82 Blank 14B-44 7/82 Original 14A-3 7/82 Original 14B-45 7/82 Original 14A-4 7/82 Blank 14B-46 7/82 Blank 14A-5 7/82 Original 14B-47 7/82 Original 14A-6 7/82 Original 14B-48 7/82 Original 14A-7 7/82 Original 14B-49 7/82 Original 14A-8 7/82 Blank 14B-50 7/82 Original 14A-9 7/82 Original 14B-51 7/82 Original 14B-i 7/82 Original 14B-52 7/82 Original 14B-ii 7/82 Blank 14B-53 7/82 Original 14B-iii 7/82 Original 14B-54 7/82 Original 14B-iv 7/82 Original 14B-55 7/82 Original 14B-v 7/82 Original 14B-56 7/82 Original 14B-vi 7/82 Original 14B-57 7/82 Original 14B-vii 7/82 Original 14B-58 7/82 Original 14B-l 7/82 Original 14B-59 7/82 Original 14B-2 7/82 Original 14B-60 7/82 Original 14B-3 7/82 Original 14B-61 7/82 Original 14B-4 7/82 Original 14B-62 7/82 Original 14B-5 7/82 Original 14B-63 7/82 Original 14B-6 7/82 Original 14B-64 7/82 Original 14B-7 7/82 Original 14B-65 7/82 Original 14B-8 7/82 Original 14B-66 7/82 Original 14B-9 7/82 Original Figure 14B-l 7/82 Original 14B-10 7/82 Original Figure 14B-2 7/82 Original 14B-ll 7/82 Original Figure 14B-3 7/82 Original 14B-12 7/82 Original Figure 14B-4 7/82 Original 14B-13 7/82 Original Figure 14B-5 7/82 Original 14B-14 7/82 Original Figure 14B-6 7/82 Original 14B-15 7/82 Original Figure 14B-7 7/82 Original 14B-16 7/82 Original Figure 14B-8 7/82 Original 14B-17 7/82 Original Figure 14B-9 7/82 Original 14B-18 7/82 Original Figure 14B-10 7/82 Original 14B-19 7/82 Original Figure 14B-ll 7/82 Original 148-20 7/82 Original Figure 148-12 7/82 Original 14B-21 7/82 Original Figure 148-13 7/82 Original 148-22 7/82 Blank Figure 14B-14 7/82 Original 148-23 7/82 Original Figure 148-15 7/82 Original 14B-24 7/82 Original Figure 148-16 7/82 Original 14B-25 7/82 Original Figure 148-17 7/82 Original 14B-26 7/82 Original Figure 148-18 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-22 LIST OF EFFECTIVE PAGES ~ 12...il....e. Revision ~ J;ta..t.e. Revision VOLUME V (continued) 15.2-22 6/83 1 15.2-23 7/82 Original Figure 14B-19 7/82 Original 15.2-24 7/82 Original Figure 14B-20 7/82 Original 15.2-25 7/82 Original Figure 14B-21 7/82 Original 15.2-26 7/82 Original Figure 14B-22 7/82 Original 15.2-27 7/82 Original Figure 14B-23 6/85 3 15.2-28 7/82 Original 15.2-29 7/82 Original VOLUME VI 15.2-30 7/82 Original 15.2-31 6/90 9 Chai;2tflr 15 15.2-32 7/82 Original 15.2-33 7/82 Original 15-i 7/82 Original 15.2-34 6/86 4 15-ii 7/82 Original 15.2-35 6/86 4 15-iii 7/82 Original 15.3-1 7/82 Original 15-iv 7/82 Original 15.3-2 7/82 Original 15-v 8/89 8 15.3-3 7/82 Original 15-vi 7/82 Original 15.4-1 7/82 Original 15-vii 7/82 Original 15.4-2 7/82 Original 15-viii 6/84 2 15.4-3 7/82 Original 15.1-1 8/89 8 15.4-4 7/82 Original Figure 15.1-1 7/82 Original 15. 4-5 7/82 Original Figure 15.1-2 3/88 7 15.4-6 7/82 Original Figure 15.1-3 6/87 6 15.4-7 7/82 Original Figure 15.1-4 6/87 6 15.4-8 7/82 Original Figure 15.1-5 6/87 6 15.4-9 7/82 Original Figure 15.1-6 6/87 6 15.4-10 7/82 Origina Figure 15.1-7 6/87 6 15.4-11 7/82 Original Figure 15.1-8 6/87 6 15.4-12 7/82 Original Figure 15.1-9 6/87 6 15.4-13 7/82 Original Figure 15.1-10 6/87 6 15.4-14 7/82 Original Figure 15.1-11 6/87 6 15.4-15 7/82 Original Figure 15.1-12 6/87 6 15.4-16 7/82 Original Figure 15.1-13 6/87 6 15.4-17 7/82 Original Figure 15.1-14 6/87 6 15.4-18 7/82 Original Figure 15.1-15 6/87 6 15.4-19 7/82 Original Figure 15.1-16 6/87 6 15.4-20 7/82 Original 15.2-1 7/82 Original 15.4-21 7/82 Original 15.2-2 7/82 Original 15.4-22 7/82 Original 15.2-3 7/82 Original 15.4-23 7/82 Original 15.2-4 7/82 Original 15.4-24 7/82 Original 15.2-5 7/82 Original 15.4-25 7/82 Original 15.2-6 7/82 Original 15.4-26 7/82 Original
- 15. 2-7 7/82 Original 15.4-27 7/82 Original 15.2-8 7/82 Original 15.4-28 7/82 Original 15.2-9 6/86 4 15.4-29 7/82 Original 15.2-10 7/82 Blank 15.4-30 7/82 Original
- 15. 2-11 7/82 Original 15.4-31 7/82 Original 15.2-12 7/82 Original 15.4-32 7/82 Original
- 15. 2-13 7/82 Original 15.4-33 7/82 Original 15.2-14 7/82 Original 15.4-34 7/82 Original 15.2-15 7/82 Original 15.4-35 7/82 Original 15.2-16 7/82 Original 15.4-36 7/82 Original 15.2-17 7/82 Original 15.4-37 7/82 Origina.
15.2-18 7/82 Original 15.4-38 7/82 Origina 15.2-19 7/82 Original 15.4-39 7/82 Original 15.2-20 7/82 Original 15.4-40 7/82 Original 15.2-21 7/82 Original 15.4-41 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-23 LIST OF EFFECTIVE PAGES
~ ~ Revision ~ ~ Revision VOLUME VI (continued) 15.5-47 8/89 8 15.5-48 8/89 8 15.4-42 7/82 Original 15.5-49 7/82 Original 15.4-43 7/82 Original 15.5-50 7/82 Original 15.4-44 7/82 Original 15.5-51 7/82 Original 15.4-45 7/82 Original 15.5-52 7/82 Original Figure 15.4-1 7/82 Original 15.5-53 7/82 Original Figure 15.4-2 7/82 Original 15.5-54 7/82 Original Figure 15.4-3 15.5-55 7/82 Original (Sheet 1) 7/82 Original 15.5-56 7/82 Original (Sheet 2) 7/82 Original 15.5-57 7/82 Original (Sheet 3) 7/82 Original 15.5-58 7/82 Blank 15.5-1 7/82 Original 15.5-59 7/82 Original 15.5-2 7/82 Original 15.5-60 7/82 Original 15.5-3 7/82 Original 15.5-61 7/82 Original 15.5-4 7/82 Original 15.5-62 7/82 Original 15.5-5 7/82 Original 15.5-63 7/82 Original 15.5-6 7/82 Original Figure 15.5-1 7/82 Original 15.5-7 7/82 Original Figure 15.5-2 7/82 Original 15.5-8 7/82 Original Figure 15.5-3 7/82 Original 15.5-9 7/82 Original Figure 15.5-4 7/82 Original 15.5-10 7/82 Original Figure 15.5-5 7/82 Original
- 15. 5-11 7/82 Original Figure 15.5-6 7/82 Original 15.5-12 7/82 Original Figure 15.5-7 7/82 Original 5.5-13 7/82 Original Figure 15.5-8 7/82 Original 15.5-14 7/82 Original Figure 15.5-9 7/82 Original 15.5-15 7/82 Original Figure 15.5-10 7/82 Original 15.5-16 7/82 Original Figure 15.5-11 7/82 Original 15.5-17 7/82 Original Figure 15.5-12 7/82 Original 15.5-18 7/82 Original Figure 15.5-13 7/82 Original 15.5-19 7/82 Original Figure 15.5-14 7/82 Original 15.5-20 7/82 Original Figure 15.5-15 7/82 Original 15.5-21 7/82 Original Figure 15.5-16 7/82 Original 15.5-22 7/82 Original Figure 15.5-17 6/84 2 15.5-23 7/82 Original Figure 15.5-18 6/84 2 15.5-24 7/82 Original Figure 15.5-19 6/84 2 15.5-25 7/82 Original 15.6-1 7/82 Original 15.5-26 8/89 8 15.6-2 7/82 Original 15.5-27 7/82 Original 15.6-3 7/82 Original 15.5-28 7/82 Original 15.6-4 8/89 8 15.5-29 7/82 Original 15.6-4a 8/89 8 15.5-30 7/82 Original 15.6-4b 8/89 8 15.5-31 7/82 Original 15.6-5 8/89 8 15.5-32 7/82 Original 15.6-5a 8/89 8 15.5-33 7/82 Original 15.6-6 8/89 8 15.5-34 7/82 Original 15.6-6a 8/89 8 15.5-35 7/82 Original 15.6-7 8/89 8 15.5-36 7/82 Original 15.6-8 8/89 8 15.5-37 7/82 Original 15.6-9 8/89 8 15.5-38 7/82 Original 15.6-10 7/82 Original 15.5-39 7/82 Original 15. 6-11 8/89 8 15.5-40 7/82 Original 15.6-12 8/89 8 15.5-41 7/82 Original 15.6-13 8/89 8 15.5-42 7/82 Original 15.6-14 8/89 8 15.5-43 7/82 Original 15.6-15 7/82 Original 15.5-44 7/82 Original 15.6-16 7/82 Original 15.5-45 7/82 Original 15.6-17 7/82 Original 15.5-46 8/89 8 Figure 15.6-1 7/82 Original
REVISION 9 6/90 SPS UFSAR LEP-24 LIST OF EFFECTIVE PAGES ~ ~ Revision ~ Revision VOLUME VI (continued) 15A-40 7/82 Original 15A-41 7/82 Original Figure 15.6-2 8/89 8 15A-42 8/89 8 Figure 15.6-3 3/88 7 15A-43 8/89 8 Figure 15.6-4 3/88 7 15A-44 8/89 8 15.7-1 7/82 Original Figure 15A-l 7/82 Original 15.7-2 6/83 1 Figure 15A-2 7/82 Original 15.7-3 6/83 1 Figure 15A-3 7/82 Original 15A-i 7/82 Original Figure 15A-4 7/82 Original 15A-ii 7/82 Blank Figure 15A-5 7/82 Original 15A-iii 7/82 Original Figure 15A-6 7/82 Original 15A-iv 7/82 Original Figure 15A-7 7/82 Original 15A-v 8/89 8 Figure 15A-8 7/82 Original 15A-vi 7/82 Original 15A-l 7/82 Original Ch1n;2:ter 16 15A-2 7/82 Blank 15A-3 7/82 Original 16-1 7/82 Original 15A-4 7/82 Original 16-2 7/82 Original 15A-5 8/89 8 15A-6 8/89 8 Chai;i:ter 17 15A-7 7/82 Original 15A-8 7/82 Blank 17-1 6/83 1 15A-9 7/82 Original 15A-10 7/82 Original 15A-ll 7/82 Original 15A-12 8/89 8 15A-12a 8/89 8 15A-12b 8/89 8 15A-13 8/89 8 15A-14 8/89 8 15A-15 8/89 8 15A-16 8/89 8 15A-16a 8/89 8 15A-16b 8/89 8 15A-17 8/89 8 15A-18 8/89 8 15A-19 7/82 Original 15A-20 8/89 8 15A-21 7/82 Original 15A-22 8/89 8 15A-23 8/89 8 15A-24 8/89 8 15A-25 7/82 Original 15A-26 7/82 Blank 15A-27 8/89 8 15A-28 7/82 Original 15A-29 8/89 8 15A-30 7/82 Original 15A-31 7/82 Original 15A-32 8/89 8 15A-33 8/89 8 15A-33a 8/89 8 15A-34 8/89 8 15A-35 7/82 Original 15A-36 7/82 Original 15A-37 7/82 Original 15A-38 7/82 Original 15A-39 7/82 Original __J
SPS UFSAR TOC-ii SURRY POWER STATION UPDATED FINAL SAFETY ANALYSIS REPORT TABLE OF CONTENTS (continued) Section Title VOLUME II (continued) 5.4 CONTAINMENT DESIGN EVALUATION. . ...... 5.4-1 5.5 CONTAINMENT TESTS AND INSPECTIONS .... ...... 5.5-1 6 ENGINEERED SAFEGUARDS ..... .. 6.1-1 6.1 GENERAL DESCRIPTION. . . ... . . .... 6.1-1 6.2 SAFETY INJECTION SYSTEM . 6.2-1 6.3 CONSEQUENCE-LIMITING SAFEGUARDS . 6.3-1 7 INSTRUMENTATION AND CONTROL. .... ... 7.1-1 7.1 GENERAL DESIGN CRITERIA. .... .. ..... 7.1-1 7.2 REACTOR PROTECTION SYSTEM. . ..... . 7.2-1 7.3 REACTOR CONTROL SYSTEM. ... .. .. 7.3-1 7.4 NUCLEAR INSTRUMENTATION SYSTEM. . 7.4-1
- 7.5 7.6 7.7 7.8 7.9 7.10 ENGINEERED SAFEGUARDS INCORE INSTRUMENTATION.
OPERATING CONTROL STATIONS. AUTOMATIC LOAD CONTROL SYSTEM. COMPUTER SYSTEM. .... SUBCOOLING MONITOR SYSTEM. 7.5-1 7.6-1 7.7-1 7.8-1 7.9-1 7 .10-1 VOLUME III 8 ELECTRICAL SYSTEMS. ...... 8.1-1 8 .1 GENERAL DESCRIPTION AND
SUMMARY
. ...... 8.1-1 8.2 DESIGN BASES. .... . . ... . ... 8.2-1 8.3 SYSTEM INTERCONNECTIONS . ......... 8.3-1 8.4 STATION SERVICE SYSTEMS . .... 8.4-1 8.5 EMERGENCY POWER SYSTEM. 8.5-1 8.6 TESTS AND INSPECTIONS .. ..... 8.6-1 9 AUXILIARY AND EMERGENCY SYSTEMS ... ..... 9.1-1 9.1 CHEMICAL AND VOLUME CONTROL SYSTEM. . ... . 9.1-1 9.2 BORON RECOVERY SYSTEM . . . . 9.2-1 9.3 RESIDUAL HEAT REMOVAL SYSTEM. .... 9.3-1 9.4 COMPONENT COOLING SYSTEMS . . . . . 9.4-1 9.5 FUEL POOL COOLING SYSTEM. . . . . . ..... 9.5-1 9.6 SAMPLING SYSTEM. . . . . . .... . .. . 9.6-1
SPS UFSAR TOC-ii SURRY POWER STATION UPDATED FINAL SAFETY ANALYSIS REPORT TABLE OF CONTENTS (continued) Section Title VOLUME II (continued) 5.4 CONTAINMENT DESIGN EVALUATION. . .... ... 5.4-1 5.5 CONTAINMENT TESTS AND INSPECTIONS . ..... 5.5-1 6 ENGINEERED SAFEGUARDS ................ 6.1-1 6.1 GENERAL DESCRIPTION . . . . . . .. 6.1-1 6.2 SAFETY INJECTION SYSTEM. ....... ..... 6.2-1 6.3 CONSEQUENCE-LIMITING SAFEGUARDS . ... 6.3-1 7 INSTRUMENTATION AND CONTROL ............. 7.1-1 7.1 GENERAL DESIGN CRITERIA. ....... 7.1-1 7.2 REACTOR PROTECTION SYSTEM. .. ... 7.2-1 7.3 REACTOR CONTROL SYSTEM. .... ... 7.3-1 7.4 NUCLEAR INSTRUMENTATION SYSTEM. .. .. ... 7.4-1
- 7.5 7.6 7.7 7.8 7.9 7 .10 ENGINEERED SAFEGUARDS INCORE INSTRUMENTATION.
OPERATING CONTROL STATIONS. AUTOMATIC LOAD CONTROL SYSTEM. COMPUTER SYSTEM. .... SUBCOOLING MONITOR SYSTEM. .. 7.5-1 7.6-1 7.7-1 7.8-1 7.9-1 7.10-1 VOLUME III 8 ELECTRICAL SYSTEMS. ...... .. .... .. 8.1-1 8.1 GENERAL DESCRIPTION AND
SUMMARY
. . . . . .. .... 8.1-1 8.2 DESIGN BASES. .............. . . .. 8.2-1 8.3 SYSTEM INTERCONNECTIONS .... .. . 8.3-1 8.4 STATION SERVICE SYSTEMS . ...... 8.4-1 8.5 EMERGENCY POWER SYSTEM. . .... . .. 8.5-1 8.6 TESTS AND INSPECTIONS ... ... 8.6-1 9 AUXILIARY AND EMERGENCY SYSTEMS . .
. . . . .... 9.1-1 9 .1 CHEMICAL AND VOLUME CONTROL SYSTEM. . . . . . ... 9. 1-'i 9.2 BORON RECOVERY SYSTEM ..... . . . 9.2-1 9.3 RESIDUAL HEAT REMOVAL SYSTEM. .... .
. ...... 9.3-1 9.4 COMPONENT COOLING SYSTEMS . . . .. 9.4-1 9.5 FUEL POOL COOLING SYSTEM. . . . ... . .... 9.5-1 9.6 SAMPLING SYSTEM ........ .. 9.6-1
SPS UFSAR TOC-ii SURRY POWER STATION UPDATED FINAL SAFETY ANALYSIS REPORT TABLE OF CONTENTS (continued) Section Title VOLUME II (continued) 5.4 CONTAINMENT DESIGN EVALUATION. . ..... 5.4-1 5.5 CONTAINMENT TESTS AND INSPECTIONS .... ..... 5.5-1 6 ENGINEERED SAFEGUARDS. .. .... 6.1-1 6.1 GENERAL DESCRIPTION. .. ... 6.1-1 6.2 SAFETY INJECTION SYSTEM. ......... 6.2-1 6.3 CONSEQUENCE-LIMITING SAFEGUARDS . ... .. 6.3-1 7 INSTRUMENTATION AND CONTROL ....... .. 7.1-1 7.1 GENERAL DESIGN CRITERIA. ...... .. 7.1-1 7.2 REACTOR PROTECTION SYSTEM. ...... .. 7.2-1 7.3 REACTOR CONTROL SYSTEM. ..... 7.3-1 7.4 NUCLEAR INSTRUMENTATION SYSTEM. .. 7.4-1
- 7.5 7.6 7.7 7.8 7.9 7 .10 ENGINEERED SAFEGUARDS
- INCORE INSTRUMENTATION. . ..
OPERATING CONTROL STATIONS. AUTOMATIC LOAD CONTROL SYSTEM. COMPUTER SYSTEM. .... SUBCOOLING MONITOR SYSTEM. .. 7.5-1 7.6-1 7.7-1 7.8-1 7.9-1 7.10-1 VOLUME III 8 ELECTRICAL SYSTEMS. . ... .. ........ 8.1-1 8.1 GENERAL DESCRIPTION AND
SUMMARY
...... 8.1-1 8.2 DESIGN BASES. ........... 8.2-1 8.3 SYSTEM INTERCONNECTIONS .. .... 8.3-1 8.4 STATION SERVICE SYSTEMS . . . . . . 8.4-1 8.5 EMERGENCY POWER SYSTEM. . . . . . .. . . 8.5-1 8,6 TESTS AND INSPECTIONS .. ........ 8.6-1 9 AUXILIARY AND EMERGENCY SYSTEMS ..... 9.1-1 9 .1 CHEMICAL AND VOLUME CONTROL SYSTEM. .... 9.1-1 9.2 9.3 BORON RECOVERY SYSTEM. . . . ..
RESIDUAL HEAT REMOVAL SYSTEM.
.. .. 9.2-1 9.3-1 9.4 COMPONENT COOLING SYSTEMS . ...... 9.4-1 9.5 FUEL POOL COOLING SYSTEM. ...... ... 9.5-1 9.6 SAMPLING SYSTEM, .... .... 9.6-1
SPS UFSAR TOC-ii SURRY POWER STATION UPDATED FINAL SAFETY ANALYSIS REPORT TABLE OF CONTENTS (continued) Section Title VOLUME II (continued) 5.4 CONTAINMENT DESIGN EVALUATION. .. .... 5.4-1 5.5 CONTAINMENT TESTS AND INSPECTIONS. . ....... 5.5-1 6 ENGINEERED SAFEGUARDS 6.1-1 6.1 GENERAL DESCRIPTION. ... ....... ........ 6.1-1 6.2 SAFETY INJECTION SYSTEM. ... 6.2-1 6.3 CONSEQUENCE-LIMITING SAFEGUARDS . ......... 6.3-1 7 INSTRUMENTATION AND CONTROL. .... 7.1-1 7.1 GENERAL DESIGN CRITERIA, . . . .. 7.1-1 7.2 REACTOR PROTECTION SYSTEM. . . . . . ...... 7.2-1 7.3 REACTOR CONTROL SYSTEM. ........ ... 7.3-1 7.4 NUCLEAR INSTRUMENTATION SYSTEM. ........ 7.4-1
- 7.5 7.6 7.7 7.8 7.9 7 .10 ENGINEERED SAFEGUARDS . ......
INCORE INSTRUMENTATION. . . . . OPERATING CONTROL STATIONS. . AUTOMATIC LOAD CONTROL SYSTEM. COMPUTER SYSTEM. . . . . SUBCOOLING MONITOR SYSTEM * . . 7.5-1 7.6-1 7.7-1 7.8-1 7.9-1 7 .10-1 VOLUME III 8 ELECTRICAL SYSTEMS. .... . . ..... . . . .. 8.1-1 8.1 8.2 GENERAL DESCRIPTION AND
SUMMARY
DESIGN BASES. ...... 8.1-1 8.2-1 8.3 SYSTEM INTERCONNECTIONS . .... 8.3-1 8.4 STATION SERVICE SYSTEMS . . ... 8.4-1 8.5 EMERGENCY POWER SYSTEM. . .... 8.5-1 8.6 TESTS AND INSPECTIONS .. ... . ... 8.6-1 9 AUXILIARY AND EMERGENCY SYSTEMS . . . .. 9.1-1 9.1 CHEMICAL AND VOLUME CONTROL SYSTEM. . . . ... 9.1-1 9.2 BORON RECOVERY SYSTEM. . . . . .... 9.2-1 9.3 RESIDUAL HEAT REMOVAL SYSTEM. . 9.3-1 9.4 COMPONENT COOLING SYSTEMS . . . . 9.4-1 9.5 FUEL POOL COOLING SYSTEM. .... . ..... 9.5-1 9.6 SAMPLING SYSTEM. . . . . . . . . . 9.6-1
SPS UFSAR
- 1.4-7 1.4.7 SUPPRESSION OF POWER OSCILLATIONS The design of. the reactor core with its* related controls and protection systems ensures that power oscillations, the magnitude of which could cause damage in excess of acceptable fuel damage limits, are not possible or can be readily suppressed.
The design of the reactor core and related protection systems ensures that power oscillations that could cause fuel damage in excess of acceptable limits are not possible or can be readily suppressed. The potential for possible spatial oscillations of power distribution for this core are reviewed as part of the core stability evaluation described in Section 1.6. Ex-core instrumentation is provided to obtain necessary information concerning axial and azimuthal power distributions. This instrumentation is adequate to enable the operator to monitor and control xenon-induced oscillations. Based on the deviations detected by the long ion chambers, provisions in the reactor control and protection system reduce trip setpoints and if necessary initiate load runback to maintain margin to departure from nucleate boiling as a result of these potential oscillations in power distribution. Incore instrumentation is used to periodically calibrate and verify the information provided by the ex-core instrumentation. The geueral conclusion based on experimental results from SENA and San Onofre is that the ex-core instruments do give an accurate indication of the fact that power redistribution is taking place. This has been confirmed by a comparison with incore instrumentation results. The temperature coefficient in the power operating range was maintained zero or negative by the inclusion of burnable poison shims in the initial core loading.. Burnable poison shims can also be used in subsequent core loadings if necessary. The reference chapters are as follows: Title Chapter Reactor 3 Instrumentation and Control 7
SPS UFSAR 1.4-8 1.4.8 OVERALL POWER COEFFICIENT The reactor is designed so that the overall power coefficient in the power operating range is not positive. The overall power coefficient is negative under normal operating conditions throughout core life. The reference chapter is as follows: Title Chapter. Reactor 3 1.4.9 REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary is designed, fabricated, and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime. The reactor coolant system, in conjunction with its control and protective provisions, is designed to accommodate the system pressures and temperatures attained under all expected modes of unit operation or anticipated system interactions, and to remain within the applicable code stress limits. The fabrication of the components that constitute the pressure-retaining boundary of the reactor coolant system is carried out in strict accordance with the applicable codes. In addition, there are areas where equipment specifications for reactor coolant system components are more restrictive than applicable codes. The materials of construction of the pressure-retaining boundary of the reactor coolant system are protected by the control of coolant chemistry so as to prevent corrosion phenomena that might otherwise reduce the system structural integrity during its service lifetime.
SPS UFSAR 3.3-11
- power, the nuclear heat flux hot-channel factor, FN, specified in Table 3.3-1, line 20.
specified in Table 3.3-1, line 21. q was established as For the hottest channel at rated power, the nuclear enthalpy rise hot-channel factors, F~H' was established as These values are specific to the initial cycle. Power capability of a PWR core is determined largely by consideration of the power distribution and its interrelationship to limiting conditions involving
- 1. The linear power density.
- 2. The fuel cladding integrity.
- 3. The enthalpy rise of the coolant.
To determine the core power capability, local as well as gross core neutron flux distributions have been determined for various operating conditions at different times in.core life .
- The presence of control rods, burnable poison rods, and chemical shim concentration all play significant roles in establishing .the fission power distribution, in addition to the influence of thermal/hydraulic and temperature feedback considerations. The computer programs used to determine neutron flux distributions include a model to simulate nonuniform water (and chemical shim) density distributions.
Thermal/hydraulic feedback considerations are especially important late in cycle life, when the magnitude of the flux redistribution and reactivity change with change in core power or control rod assembly movement are strongly influenced by enthalpy rise up the core and by the fuel burnup distribution. Consequently, extensive X-Y and Z power distribution analyses have been performed to evaluate fission power distributions. Typical X-Y power distributions are presented in Figures 3.3-2 and 3.3-3 to illustrate the combined effect of a control rod assembly group upon assembly average power density at different elevations in the core. Incore instrumentation is employed to evaluate the core power distributions throughout core lifetime to ensure that the thermal design criteria are met.
SPS UFSAR 3.3-12 The ex-core nuclear instrumentation system supplies the necessary information for the operator to control the core power distribution within the limits established for the protection system design. This information consists of a two-pen recorder for each long ion chamber, which displays the upper and lower ion chamber currents, and an indicator that gives the difference in these two currents for each long ion chamber. These ion chamber currents to the recorders and indicators are calibrated against incore power distribution obtained from the movable detector system generated in the adjacent section of the core. This essentially divides the core into eight sections, four in the upper half and four in the lower half. The relationship between core power distribution and ex-core nuclear instrumentation readings was established during the startup testing program (Chapter 13). Incore flux measurements were made for reactor power in the range of 25 to 100%. These measurements, together with long ion chamber currents, were processed to yield the relationships between core average axial power generation, axial peaking* factor, and axial offset as indicated by the ex-core nuclear instrumentation. These relationships can be checked during operation to assess the effect of core burnup on the s*ensitivity between incore power distribution and ex-core readings.
- The reactor core may be subject to axial xenon oscillations at the end of a fuel cycle life. The axial instability is due principally to the negative moderator temperature coefficient of reactivity that exists at end of life.
Since the moderator coefficient at beginning of life is small, stability against axial oscillations is greatly increased at beginning of life. Consequently, stability margin experiments would not be informative at beginning of life. A more detailed discussion of the background, analytical, and experimental data which form the basis for this approach is given in 4 WCAP-7208.
REVISION 1 6/83 SPS UFSAR 4-:iii e CHAPTER 4: REACTOR COOLANT SYSTEM TABLE OF CONTENTS (continued) Section Title Page 4.2.9.5 Pump Power-Differential Pressure. 4.2-30 4.2.9.6 Experience.
- 4.2-32 1 4.2.9.7 Low Flow Trip Setpoint ** 4.2-33 4.2.10 Loose Parts Monitoring System 4.2.33 4.2.10.1 Design Criteria . . . 4.2-34 4.2.10.2 System Description. 4.2-34 4.2 References . . * . * . . 4.2-36
- 4. 3 SYSTEM DESIGN EVALUATION. . . * * * . . . * . . . . . . . . *
- 4.3-1
- 4.3.1 4.3.1.1 4.3.1.2 4.3.1.3 Safety Factors . .
Reactor Vessel * . Steam Generators .
- Steam-Generator Tube Vibration Considerations .
4.3-1 4.3-1 4.3-7 4.3-9 4.3.1.4 Piping Quality Assurance . . . . * . 4.3-12 4.3.2 Reliance on Interconnected Systems. 4.3-13 4.3.3 System Integrity. 4. 3-13 4.3.3.1 Reactor Coolant Pump Flywheel Integrity . . 4.3-14 4.3.4 Overpressure Protection. 4.3-16 4.3.4.1 Operational Conditions. 4.3-17 4.3.4.2 Air Supply * . 4.3.5 System Accident Potential 4.3.6 Redundancy . . 4.3-18
- 4. 3 Refere1.1ces. . 4. 3-20 11 4.4 TESTS AND INSPECTIONS . . . . . . . . . . . . * . . . . . . . . 4.4-1 4.4.1 Reactor Coolant System Inspection . . . . . . . * * . .
- 4.4-1 4.4.1.1 Nondestructive Inspection of Materials and Components . 4.4-1
REVISION 1 6/83 SPS UFSAR 4-iv CHAPTER 4: REACTOR COOLANT SYSTEM TABLE OF CONTENTS (continued) Section Title 4.4.1.2 Electroslag Weld Quality Assurance. 4.4-3 4.4.1.2.1 Piping Elbows
- 4.4-3 4.4.1.2.2 Reactor Coolant Pump Casings * *
- 4.4-4 4.4.1.3 Reactor Coolant System Field Erection and Welding . 0 4.4-6 4.4.1.4 Reactor Coolant System Cleanliness. . .... 4.4-7 4.4.1.5 Reactor Coolant System Testing Following Opening. 4.4-7 4.4.1.6 Inservice Inspection Capability .......... 4.4-7
REVISION 1 6/83 SPS UFSAR 4-v CHAPTER 4: REACTOR COOLANT SYSTEM LIST OF TABLES Table Title 4.1-1 Reactor Coolant System Design Pressures . 4.1-21 4.1-2 Reactor Vessel Design Data . . 4.1-22 4.1-3 Pressurizer and Pressurizer Relief Tank Design Data * . . . 4.1-23 4.1-4 Steam-Generator Design Data. 4.1-25 4.1-5 Reactor Coolant Pump Design Data. 4.1-26 4.1-6 Reactor Coolant Piping Design Data . . 4.1-27 4.1-7 Loop Stop Valves . . 4.1-27 4.1-8 Thermal and Loading Cycles. 4.1-28 4.1-9 Reactor Coolant System - Code Requirements. 4.1-29 4.1-10 Capsule Specimen Data, First Four of Eight
- 4.1-11 4.1-12 Capsules . . . . . . . . * . . * .
Capsule Specimen Data, Second Four of Eight Capsules . . . . . .
- Relative Exposures of Capsules and Reactor 4.1-30 4.1-31 Vessel Wall * . . . . 4.1-31 4.1-13 Reactor Pressure Vessel NDTT Data, Surry Unit 1. 4.1-32 4.2-1 Reactor Coolant System Materials of Construction. . 4.2-37 4.2-2 Reactor Coolant System Water Chemistry Specification. 4.2-39 4.2-3 Radiation-Induced Increase in Transition Temperature for A302B Steel. 4.2-40 1 4.2-4 Reactor Coolant System Flow Data from San Onofre Unit 1 . 4.2-40 4.3-1 Summary of Estimated Primary-Plus-Secondary Stress Intensity for Components of the Reactor Vessel . . . . . . * . . 4.3-23
REVISION 1 6/83 SPS UFSAR 4-vi CHAPTER 4: REACTOR COOLANT SYSTEM e LIST OF TABLES (continued) Table Title 4.3-2 Summary of Estimated Cumulative Fatigue Usage Factors for Components of the Reactor Vessel. 4.3-23 4.3-3 Ratio of Allowable Stresses to Computed Stresses 1 for a Steam-Generator Tube Sheet Pressure Differential of 2485 psig * . . . . . . . 4.3-24 ..4.4-1 Reactor Coolant System Quality Assurance Program 4.4-9 e __J
SPS UFSAR 4.1-9
- d. Prior to placing the vessel in service, an ultrasonic examination of pressure boundary welds was performed in accordance with the ASME Code for inservice inspection of nuclear reactor coolant systems.
- e. In addition to the ASME Code-required third-party inspector, Westinghouse provided full-time quality assurance coverage in the Babcock & Wilcox Co. shop, and also in the Rotterdam Dockyard (RDM) shop during the fabrication of the Surry vessels.
- f. In addition to the quality assurance coverage provided by Westinghouse, Vepco and the architect-engineer, Stone & Webster, performed regular quality assurance audits of reactor vessel fabrication as performed by RDM.
4.1.7 IRRADIATION SURVEILLANCE PROGRAM
- 4.1.7.1 General Description In the surveillance programs, the evaluation of the radiation damage is based on pre-irradiation testing. of Charpy V-notch and tensile specimens, and post-irradiation testing of Charpy V-notch, tensile, and wedge opening loading (WOL) fracture mechanics test specimens. These programs are directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels based on the transition temperature approach and the fracture mechanics approach, and is in accordance with ASTM-E-185, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors."
Steel alloy specimens of a comparatively low melting point are included as thermal control specimens. They provide indication that the area of surveil-lance has exceeded a given temperature. The reactor vessel surveillance program uses eight specimen capsules, more than twice the minimum number recommended by ASTM-E-185. The capsules are located about 3 in. from the vessel wall, directly opposite the center portion of the core. Elevation and plan views showing the location and dimensional spacing of the capsules with relation to the core, thermal shield,
SPS UFSAR 4 .1-10 and vessel and weld seams are shown in Figures 4.1-1 and 4.1-2, respectively. The capsules can be removed when the vessel head is removed, and can be replaced when the internals are removed. The capsules contain similar material (steel) to that found in the Surry reactor vessels. These specimens resemble material from the shell plates located in the core region of the reactor and associated weld metal and heat-affected-zone metal. (As part of the surveillance program, a report of the residual elements in weight percent to the nearest 0.01% will be made for surveillance material base metals and as-deposited weld metal.) In addition, 32 correlation monitors made from fully documented specimens of SA302 Grade B material obtained through Subcommittee II of ASTM Committee ElO, Radioisotopes and Radiation Effects, are inserted in the capsules. The eight capsules contain approximately 32 tensile specimens, 240 Charpy V-notch specimens (which include weld metal and heat-affected-zone material), and 40 ~OL specimens. The dosimeters permit evaluation of the flux seen by the specimens and vessel wall. In addition, thermal monitors made of low-melting alloys are included to monitor temperature of the specimens. thermal conductivity. The specimens are enclosed in a tight-fitting stainless steel sheath to prevent corrosion and ensure good The complete capsule is helium leak tested. Vessel material sufficient for at least two capsules will be kept in storage should the need arise for additional replacement test capsules in the program_. The anticipated degree to which the specimens will perturb the fast neutron flux and energy distribution will be considered in the evaluation of the surveillance specimen data. Verification and possible readjustment of the calculated wall exposure will be made by use of data on all capsules withdrawn. Specimen data for four capsules are given in Table 4.1-10. Speciman data for an additional four capsules are given in Table 4.1-11. The tentative schedule for removal of capsules is as follows: Capsule 1 Replacement of first region of core
SPS UFSAR 4.1-11
- Capsule 2 Capsule 3 5 years 10 years Capsule 4 20* years Capsules 5, 6, 7, and 8 Extra capsules for complementary or duplicate testing Irradiation of the specimens is higher than the irradiation of the adjacent vessel wall because they are closer to the core than the vessel itself. Since these specimens experience higher irradiation and are actual samples from the materials used in the vessel, the NDTT measurements are representative of the vessel at a later time in life. Data from fracture toughness samples (WOL) are expected to provide additional information for use in determining allowable stresses for irradiated material.
- The calculated maximum fast-neutron exposure (nvt) at the vessel wall is 4.1 x 10 19 nvt greater than 1 MeV. The reactor vessel surveillance capsules are located at 15, 25, 35, and 45 degrees, as shown in Figure 4.1-2.
relative exposures of the capsules and the adjacent vessel wall, and the The vessel maximum, are listed in Table 4.1-12. The observed shifts in NDTT of the core region materials with irradiation will be used to confirm the calculated limits so that full operating pressure is not obtained until the affected vessel material is above the now higher design transition temperature in the ductile material region. The pressure during startup and shutdown at the temperature below NDTT is maintained below the threshold of concern for safe operation. The design transition temperature is a minimum of NDTT plus 60°F, and 9ictates the procedures to be followed in the hydrostatic test and in station operations to avoid excessive cold stress. The value of the design transition temperature is increased during the life of the station as required by the expected shift in the NDTT, and as confirmed by the experimental data obtained
SPS UFSAR 4.1-12 I from irradiated specimens of reactor vessel materiale; during the station lifetime. To define permissible operating conditions below design transition temperature, a pressure range is ~stablished which is bounded by a lower limit for pump operation and an upper iimit that satisfies reactor vessel stress criteria, to allow for thermal stresses during heatup or cooldown as a function of rate of change of coolant temperature. Since the normal operating temperature of the reactor vessel is well above the maximum expected design transition temperature, brittle fracture during normal operation is not considered to be a credible mode of failure. 4.1.7.2. Measurement of Integrated Fast Neutron (E Greater than 1.0 MeV) Flux at the Irradiation Samples Information on the spectrum of neutron fluxes at the location of the 1 irradiation samples is obtained from the multigroup diffusion code PlMG. Dosimeters including U-238, Np-237, Co-Al, Cu, Ni, Cd, shielded Co-Al, and Fe from specimens are contained in the capsule assemblies. The procedure for measurement of fast neutron flux by the 54 Fe (n,p) 54 Mn reaction is described below. The measurement technique for the other dosimeters, which are sensitive to different portions of the neutron spectrum, is similar. 54 The Mn product of this reaction has *a half life of 314 days and emits gamma rays of O. 84-MeV energy, which are easily detected using a Na! 54 scintillator. In irradiated steel samples, chemical separation of the Mn may be performed to ensure freedom from interfering a~tivities. This 54 separation is simple and very effective, yielding sources of very pure Mn activity. In some samples all the interferences may be corrected for the gamma spectrometric methods without any chemical separation. The count data 54 are used to give the specific activity of Mn per. gram of iron. Because of 54 the relatively long half life of Mn, the flux may be calculated for irradiation periods up to about 2 years. Beyond this time the dosimeter begins to reflect the later stages of irradiation. Calculation of total ~ose I _J
SPS UFSAR 4.1-15
- 4 .1. 7. 3 Calculation of Integrated Fast Neutron (E Greater than at the Irradiation Samples LO MeV) Flux The method to be described here is an approximation to the ideal three-
- t dimensional neutron transport . solution, but correlations between its predictions and measurements on samples irradiated in the Yankee and Saxton cores indicate good agreement.
The spectrum of neutron fluxes at the capsule location is obtained from 1 the one-dimensional multigroup diffusion code PlMG for the array of annular
*. shields surrounding a cylindrical core of infinite height. The cylindrical core has a cross-sectional area equal to that of the actual core. The radial source distribution chosen for the core represents the expected average over the life of the station. The magnitude of the neutron fluxes generated by the PlMG code, which does not treat transport effects, is adjusted by application of a spatial correction factor. This factor is the ratio of the fast neutron 3
dose rate calculated by the SPIC-1 code for an all-water medium surrounding a
- typical Westinghouse pressurized-water reactor to the fast neutron dose rate obtained by PlMG in the identical geometry. The SPIC-1 fast neutron dose rate calculation uses an empirical fast neutron attenuation kernel in the form of a linear combination of single exponentials that are fitted to the experimental fast neutron dose rate distribution in pure water.
The axial and azimuthal variations of neutron flux at the capsule location are determined separately. The axial distribution is expressed as 4 the ratio of the normalized results of two calculations using PDQ4, a two-dimensional, four-group (r,z) diffusion code. In the first of these, an infinitely high equivalent cylindrical core, with a fission neutron source strength SI per unit height, is surrounded by an all-water medium containing the capsule location. In the second, the finite height is surrounded by an all-water medium. The fixed source option of the PDQ4 code is selected so
*:**. that the axial variation of source strength in the core represents a good ;..'.approximation to the average over the core life. The radial distribution is identical to that chosen for.PlMG. The ratio
SPS UFSAR 4.1-16 'where subscripts F and I denote finite and infinite core representations, respectively, is the required axial correction term. The azimuthal distributions of neutron fluxes at the sample location are derived from a comparison of the results of the two-dimensional, four-group 5 6 (x,y) code PDQ3 and the one-dimensional, four-group diffusion program AIM-5. In the PDQ3 calculation, the core, whose shape can be specified exactly, is surrounded by an all-water medium. The radial and azimuthal source distributions in the core are both reasonable approximations to the averages expected during the core life. The radial source distribution in the AIM-5 calculation, in which the equivalent cylindrical core is surrounded by an all-water medium, is identical to that chosen for PlMG. The product of the spatially corrected PlMG results, the axial correction term, and the azimuthal correction term defines the three-dimensional variation of neutron flux at the sample locations. The technique described above overpredicts the Saxton-measured values by 30% and the Yankee-measured values by 14%. In both reactors the measured results were averages for a set of specimens in a capsule located outside the thermal shield opposite a core corner. It also gives excellent agreement with measured data reported for the PM2A reactor. Based on the above evidence, it is concluded that the PlMG calculation, corrected as described, is conservative by approximately 20%. 4.1.7.4 Measurement of the Initial NDTT of the Reactor Pressure Vessel Baseplate and Forgings Material The unirradiated or initial NDTT of pressure vessel baseplate and forgings material is presently measured by two methods. These methods are the drop weight test per ASTM E208 and the Charpy V-notch impact test (Type A) per ASTM E23. Surry Unit 1 reactor pressure vessel NDTT data are presented in Table 4. 1-13. The NDTT is defined in ASTM E208 as "the temperature at which a specimen is broken in a series of tests in which duplicate no-break performance occurs
SPS UFSAR 4.1-17 - at 10°F higher temperature." is a certain "fixed" value. Using the Charpy V-notch test, the NDTT is defined as the temperature at which the. energy required to break the spe*cimen For SA 533B:c1ass 1 and A 508 Class 2 steel, the ASME III Table N-421 specifies an energy value of 30 ft-lb. This value is based on a correlation with the drop weight test and is referred to as the "30 ft-lb-fix." A curve of the temperature versus energy absorbed in breaking the specimen is plotted. To obtain this curve, 15 tests are performed, which include three tests at five different temperatures. The intersection of the energy versus temperature curve with the 30 ft-lb ordinate is designated as the NDTT. The available data indicate differences as great as 40° between curves plotted through the minimum and average values, respectively. The determination of NDTT from the average curve is considered represent~tive of the material and is consistent with procedures as specified in ASTM E23. In assessing the NDTT shift due to irradiation, the translation of the average ."i curve is used. I I
- As part of the Westinghouse surveillance program referred to above, Charpy V-notch impact tests, tensile tests, and fracture me*chanics specimens are taken from the core region plates and forgings, and core region weldments including heat-affected-zone material. The test locations are similar to
'.j those used in the tests by the fabricator at the plate mill.
The uncertainties of measurement of the baseplate NDTT are
- 1. Differences in Charpy V-notch ft-lb values at a given temperature between specimens.
- 2. Variation of impact properties through plate thickness.
The fracture toughness technology for pressure vessels and correlation with service failures based on Charpy V-notch impact data is based on the averaging of data. The Charpy V-notch 30-ft-lb-fix temperature is based on multiple tests by the material supplier, the fabricator, and by Westinghouse as part of the surveillance program. The average of sets of three specimens at each test temperature is used in determining each of five data points
SPS UFSAR 4.1-18 (total of 15 specimens). In the review of available data, differences of 0°F to approximately 4;0°F are observed in comparing curves plotted t,hrough the minimum and average values, respectively. The value of NDTT derived from the average curve is judged to be representative of the material because of the averaging of at least 15 data points, consistent with the specified procedures of ASTM E23. In the case of the assessment of NDTT shift due to fast neutron flux, the displacement of transition curves is measured. The selection of maximum, minimum, or average curves for this assessment is. not significant since like curves are used. There are quantitative differences between the NDTT measurements at the surface, one-quarter thickness (1/4T), or center of a plate. The NDTT from 1/4T to the center in heavy plates is observed to vary from improvement in the NDTT to increases up to 85°F *. The NDTT at the surface is measured to be as much as 85°F lower than at 1/4T. The 1/4T location is considered conservative since the enhanced metallurgical properties of the surface are not used for the determination of NDTT. In addition, the limiting NDTT for the reactor vessel after operation is based on the NDTT shift due to irradiation. highest at the inner surface, Since the fast neutron dose is usage of the 1/4T NDTT criterion is conservative. Data have been accumulated on the variation of NDTT across heavy section steels at Westinghouse, Nuclear Energy Systems. Similarly, an evaluation of properties of pressure vessel steels in plates 6 to 12 in. thick has been sponsored by the Pressure Vessel Research Committee. Data show NDTT differences between 1/4T and center of less than 20°F. The criterion of using NDTT + 60°F at the 1/4T location without taking advantage of the enhanced properties at the surface of reactor vessel plates is conservative. To assess any possible uncertainties in the consideration of NDTT shift for welds, heat-affected zones, and base metals, test specimens of these three material types are included in the reactor vessel surveillance program. Additional information is available in References 7 and 8, both of which e were submitted to the NRC by Vepco letter dated January 23, 1978.
SPS UFSAR 4.1-19 4
- 1 REFERENCES e
- 1. H. Bohl, Jr., et al., PlMG, a One-Dimensional Multigroup Pl Code for the IBM-704, WAPD-TM-135, 1959.
- 2. K. Shure, "Radiation Damage Exposure and Embrittlement of Reactor Pressure Vessels," Nuclear Applications, Vol. 2, April 1966.
- 3. P.A. Gillis et al., SPIC-I, an IBM-704 Code to Calculate the Uncollided Flux Outside a Right Circular Cylinder, WAPD-TM-176, 1959.
- 4. W.R. Cadw~ll, PDQ4, a Program for the Solution of the Neutron Diffusion Equations in Two Dimensions on the Philco-2000, WAPD-TM-230, 1961.
S. W.R. Cadwell, PDQ3, a Program for the Solution of the Neutron Diffusion Equations in Two Dimensions on the IBM-704, WAPD-TM-179, May 1960.
- 6. H. P. Flatt and D. C. Baller, AIM-5, a Multigroup One-Dimensional Diffusion Equation Code, NAA-SR-4694, March 1960.
- 7. Virginia Electric and Power Company, Surry Unit No. 1, Reactor Vessel Radiation Surveillance Program, WCAP-7725.
- 8. Virginia Electric and Power Company, Surry Unit No. 2, Reactor Vessel Radiation Surveillance Program, WCAP-8805.
SPS UFSAR 4.1-31 Table 4 .1-11 CAPSULE SPECIMEN DATA, SECOND FOUR OF EIGHT CAPSULES No. No. No. Material Charpy Tensile WOL Plate Ill (high NDT) 8 2 2 Plate 112 8 Weld metal 8 2 2 Heat-affected-zone metal 8 Dosimeters Pure Cu Pure Fe Pure Ni CoAl (0.15% Co) CoAl (cadmium shielded) ~ U238 NP237 Thermal Monitors*
.97 .5 Pb, 2.5 Ag (579°F melting point) 97.5 Pb, 1.75 Ag, 0.75 Sn (590°F melting point)
Table 4 .1-12 RELATIVE EXPOSURES OF CAPSULES AND REACTOR VESSEL WALL Lead Adjacent Vessel Wall Lead Vessel Maximum Capsule Location by a Multiplying Factor of by a Multiplying Factor of 15 degrees 2.7 1.8 25 degrees 2.7 1.1 35 degrees 2.7 0.8 45 degrees 2.5 0.6
Table 4.1-13 REACTOR PRESSURE VESSEL NDTT DATA, SURRY UNIT 1 Drop Wt. RT Shelf NDTT Energy Lat. Exp. NDT *Energy Component Grade Cu(%) (OF) (ft-lb) (mils) (OF) (ft-lb) Inter shell (plate 1) A533B Cl.1 0.11 10 62 53 10 115 Lower shell (plate 1) A533B Cl.l 0.11 20 64 53 20 103 Core weldment Weld metal 0.25 50 46 0 70 Heat attached zone 54 44 0 81 Note: Charpy V-notch data for shell plates obtained from full Charpy curves for specimens oriented normal to the principal rolling direction. Where drop weight NDTT is not indicated, an NDTT maximum of 0°F was used. Inter shell (plate 2) A533B Cl.1 0.11 0 65 42 0 143 Lower shell (plate 2) A533B Cl.1 0.11 0 76 57 0 123 Note: Charpy V-notch data obtained from full Charpy curves for specimens oriented in principal rolling direction. Closure head dome A533B Cl.l 0 53 38 0 115 Head flange A50B Cl.2 115 77 60 192 Vessel flange A50B Cl.2 101 63 60 144 Inlet nozzle A50B Cl. 2 67 53 60 98 Inlet nozzle A50B Cl. 2 67 55 60 97 Inlet nozzle A50B Cl.2 101 74 60 104
- Outlet nozzle A50B Cl.2 110 85 60 131 Outlet nozzle A50B Cl.2 95 73 60 110 Outlet nozzle A50B Cl. 2 90 71 60 105 Upper shell A50B Cl.2 40 98 75 40 127 Bottom head ring A50B Cl. 2 50 118 74 50 132 Bottom head dome A533B Cl.1 0 60 41 0 130 .
.i:--
I-' I Note: Where drop weight NDTT is not indicated, an NDTT maximum of 60°F was used. Charpy V-notch data obtained from w N full Charpy curves for specimens .oriented in principal rolling or working direction
- e
REVISION 1 6/83 SPS UFSAR 4. 2-23 \1 All external insulation of reactor coolant system components is compatible with the component materials. The cylindrical shell exterior and closure flanges to the reactor vessel are insulated with metallic reflective insulation. The closure head is insulated with low halide-content insulating material. All other external corrosion-resistant surfaces in the reactor coolant system are insulated with inhibited low-halide or halide-free insulating material as required. The stress limits established for the reactor vessel are dependent upon the temperatures at which the stresses are applied. As a result of fast neutron irradiation in the region of the core, the material properties change, including an increase in the NDTT. This was discussed in Section 4.1.7. An NDTT no greater than 40°F was set as the design limit. The material was tested to verify conformity to specified requirements and to determine the actual NDTT value. In addit_ion, this plate was 100% volumetrically inspected by ultrasonic test using both longitudinal and shear wave methods .
- The remaining material in the reactor vessel and other reactor coolant
-system components meets the appropriate design code requirements and specific component function.
The reactor vessel material was heat-treated specifically to obtain good notch-ductility. This ensured a low NDTT, and thereby gave assurance that the finished vessel could be initially hydrostatically tested and operated as near to room temperature as possible without restrictions. The techniques used to measure and predict the integrated fast neutron (E greater than 1 MeV) fluxes at the sample locations are described in Section 4 .1. 7. The calculation method used to obtain the maximum neutron (E greater than 1 MeV) exposure of the reactor vessel is identical to that described for the irradiation samples. Since the neutron spectra at the samples are applied with confidence to the adjacent section of reactor vessel, the maximum vessel exposure is obtained from the measured sample exposure by appropriate application of the calculated azimuthal neutron flux variation. The estimated maximum integrated fast neutron (E greater than 1 MeV) 19 2 exposure of the vessel is computed to be 4.1 x 10 n/cm for 40 years'
REVISION 1 6/83 SPS UFSAR 4.2-24 19 2 operation at 2441 MWt (4.2 x 10 at 80% load factor. n/cm for 40 years' operation at 2546 MWt) Based on Table 4.2-3, the predicted NDTT shift for this e exposure is 290°F, the value obtained from the curve shown in Figure 4.2-9 with irradiation at 550°F. 19 2 The calculated neutron exposure exceeds the value of 2.8 x 10 n/cm (E greater than 1 MeV) reported in the Prelimina~y Safety Analysis Report. The reasons for the increase are as follows.
- 1. Core design refinements leading to the adoption of a radial power distribution that includes an increased energy generation in the peripheral assemblies of the core.
- 2. The associated effect on the azimuthal variation of fast neutron fluxes at the reactor vessel inner surface.
To evaluate the NDTT shift of welds, heat-affected zones, and base material for the vessel, test coupons of these material types are included in the reactor vessel surveillance program described in Section 4.1.7, along with the methods used to measure the initial NDTT of the reactor vessel baseplate material. 4.2.6 MAXIMUM HEATUP AND COOLDOWN RATES The reactor coolant system operating cycles used for design purposes are given in Table 4.1-8 and described in Section 4.1.4. During unit heatup and cooldown, the rates of temperature and pressure changes are limited. The system design heatup and cooldown rate of 100°F/hr satisfies stress limits for cyclic operation (ASME Vessel Code, Section III) and is consistent with the expected number of cycles. However, the normal system heatup and cooldovm rate is conservatively set at 50°F/hr. Sufficient electrical heaters are installed in the pressurizer to permit an adequate heatup rate of 55°F/hr when starting with a minimum water level. This rate takes into account the small continuous spray flow provided to maintain the pressurizer liquid homogeneous with the coolant. e
SPS UFSAR 4.3-15 specified. The finished flywheels are subjected to 100% volumetric ultrasonic inspection. The finished machined bores are also subjected to magnetic particle or liquid penetrant examination, These design-fabrication techniques result in flywheels with primary stress at operating speed (shown in Figure 4.3-5) le-ss than 50% of the minimum specified material yield strength at room temperature (100° to 150°F), Bursting speed of the flywheels has been calculated on the basis of 14 Griffith-Irwin's results to be 3900 rpm, more than three times the operating speed. A fracture mechanics evaluation was made of the reactor coolant pump flywheel. This evaluation considered the following assumptions:
- 1. Maximum tangential stress at an assumed overspeed of 125%.
- 2. A crack through the thickness of the flywheel at the bore.
- 3. 400 cycles of startup operation in 40 years .
- Using critical stress intensity factors and crack growth data attained on flywheel material, the critical crack size for failure was greater than 17 in.
radially, and the crack growth data was O. 030 in. to O. 060 in. per 1000 cycles. 15 An ultrasonic inspection capable of detecting at least O. 5-in. -deep cracks from the ends of the flywheel, and a dye penetrant or magnetic particle test of the bore, both at the end of 10 years, will be more than adequate as part of a unit surveillance program. The design specifications for the reactor coolant pumps include as a design condition the stresses generated by a design-basis earthquake ground acceleration of 0.15g. The pump would continue to run unaffected by such conditions. In no case does any bearing stress in the pump exceed or even approach a value that the bearing could not carry. In order to preclude undetected flywheel deterioration during plant life, even though such deterioration is not expected, the ultrasonic inspections are repeated at regular intervals during station life.
REVISION 1 6/83 SPS UFSAR 4.3-16 Following a hypothetical bearing seizure, the flywheel is not expected to twist off. Therefore, the reactor coolant pumps are not considered sources of missiles and the engineered safeguards are not in jeopardy. 4:~3. 4 OVERPRESSURE PROTECTION The reactor coolant system is protected against overpressure by code safety valves and power-operated relief valves located on the top of the pressurizer. The safety valves on the pressurizer are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code. The capacity of the pressurizer code safety valves and power-operated relief valves is determined from considerations of the reactor protection system, and accident or transient conditions that may cause overpressure. All analyses for Surry have been done on the basis of satisfying 10 CFR 50, Appendix G limits over a 40-year plant life. Without regard for the temperature, the Appendix G limit will never be less than 500 psig during the entire 40-year plant life. Therefore, 500 psig has been used as an absolute pressure limit in determining the valve setpoints.
- Vepco has submitted plant specific data to the NRC for safety and relief valve testing and for the power-operated relief valve block valve testing required by NUREG-0737, Item II.D.l. This information is provided in Refer-ences 16 and 17, respectively. 1 In Reference 16 it was concluded that the Surry Units 1 and 2 safety and relief valves, piping arrangements, and fluid inlet conditions were bounded by the valve and test parameters of the EPRI safety and relief valve test program and that the EPRI tests confirmed the ability of the safety and relief valves to open and close under the expected operating fluid conditions. In Reference 17 it was concluded that the block valve tested by EPRI was similar in design to the Surry block valves and that the valve successfully completed the *evalu-ation and test program, fully opening and closing on demand.
Mass input and heat input pressure transients were assumed, and calculations were performed to determine the overpressure protection setpoints J
REVISION 1 6/83 SPS UFSAR 4.3-17
- for the power-operated relief valves. 18
- 19 For the limiting pressure transient (mass input case), the setpoints should be as follows:
PORV 1 - 410 psig open, 400 psig reset PORV 2 - 425 psig open, 415 psig reset This provides ensurance that the 500-psig Appendix G limit will not be exceeded during a limiting pressure transient. Sufficient safety margin is provided by these setpoints, since the transient analysis shows that only one power-operated relief valve, set to open at 435 psig, is required to prevent exceeding the Appendix G limit. The power-operated relief valves are set to operate at the above values whenever the reactor coolant system is less than 350°F and the reactor head is bolted. 4.3.4.1 operational conditions The operational conditions requiring overpressure protection are controlled by administrative procedures as follows:
- 1. While water-solid, only one charging pump is permitted to be operational.
- 2. The safety injection accumulators must be isolated so that they do not discharge into the reactor coolant system while water-solid.
3; The safety injection logic must be blocked while water-solid.
- 4. The temperature of the reactor coolant system must not be more than 50°F cooler than the bulk water in the steam generators before starting the first reactor coolant pump.
4.3.4.2 Air Supply A backup air supply is provided to ensure power-operated relief valve
- operability in the event of a loss of the primary air supply and/or a loss of offsite power. The sizing of the redundant air supply considered both
REVISION 8 8/89 SPS UFSAR 4.3-18 power-operated relief valve response times with the assumed that operator action does not occur for 10 min. above setpoints, response is such that the valves may be caused to cycle at a rate of 6 sec per cycle. This would require an air supply capable of 100 cycles. and The fastest system Four high-pressure air bottles are provided for each valve, with each bottle capable of 31 cycles. The air bottles are 1.74 ft 3 and contain a nominal pressure of 2200 psig. The minimum pressure required to fully stroke the valve is 61 psig. Thus, for 31 cycles of operation, assuming a small amount of air loss per cycle, the required air pressure is 1000 psig. The containment instrument air system header to the power-operated relief valves contains a solenoid valve and a backup high-pressure air supply in the line to each relief valve. Two reactor coolant system transmitters are used to provide high-pressure alarms and to control the valves. Two key-lock switches are installed in the main control board (vertical section) to permit administrative control of the system at the appropriate point during cooldown or heatup. 4.3.5 SYSTEM ACCIDENT POTENTIAL The potential of the reactor coolant system as a cause of accidents was evaluated by investigating the consequences of certain credible types of component and control failures, as discussed in Sections 14. 2 and 14. 3. Reactor coolant pipe rupture is evaluated in Section 14.5. As evaluated in Section 14.2, no credible component or control failure results in a DNBR less than 1.3. Sections 14.3 and 14.5 show that for breach of the reactor coolant system boundary resulting from a steam-generator tube rupture, a rod-ejection accident, or a pipe break up to and including the double-ended rupture of a reactor coolant pipe*, the consequences in terms of activity releases are within the guidelines of 10 CFR 100. 4.3.6 REDUNDANCY Each loop of the reactor coolant system contains a steam generator and a reactor coolant pump. Operation at reduced reactor power is possible with one
- As discussed in Section 15.6.2, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop piping; the potentially governing pipe ruptures remaining are discussed in Section 15.6.2.
REVISION 1 6/83 SPS UFSAR 4.3-21
4.3 REFERENCES
(continued)
- 11. D. Burgreen, J. J. Byrnes, and D. M. Benforado, "Vibration of Rods Induced by Water in Parallel Flow," Transactions ASME 80, pp. 991-1003, 1958.
- 12. TEM, Tubular Heat Exchange Manufacturers Association Manual.
- 13. R. L. Chait, C. C. Peake, and A. Lohmeier, "Design and Manufacture of Large Surface Condenser - Problems and Solutions," Proceedings, American Power Conference 28, pp. 469-483, 1966.
- 14. E. L. Robinson,. "Bursting Tests of Steam-Turbine Disk Wheels,"
Transactions ASME, July 1944.
- 15. D. H. Winne and B. M. Wundt, "Application of the Griffith-Irwin Theory of Crack Propagation to the Bursting Behavior of Discs, Including Analytical and Experimental Studies," Transactions ASME, December 1, 1957 .
- 16. Letter from R.H. Leasburg, Vepco, to H. R. Denton, NRC,
Subject:
Response to NUREG-0737 Post-TMI Requirement - Item II.D.l, Relief, Safety, Block Valve Test and Discharge Piping Analysis Requirements, 1 Plant Specific Report, dated July 1, 1982 (Serial No. 392).
- 17. Letter from R.H. Leasburg, Vepco to H. R. Denton, NRC,
Subject:
Response to NUREG-0737 Post-TMI Requirement - Item II.D.l, Relief, Safety, Block Valve Test and Discharge Piping Analysis Requirements, Block Valve Reports, dated September 1, 1982 .(Serial No. 514).
- 18. Virginia Electric and Power Company, Pressure Mitigating System Transient Analysis Results, July 1977.
- 19. Virginia Electric and Power Company, Pressure Mitigating System Transient Analisis Results, Supplement to Juli 1977 Re:eort, September 1977.
SPS UFSAR 5.1-1 Chapter 5 CONTAINMENT SYSTEM This section describes the containment system for either unit. The containment systems for the two units are similar and completely independent. 5.1 GENERAL DESCRIPTION The containment system, together with the engineered safeguards (Chapter 6), is designed to limit radiation doses under conditions resulting from the design-basis accident (Chapter 14) to less than the 10 CFR 100-suggested criteria at the site boundary and beyond. The steel-lined, reinforced-concrete containment structure, including foundations, access openings, and penetrations are designed and constructed to maintain full containment integrity when subjected to the temperatures,
- pressures, potential missiles resulting from the design-basis accident, and the earthquake conditions and tornados described in Chapter 2. Systems are provided to remove heat from the containment and to ensure against breaching containment integrity at the time of, or following, the design-basis accident, or any lesser accident.
The containment concept includes provisions for routine operation at a reduced internal pressure in which the air partial pressure varies between about 9.0 and 11.0 psia, and for the return to subatmospheric pressure within 60 min after the design-basis accident through the use of multiple spray sys-tems. This concept provides for positive termination of outleakage of fission products from the containment, since the containment is maintained at subatmospheric pressure after depressurization. The pressure following depressurization is maintained at no greater than 13.9 psia. Provisions have been made for the leak testing of liner seams during construction; for air pressure and leak testing of the containment structure at the completion of construction; for leak testing of the penetrations and access openings at any time; for continuous leak monitoring of the containment
SPS UFSAR 5.1-2 structure while at subatmospheric pressure; and for periodic pressure testing of-the containment structure throughout station life. Details of containment structural design are given in Chapter 15, and details related to performance during postulated accident situations are given in Chapter 14.
REVISION 3 6/85 SPS UFSAR 5.3-9
- between the ILRT and the supplemental test. The results from the supplemental test are acceptable if the difference between the supplemental test data and the data obtained from either the reference volume test method or the absolute test method is within .25 La, where La is the ma~imum allowable leakage rate at the- calculated peak containment pressure.
The containment weld channels have threaded plugs that- were used for-the
- leak test;lng of welds of the containment liner plate during construction
- These plugs are removable to allow leak testing d~ring the operational phase of the plant.
Weld channels of the dome of the containment are located outside the liner plate, with the test hole tapped and plugged from the inside; this allows the strength weld between plates to be leak tested. Weld channels of the floor liner plate are piped through the concrete, which covers the floor,- to test port panels, and plugged for future testing requirements. Weld channels on the straight side walls of the containment are located inside the liner plate, which allows the weld between the channel and containment liner to be leak tested. 5.3.2.3 Design Evaluation Periodic leakage monitoring can be performed by two different methods, the absolute method and the reference volume method. Either method is verified by a supplemental test. The failure of one system does not preclude the monitoring of containment leakage by the other system. Either
REVISION 3 6/85 SPS UFSAR 5.3-10 the absolute method or the reference volume method is sufficiently accurate to establish that the containment leakage rate is less than O.1% of the containment volume per day. As part of the containment isolation system (Section 5.2), each open leakage-monitoring line penetrating the*containment structure is provided with two automatic trip valves and each closed line is provided with one automatic trip valve. In the event of an incident, these lines* are. closed and no leakage to the environment occurs. The containment leakage-monitoring system tubing, as an extension of the containment, is designed to withstand the pressure and temperature expected during an incident. 5.3.2.4 Tests and Inspections All instruments, including manometers., resistance thermometers, and dewpoint cells are calibrated before each Class A test. Before the initial Class A test, the sealed-bulb system is filled with pure Freon and tested for leakage with a halogen detector. The Freon is evacuated by the leakage-monitoring, sealed-system vacuum pump before dry air is added. Over the economic lifetime of the units, the leakage data acquired from these various test methods may be used to justify increasing the period over which integrated leak rate tests are conducted. 5.3.3 SPRAY SYSTEMS The containment-spray systems, which consist of containment-spray subsystems and recirculation-spray subsystems, are described in detail in Sections 6.3.1 and 6.3.2. The containment-spray subsystems operate during the depressurization period after a LOCA.
SPS UFSAR 6. 'i.-7 reflooded and that the reactor vessel is flooded at least to the nozzle. The e core decay heat is removed by boiloff of the injected water, and ultimately the core is subcooled. The low-head pumps will recirculate the sump water, either directly to the reactor coolant loops for large breaks, or to the suction of the high-head pumps for small breaks, to ensure continued long-term cooling of the core. Hot-leg connections for the low-head systems were selected to provide the optimum performance for the above functions and achieve a diversity of injection locations and flexibility to meet all long-term cooling requirements. The cold-leg break is the most limiting, since the flow from one of the three accumulators is lost through the break, the steam-binding problem is more severe, and the clad temperatures at the end of blowdown are higher than for a comparable hot-leg break. The combination of these factors requires a larger flow for the cold-leg break. The two unbroken cold-leg lines will deliver 340 gpm.frcim the high-head pump, with allowance for part of the flow to spill through the break in the cold leg. The low-head pumps provide the means to recirculate the sump water cooled by the spray heat exchangers and continue cooling the core through several alternate flow paths. The flow provided is in excess of that required to replace boiloff with allowance for spilling injection flow where applicable. If the loss of coolant occurred on the cold leg of one of the loops, the injected water would. pass through the core to the break and ultimately subcool the core in the forced-circulation mode. Motor-operated valves of the safety injection system that are under manual control, that is, valves that normally are in their ready position and do not receive a safety injection signal, have their positions indicated on a common portion of the control board. At any time during operation, if one of these valves is not in the ready position for injection, it is shown visually on the board. In addition, an audible alarm alerts the operator to the condition.
SPS UFSAR 6.2-8 A detailed listing of the instrumentation readouts on the control board that the operator can monitor during initial injection is given in Table 6.2-2.
- 6. 2-. 2 .1. 2 Changeover from Injection to Recirculation The transfer of the safety injection suction lineup from the refueling water storage tank to the containment sump takes place automatically. The automatic transfer is initiated by a 2/4 matrix involving a refueling water storage tank level coincident with the two position key switches being in the recirculation mode transfer position (one key switch for each train).
The valve alignment sequence begins with MOV-863A and B opening while MOV-885A, B, C, and D close. The maximum closing time for these valves is 2 min. MOV-836A and B supply suction to the high-head safety injection pumps from the discharge of the low-head safety injection pumps. MOV-885A and D and MOV-885B and C are series valves isolating each of the two low-head safety injection pump recirculation lines to the refueling water storage tanks. minutes after the start of the valve positioning sequence, MOV-860A and B start to open, while MOV-862A and Band LCV-115B and D start to close. maximum closing time for these valves is 2 min. The Two MOV-860A and B supply suction to the low-head safety injection pumps from the containment sump. MOV-862A and B isolate suction to . the low-head safety injection pumps from the refueling water storage tank. LCV-115B and D isolate the high-head safety injection charging pump suction from the tank. Existing check valves in these lines prevent cross connecting the containment sump with the refueling water storage tank. The entire sequence of the automatic transfer takes 4 min. (Note: Preface each MOV number with a 1 or 2 depending upon the unit, i.e., MOV-1885A Unit 1, MOV-2885A Unit 2.) The transfer system incorporates several functions to allow manual ini'tiation of the switchover, automatic switchover as described above, bypass of system functions during refueling operations, and switchover by the existing method of individual valve alignment. The functions use three push buttons and one 2-position key switch. Push buttons 1 and 2 are for MANUAL e
REVISION 3 6/85 SPS UFSAR 6.2-17
- under the seat to prevent the leakage of recirculated water through the valve stem packing. Relief valves are 'totally enclosed. All modulating control valves, of all sizes that are exposed to normally radioactive fluid are provided with double-packed stuffing boxes and stem leakoff connections that are piped to the vent and drain system.
The check valves that isolate the safety injection system from the re~ctor coolant system are installed adjacent to the reactor coolant piping to reduce the probability of a safety injection line rupture causing a LOCA. The gas relief valves on the accumulators protect them from pressures in excess of the design value.
- 6.2.2,2.4 Motor-Operated Valves The pressure-containing parts (body, bonnet~- and disks) of the valves employed in the safety injection system are designed according to criteria established by the USAS Bl6.5 or MSS SP-66 specifications. The materials of construction for these parts are procured as per ASTM Al 82, F316, or A351, Grade CF8M or CF8.
Radiographic inspection is conducted in accordance with the procedure outlined in ASTM E-94. Radiographic acceptance standards are outlined in ASTM E-71, E-186, or E-280, whichever is applicable, and meet the requirements of severity level 2, except that D, E, F, and G defects are not permissible. The body, bonnet, and disk are liquid penetrant inspected in accordance with ASME Code, Section III, paragraph N-627. When a gasket is employed, the body-to-bonnet joint is designed as per ASME Code, Section VIII, or USAS Bl6.5, with a fully trapped, controlled compression, spiral-wound asbestos gasket with provisions for seal welding, or of the pressure seal design with provisions for seal welding. The body-to-bonnet bolting and nut materials are procured per ASTM Al93 and Al94, respectively
- SPS UFSAR 6.2-18 The entire assembled unit is hydrotested as outlined by Manufacturers Standardization Society in the Valve and Fitting Industry, Specification 61 (MSS SP-61), with. the exception that the test is maintained for a minimum period of 30 min. Failure of the test is cause for rejection. The seating design is of the Darling parallel disk design, the Crane flexible wedge design, or the equivalent. These designs have the feature of releasing the mechanical holding force d~ring the first increment of trave!. Thus, the motor operator has to work only against the frictional component of the hydraulic imb~lance on the disk and against the packing box friction. The disks are guided throughout their full travel to prevent chattering and provide ease of gate movement. The seating surfaces are hard-faced (Stellite No.~ or equivalent) to prevent galling and reduce wear.
The stem material is AS'l'M A276, Type 316, condition B, or precipitation-hardened 17-4 PR stai~less, procured and heat-treated to Westinghouse specifications. These materials are selected because of their corrosion resistance, high-tensile properties, and their resistance to surf ace scoring by the packing. The valve stuffing box is designed with a lantern ring leakoff connection with a minimum of a full set of packing below the lantern ring and a maximum of one-half of a set of packing above the lantern ring; a full set of packing is defined as a depth of packing equal to 1. 5 times the stem diameter. The experience with this stuffing box design and the selection of packing and stem materials has been very favorable in both conventional and nuclear power stations. The motor operator is extremely rugged. The unit incorporates a hammer-blow feature that allows the motor to impact the disks away from the fore- or backseat upon opening or closing. The hammer-blow feature not only impacts the disk but allows the motor to rapidly attain its operational speed. The valve is assembled, hydrostatically tested, seat-leak.age tested (fore and back), operationally tested, cleaned, and packaged.as per specifications. In some cases, extension stems are used on motor-operated valves. These valves would be operationally tested initially without the extension stems,
REVISION J 6/85 SPS UFSAR 6.2-23
- operate with full system differential pressure. The isolation valve is normally blocked opened by de-energizing the valve motor operators when the reactor coolant pressure exceeds 1000 psig. An alarm in the control room sounds if the valve is inadvertently closed. It is not. expected that the isolation valve will have to be closed because of excessive leakage through the check valves.
When the reactor coolant system is being pressurized during the normal unit heatup- operation, the check valves are tested for leakage as soon as . there is about 100 psi differential across the valve. This test confirms the seating of the disk and whether or not there has been an increase in the leakage since the last test. When this test is completed, the discharge line test valves -are opened and the reactor coolant system pressure increase continued. There should be no increase in leakage from this point on, since increasing reactor coolant pressure increases the seating force and decreases the probability of leakage *
- 6.2.2.2.8 Relief.Valves The accumulator relief valves are s.ized to pass nitrogen at a rate in excess of the accumulator gas fill-line delivery rate. The relief valves also pass water in excess of the expected leak rate, but this is not necessary, because the time required to fill the gas space gives the operator ample opportunity to correct the situation. For an inleakage rate 15 times the manufacturing test rate, there are about 1000 days before water reaches. the relief valves. Prior to this, level and pressure alarms would have been actuated.
6.2.2.2.9 Valve Leakage Limitations Exceptional tightness is specified for all valves, and packless diaphragm valves are used where possible (such as for instrument valves). Normally open valves have backseats that limit leakage to less than 3 1 cm /hr/in. of stem diani.eter, assuming no credit for packing in the valve .
REVISION 3 6/85 SPS UFSAR 6.2-24 Normally closed globe valves are installed with pressure under the seat to prevent stem leakage from the more radioactive fluid side of the seat.
- Motor-operated valves exposed to recirculation flow are provided with double-packed stuffing boxes and stem leakoff connections, which are piped to the vent and drain system.
The specified leakage across the valve disk required to meet the equipment specification and hydrotest requirements is as follows:
. 3
- 1. Conventional globe valves - 3 cm /hr/in. of nominal pipe size.
3
- 2. Gate valves - 3 cm /hr/in. of nominal pipe size.
Motor-operated gate valves - 3 cm3 /hr/in. o{ nqminal pipe size.
- 4. Check valves - 3 cm3 /hr/in. of nominal pipe size; 10 cm3 /hr/in. for 5.
300-lb and 150-lb USA standards.
.Accumulator check valves - 2 cm3 /hr I in. of nominal pipe size.
Leakage from components of the recirculation loop, including valves, is given in Table 6.2-7. 6.2.2.2.10 Pump and Valve Motors Electrical insulation systems for motors outside containment are supplied and tested in accordance with USASI, IEEE, and NEMA standards. Temperature rise design selection is such that normal long life is achieved even under accident loading conditions. Motors for the valves inside the containment* are designed to withstand containment environment conditions following the LOCA so that the valves can perform the required function during the recovery period.
REVISION 3 6/85 SPS UFSAR 6.2-61
- Table 6.2-5 DELETED
SPS UFSAR 6.2-62 Table 6.2-6 PUMP PARAMETERS
- Safety Injection Charging Pumps Number of pumps (per unit) 3 Design pressure, discharge, psig 2750 Desi_gn pressure, suction, psig 250 Design temperature, °F 250 Design flow rate, gpm 150 Maximum flow rate, gpm 600 Design head, ft 5800 Type Horizontal centrifugal Low-Head Safety Injection Pumps Number of pumps (per unit)
Type Design pressure, discharge, psig 2 Vertical centrifugal 300 (150 lb ASA discharge flange) Design temperature, °F 300 Design flow, gpm 3250 (injection) Design head, ft 225 Maximum flow rate, gpm 4200
REVISION l 6/83 SPS UFSAR 6. 3-3 I1 The systems operate at a relatively low pressure of 100 psi gauge and are not highly stressed during operation, so that the inducement toward cracking is reduced. Because the pH of the containment spray solution is above 8.0 and the recirculation spray solution pH is essentially 8.0, the potential for caustic stress corrosion cracking in the containment spray system and recirculation spray system is virtually nonexistent. 6.3.1.2.2 Motors Electrical insulation for motors located outside containment is in accordance with ANSI, IEEE, and NEMA standards, and is tested as required by these standards. Temperature rise design is such that normal long life is achieved even under accident loading conditions. Winding insulation has been developed that operates at temperatures well in excess of those calculated to occur under design-basis accident conditions . This type of insulation is used in motors located inside containment. The containment motors have been selected to ensure operation during LOCA conditions. Motor electrical insulation is in accord_ance with ANSI, IEEE, and NEMA standards. The motors are tested as required by these standards. Bearings are antifriction type, silicone grease lubricated. Bearing loading and high-temperature tests have been performed, and the expected bearing life equals, or exceeds, that specified by the American Federation of Bearing Manufacturers (AFBM). 6.3.1.2.3 Piping Piping fabrication, installation, and testing are in accordance with the Specification for Power Plant Piping, ANSI B31. 1, with supplemental requirements and inspections as necessary for use in nuclear applications. Pipe routing and supports are such that missiles generated from postulated events or the effects of LOCAs do not impair the operation of spray systems. e
REVISION 1 6/83 SPS UFSAR 6. 3-4 I1 6.3.1.2.4 Heat Exchangers and Vessels Heat exchangers and vessels are designed to ASME Code, Section III C, and have been radiographically inspected to ensure their structural integrity. Heat exchangers and vessels are of welded construction to preclude leakage. The design data of the spray system components are given in Table 6. 3-1. 6.3.1.3 Description The spray system consists of two separate but parallel containment spray headers, each of 100% capacity, and four separate but parallel recirculation
- spray headers, each of 50% capacity.
An additional ring header common to both containment spray trains is installed at elevation 95 ft 6 in. outside the crane wall. *check valves are installed in each branch connection from the riser to the common header to limit fill time, should one containment spray pump train fail to start. The containment spray subsystem is shown in Figure 6. 3-1, and the recirculation spray subsystem is shown in Figure 6.3-2. Elevations of all piping and components of these subsystems are shown in Figure 6.3-3. Each of the containment spray headers draws water independently from the refueling water storage tank; The sodium hydroxide solution used for iodine removal from the containment atmosphere is added to the containment spray water by a balanced gravity feed from the chemical addition tank. The refueling water storage tank is a vertical cylinder with a flat bottom and a dome top, and is secured to a reinforced-concrete foundation. The refueling water storage tank is fabricated of ASTM A240, Type 304 stainless steel, in accordance with API STD-650. This standard has been upgraded to provide requirements for welding, welding procedures, welder qualification, weld joint efficiency, and weld inspection in accordance with Section III of the ASME Code. The chemical addition tank is a vertical cylindrical vessel with flanged and dished heads mounted on a skirt and secured to the reinforced-concrete foundation. The chemical addition tank is fabricated of ASTM A240, e Type 304 stainless steel in accordance with Section VIII of the ASME Code.
SPS UFSAR 6.3-21 Table 6.3-1 (continued) SPRAY SYSTEM COMPONENT DATA
- Recirculation Spray Pump Motor (outside containment)
Number 4 (2 per unit) Horsepower, hp 300 Electrical characteristics 460 V, 3 phase, 60 cycle Service factor 1.15 Insulation Encapsulated Recirculation Spray Coolers Number 8 (4 per unit) Design duty, Btu/hr, each 55,534,520 Shell Tube Fluid flowing Recirculation Service water spray water Design pressure, psig 150 150 Design temperature, °F 250 250 Operating pressure, psig 100 20 Operating temperature, °F 200 95 Material Cupro nickel Cupro nickel Chemical Addition Tank Number 2 *(1 per unit) Type Vertical cylindrical Capacity, gal 4200 Design pressure, psig 25 Design temperature, °F 150 Material ss 304 Design code ASME Section VIII Operating pressure, psig Atmospheric Operating temperature, °F Ambient NaOH concentration,% 17-18
SPS UFSAR 6.3-22 Table 6.3-1 (continued) SPRAY SYSTEM COMPONENT DATA Chemical Addition Tank Pump Number 2 (1 per unit) Type Vertical centrifugal Rated flow, gpm 50 Rated head, ft 7 Brake horsepower, hp 0.1 Seal Mechanical Design pressure, psig 225 Material Pump casing ss 316 Shaft SAE 4140 Impeller ss 316 Piping Piping is designed to the Code for Pressure Piping, ANSI B31.1. Valves Valves are designed in accordance with the Code for Steel Piping Flanges and Flanged Fittings, ANSI B16.5. __J
SPS UFSAR 7-vii e CHAPTER 7: INSTRUMENTATION AND CONTROL TABLE OF CONTENTS (continued) Section Title 7.9 COMPUTER SYSTEM. * * * * * * * * * * * * * * * * * *
- 7.9-1
)
7.9.1 Design Bases .... ...... 7.9-1 7.9.2 System Description ...... 7.9-1 7.9.2.1 Analog Scanning. . .. . . . . . . 7.9-1 7.9.2.2 Alarming . . . . . . . .. . . ... 7.9-2 7.9.2.3 Alarm Review .. . ... . . .
. 7.9-2 7.9.2.4 Analog Trend . .... . ...
. 7.9-2 7.9.2.5 Digital Trend on Typewriter. . .... . ...
. 7.9-2 7.9~2.6 Digital Display. . . . .... . ...
. 7.9-3 7.9.2.7 Sequence of Events ... . 7.9-3
~
7.9.2.8 Normal and Summary Logging . ..... 7.9-3 7.9.3 System Evaluation. . . . . . . ..... . 7.9-3 7.10 SUBCOOLING MONITOR SYSTEM * * * * * * * * * * * * * * * * * *
- 7~10-1 7.10.1 Design Bases * *
- 7.10-1 7 .10. 2 System Description * * *
- 7 .10-1 7.10.3 System Evaluation ** *
- 7 .10-2 7.10 References * * * * * * . .. *. 7 .10-4
l SPS UFSAR 7-viii CHAPTER 7: INSTRUMENTATION AND CONTROL LIST OF TABLES Table Title Page 7.2-1 Reactor Trips * * * * * * * * .7.2-41 7.2-2 Logic Symbols. * * ** * * * * * * * *
- 7.2-44 7.2-3 Protection Interlocks.
- 7.2-47 7.2-4 Rod Stops ** * *
- 7. 2-50 7.4-1 Source Range Signals * * ** ** 7.4-27 7.4-2 Intermediate Range Signal. . * * * * * . 7.4-28 7.4-3 Power Range Signals. * *
- 7.4-29 7 .5-1 Emergency Safeguards Actuation Functions * * * * *
- 7.5-23 7.5-2 Valves/Dampers Actuated by Emergency Safeguards ~
Signals * * *
- 7.5-25 7.5-3 Safety-Related Systems *
- 7.5-36 7 .10-1 Subcooling Meter Design Data. *
- 7 .10-5 7 .10-2 Subcooling Monitor System Temperature and Pressure Inputs **
- 7.10-7
SPS UFSAR 7-ix - CHAPTER 7: INSTRUMENTATION AND CONTROL LIST*OF FIGURES Figure Title 7.2-1 Typical Illustration of ~T - T Protection avg 7.2-2 Reactor Trip Signals 7.2-3 Logic Diagram for Low Reactor Coolant Flow Trips 7.2-4 Design to Achieve Isolation Between Channels 7.2-5 Basic Elements of an Analog Protection Channel 7.2-6 Trip Logic Channels 7.2-7 Analog Channels
- 7. 2-8
- Logic Channel Test Panels 7.2-9 Con_trol Group Rod Insertion Monitor 7.2-10 T - ~T Protection avg
- 7. 2-11 Pressurizer Pressure Control and Protection
- 7 .2-12 Pressurizer Level Control and Protection
- 7.2-13 7.3-1 7.3-2 7.3-3 Steam-Generator Level Control and Protection Simplified Block Diagram of Reactor Control System T
avg Control System Power Supply to Rod Control Equipment and Control Rod. Drive Mechanisms 7.4-1 Neutron Detector Ranges of Operation 7.4-2 Nuclear Instrumentation System 7.4-3 Neutron Detector Locations 7.5-1 Safety Injection System Actuation 7.5-2 Engineered Safeguards Actuation Circuits 7 .5-3 Simplified Diagram for Overall Logic Relay Test Scheme 7.5-4 Simplified Diagram for.Relay Logic Channel Testing 7.6-1 Incore Instrumentation - Details 7.6-2 Incore Mechanisms 7:7-1 Arrangement - Main Control Room 7.8-1 Automatic System Load Control
. 7 .8-2 Simplified Unit-Load Control Interconnection
SPS UFSAR 7.4-1 7.4 NUCLEAR INSTRUMENTATION SYSTEM e 7.4.1 DESIGN BASES
.* 7. 4 .1.1 Fission Process Monitors and Controls The nuclear instrumentation system is used primarily for reactor protection. It permits monitoring of neutron flux and generates appropriate trip and alarm functions for various phases of reactor operating and shutdown conditions. It also provides a secondary control function, and indicates .:,.reactor status during
- startup and power operation. The nuclear instrumentation system uses information from the three separate types of instrumentation channels to provide three discrete protection levels. Each range of instrumentation (source, intermediate, and power) provides the necessary overpower reactor trip protection required during operation in that range. The overlap of instrument ranges provides reliable continuous protection from source to the intermediate and low power ranges. As the
- reactor power increases, the overpower protection level is increased (administratively) after satisfactory higher-range instrumentation operation is obtained. Automatic reset to more restrictive trip protection is provided when reducing power.
Several types of. neµtron detectors, with appropriate solid-state electronic circuitry, are used to monitor the leakage neutron flux from a
- completely shut down condition to 120% of full power. The power range channels are capable of recording overpower excursions up to 200% of full
., .power.
The neutron flux covers a wide range between these extremes. Therefore, monitoring with several ranges of instrumentation is necessary. The lowest range (source range) covers six decades of leakage neutron flux. The lowest observed count rate depends on the strength of the neutron sources in the core and the core multiplication associated with the shutdown reactivity. This is generally greater than one count per sec. The next range (intermediate range) covers approximately eight decades. Detectors* and
SPS UFSAR 7.4-2 instrumentation are chosen to provide overlap between the higher portion of the source range and the lower portion of the intermediate range. The highest range of instrumentation (power range) covers slightly more than two decades of the total instrumentation range. This is a linear range that overlaps with the higher portion of the intermediate range (the intermediate range monitors go off-scale at about 70% of rated power). The overlap for all detector ranges is shown in Figure 7.4-1 in terms of leakage neutron flux for a typical PWR plant. Startup-rate indication for the source and intermediate range channels is provided at the control console. The system described above provides control room indication and recording of reactor neutron flux during core loading, shutdown, startup, and power operation, as well as during subsequent refueling. Reactor trip and rod-stop control and alarm signals are provided by this sys*tem for safe plant operation. Control and permissive signals are transmitted to the reactor control and protection system for automatic plant control. Equipment failures and test status information are annunciated in the control room. 7.4.2 SYSTEM DESCRIPTION The nuclear instrumentation system (Figure 7. 4-2) *consist of eight independent channels: two of these are the source range, two are the intermediate range, and four are the power range channels. In addition, there are three auxiliary channels: the visual-audio count rate channel, the comparator channel, and the startup rate channel. The various detectors associated with the eight primary channels ,are shown in relative position with respect to the core configuration on Figure 7.4-3. 7.4.2.1 Protection Philosophy Nuclear unit protection assurance is obtained from the three ranges of ex-core nuclear instrumentation. Separation of redundant protective channels is maintained from the neutron sensor with its associated cables to the signal conditioning equipment in the control room with its associated output wiring, indicating or recording devices, and protective devices. Where redundant J
SPS UFSAR 7.4-21 function, concerned with turbine load cutback and rod stop, for the channel under test. The panel i~ located in nuclear instrumentation rack No. 4. Switches are also provided on this panel to permit a failed power range
~hannel's overpower-rod-stop function to be bypassed, and its average power signal to the reactor coolant system to be replaced by a signal derived from an active channel. This allows normal power operation to continue while the failed channel is repaired.
7.4.3.9 Output Information Tables 7.4-1, 7.4-2, and 7.4-3 provide the nuclear instrumentation system control and indication output information for the source, intermediate, and power ranges, respectively. 7.4.4 SYSTEM EVALUATION 7.4.4.1 Philosophy and Setpoints During plant shutdown and operation, three discrete, independent levels of nuclear protection are provided from the three ranges of ex-core .nuclear instrumentation. The basic protection philosophy is that the'level protection is present in all three ranges to provide a reliable, rapid, and restrictive protection system that is not dependent upon operation of higher-range instrumentation. Reliability is obtained by providing redundant channels that are physically and electrically separated. Fast trip response is an inherent advantage of using level trip protection in lieu of startup-rate protection
.,(with a long time constant) during plant startup.. More' restrictive operation
. ~
is an inherent feature, since an increase in power cannot be performed until satisfactory operation is obtained form higher-range instrumentation, which permits administrative bypass of the lower-range instrumentation. On decreasing power level, protection is automatically made more restrictive. Startup accidents while in the source range are rapidly terminated without
SPS UFSAR 7.4-22 significant increases in nuclear flux, and with essentially no power generator or reactor coolant temperature increase. The indications and administrative actions required by this protection system are readily available to the operator and should result in a safe, uncomplicated increase of power. 7.4.4.2 Reactor Trip Protection During reactor startup, the operator is made aware of satisfactory operation of one or more intermediate range channels by annunciation (audible and visual) at the control board. The source and i~termediate range flux level information is also readily available on recorders and indicators at the control console. At this time, if both intermediate range channels* are functioning properly, the operator would depress the two manual block switches associated with the source range logic circuitry, thus causing cutoff of source range detector voltages and blocking the trip logic outputs. The manual block should not be initiated, however, until at least one decade of satisfactory intermediate range operation is obtained. performed if desired. If one intermediate range channel is not functioning, normal power increase could still be The permissive P-6 annunciation is continuously displayed by the control board status lights. Continuation of the startup procedure in the intermediate range* would result in a normal power increase and the receipt of a permissive signal from the power range channels when two out of four channels exceed 10% of full power. The operator is alerted to this condition by a control board permissive* status light. Indicators (one per channel) and a recorder also indicate unit status in terms of percent full power~ If the operator does not block the intermediate range trip and continues the power increase, a rod stop will automatically occur from either of the intermediate range channels. The operator should depress the momentary "Manual Block" pushbuttons associated with the intermediate range rod stop and reactor trip logic. This would transfer protection to the low-range trips for the four power range channels. The permissive P-10 status light would be continuously displayed, as was P-6. The two low-range manual block switches must be depressed to initiate blocking /'
SPS UFSAR Figure 7.4-3 PROPORTIONAL COUNTER & LONG LONG COMPENSATED IONIZATION CHAMBER ION ION CHAMBER 0 CHAMBER FLAT CORE CORNER SPARE
- SPARE WELL Q 0 WELL 0
LONG PROPORTIONAL COUNTER LONG ION COMPENSATED IONIZATION CHAMBER ION CHAMBER CHAMBER NEUTRON DETECTOR LOCATIONS
SPS UFSAR 7.6-1 7.6 INCORE INSTRUMENTATION 7.6.1 DESIGN BASIS The incore in.strumentation is designed to yield information on *the neutron flux distribution and fuel assembly outlet temperatures at selected core locations. Using the information thus obtained, it is possible to confirm the reactor core design parameters. The system provides means for acquiring data only, and performs no operational unit control. 7.6.2
- SYSTEM DESCRIPTION 7.6.2.1 General The incore instrumentation system consists of thermocouples positioned to measure fuel assembly coolant outlet temperature at preselected locations, and flux thimbles, which run the length of selected fuel assemblies to measure the neutron flux distribution within the reactor core. The initial design called for 51 thermocouples and 50 flux thimbles. The high-pressure seals for the thermocouples and flux thimbles are shown on Figure 7.6-1.
The experimental _data obtained from the incore temperature and flux distribution instrumentation ~ystem, in conjunction with previously determined analytical information, can be used to determine the fission power distribution in the core at any time throughout core life. This method is more accurate than using calculational techniques alone. Once the fission power distribution has been established, the maximum power output is primarily determined by thermal power distribution and thermal/hydraulic limitations, which determine the maximum core capability. The incore instrumentation provides information that may be used to calculate the coolant enthalpy distribution, and the fuel burnup distribution, and to estimate the coolant flow distribution. e
SPS UFSAR 7.6-2 Both radial and azimuthal symmetry of power distributions may be evaluated by comparing the detector and thermocouple information from one e quadrant with similar data obtained from the other three quadrants. 7 *.6. 2. 2 Thermocouples Chromel-alumel thermocouples are threaded into guide tubes that penetrate the reactor vessel head through seal assemblies, and terminate at the exit flow end of the fuel assemblies. The thermocouples are enclosed in stainless steel sheaths within the guide tubes to facilitate replacement when necessary. Thermocouple readings are monitored by the computer, with backup readout provided by a precision indicator with manual point selection. The support of the thermocouple guide tubes in the upper core support assembly is described in Chapter 3. 7.6.2.3 Movable Neutron Detectors 7.6.2.3.1 -Mechanical Configuration Miniature neutron flux detectors, remotely positioned in the core, provide remote. readout for flux mapping. The basic system for the insertion of these detectors is as shown in Figure 7. 6-2. Retractable thimbles, into which the miniature detectors are driven, are pushed into the reactor core through conduits that extend from the bottom of the reactor vessel down through the concrete shield area, then to a thimble seal table. The thimbles are closed at the leading ends, are dry inside, and serve as the pressure barrier between the reactor water pressure and the atmosphere. Mechanical seals between the retractable thimbles and the conduits are provided at the seal line. During reactor operation, the retractable thimbles are stationary. They are extracted downward from the core during refueling to avoid interference within the core. A space above the seal line is provided for the retraction operation.
SPS UFSAR 7.6-3 The drive system for the insertion of the miniature detectors consists of.* a combination of drive assemblies, five-path rotary transfer devices, and ten-path rotary transfer devices, as shown in Figure 7.6-2. The drive system pushes hollow helical-wrap drive cables into the core. I Miniature detectors are attached to the leading ends of the cables, and small-diameter sheathed coaxial cables are threaded through the hollow centers back to the ends of the drive cables. Each drive assembly consists of a gear motor that pushes a helical-wrap drive cable and detector through a selective thimble path by means of a special drive box, and includes a storage device that accommodates the total drive cable length. Further information on mechanical design and support is provided in Chapter 3. 7.6.2.3.2 Control and Readout Description The control and readout system provides means to rapidly traverse the miniature neutron detectors to and from the reactor core at 72 ft/min, and to traverse the reactor core at 12 ft/min while plotting the thermal neutron flux versus detector position. The control syste~ consists of two sections, one physically mounted with the drive units, and the other contained in the control room. Limit switches in each tubing run provide signals to the path display to indicate the active detector path during the flux mapp_ing operation. Each gear box drives an encoder for position indication. One five-path group path selector is provided for.each drive unit to route the detector into one of the flux thimble groups or to storage. A ten-path rotary transfer assembly is used to route a detector into any one of up to ten selectable thimbles. Manually operated isolation valves on each thimble allow free passage of the detector and drive cable when open. When closed, these valves prevent steam leakage from the core in case of a thimble rupture. Provision is made to separately route each detector into a common flux thimble to permit cross calibration of the detectors. The control room contains the necessary equipment for control, position indication, and flux recording. Panels are provided to indicate the position of the detectors, and for plotting the flux level versus the detector position. Additional panels are provided for such features as drive motor controls, core path selector switches, plotting, and gain controls. A
SPS UFSAR 7.6-4 flux-mapping operation consists of selecting (by panel switches) flux thimbles in given fuel assemblies at various core locations. The detectors are driven to the top of the core and stopped automatically. An x-y plot (position vs flux level) is initiated with the slow withdrawal of the detectors through the core from the top to a point below the bottom. In a similar manner, other core locations are selected and plotted. Each detector provides axial flux distribution data along the center of a fuel assembly. Various radial positions of detectors are then compared to obtain a flux map for a region of the core. 7 .6 .* 3 SYSTEM EVALUATION The thimbles are distributed nearly uniformly over the core, with about the same number of thimbles in each quadrant. The number and location of these thimbles have been chosen to permit measurement of local to average peaking factors to an accuracy of +/-10% (95% confidence). Measured nuclear peaking factors are increased by 10% to allow for possible instrument error
- The DNBR calculated with the measured hot-channel factor is compared to the DNBR calculated from the design nuclear hot-channel factors. If the measured power peaking is larger than expected, reduced power capability is indicated.
The units have the capability for using fixed incore detectors, if required. Any decision to employ fixed incore detectors would be based upon operating experience in similar plants.
- i .*.,
1* *
- I, ,
rO
\~ :
\\.\
.,HIM~L.E. <:>E .... \.. '?l..UGt.
(HYDR.Q'ia,,A,,C. ,Esr) TO 0.A."T'A. LOG.(;E,~ I
~-OCO~PLE I
IHE.RMOC.OUPLE. 10 C.O"""'Dµ1T ,':>E,AL. ANO D1';;)(.QN"'1.E.C..'1' PLUC2, 5
=
/ *,
'-~
THIMBL
' .~~---,
~ LOW PRE.~':i.U'l.E 'EaEA.~
( RE:FL.ELoNG, ONLY) iHIMBLE. C.ONOUII WE.LD UNION znH 0
- d t
- c:I H
z ti) t-3
; t:c:I ti) l"tj zt-3 ti)
- > C:
t-3 tzj H ti) 0 z ~ tzj t::1 I-'* t,:j OQ t-3 i::
- > ~
H (D
~
ti)
"'I I-'
SPS UFSAR 7 .10-1 7.10 SUBCOOLING MONITOR SYSTEM 1 In response to NUREG-0578, instrumentation to detect inadequate core cooling has been installed at Surry Units 1 and 2. 7.10.1 DESIGN BASES The subcooling monitor system was designed by Westinghouse, and meets all 2 the requirements of Regulatory Guide 1. 97. Redundant primary coolant saturation meters are provided, with each meter consisting of a calculator and continuous control room display. Meter input consists of signals from eight core-exit thermocouples ( two from each core quadrant) , three. resistance temperature detectors (reactor coolant system hot and cold legs), and reactor coolant system pressure transmitters. The subcooling monitors are fed from emergency power. The safety-grade signal inputs and the calculational devices are not 3 4 qualified to IEEE-323 or IEEE-344. Design data are given in Tables 7.10-1 and 7.10-2. 7.10.2 SYSTEM DESCRIPTION The core subcooling monitor utilizes inputs from the reactor coolant loop resistance temperature detectors, reactor coolant system pressure, and selected incore thermocouples. A microprocessor is employed to calculate saturation temperature for the existing reactor coolant system pressure, and determine the margin to saturation based on the various temperature inputs. The information display consists of a main control board indication of the margin to saturation and expanded information at an electronics drawer located with plant protection and control equipment. The system is divided into two redundant channels. The redundant analog meters mounted on the main control board have nonlinear scales, expanded near saturation, with the range extended to the limits of the incore thermocouples.
SPS UFSAR 7.10-2 In addition to the numerical indication, the meter faces*are color-coded to increase visual acuity for the operators. The auctioneered low reactor coolant system pressure is utilized by the micropro*cessor to calculate the saturation temperature for the existing system pressure. By subtracting the auctioneered high incore thermocouple signal from the calculated saturation temperature, the current margin to saturation is calculated. This information is then displayed on the main control board meters. The margin to saturation from the auctioneered loop temperature signals is also available on the main control board meters by mea~s of a two-position
- switch to be located on the main board.
Two levels of alarm are provided, the first to indicate the development of off-normal conditions, the second the approach to loss of core subcooling. The actual setpoint for either alarm is controlled by the microprocessors, and is easily modified by keyboard entries at the main processing unit. Expanded information is displayed on the front of the electronics drawer to enable the operator to determine core cooling conditions. In addition to the digital displays, a graphic color display of the core thermocouples is provided on the electronics drawer. If all temperatures are sufficiently below saturation, the lamps will be green. If any individual. temperatures reaches an off-normal condition, the lamp display changes to yellow and alarm contacts for an annunciator are closed. Should conditions continue to degrade to the preset margin to saturation, the lamp display changes to red and an addition set of alarm contacts are closed for annunciation. 7 ** 1'0. 3 SYSTEM EVALUATION The subcooling monitoring system is isolated from the reactor protective instrumentation. The input signals are passed through isolation amplifiers that prevent erroneous signals generated in the subcooling instrumentation from backfeeding into reactor protective circuits. The core sub cooling monitoring system will function with no effect from the loss of the plant computer.
SPS UFSAR 7 .10-3 The design of the subcooling monitoring system allows alarm actuation e regardless of the meter input position switch. The internal circuitry of the system determines the margin to saturation based on the least conservative inputs from the resistance temperature detectors or thermocouples and the auctioneered pressure inputs. Alarm actuation is based on this determination and not on the meter position switch. Changes to emergency procedures have been made to emphasize the need to ensure adequate coolant flow and to ensure that the reactor coolant temperature and pressure is maintained or immediately restored to achieve an appropriate margin to saturation. The procedures include Westinghouse guidelines for the identification of and recovery from inadequate core cooling conditions. All licensed reactor operators and trainees receive special instruction with particular emphasis on.the use of existing instrumentation ta determine core conditions. The plant computer is used to monitor plant saturation conditions. The
- operator can initiate a trend printout of the system saturation temperature and one or more of the critical reactor coolant system*temperatures.
A saturation curve is posted on the control board and provided in current procedures as a backup to the subcooling monitor system. This curve, combined with the nearby indications of reactor coolant system temperatures and pressure, enables the operator to quickly determine the system's margin to saturation.
SPS UFSAR 7.10-4 7.10 REFERENCES
- 1. U. S. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," NUREG-0578, July 1979.
- 2. U.
- S. Nuclear Regulatory Commission, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Regulatory Guide 1.97, December 1980.
- 3. IEEE Standard 323-1974, IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations, 1974.
- 4. IEEE Standard 344-1975; Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations, 1975.
__J
SPS UFSAR 7.10-5 Table 7 .10-1 SUBCOOLING METER DESIGN DATA I. Display
- 1. Information displayed (T-Tsat, Margin to saturation in PSI Tsat, Press, etc.)
Degrees superheat Invalid results are indicated if inputs are invalid or if unit fails
- 2. Display type Analog, continuous
- 3. Redundancy Yes
- 4. Location Control room
- 5. Alarms Caution.,. 25°F subcooled based on loop RTDs
- 15°F subcooled based on incore T/Cs
- 6.
7. Uncertainty (°F) Range Alarm - 0°F subcooled based on loop 5 RTDs and T/Cs 1000 psi to O psi margin to saturation 0°F superheat to 2000°F superheat II. Calculator
- 1. Type Microprocessor
- 2. Availability (% of time) 100% microprocessor
- 3. Redundancy Yes
- 4. Selection logic Highest temperature, lowest pressure
- 5. Calculational technique Microprocessor
- 6. Information displayed Saturation margin, TSAT, PSAT
SPS UFSAR 7 .10-6 Table 7 .10-1 (continued) SUBCOOLING METER DESIGN DATA III. Input
- 1. Temperature (RTDs or T/Cs) RTDs - Hot- and cold-leg wide-range temperatures and incore T/C reference junction box bar T/Cs - incore thermocouples
- 2. Temperature sensors Hot - and cold-leg temperatures (two per meter)
Incore T/Cs (ei$ht per meter) Incore reference junction box temperature (one per meter)
- 3. Range of temperature sensors See Table 7 .10-2
- 4. Uncertainty of temperature See Table 7.10-2 sensors 5.
6. Pressure Pressure sensors Electronic pressure transmitter Reactor pressure (one per meter) Pressurizer pressure (one per meter)
- 7. Range of pressure sensors See Table 7 .10-2
- 8. Uncertainty of pressure sensors See Table 7.10-2 e
REVISION 4 6/86 SPS UFSAR 7.10-7
- Table 7 .10-2 SUBCOOLING MONITOR SYSTEM TEMPERATURE AND PRESSURE INPUTS I. Wide-Range Pressure Signalsa 0-3000 psig, 1-5 V, linear Manufacturer Rosemont I
Accuracy +/-0.5% of calibrated span Repeatability 0.1% of calibrated span Deadband 0.1% of calibrated span Temperature rating (limits) from -40° to +212°F Protection Inside vapor container Transmitter output range 4-20 ma de Location Loop B hot leg Loop C hot leg II. Pressurizer Pressure Control Signals 1700-2500 psig, 1-5 V Manufacturer Fischer & Porter Co. Accuracy +/-0.5% of calibrated span Repeatability 0.1% of calibrated span Deadband 0.1% of calibrated span Temperature rating (limits) from -40° to +212°F Protection Inside vapor container Transmitter output range 4-20 ma de b Location Pressurizer pressure control loop Condensing pot loopc III. Wide-Range Loop Temperature RTDs 0-700°F, 1-5 V, linear Manufacturer Weed Instrument Model N9001D-2B Type Platinum aSubcooling monitor channels A and B have input from separate safety-grade wide-range pressure transmitters. b Input to channel A of the subcooling monitoring system. C Input to channel B of the sub cooling monitoring system.
SPS UFSAR 7 .10-8 III. Table 7.10-2 (continued) SUBCOOLING MONITOR SYSTEM TEMPERATURE AND PRESSURE INPUTS Wide-Range Loop Temperature RTDs (continued) Accuracy of calibration +/-O.l°F Temperature range -30°F to 650°F Pressure range 6000 psi maximum Radiation 25 rad/hr Repeatability The sensor will withstand 10 consecutive temperature shocks within the design range of 2.1 without shifting ice point resistance by more than an amount equivalent to 1.0% of the actual span.
. d L ocation Reactor coolant loop hot and cold legs IV. Incore Thermocouples 0-2300°F, type K, 1/8-in. diameter Protection Stainless steel sheathed, aluminum oxide insulated RTD composition Chromel-alumel Tenperature Conforms to the "K" calibration curve with +/-2°F Range and accuracy From Oto 530°F, and within
+/-3.8% of point from 530 to 700°F Manufacturer Thermo Electric Co., Inc.
dThe RTD inputs come directly from the loops and not from the bypass loops, and are not force-flow-dependent. Therefore, even under natural circulation conditions, accurate inputs are provided for the subcooling monitor system. There is a hot leg and a cold-leg RTD in each loop. They input to the subcooling monitor system as follows: Loop A to Channel A, Loop B to Channels A and B, and Loop C to Channel B. J
SPS UFSAR 7 .10-9 Table 7.10-2 (continued) SUBCOOLING MONITOR SYSTEM TEMPERATURE AND PRESSURE INPUTS IV. Incore Thermocouples (continued) Environmental conditions Ambient air, 40° to 200°F and 100% relative humidity at electrical connection end 9 Radiation 10 h/hr maximum (neutron and gamma) Fluid Radioactive water or steam, 40° to 700°F Pressure (design) 2500 psig
. e L ocat1.on Upper core locations, each thermocouple has a specific length
- v. Reference Junction Box f Temperature Thermocouples 50-350°F, normal temperature at 160°F 2
°F = 104.245 + 4.53464X + 0.00220989X (Xis in millivolts)
Manufacturer C. J. Enterprises Model BRJW14S-36TT-80705 Composition Platinum alloy sensing probes Protection Weatherproof housing Power supply 115 V, 60 Hz Location Inside containment (outer.crane wall area) eThere are 16 individual thermocouples used for inputting into the subcooling monitor system, with eight thermocouples per channel. Two thermocouples per quadrant from diverse power supplies are fed into each channel.
£The reference junction boxes (RJB) do not provide direct signal inputs to the subcooling monitor system. The temperature inputs can be read on the process computer. The RJBs influence the actual thermocouple inputs by compensating for containment temperature. RJB #1 modifies the incore thermocouple inputs to Channel A and RJB #2 the incore thermocouple inputs to Channel B of the SMS.
REVISION 9 6/90 SPS UFSAR 8.2-3 standards: In the emergency switchgear. room,- some of the cables*have been run in trays. These ladder type trays have solid covers placed directly on the
<top of the trays and may have a so*lid tra_nsite or *asbestos blanket {see NUS-357, Criteria for installation and Identification of Eiectrical Cables for Surry Power Station - 1972 Extension) placed on the bot to~ of the trays prior to* installation of cables. This
- installation has the. same protection integrity as cable. in -conduit and* facilities installation and inspection o*f
- these critic~l cables.
Power and control cables are distributed from the switchgear and control area by means of _rigid metal conduits or ladder type cable trays._ Control cables are of single or multitonductor constructio~ witQ {nsulation rated at 600 or .1000 V and with overall flame-retardan*t jackets. Low-voltage
'instrument conn'ections are made using . flame-retardant insulated cables, rated at 300 or 600 V. These cables are provided with electrostatic *shield and an overall flame-retardant, jacket_.
a total coverage r Thenormal_current rating of all insulated conductors is limited to that continuous value which dcies not cause excessive insulation -deterioration from heating. Selection of conduct-or. s.i.zes -are based. on Power Cable Ampacities,
- published by the IPCEA.
Fire-resisting fillers, tapes, binders, and -jackets w_ere specified for all cabl~ coristructi6n.: All cable tray installations have approved fire stop~ in both horizontal and vertical runs arid ~re provided with a solid ra{sed cover or a corrugated solid aluminum cover. Covers may be omitted on top trays run under solid floors. All conduit installations consist of plastic conduit encased_ in c*oncrete or exposed rigid metal conduit. All:electrical-equipment and cables are designed .to operate within their normal rating or temperature rise. The service factor or*overload rating will not be used during normal operating conditions. All connections at the 22-kv*voltag~ level are made with isolated phase construction_ ~esigned for forced~air cooling. The station batteries are* sized to .operate circuit breaker controls,* / turbi'ne shutdown oil pumps, -instrumentation, emergency lighting, and vital
SPS UFSAR 8.2-4 nuclear* channels for- 2 hr without benefit. o( any __ station power. The battery chargers,are s;onnected to the emergency bus andprovide_charging current t:o the battery and *load when the emergency *bus is-energized. Lighting distribution and intensities have been selected in acco*r.dance with the rec:_ommendations of t:he*Illumination Engineering Society (IES).
**.1
*. I
. .,,.-.__ *~ ~- '
REVISION 9 6/90 SPS UFSAR 8.5-~ 3* High-High consequences limiting safeguards.
- 4 .. Manual.*
Cond-itions that render the diesel generator* incapable of responding t9 an automatic emergency start signal.are:
- 1. Diesel-generator output differential current fault.
- 2. Diesel-generator output overcurrent fault.
- 3. No diesel-generator field;
- 4. Overspeed.
(The above four conditions must be reset prior to*any start).
- 5. Man.ual stop from local or remote locations.
6~ Control r6om switch in "exercise" instead of "autb"r
- 7. Engine control cabinet (at the diesel)-. switch in* ..".local start" ..
- 8. Nece~sary circuit breakers in "off~ Position.
- 9. Low starting air pressure.
- 10. Less tha.n required fuel inventory.*
- 11. Low battery voltage.
Alarms and annunciators actuate in the control room and the local .. diesel-generator control panel when a fault condition associated with the diesel generator exists.* An .emergency diesel auto.:..start-disabled alarm is o~~ained in. the.control room whenever the local diesel control paneL selectoi
~witch is in the "local start" poiition or when the*"auto-exercis~" switch cin the remote control room pane.l is in the "exercise" position. . If any - of . the other disabling conditions exist, an emergency generator trouble alarm will be received in the control room.
Each diesel-generator set is sized to start and accept the .post-design-basis-acc~dent load in about 10 sec and be *loaded with the requisite loads within about 25 sec. The starting load capacity is 12,500 .kVA, the 2000-hr rat.ing is 2750 kW,- and the 2-hr rating is 2850 kW. The actual loads, using conservative rating~ for accident conditions* at Surry, require approximately 87-50
- kVA for starting* and 2320 kVA for running. The starting, accelerating, and loading* times of* the diesel generato*rs using simulated loads .were
REVISION 1 6/83 SPS UFSAR 8.5-6 .witnessed and checked befbre the units *were. accepted from the .engine manufacturer. The continued ability to accept load is tested as described in Section 8.6. If required, the emergency buses can be powered from the onsite diesel generators. The buses are protected from both a loss-of-voltage and a degraded grid voltage condition. The voltage of each bus _is monitored ori each phase with separate loss-of-voltage and degrade_d voltage relays. Each separate set of relays will provide the input to a*coincident two-out-of..:.three logic* scheme. Theundervoltage setpoints (trip values) for these two protection schemes are provided in Table 8.5-1. The system operation is described below: o On a loss-of-voltage condition (below 75+/-1% of rated voltage) the separate relays will trip and, after a time delay,* initiate an automatic transfer of the Class lE emergency buses from the offsite
. sourc~ to the diesel generator-. The time delay is 2 sec, +5 sec, 1.
* -0.1 sec and is discussed below. The emergency bus .is loaded onto
- the* diesel generator after th~ generato*r has* achieved 95% of rated volta~e an~ synchrono~s s~eed *. This is req~ired to occur within 10 sec after the signal is received.
o When a degraded voltage exists (90% to 7 5% of rated voltage), the two~out-of-three iogic scheme will initiate an alarm in the control room at. 10 sec, start the diesel generators at 50 sec and transfer the *'deenergized Class lE emergency buses from the offsite source to the diesel generators at. _60 sec.
- If a *safety:-inj ection or consequence-limiting safequards signal is concurrent with the degraded voltage, the 10-, 50-, and 60-sec time delays are bypassed and. a 7-sec time delay is used. At 7 sec . the. diesel generators.
receive a start signal and the transfer from offsite to onsite power is initiated. Upon transfer initiation for* a safety_--inject_i,_Q.!L or consequen~e-lJ.mi_ti~~L1>afeguards condition, ~he offsite source . feeder breakers to the Class lE buses, the stub bus tie breaker, the residual heat: removal and component c*ooling pumps, and a . charging pump are automatically tripped.
REVISION 1 6/83 SPS UFSAR 8.5-11 - place on failure of the normal source. The following instruments are provided in the control room to monitor emergency bus voltage performance:
- 1. Battery voltage indication.
- 2. Battery ground indication.
- 3. Emergency bus voltage.
- 4. Emergency bus frequency.
- 5. Emergency bus undervoltage alarm.
- 6. Low battery voltage alarm.
All routine control of normal and standby electrical power is from the control room. However, essential loads can also be controlled from the emergency switchgear located below the control room. The emergency switchgear is designed so that local operation is possible with or without control power. All control switches on the main control board are clearly identified by system. Emergency switchgear and control centers are identified as control devices for essential components
- The diesel-generator panel contains instruments and controls to serve the emergency bus. Provisions for synchronizing the diesel generator manually with the reserve station service power systems are also provided. The generators are manually synchronized with the system and loaded for periodic load tests. Automatic synchronization is not used.
The diesel generators and associated equipment are located in a Class I and tornado-protected structure. Each generator and its associated equipment will withstand, without loss of function, either the design-basis earthquake or the atmospheric pressure drop associated with the design tornado. Emergency switchgear is located in the shielded control area below the control room. Status of the emergency power bus can be determined at the emergency switchgear. Emergency distribution air circuit breakers can be manually operated at the switchgear (Section 7.7).
REVISION 1 6/83 SPS UFSAR 8.5-12 All switchgear associated with the electrical feeds to the emergency buses _are enclosed in metal housings and protected from the weather. These enclosures are equipped with thermostatically controlled electric heaters to prevent condensation. The majority of lines and valves required for containment isolation or engineered safeguards operation are located in protected areas. All essential electrical components and circuits are located and distributed within protected zones. All cables, conductors, motors, pumps, control stations, etc., are identified by a mark number or by function. The markings consist of painted stencils or marked tags applied or attached to each component. All lines or valves subject to freezing or crystallization of boron are electrically heat traced and insulated. The heat tracing is automatically energized when the temperature drops below 35°F for water lines or 170°F for boron lines. Therefore, icing or crystallization could not interfere with the isolation of containment or injection of coolant during accident conditions.
SPS UFSAR 9-xi
- CHAPTER 9: AUXILIARY AND EMERGENCY SYSTEMS TABLE OF CONTENTS (continued)
Section Title 9.14.3 Design Evaluation * *
- 9.14-2 9.14.4 Tests and Inspections. 9.14-3 9.15 INDEPENDENT FUEL EVALUATION SYSTEM * * * * * * * * * * * * * *
- 9.15-1 9.15.1 Design Basis. 9.15-1 9.15.2 Description. * * **** 9.15-1 9.15.2.1 Fuel Assembly Examination System ** 9.15-1 9.15.2.2 Fuel Rod Examination System * *
- 9.15-1 9.15.2.2.1 Support Stand * * * * * *** 9.15-2 9.15.2.2.2 Fuel Assembly Storage Rack. 9.15-2
- 9.15.2.2.3 9.15.2.2.4 9.15.2.2.5 9.15.2.2.6 Fuel Rod Storage Rack * * *
- Fuel Rod Visual Examination Fuel Rod Cleaning **
Fuel Rod Profilometer. 9.15-3 9.15-3 9.15-3 9.15-4 9.15.2.2.7 Fuel Rod Length Measuring Subsystem. 9.15-4 9.15.2.2.8 Fuel Rod Eddy Current Test Subsystem ** 9.15-4 9.15.2.2.9 Fuel Rod Gamma Scan. * ******** 9 .15,-4 9.15.2.3 Fuel Rod Handling System. 9 .15-5 9.15.2.4 Fuel Examination Bridge System ** 9.15-5 9.15.3 Evaluation * * * * * * *
- 9 .15-6 Appendix 9A High-Density Spent-Fuel Storage Rack Design ** 9A-l Appendix 9B Movement of Heavy Loads. * * **** *
- 9B-l
.Appendix 9C Flood Control System. * * ******** 9C-l
SPS UFSAR 9-xii CHAPTER 9: AUXILIARY AND EMERGENCY SYSTEMS LIST OF TABLES Table Title 9.1-1 Chemical and Volume Control System Code Requirements
- 9.1-39 9.1-2 Chemical and Volume Control System Principal Component Data Summary 9.1-40 9.1-3 Chemical and Volume Control System Performance Requirements. 9.1-43 9.1-4 Fission Product Concentrations in the Reactor Coolant with Small Cladding Defects in 1% of the Fuel Rods * * * * * *
- 9.1-45 9.1-5 Parameters Used in the Calculation of Reactor Coolant Fission Product-Activities 9.1-47 9.1-6 Maximum Volume Control Tank Noble Gas Concentration in Vapor Phase with Small Cladding Defects in 1% of the Fuel Rods * * *
- 9.1-49 9.1-7 9.1-8 9.1-9 Tritium Sources in Reactor Coolant Operation.
Boric Acid Storage Tank Water Chemistry. Consequences of Failures or Malfunctions of the Chemical and Volume Control System within the Reactor Containments. 9.1-50 9 .1-51 9.1-52 9.2-1 Boron Recovery System Component Design Data. 9.2-7 9.2-2 Boron Recovery System Malfunction Analysis ** 9.2-21 9.3-1 Residual Heat Removal System Code Requirements 9.3-9 9.3-2 Residual Heat Removal System Design Data * *
- 9.3-10 9.3-3 Residual Heat Removal System Chemistry Guidelines. 9.3-12 9.3-4 Residual Heat Removal Loop Malfunction Analysis * *
- 9.3-13 9.4-1 Component Cooling Water System Component Design Data. 9.4-21 9.4-2 Chilled Water System Components, Materials, and Design Conditions * * * *
- 9.4-23 9.4-3 Chilled Water Circulation Pump Data ** 9.4-27 9.4-4 Chilled Water Surge Tank Data * * * *
- 9.4-28 9.4-5 Chilled Water System Piping and Valve Data ** 9.4-29 9.4-6 Neutron Shield Tank Cooling Water System Component 9.4-7 Design Data.
- Charging Pump Cooling Water System Component Design Data.
9.4-30 9.4-32 -
REVISION 5 12/86 SPS UFSAR 9.3-3 line to the residual heat removal system is located in the hot leg of reactor coolant loop Number 1 between the main loop stop valve and the reactor core . The return line connects to the cold legs of two loops through the safety injection system. The heat loads are transferred by the residual heat exchangers to the component cooling water in the component cooling system (Section 9.4). During unit cooldown, the cooldown rate of the reactor coolant is controlled by regulating the flow through the tube side of the residual heat exchangers. A single bypass line and a remotely operated control valve around both residual heat exchangers are used to maintain a constant coolant flow through the residual he*at . removal system while controlling coolant temperature. The entire residual heat removal system is located inside the containment, with the exception ~f the line penetrating the containment that connects to the refueling water storage tank. The residual heat removal pumps are normally controlled from the control
- room. In the event of a control room evacuation, pumps can be operated at the switchgear in the emergency switchgear room.
on compliance with 10CFR50 Appendix R. See Section 7.7.2 for discussion During refueling, the water level in the reactor cavity is lowered by opening a valve at the residual heat removal pump discharge and then pumping the water into the refueling water storage tank. The normal residual heat removal line is closed during the transferral. The RHR purification flow isolation valves are equipped with quick-disconnect instrument air fittings to provide a method to locally operate the valves with a portable air source. The operation of these valves is required for decay heat removal during hot shutdown following a postulated fire in accordance with the requirements of Appendix R to 10CFRSO. The residual heat removal system is not an engineered safeguards system.
- 9.3.2.2 Components Residual heat removal system component design data are listed in
REVISION 5 12/86 SPS UFSAR 9.3-4 Table 9.3-2. 9.3.2.2.1 Residual Heat Exchangers
- The residual heat exchangers are of the shell and U-tube type, with the tubes welded to the tube sheet. Reactor coolant circulates through the tubes while component cooling water circulates through the shell side. The tubes and other surfaces in contact with reactor coolant are austenitic stainless steel, and the shell is carbon steel.
9.3.2.2.2 Residual Heat Removal Pumps The two 50%-capacity residual heat removal pumps are in-line vertical centrifugal units with special seals to prevent reactor coolant leakage. All pump parts in contact with reactor coolant are austenitic stainless steel or adequate corrosion-resistant material. 9.3.2.2.3 Residual Heat Removal System Valves The valves used in the residual heat removal system are constructed of austenitic stainless steel or other adequate corrosion-resistant materials, such as Haynes alloy 25 and 17-4 PH stainless steel. Manual stop valves are provided to isolate equipment for maintenance. Throttle valves are provided for remote and manual control of residual heat exchanger tube-side flow. Check valves prevent reverse flow through the residual heat*temoval pumps. Isolation of the residual heat removal system is achieved with two' remotely operated stop valves in series in the pipe from a reactor hot leg to the suction side of the residual heat removal pump, and by a check valve (located in the safety injection system) in series with a remotely operated stop valve in each line from the residual heat removal pump discharge on two reactor cold legs. Overpressure in the residual heat removal system is relieved through a relief valve to the pressurizer relief tank in the reactor coolant system.
SPS UFSAR 9.4-17 is then returned to the liquid waste disposal system (Section 11.2.3) via the sump pump provided within the curbed area, Welded construction is used almost exclusively throughout the system to minimize possibility of leakage from pipes, valves, and fittings. Small leakage inside the containment is not considered to be objection-able. Contamination could result from the following: side-to-side leakage in a heat exchanger in the chemical and volume control, residual heat removal, or sampling systems, or a leak in the thermal barrier of a reactor coolant pump. Leakage from the system is primarily detected by falling _surge tank level. Temperature, level, and flow indicators in the control room may be used to detect leakage at certain points. Elsewhere, leaks can be located by inspection or isolation. 9.4.4.3 Incident Control The piping mains have the following valves at the containment walls: shutoff valves outside containment and check valves inside containment in supply lines; trip and shutoff valves outside containment in return lines. The trip valves close upon receiving the containment isolation signal. Piping for the reactor coolant pumps, reactor shroud cooling coils, and containment recirculation air coolers is valved in an identical manner (Section 5.2.2). The temperature of cooling water supplied to the reactor containment recirculation air coolers should not exceed 70°F; this means that during periods of warmer river water, these coolers use chilled water. The cooler inlet and outlet lines are sized large enough so that abnormally high flows of regular component cooling water can be accommodated with the globe valves wide open. With this added flow, the coolers can control containment pressure rise caused by a minor incident or pipe break, thus avoiding use of the safeguards system (Chapter 6.0). The transfer, or supply and return between the two systems, is accomplished by the use of air-operated flow-diverting valves. These valves are operated remote manually by means of a switch mounted on the ventilation panel in the control room. In warmer weather, the containment air coolers remain on chilled water supply unless a minor incident occurs.
SPS UFSAR 9.4-18 9.4.4.4 Component Cooling Makeup Water Makeup for the component cooling water system is provided from the discharge of the main condensate pumps, which draw on the condenser hotwell. Operation of engineered safety features will not be affected by loss of makeup water to the component cooling water system if offsite power is lost, since the component cooling water system is not required for operation of engineered safety features. During normal prolonged outages of both units (with station power available), a separate makeup pump supplies makeup water to the closed-loop component cooling and bearing cooling systems. To maintain component cooling water to the fuel pool cooling system (Section 9.5), a component cooling water pump can also be operated from the emergency electrical bus during a complete loss of off site power. The component cooling water system is closed, and leakage from this system will be at a very slow rate. In addition, the component cooling water surge tank is a source of reserve water, which must be exhausted before makeup to the component cooling water system is required. If the component cooling water system requires makeup before offsite power is restored, a portable pump can be connected to supply makeup. 9.4.4.5 Cooling Water Support for Other Systems In the event of a single failure in the component cooling water system (e.g., at the discharge header), cooling water for the reactor coolant pumps, the excess letdown heat exchanger, the residual heat removal system, and the nonregenerative heat exchanger would be lost. The charging pumps would not be affected because they have been provided with a separate system. In the unlikely event of a total loss of component cooling water, the operator would bring the reactor to a safe shutdown, or hot standby condition, with all reactor coolant pumps tripped and letdown flow discontinued, but with charging pumps operating to supply seal-water injection flow to the reactor coolant pump seals. This condition can be maintained until the pressurizer
-* SPS UFSAR 9.9-3
- 3. * !n the event of a loss -of station power in two units, the total flow*
o*f 18 ~ 000 gpni ~ould be required to cool down the units. This would require two of the three pumps to* be operated *. 9.
9.2 DESCRIPTION
Service water is supplied from the circulating* water system
.(S_ection 10.3~4)_ by gravity* fiow between the high-:-level int.ake
- canal and discharge canal **seal pit". During normal. operation, the water level* i~ the intake canal is approximately_ 20 ft above the* l_evel in the* seal_ pit at* the di~charge canal. This* differential head supp:J.ies - the service ~ater_. to
- par:allel flow .. loops through the bearing cooling water h_eat exchangers,
- . component cooling heat exchangers, ind recirculati~n spray- heat exchangers, . * * -
which are also .in parallel with tp.e main condenser *
. Remotely operated b~tter:fly valv:es are installed at the four inlets* and
_- outiets oL.each main condenser a~d in the supply._: lines to. the* bearing coo~ing
.water heat e~changer~, the :.component cooling ~eat e~changers, a~d the re~ircul~tio~: ;pray h~at excha~gers ~ The oper~tion of th~se* val;es i~ listed _ . iri -Tabie- 9. '~-1. Thes_e motor-operated valves - are positioried, atitoma.tical.ly'*:_for various *incident c~nditibns,* to conserve*approximately 25,000,000.gal of water in the intake canal for _"critical services.*
Power for these ~aives. is. normally . fro~ th~: static~. power supply;
*howe~er,
- in th~* event of *a loss :of power, !)rovis'ions have been made. to supply.
the: valves .. with power: f ram the emergency gene_ra tors.
-Trash* 'racks* have. been installed to prevent large pieces *_of trash from" ent_ering the fattak~ struc-ture which couid ad~e;sely affect: the o~eration *of .. th~. emergen~y serv'ice*. water pumps~ . ?irice. each bay of the intake structur:e i's
. -~-ized for a.. t~tat.flow o_f 210,000 gpm and each emergency servic~ wat~r** pump.* delivers* 15,000 gp~~ sufficient water will be provided to .. the emergency water.*
- _ pumps .as -long a_s *approximately 7 .5%* of the flo.w area of the .. racks remains
**clear.
REVISION 9 6/90 SPS UFSAR 9.9-4 In the event of a. power failure simultaneous with the accumulation of trash at the trash racks, accumulated trash can be removed from the screens of the station intake .by manual raking_. This procedure could be done indefinitely if necessary although it is expected that the duration of the loss of power would be relatively short, i.e., less than 1 week. The maximum service water requirements of the system during normal operation are given in Table 9.9-2. 9.9.3 DESIGN EVALUATION The following ~omponents of the auxiliary cooling systems are required f6i p~rformance of.the engineered safety features: MOV.,-SW-103A, B~ C, & D -
- Motor-operated _vcl,lves . that* admit service water to .the recirculation spray cooleis.
MOV-CW-106A, B~ C, & D - Motor.--opeiated _valves that. stop water flow MOV-CW-lOOA, B, C, & D. - MOV-SW-102A & B to the main condenser. Moto~-operated valves that stop water flow from the main tondenser diBcharge. Motor-operated valves that stop service water to. component cooling_ water heat exchangers. MOV-$W-101A & B* Mo.t~r-operateci. valves
- that .stop service water to the
- bearing cooling water heat excJ:)angers.
SW-P-lOA & b Charging pump service water pumps to supply cooling water to the charging pump cooling water system. CC-P-2A & B Charging pump cooling water pumps* that circulate the component cooling water of* the charging pump co'oling water system. 1-cw-Ls~102 & 103 Canal level switches which provide a signal to 1solate non-essential flows from the intake c*anal .. 2-CW-LS-202 & 203 Canal level switches which provid~ a s~gnal tc is::,.1..a,te non-essentia*l flows from the intake canal,.
REVISION l 6/83 SPS UFSAR' 9.9-5 The . associated in.strumentation *and*_ power. s*ystems for the operation of
- these *components _are redundant,* an.d have protected* power and. c*ontrol circuits in.conformance to IEEE-279 and 10 CFR 50 General Design Criteria.
The components* themselves a,re redundant *except for .motor-9perated *valves
. ' ~ .
* *MoV-SW-102A & B and MOV-SW-lOlA & B, which. stop. service _water. to the -component cooling water heat exchangers ari.d the bearing ~ooling water h~at * *e~changers *.
- M~tor-operated valves MOV-SW-192.A & Bare.in parallel pipelines, as aremotor-operated valves M~V-SW-lOlA & B. Failure of one of these valves will allow service water t*c;:, esca,pe from* the service water canal* through. the -~ompon~nt _*
cooling water heat exchc1:ngers or the. be_ar:i.ng cooling water he~t exchanger.s.
-.. However, in the event of failure oi _one of these motor-:_op~r:ated valves,. manual
- valves that are accessible immediately following a design~basis
- accident are provided* to. isolate* the service water pipel_ines *to the _bearing_ cooling water*
'. 'heat exchangers *and the components cooling heat exchaI?,gers. thereby conl;erving
-.'ali the.water in the*s_e.rviC:e water cana) for'*the r_ecirculatio*n_ sp:ray coolings.
In the *event of the.design-basis* accident, the valves fn the. supply lines to the bearing .cbo_ling. ~ater 'heat exchange_rs_ and comp,onen1: cooling .heat.
. *exchangers may be reopened remote-manually_ from 'the control room .j,f SE:rvi~e.
-water to these_systems is considered necessary.
Automatic* _*temperature control of. the* charging pump lube_.oil _systems_- is: provided by the* use* of . air-operated: contra*1 . .* vaives. 1:hes~ va1v~s are _,* '. _installed ,in* the service. wate*r oudet of ~a~h* lube-o:!-1 cooler, . Capillary type.. .1.
*thermal elements* are installed in. the oil lin,es which provide the Si!;nal. t_b a ..
. pneumatic-indicating temperature . controller with* the _output signal _operating the* co'ntrolvalve.
The
- piping and equipment movements at 'the: recirculation.-* spray heat*.
exch~rtgers have' beeri analyzed in ac~ordan.ce with earthquake design *criteria '
.and have been .installed to ensure that no 'undue forces* are ex.erted . 0~ piping or equipment nozzles *.
The gravity flow of* service *water .. from .the. intake
- canai ensures adi::quate
. co.~l:!-~g water* to the recirculation spray *heat exchangers. for m~re tha~, 24 \1r
* ** in ,the case :of the design:.:.basis accident;* if *water is not useci for *component.
-~ . -
REVISION 9 6/90 SPS UFSAR 9.9-6 or bearing cooling. This supply of cooling. water is based operation_ with two half-*c_apacity recirculation spr_ay heat exchangers. on expected that before the end of this period, normal power will be restored to the circulating water pumps to replenish the intake canal water supply. one unit It is Three diesel-driven emergency service* water pumps are provided in the event that power is.not restored within the 24~hr period. If the second reactor unit is being*refueled and the reactor head is not in place, continued core ~ooling is required in the event of a loss of statioh power. Service water flow: to one component cooling heat exchanger will be co*ntinµed and the _cooling water supply in the intake will* last for approximately 12 hr without using 'diesel-driven emergency *service water pumps .. A 48-gal diesel fuel-oil storage tank provides sufficient fuel to operate all three pumps for 125 hr. One.'diesel-o.riven pump is also electric-mo.tor-driven. All pumps can be started_ locally. The electric-motor-driven pump is normally .. used to maintain the intake* canal full of water when. both units are shut down~ water The possibility o*f leakage from th~ reactor containment into t,he *service through* the recircu1ation _spray heat exchangers after a LOCA is discussed in Section 6.3,1. 9.9.3.1 System Reliability A double set of normally closed parallel motor-operated valves control the service water supply to the recirculation spray heat exchangers, thus providing positive assurance that cooling water can r~ach the exchangers in the event of a malfunctioning valve. Three diesel-driven emergency service water pumps_ are fu*rnished to provide makeup to the intake canal during a loss of off-site power. The three diesels are each equipped with an alternator capable of .maintaining a float charge on the starting batteries during a loss of_ off-site power. Blocking diodes are provided to isolate the safety rel~ted batteries from the non...: safety related battery charg~rs. J
REVISION 7 3/88 SPS UFSAR 9 .10-9b 9.10.2.3 Ventilation Systems ~ 9.10.2.3.1 Smoke Removal No special smoke-exhausting systems are provided at the plant. The normal ventilation systems can be used for smoke removal for some types of fires, even though they are not specifically designed for this purpose. When normal ventilation systems .cannot be used, the fire brigade will use portable ventilation units with flexible ducting for smoke removal. 9.10.2.3.2 Filters Charcoal filters are used in the auxiliary building ventilation system filters, the containment iodine charcoal filter units, the gaseous waste disposal system, and the control room emergency ventilation system. The auxiliary building ventilation system contains redundant trains of charcoal filters housed in separate meta 1 cabj nets enclosed in separate concrete
- cubicles. The inlet and outlet dampers on these filter units can be shut to prevent radiation rel ease from a damaged unit and the redundant unit can continue to operate. These units are currently protected by a manually actuated carbon dioxide suppression system. The containment iodine charcoal filter units are enclosed in separate structures with 18-in.-thick concrete wa 11 s and roof. A fire in these filters would have no direct effect on safety-related equipment or cables because of the intervening distance and barriers. The gaseous waste disposal system charcoal filters are housed in a metal enclosure away from safety-related equipment and cables. These filters
REVISION 1 6/83 SPS UFSAR 9.10-10 can be isolated and air-cooled if subjected to excessive decay heating. The control room emergency ventilation system charcoal filters are located in the turbine building in a metal enclosure. These filters are isolable from the control room by a normally shut motor-operated valve in the supply pipe to the control room. The only safety-related cable located near the control room emergency ventilation charcoal filter is the power feed to the respective fan motor; therefore, a fire would not affect safe shutdown of the plant. 9.10.2.3.3 Breathing Equipment There are 20 self-contained air-breathing apparatuses dedicated at all times to fire brigade use. There are in excess of 50 spare air cylinders of 30-min capacity available in the plant. Self-contained units are distributed at various locations throughout the plant, with four sets kept at the control room. Air recharging for the fire brigade cylinders is from a six-cylinder j1 cascade system that is recharged by an air compressor designed for that purpose. Recharging is carried out at a location in the south annex building outside the main security area of the plant. 9.10.2.4 Floor Drains In general, measures have been taken to prevent the spread of combustible liquids in the event of leakage from reservoirs and piping. The lube-oil reservoir, lube-oil cooler, and high-pressure control fluid reservoir for each unit are located in a diked area without floor drains. The hydrogen seal oil unit is surrounded by a spillage trench to prevent the spread of lube oil. The two lube-oil storage tanks are located together in a diked area without drains. The drains in the storage tank room outside the dikes are plugged. The fuel tank for the diesel-driven emergency service water pumps is located in a separate diked room, without floor drains, in the intake structure. These measures are adequate to contain leakage to the area of origin.
REVISION 6 6/87 SPS UFSAR 9.12-15 e 4. The manipulator crane is parked and secured.
- 5. The reactor vessel head is picked up by the reactor containment polar crane and positioned over the reactor vessel.
- 6. The reactor vessel head is slowly lowered as the water level is lowered. The water level is lowered by opening a valve at the residual heat removal pump discharge, and water is pumped from the reactor cavity into the refueling water storage tank. The normal residual heat .removal line is closed while pumping to the tank.
- 7. When the reactor vessel head is about l ft above the flange, the reactor cavity is drained. When the water in the reactor *cavity is slightly below the vessel flange lever, the valve to the refueling water storage tank at the residual heat removal pump discharge is closed. The normal residual heat removal operation is restored, and most of the remaining water in the. reactor cavity is pumped to the storage tank by the reactor cavity purification pumps. The small amount of water remaining is drained to the containment sump and then pumped to the liquid waste disposal system.
- 8. The reactor vessel head is seated.
- 9. The guide studs are removed to their storage rack. The stud hole plugs are removed.
- 10. The head studs are replaced and retensioned.
- 11. The fuel transfer tube cover is replaced.
- 12. Electrical leads and cooling air ducts are reconnected to the control rod assembly drive mecha~isms.
-- 13. 14. Vessel head insulation and instrumentation leads are replaced. A hydrostatic test is performed on the reactor coolant system.
REVISION 1 6/83 SPS UFSAR 9.12-16 11 15. 16. Control rod assembly drive operation is checked. The control rod assembly drive mechanism missile shield is picked up with the reactor containment*crane and replaced.
- 17. Preoperational startup tests are performed.
9.12.S.S Handling of Failed-Fuel Assemblies A fuel assembly considered to be leaking is placed in a failed-fuel can located in the spent-fuel pool, and the can is sealed to provide an isolated chamber to test for the presence of fission products. The failed-fuel cans are stainless steel, with lids that can be bolted in place remotely. An internal gas space in the lid provides for water expansion and for collection and sampling of fission product gases. Various remotely operable quick-disconnect fittings permit connections of the can to sampling loops for continuous circulation through the can. If sampling confirms the presence of fission products indicative of a cladding failure, the sampling lines are closed off by valves on the can, and the encapsulated fuel assembly is stored in the spent-fuel storage pool to await shipment. Design of the can complies with shipping cask design requirements, so that the defective fuel can be stored and shipped while sealed in the failed-fuel can. 9.12.6 FUEL HANDLING SYSTEM DESIGN EVALUATION Underwater transfer of spent fuel provides essential simplicity and safety in handling operations. Water is an effective, economic, and transparent radiation shield, and a reliable cooling medium for removal of decay heat. Basic provisions to ensure the safety of refueling operations include the following:
SPS UFSAR 9.14-1 9.14 DECONTAMINATION FACILITY The decontamination facility (Figure 9.14-1) is a poured-concrete and concrete-block structure on the north side of the fuel building, under the fuel cask trolley rails. This location makes it accessible for transporting in and out of the building major items to be decontaminated. Roof hatches and a rolling steel door provide access for equipment. 9.14.1 DESIGN BASES The facility is designed to provide an area in which equipment can be decontaminated without releasing activity to the environment in an uncontrolled manner. Decontamination procedures are specified to reduce surface contamination to a level such that the components can be handled in a safe manner. 9.
14.2 DESCRIPTION
- The decontamination building is a poured-concrete and concrete-block building abutting the east end of the fuel building's north wall. A 125-ton trolley runs through a high-bay portion of the building immediately adjacent to the fuel building, and over the roof for the remainder of the decontamina-tion building. Three roof hatches permit casks or other objects to be lowered from the trolley into the building. A tramrail in the building permits the movement of small parts between ~ork areas and tanks with minimum personnel exposure. AT-shaped rolling steel door encloses the high-bay area from the outside when the trolley is not in use. The fuel building and decontamination building are separated by a weathertight structural gap to permit independent motion of the buildings in the event of an earthquake.
Ventilation air is exhausted from the decontamination building through the monitored ventilation vent system. On a high alarm by ventilation vent monitors, the decontamination exhaust is remote-manually diverted through charcoal filters as described in Section 9 .13. The exhaust capacity is greater than the supply capacity of this system, thus producing a slightly negative pressure in the buildings, so that all air leakage is inwards.
SPS UFSAR 9.14-2 Personnel normally enter the decontamination building by passing through the fuel building from the auxiliary building. An emergency exit from the decontamination building is provided. Liquid wastes from decontamination work are piped to the liquid waste disposal system (Section 11.2.3) for processing. The interior surfaces of the building are covered with suitable materials to permit easy decontamination. A stainless steel pad is provided to protect the floor under heavy objects. Hose connections are provided for compressed air and primary-grade water at each work area. An ultrasonic cleaning tank is provided for immersion cleaning of small components. This equipment permits decontamination of equipment by wash-down combined with scrubbing and soaking with or without agitation. The various decontamination methods provide a flexibility that will give the best decontamination for a specific job, minimize personnel exposure, and limit the release of radioactive material to the environment. Technical information on the equipment provided in the facility is given in Table 9.14-1. Working personnel will be monitored by health physics monitors to ensure that established exposure limits are not exceeded. A contaminated solution holdup tank is provided to receive spillage from equipment, runoff from cleaning operations, and disposal of cleaning solutions. Jhis tank has a pump for transferring liquid to the liquid waste disposal system. A filter is provided for preliminary cleanup of the fluid prior to pumping out. 9.14.3 DESIGN EVALUATION The facility provides a contained area with all discharges controlled to prevent the inadvertent release of activity to the environment. In the event of leakage from piping or equipment, all areas of the building are provided with sumps to which fluids will drain. The sumps discharge to the liquid waste disposal system. Airborne particulate matter is retained within the building because of the slightly subatmospheric pressure, and is discharged in a controlled manner through the monitored ventilation vent.
SPS UFSAR 9.14-3 9.14.4 TESTS AND INSPECTIONS Periodic tests are conducted on the radiation detection equipment in the ventilation system. Operating equipment and storage tanks are subjected to periodic visual inspections
- SPS UFSAR 9.14-5 Table 9 .14-1 DECONTAMINATION FACILITY COMPONENT DATA Contaminated solution hold-up tank Number 1 Capacity, gal 2,000 Design pressure, psig 30 Design temperature, °F 150 Operating pressure Atmospheric Operating temperature Ambient Material ss 3161 Design code ASME VIII Contaminated solution hold-up tank filter Number 1 Retention size, m 10-25 Filter element material Synthetic fiber or metallic Capacity normal, gpm 20 Capacity maximum, gpm 20 Housing material ss 3161 Design pressure, psig 150 Design temperature, °F 150 Design code ASME VIII Contamination solution hold-up tank pump Number 1 Type Centrifugal Motor horsepower, hp 1 Seal type Mechanical Capacity, gpm 20 Head at rated capacity, ft 50 Design pressure, psig 100 Materials Pump casing ss 316 Shaft ss 316 Impeller ss 316
SPS UFSAR 9A-iii APPENDIX 9A: HIGH-DENSITY SPENT-FUEL STORAGE RACK DESIGN TABLE OF CONTENTS Section Title 9A. l DESIGN BASES . * . * * . * * * * * * * * * . . * * * * * * .
- 9A-l 9A.2 STORAGE RACK DESCRIPTION. * * * . . * * * . * * * . *
- 9A-3 9A.3 STORAGE RACK EVALUATION. * * * * * * * * * . * * * * *
- 9A-7 9A.3.l Structural and Seismic Analysis ** 9A-7 9A.3.l.l Applicable Codes, Standards, and Specifications. 9A-8 9A.3.l.2 Loads and Load Combinations. 9A-9 9A.3.l.3 Design and Analysis Methods. 9A-10 9A.3.l.3.l Static Analysis ** 9A-10 9A.3.l.3.2 Dynamic Analyses 9A-ll 9A.3.l.3.3 Thermal Growth. 9A-13 9A.3.l.4 Structural Acceptance Criteria. 9A-13 9A.3.l.5 Results of Analysis ** 9A-14 9A.3.2 Nuclear Analysis * * * . 9A-15 9A.3.2.l Design Criteria and Assumptions .
- 9A-16 9A.3.2.2 Configurations Analyzed * * . 9A-17 9A.3.2.2.l Normal Configurations. 9A-17 9A.3.2.2.2 Abnormal Configurations ** 9A-17 9A.3.2.3 Calculational Methods ** 9A-17 9A.3.2.3.l Analysis of Cases . . 9A-17 9A.3.2.3.2 Codes.
- 9A-18 9A.3.2.3.3 Benchmark Calculation for Diffusion Theory 9A-19 9A.3.2.4 Results of Analysis * *
- 9A-19 9A.3.2.4.l Normal Configurations. 9A-19 9A.3.2.4.2 Abnormal Configurations *. 9A-20 9A.3.2.4.3 Transport Theory Check 9A-21 9A.3.2.4.4 Maximum keff Values *. 9A-21 9A References . * . * * . * * . . 9A-22
SPS UFSAR 9A-iv APPENDIX 9A: HIGH-DENSITY SPENT-FUEL STORAGE RACK DESIGN LIST OF TABLES Table Title 9A-l Combined Stress Surrnnary - Fuel Racks. 9A-23 9A-2 Summary of Stresses * * * . ... ... 9A-24 J
SPS UFSAR 9A-1 Appendix 9A HIGH-DENSITY SPENT-FUEL STORAGE RACK DESIGN 9A.1 DESIGN BASES The high-density spent-fuel storage racks are designed to provide storage locations for up to 1044 fuel assemblies, and are designed to maintain the stored fuel, having an equivalent uranium enrichment of 3.5 wt% U-235 in uo 2 , in a safe, coolable, and subcritical configuration under all conditions
- SPS UFSAR 9A-3 9A.2 STORAGE RACK DESCRIPTION Each fuel storage rack consists of a 6 x 6 array of fuel storage cells, which are square stainless steel boxes spaced nominally 14 in. on centers.
The rack is shown on the general arrangement drawing, Figure 9A-l. The fuel storage rack has two basic components: the support structure and the fuel storage cell. The support structure consists primarily of the four corner storage cells, which interface with the spent-fuel pool floor pads, and two horizontal grid members, which are supported by the four corner cells and which maintain the horizontal position and vertical alignment of the remaining 32 (inner) storage cells. The inner storage
- cells rest directly on the spent-fuel pool floor. Diagonal bracing is provided on the structure to accommodate the loads imposed by rack installation, by fuel handling, and by seismic events.
Horizontal seismic loads are transmitted from the rack structure to the spent-fuel pool floor through restraint devices that capture the existing
- spent-fuel pool floor pads and mate with the fuel rack structure corner cells.
The restraint devices also permit leveling of the fuel racks, and require no modification of the existing fuel rack support pad. The vertical seismic loads are essentially transmitted directly to the pool floor by each storage cell. No bracing to the pool wall is required to support the racks during a seismic event. The racks, however, are connected to each other at the top grid to preclude potential uplift. 2 Each corner storage cell is nominally 9.44 in. (o.d.) by approximately 172 in. long, with 0.250-in. walls. Each of the 32 inner storage cells is 2 nominally 9.12 in. (o.d.) by approximately 170 in. long, with 0.090-in. walls. The cells are flared at the top to aid in insertion of the fuel assembly into the cell. Attached to the bottom of each cell are four stainless steel posts that support the fuel assembly. The posts attached to the 32 inner cells rest directly on the floor of the spent-fuel pool and space the cells off the pool floor a sufficient distance to ensure adequate area for cooling flow. To accommodate any unlevelness in the pool floor liner, the rack is designed to permit the inner cells to move vertically within the rack
SPS UFSAR 9A-4 structure (a +/-1-in. motion is provided). The inner cells, however, are positively locked into the support structure so that they cannot be inadver-tently lifted out of the rack. The corner cells rest on adapter plates. The adapter plates are keyed to the existing rack stops, and the corners of the fuel storage cells are keyed to the adapter plates through 1-5/8-in.-diameter restraint pins. For installation purposes, a nominal clearance of 1/16 in. is provided all around between the restraint pin hole in the corner storage cells and the restraint pin, and between the clearance cutouts in the adapter plates and the existing rack stops. The clearance also provides sufficient allowance for thermal expansion. Horizontal seismic loads are transmitted from the rack structure to the existing rack stops at each corner of the rack through the adapter plates and pins. The racks cannot slide during any design-basis seismic event. There is no interference between the spent-fuel storage racks and the gates and tools in storage within the pool. All of the equipment stored within the fuel pool, except the fuel pool gates, weighs less than a fuel assembly. Therefore, any possible interaction between these tools and the fuel racks would be less severe than interaction between a fuel assembly and the spent-fuel storage racks, which has been analyzed. The fuel pool gates, however, weigh 3300 lb, which is more than a fuel assembly. These gates are stored so as to be captured at both the top and the bo~tom, making interaction between the gate and the spent-fuel storage racks ve~y unlikely. The rack grids maintain the horizontal position of the inner cells relative to each other and the corner cells so that impact between inner cells and/or corner cells is not possible. Each grid consists of welded 4 in. by 1.5 in. by 3/16 in. channels forming square openings in which the inner cells are placed. The grids are welded to the top and bottom ends of the heavy wall (O. 25-in. thick) corner storage cells to form the basic rack structure. Diagonal bracing welded to the corner storage cells completes the rack structure and provides the lateral and torsional rigidity to accommodate seismic and installation loads.
SPS UFSAR 9A-5 At each grid elevation, four angle clips capture the corners of each e inner cell. These clips are welded to the channel members of each grid to maintain pitch and vertical plumbness. A slight clearance is provided between the clips and
- the cells (1/64-in.
- maximum for each clip) to facilitate fabrication and to permit vertical movement of the inner cells. Such vertical movement does not introduce any stresses/deformations in the rack structure or the inner storage cells, since each inner cell can move freely past the grid retaining clips to sit directly on the pool floor. The design permits the vertical loads for each inner cell to be transmitted to the pool floor. It is necessary to limit the vertical travel of the inner storage cells to prevent (1) removal of a cell during fuel-handling operations (e.g., stuck fuel assembly load case) and (2) a cell dropping out of the rack during rack installation/removal. Mechanical stops welded to each inner cell limit the total vertical travel to about 2 in. (+ 1 in.). These stops will support the weight of the fuel cell plus a fuel assembly if necessary.
A fuel assembly guard structure is provided to prevent a fuel assembly from being brought up against the side of the peripheral fuel racks wherever the space between the fuel racks and the fuel pool walls is sufficient to insert an assembly. The structure is a 4 in. by 2 in. by 3/16 in. angle welded to the outside channel of the upper grid. With this structure ~n place it will not be possible to move a fuel assembly closer than approximately 8 in. to stored fuel, thereby maintaining a pitch in excess of 17 in. for this condition. The guard structures are required on the east and west sides of the storage rack array, and on the two racks adjacent to the Unit 2 refueling canal. The space between the fuel racks and the north or south walls is not sufficient to insert a fuel assembly.
SPS UFSAR 9A-15 The following load combinations were considered:
- 1. Hydrostatic+ dead load+ live load.
- 2. Hydrostatic+ dead load+ live load+ operating-basis earthquake.
- 3. Hydrostatic+ dead load+ live load+ safe-shutdown earthquake.
- 4. Hydrostatic+ dead load+ live load+ high-density racks.
The allowable stresses are based on the minimum sampled coupon strength of 43,600 psi and the acceptance criteria stated in ACI 318-63. It should be noted that with the new high-density fuel storage racks, the mat loadings are lower than those originally calculated. This is due to the different analytical model used. For the high-density fuel storage rack loadings, the model accounted for the detailed location of both the pilings and the fuel rack embedments in the mat. This resulted in a significant portion of the load due to spent fuel being transmitted to the pilings without inducing mat bending. In the analysis for the original loading, the rack loads were spread uniformly over the mat, and the pilings were lumped at discrete locations that were further apart than the actual pile spacing. The method used to calculate the mat loadings from the new high-density fuel storage racks represents the as-built condition at Surry. As given in Section 9. 5, the spent-fuel pool temperature will be maintained at or below the original limits of 140°F (normal case) and 170°F (abnormal case). Therefore, the maximum temperature of the thermal gradient in the pool walls and the base slab as originally designed will not be exceeded. 9A.3.2 NUCLEAR ANALYSIS A detailed nuclear analysis was performed to demonstrate that for all
, anticipated normal and abnormal configurations of fuel assemblies within the fuel storage racks, the k effective of the system is substantially below 0.95.
Certain conservative assumptions about the fuel assemblies and racks were used in the calculations. These assumptions are described in Section 9A.3.2.l.
SPS UFSAR 9A-16 The principal calculational method used for the criticality analysis was diffusion theory using the HAMMER and EXTERMINATION codes. Verification calculations were done by transport theory using GGC-3 and DOT. A detailed description of the calculational method and codes is presented in Section 9A.3.2.3, together with a description of a benchmark of the diffusion theory method. The results of the criticality analysis are presented in Section 9A.3.2.4. 9A.3.2.l Design Criteria and Assumptions The criticality design criterion established for the Surry Power Station spent-fuel racks is that the multiplication constant (keff) shall be less than 0.95 for all normal and abnormal configurations, as confirmed by transport theory. The following conservative assumptions were used in the criticality calculations performed to verify the adequacy of the rack design with respect to the rack design criteria:
- 1. The pool water has no soluble poison. *
- 2. The fuel assemblies have no burnable poison.
- 3. The fuel is fresh and of a specified enrichment as high or higher than that of any fuel available.
- 4. The rack configuration is infinite in all three dimensions.
- 5. No credit is taken for structural material other than the fuel can.
- 6. All fuel cans are assumed to be .085 in. thick, the minimum allowable thickness.
SPS UFSAR 9A-17 9A.3.2.2 Configurations Analyzed The various configurations of fuel within racks that are possible are classed as either normal or abnormal configurations. Normal configurations result from the placement of fuel within racks and the variation in rack dimensions permitted in fabrication. Abnormal configurations are typically the results of accidents or malfunctions such as seismic events, malfunction of the fuel pool cooling system, etc. 9A.3.2.2.1 Normal Configurations The normal configurations analyzed were: a reference configuration at the nominal conditions (except wall thickness, which was specified at the minimum value permitted in fabrication), one case with reduced cell pitch, and an eccentric assembly positioning case. A transport theory analysis of the reference configuration was done to establish a diffusion to transport theory bias .
- 9A.3.2.2.2 Abnormal Configurations The abnormal configurations analyzed were a variation fro~ the maximum water density at near 40°F to a partly voided situation at 250°F ( the approximate boiling point near the bottom of the rack), and a configuration where a complete can and fuel assembly were displaced due to failure of the retaining clips. Inadvertent placement of a fuel assembly adjacent to a fuel rack was not analyzed, since structure is provided on the peripheral racks where required to maintain a center-to-center spacing in excess of 17 in., a clearly acceptable value (keff is less than 0.87).
9A.3.2.3 Calculational Methods 9A.3.2.3.1 Analysis of Cases The reference configuration was analyzed by representing a rack with 14-in. cell pitch, 3.5 wt% fuel assemblies, .085-in. wall stainless steel cans, and a 68°F water temperature, by a quarter assembly configuration with
SPS UFSAR 9A-18 reflecting boundary conditions. The configuration with the minimum pitch of 13-15/16 in. was also analyzed using the quarter assembly configuration. The effect of pool temperature variation was analyzed by running the reference configuration at temperatures of 40°F, 90°F, 150°F, 212°F, 250°F, and voided at 250°F. The eccentric configuration was studied by a full assembly problem representing an assembly placed in the corner of a can. This is a conservative representation because it actually represents the entire rack array having all groups of adjoining assemblies placed as close as possible to one another. The effect of displacement of a single can due to a broken clip was analyzed by calculating keff for an array where every other assembly in every other row is displaced approximately 0.25-in. along the row. Cross sections for these cases were obtained by calculating fuel lattice cell and guide tube lattice cell problems using HAMMER for each temperature and water density required. The stainless steel cross sections were derived from microscopic cross sections from the moderator region of the fuel lattices. Water cross sections were obtained from water region edits in the guide tube lattices. 9A.3.2.3.2 Codes HAMMER (see Reference 1) is a multigroup integral transport theory code that is used to calculate lattice cell cross sections for diffusion theory codes. This code has been extensively benchmarked against D o and light water 2 moderated lattices with good results. EXTERMINATOR (see Reference 2) is a two-dimensional multigroup diffusion theory code used with input from HAMMER to calculate keff values. Cross sections for transport theory calculations are calculated in the fast groups by GGC-3 and in the thermal group by HAMMER. GGC-3 (see Reference 3) is a consistent Bn or P 1 code for the calculation of fast-neutron e
SPS UFSAR 9A-19 spectra and associated multigroup constants. Resonance calculations are performed by Nordheim methods. DOT (see Reference 4) is a two-dimensional multigroup discrete ordinate transport theory code with general anisotropic scattering used with GGC-3 and
.HAMMER input to calculate keff values.
9A.3.2.3.3 Benchmark Calculation for Diffusion Theory The HAMMER and EXTERMINATOR combination has been benchmarked against a cold critical experiment performed at the Lacrosse Boiling Water Reactor with excellent results (see Reference 5). The calculated keff differed from the experimental value by only 0.0017. This critical experiment was similar to the configuration used in the fuel storage racks, in that the fuel was enclosed in stainless steel shrouds and water gaps existed between these shrouds. ~ 9A.3.2.4 Results of Analysis 9A.3.2.4.1 Normal Configurations The following results were obtained for the normal configuration: Description 14-in. pitch, 68°F 0.87717 13-15/16-in. pitch, 68°F 0.87893 14-in. pitch, 68°F eccentric 0.88287 fuel placement
~k due to reduced pitch +0,00176
~k due to eccentric fuel +0.0057 placement
SPS UFSAR 9A-20 The worst-case normal configuration keff is obtained by statistically combining the effects of the normal variations.
~keff = [(.00176) 2 + (.0057) 2 ]~ = +.00597 Worst-case normal configuration keff 0.87717 + 0.00597 = 0.88314 9A.3.2.4.2 Abnormal Configurations The following results were obtained for the abnormal configurations:
DescriEtion keff 14-in. pitch, 39°F 0.87753 14-in. pitch, 90°F 0.87843 14-in. pitch, 150°F 0.88078 14-in. pitch, 212°F 0. 88310 14-in. pitch, 250°F 14-in. pitch, 250°F voided to .925 gm/cc 0.88334 0.88104
~k due to increased +0.00617 pool temperature For the displaced can problem, a different mesh arrangement was required with larger mesh sizes, so a base case was also run:
DescriEtion 14-in. pitch, 68°F, 0.87860 no displacement 14-in. pitch, 68°F, 0.87891 1 can displaced .25 in.
~k due to can +0.00031 displacement
_J
SPS UFSAR 9A-21 The worst-case abnormal configuration combines the worst-case normal configuration with the most adverse abnormal condition (pool temperature rise). Worst-case abnormal configuration keff = 0.88314 + 0.00617 = 0.88931 9A.3.2.4.3 Transport Theory Check The reference case at 14-in. pitch and 68°F with .085-in. cans was recalculated using GGC-3 and DOT. The keff value was .91279. This establishes a diffusion-to-transport bias of .91279 - .87717 = +.03562 k. Worst-case normal configuration keff = 0.91876 (with bias) Worst-case abnormal configuration keff = 0.92493 (with bias) ~ 9A.3.2.4.4 Maximum keff Values Using a criticality calculational uncertainty of .01 combined statistically with the worst-case abnormal configuration, the upper limit on .,I i keff becomes 0.93061. This is well under the design criterion of about 0.95 and would remain so even with much larger uncertainty values.
SPS UFSAR 9A-22 9A REFERENCES
- 1. J.E. Sutch and H. C. Honeck, DP-1064, The HAMMER System, January 1967.
- 2. T. B. Fowler et al., ORNL-4078, EXTERMINATOR-2, April 1967.
- 3. Report No. 320-3254, ISB/GGC3, IBM Scientific Center, Palo Alto, California.
- 4. WANL-TME-1982, DOT IIW User's Manual, December 1969.
- 5. R. J. Weader, NES 81A0260, "Criticality Analysis of the Atcor Vendenburgh Cask," February 1975.
e
SPS.UFSAR 10.3-2,5* switches are* mounte.d in the same. cabinets with the analog and. lo~ic *equipment ** This equ:f,pment is annunciated on the control board-when i~* test. The electrohydraulic overspeed protection system is designed- to meet nucle_ar protection system' criteria on 're_dµndancy, 'sepa'rati.on, and reliability. Any single* active component failure cannot: cause failuri to trip the' unit on overspeed. L~~s ~i~ ac :p-ower wi.il also trip th~ .unit. Fa?lt trees. are presented (Figures 10.3-15 through 10~3:..19) to show the possible combinations of component failures . that. can
- 1ead to loss of function. The overali .
--pro'bability"-_of failure to ;imit the turbine to 118% of rated speed, *con..:.
sidering only the additional ove~speed detection and tripping system, is
. computed to be 9.6 X 10.:.a. This is bas~d o~ a f,igure:_ o( 7.5 x-{0'- 7 -fo~ ~he:*_.
'p:robability -C)f :- f~ilur~ 'to' d~mp hydraulic fluid fr-om a*. single .steam valve' actuator based oti
- periodic tests :and: *ge-neric** failure rat~s for components ,
',' extracted fr.om _'various government and' industry sources.
10.3.4* CIRCULATING WATER SYSTEM The circ'ulatirig wa.ter _sy~tem;. Figure 10.3_;20, provides c.ooling water. fo~
- the main ~ohdense*rs*: and the service water systems qf both units~<
10~3.4.i * ,D.esign:Bas*is. - ..... -. To prev~~t . t,he direct - recircu'lation
- of'. the* heated circula_tiilg water .
*. disc_harge~
*.the .
_system _{s designed* to takewat_er 'from, the. James .
-River-on d:e". ** east *end of the. site - .
and to discharge to_ .the . 'Ja~es
- River' .
.on . ~he west. end .of: . *.*,;._
the site~
- _The shor~line di~tance be_t~een the in;ake .and discharge poi.n,~s -is:<,
about 5 ~ 7 mi~_es _- and the overland distari~e ac:ross *the_ peninsul~ is about l *. ~.miles.: *
- Each. uriit requires 840,000 gpm of river water :to s11pply ~ondensing. arid_,*_
service *~t:iter needs. To provide. op.erational
' . ' . :flexibility, system
. ~ .
reliability/
-_ ~nd st~tioh economy. the wa*ter requix:em.an:t,. for *each uiiit is -supplied 'by foul.'..':
- 2-10,000-gpm pumps.* -- These pump~ discharge to the. common . . high-:-level intake. .. '
*.~:cana). that* conveys. the circulating _water . to ~he statfon area.' Coarse trash *is
'removed from the ciI'culating water *by trash r~cks- at* th~ river: intake struct~;e., '
REVISION 9 6/90 SPS UFSAR l O. 3-26*' and finer trash is removed a_t the river intake and at the entry-,-bay and stati ends of the intake canal by two sets of travelling water screens. circulating water flows by gravity from the high-*level intake canal through four buried parallel lines to each condenser and then- through foui separate lines to a concre'te tunn_el for each unit. The tunnels terminate _at seal pits located at the*edge of the circulating water discharge canal, which is common to both units.
- The discharge canal conveys the flow* to the James River.. The discharge channel within .,the river is provided w'ith rock groins along each** side to control sedimentation and to maintain exit velocities of the circulating water to achieve *d~sir~d dilutio*n effects of the .heated effluent. A ci;culating water discharge canal control structure extending p~rtially across the opening between ~rains is prbvided to ensure prqper veloci~y conditions when only one uriit is operating ..
Some components of the circulating water system . are used for handling service water, and are thereiore ., designed* ~s Seismic Category I structuies a ~ components. These* components* are: 1.. The circulating water intake structure at the river.* 2 ._ High-level intake c*anal: 3.- High-level intake stiucture.
- 4.
- _Buried circulating water pipin~ and valves between the high-level intake canal and the ci'i:*culating ~ater discharge tunnel. *
- 5. Ci~cul~ting water.~ischarge tunnel.
- 6. Seal* pits.
- 7. Intak.e canal low-level isolatiqn level switches ( l-CW.,.-LS-102 & 10 3, -1 2-CW-LS-202 & 203.
10.3.4.2 Description The circulating water is withdrawn from the James River through a channel dredged in the river _bed betweer. the main river channel and the shore,- a
I REVISION 9 6/90 SPS UFSAR distance of approximately 5000 ft: The channel invert is 150 ft wide at elevation -13. 3 ft. The channel is also used for* shipping materi.als
- and equipment to. the permanent dock on the east side of the site.
The circulating water intake structure is located at the shore end of the river intake channel and is an eight-bay reinforced-concrete struct~re. The exposed deck of the structure is at elevation +12 ft; however,.* the pillbox enclosure for the emergency service water pumps is protected from flooding to* elevation +21 ft and from wave run-up to elevation +33;5 ft. The invert of the intak~* structuri is at elevation *-25.25 ft. Each bay houses one*of the eight
. circulating w_ater pumps for. the two units. These pumps
- are rated at 210,000
*gpm at 28 ft total dynamic head when running at 220 rpm. Each pump is driven by. a vertical, solid-shaft,. 2000~hp, . inductiOI1 motor. The pumps are of the no_npullout type and are serviced using mobile hoisting equipment. Before entering the. pumps; river water passes through a trash rack and travelling screen at the mouth of each bay. The travelling screens are provided with deflector flaps . and screenwash pumps for. low-pressure water spray. The deflector flaps ensure that *fish dumped from the screens are deflected. into a trough for transport via an effluent flume back to the James River. The low- . pressure* spray ensures more efficient washing of fish .from the screens into the fish collector trough. The trash rack is serviced hy a movable trash rack rake that discharges. co.llected trash to a receptacle where it accumulates until - trucked off-site .for dis~osal: in the event of a power failure, accumulated tr~sh --can be- remo~ed by manual raking. This process could continue inde~initely; however, it.is expected that .any power failure at th~ station low-level. intake would be of relatively short duration, Each circulating water pump discharge line. is a 96-in. -diameter steel pipe that conveys the water over the embankment of and into the high-level intake canal. At the crest of the canal embankment,. the crown of the pipe _is provided* with a pair of vacuum breakers and a tap* for the vacuum_ priming system. The vacuum breakers open .when the *circulating. water pump is.
deenergize.d. These vacuum breakers pre~ent loss of water from the high-level intake canal by siphoning th(ough idle pumps. A passive vacuum breake.r, designed to_ conserve intake canal water; in the. event of a failure of "the
- paired acti~e vacuum breaker valves .is located on the discharge end of each 96~
.*in. -diameter pipe. The passive vacuum breaker consists .of a 20-in. :..diameter*
REVISION 9 6/90 SPS UFSAR 10 .. 3~27a _, I low profile vertical pipe projection which extends to elevation +25 ft of ~h intake canal. The passive vacuum breakers are designed to interru*pt postulated siphon action prior to reaching the technical specification limitation for. canal water level. The vacuum priming system prevent*s air accumulation in the pump discharge line while the pump is operating. This system is isolated when the circulating water pumps are deenergized .
REVISION 1 6/83 SPS UFSAR 10.3-49
- 4. Clean and reclaim used oil from the storage tanks, pumping it from the used-oil storage tank via the purifier to the clean-oil storage tank.
10.3.7.1 Design Basis The lubricating oil system consists of a 21, 000-gal reservoir, two 22,000-gal horizontal all-welded steel storage tanks, an oil purifier, and two identical motor-driven transfer pumps. The two gear-type positive displace-ment transfer pumps are each capable of two-speed operation at 108 and 48 gpm to accomplish the various batch cleaning, transfer, and circulating opera-tion$. The three-compartment Bowser oil purifier is rated at 3000 gal/hr. 10.3.7.2 Description A turbine shaft-driven oil pump normally supplies all lubricating-oil requirements to the turbine-generator unit. An ac motor-driven turning gear oil pump is installed for supplying lubricating oil during startup, shutdown, and standby conditions. An emergency de motor-driven oil pump, operated from the station battery, is also available to ensure lubricating oil to the bearings. Cooling water from the bearing cooling water system (Section 10.3.9) is used for the turbine lube-oil coolers, which are immersed in the main oil reservoir. The two 22,000-gal storage tanks are normally designated "clean" and "used," but are interchangeable and are located inside a fireproof room equipped with water sprays and vent fans. The transfer pumps and piping are arranged so that oil can be processed from the oil reservoir or either of the two storage tanks. The processed oil can be returned to either of the other two. Vapor extractors purge oil fumes from the oil purifier and reservoir and exhaust to the atmosphere outside of the turbine building. All piping and valves in the system are of welded steel, and high-pressure bearing oil piping is enclosed in a guard pipe. 10.3.7.3 Performance Analysis Only the de motor-driven oil pump is needed to function during a LOCA or loss of station power to supply lubricating oil to the turbine-generator
REVISION 1 6/83 SPS UFSAR 10. 3'.""50 bearings. The station battery provides an uninterrupted source of power to this pump. No other part of the lubricating oil system is required to operate. 10.3.7.4 Tests and Inspections The de bearing oil pump is tested monthly. 10.3.8 SECONDARY VENT AND DRAIN SYSTEMS Because the steam and power conversion system is normally nonradioactive, vents and drains are arranged in much the same manner as those in a fossil-fueled power station. However, because air ejector vents and steam-generator blowdown can possibly become contaminated and because they discharge to the environment, they are monitored and discharge under controlled conditions as described in Chapter 11. The air ejector vent subsystem is shown in Figure 10.3-7. The steam-generator blowdown system is shown in Figure 10.3-24. The steam-generator blowdown treatment system is described in Chapter 11. 10.3.8.1 Design Basis Each of the condenser steam jet air ejectors (two per shell) is designed to remove 12.5 cfm of free air. Each ejector normally uses about 800 lb/hr of steam at 150 to 200 psig from the auxiliary steam header, while using 900 gpm of condensate for cooling. Separate hogging or vacuum priming jets are used to reduce condenser vacuum to 1 to 3 in. Hg abs during startup. 10.3.8.2 Description Generally, secondary plant piping drains to the condenser. Vent gases removed from the condensers by the air ejectors are normally discharged through a radiation monitor (Section 11.3) to the atmosphere. If a steam-generator tube ruptures, with subsequent contamination of the steam, the
REVISION 1 6/83 SPS UFSAR 10.3-53 The principal equipment served by the bearing cooling water is listed in Table 10. 3-4. The full-size 13,000-gpm motor-driven pumps circulate the cooling water through the above equipment and the bearing cooling heat exchangers. 10.3.9.2 Description The cooling water flowing through the major equipment coolers, such as the hydrogen and oil coolers, is controlled automatically to maintain constant temperature of the cooled fluid. A head tank is provided to maintain a positive pressure at all points on the system. Makeup to this tank is normally from the gland steam condensate drain pump; however, when this system is not in operation, makeup is provided from the condensate system (Section 10.3.5). Corrosion in the bearing cooling water system is inhibited by the use of morpholine and hydrazine. It is felt that the use of a filming amine (mor-pholine) provides adequate chemical protection in the.system. The morpholine 1 and hydrazine are obtained from the reservoir tank in the chemical feed system. The system design allows for continuous service to both the chemical feed system and the bearing cooling system. 10.3.9.3 Performance Analysis The instrument air compressor is required to operate when taking the plant to a cold, shutdown condition following a LOCA or a loss of station power. This requires that the bearing cooling water system operate in order to supply the instrument air compressor, Under these conditions, jacket cooling water for this compressor is supplied by well-water from the fire protection system.
REVISION 1 6/83 SPS UFSAR 10.3-54
10.3 REFERENCES
- 1. U. S. Nuclear Regulatory Commission, "Seismic Analyses for As-Built Safety-Related Piping Systems," IE Bulletin 79-14, June 2, 1979.
- 2. Letter from A. Schwencer, NRC, to W. L. Proffitt, Vepco,
Subject:
Safety Evaluation of the Steam-Generator Repair Program by the Office of Nuclear Reactor Regulation, License Nos. DPR-32 and DPR-37, dated December 15, 1978.
- 3. U. S. Nuclear Regulatory Commission, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendation, NUREG-0578, July 1979.
SPS UFSAR 11.3-13 e The use of a single detector is justified in lines used for normal releases from the plant. Each detector is checked daily by energizing its check source, checked monthly by testing its circuit and alarm functions, and calibrated each refueling outage. The liquid and gas waste tanks are sampled
;:.'., and analyzed before and during discharges . Effluent source terms are
.,- discussed in Appendix llA.
Channels monitoring Unit 1 are supplied from the emergency bus for Unit 1. Channels monitoring Unit 2 are supplied from the emergency bus for Unit 2. Channels monitoring systems or areas connnon to both units can be supplied from the emergency bus for either Unit 1 or Unit 2. The type of detector, sensitivity, range, background radiation, and other information for each channel are listed in Table 11.3-6. Counting rates are given in Table 11. 3-7. A description of each channel is included in the following text.
- 11.3.3.1 Process Vent Particulate Monitor This channel continuously withdraws a 10-scfm sample from the process vent and passes the sample through a moving filter paper with a collection.
efficiency of 99% for particle sizes greater than 1. 0 µ. The amount of deposited activity is continuously scanned by a lead-shielded beta scintilla-
*'-* tion detector. A high-activity alarm automatically initiates the closure of ,:;:. the process vent discharge line valves.
A separate isokinetic sampling nozzle is provided for each unit, as shown in Figure 11.2-8. The nozzles sample the process vent fluid to ensure that a representative sample is taken. Isokinetic sampling is achieved by locating the nozzles in a straight unobstructed piping run of at least five pipe
... diameters (30 in.). The sampling systems include a pump and a flow meter, and supply 1-cfm flow to a sample filter.
11.3.3.2 Process Vent Gas Monitor This channel takes the continuous process vent sample, after it has passed through the particulate filter paper, and draws it through a sealed
SPS UFSAR 11.3-14 system to the process vent gas monitor assembly, which is a fixed lead-shielded sampler enclosing a beta scintillation detector. The sample activity is measured, and is then returned to the process vent. A high-activity alarm automatically initiates the closure of the process vent discharge line valves. A purge system is integral with the gas monitoring system for flushing the sampler with clean air for purposes of calibration. 11.3.3.3 Ventilation Vent Particulate and Gas Monitors These two channels continuously sample the ventilation vent particulates and gas in the same way as the two channels monitor the process vent sample. All equipment is identical with that of the two process vent channels, except (1) pressure-protecting valves are not required, (2) an inline, easily removable, charcoal filter is provided between the particulate and gas monitor, and (3) a multiprobe isokinetic sampler is provided to obtain a representative sample in the 80-in. diameter duct. 11.3.3.4 Component Cooling Water Monitors These two channels continuously monitor the component cooling water by means of a gamma scintillation detector enclosed in 3 in. of lead shielding and mounted on the component. cooling piping. The complex piping arrangement of this system dictates that two detectors are required to ensure that the system is properly monitored. Activity is indicative of a leak into the component cooling system (Section 9.4) from one of the radioactive systems that exchange heat to the component cooling system. 11.3.3.5 Component Cooling Heat Exchanger Service Water Monitor This channel continuously monitors the service water effluent from the four component cooling water heat exchangers. A 2-gpm sample of the service water from each exchanger is pumped into a common discharge. A separate sample pump is installed after each heat exchanger to provide equal sampling throughout their operating range. The combined sample flow is passed through a liquid sampler, shielded by 4 in. of lead, enclosing a gamma scintillation detector. Upon indication of activity, a valving arrangement allows the heat exchangers to be individually sampled to determine the origin of the activity.
SPS UFSAR 14-xi CHAPTER 14: SAFETY ANALYSIS LIST OF FIGURES (continued) Figure Title 14.2-13 Rod Withdrawal at 10% Power 14.2-14 Rod Withdrawal at 60% Power 14.2-15 Rod Withdrawal at 100% Power 14.2-16 Response to a Dropped Control Rod Assembly of Worth 3
-2.6 x 10- delta k with Load Cutback, Nuclear Power, Heat Flux, and Pressurizer Pressure 14.2-17 Response to a Dropped Control Rod Assembly of Worth
-3 .
-2.6 x 10 delta k with Load Cutback, T , Inlet avg Temperature, and Steam Load 3
14.2-18 Response to a Dropped Rod of Worth -1.0 x 10- delta k with Load Cutback, Nuclear Power, Heat Flux, and Pressurizer
- 14.2-19 14.2-20 Pressure Response to a Dropped Rod of Worth -1.0 x 10 Load Cutback, T avg
. -3 delta k with
, Inlet Temperature, and Steam Load Variation in Reactivity Insertion Rate with Initial Boron Concentration for a Boron Dilution Rate of 165 gpm 14.2-21 Startup of an Inactive Reactor Coolant Loop, Loop Stop Valves Initially Open 14.2-22 Startup of an Inactive Reactor Coolant Loop, Stop Valves Initially Open 14.2-23 Reactor Coolant Parameters .for EOL Accidental Opening of a Loop Stop Valve 14.2-24 Time Dependent Nuclear Parameters for the EOL Accidental Opening of a Loop Stop Valve 14.2-25 Maximum Heat Flux Caused by Accidental Opening of a Loop Stop Valve 14.2-26 Feedwater Bypass Opening BOL, No Reactor Control, Sheet 1 14.2-27 Feedwater Bypass Opening BOL, No Reactor Control, Sheet 2 14.2-28 Feedwater Bypass Opening BOL, with Reactor Control, Sheet 1 14.2-29 Feedwater Bypass Opening BOL, with Reactor Control, Sheet 2 14.2-30 10% Step Load Increase BOL, No Control, Sheet 1 14.2-31 10% Step Load Increase BOL, No Control, Sheet 2
SPS UFSAR 14-xii CHAPTER 14: SAFETY ANALYSIS LIST OF FIGURES (continued) Figure Title 14.2-32 10% Step Load Increase EOL, No Control, Sheet 1 14.2-33 10% Step Load Increase EOL, No Control, Sheet 2 14.2-34 10% Step Load Increase BOL, with Control, Sheet 1 14.2-35 10% Step Load Increase BOL, with Control, Sheet 2 14.2-36 10% Step Load Increase EOL, with Control, Sheet 1 14.2-37 10% Step Load Increase EOL, with Control, Sheet 2 14.2-38 Three Loops Operating, Three Pumps Loss of Flow Incident 14.2-39 Three Loops Operating, One Pump Loss of Flow Incident 14.2-40 Two Loops Operating, No Loop Stop Valves Closed, Two Pumps Loss of Flow Incident 14.2-41 Two Loops Operating, No Loop Stop Valves Closed, One Pump 14.2-42 14.2-43 Loss of Flow Incident Two Loops Operating, Loop Stop Valves Closed in One Loop, Two Pumps Loss of Flow Incident Two Loops Operating, Loop Stop Valves Closed in One Loop, One Pump Loss of Flow Incident 14.2-44 Three Loops Operating, Three Pumps Loss of Flow Incident, Sheet 1 14.2-45 Three Loops Operating, Three Pumps Loss of Flow Incident, Sheet 2 14.2-46 Three Loops Operating, One Pump Loss of Flow Incident, Sheet 1 14.2-47 Three Loops Operating, One Pump Loss of Flow Incident, Sheet 2 14.2-48 Two Loops Operating, No Loop Stop Valves Closed, Two Pumps Loss of Flow Incident, Sheet 1 14.2-49 Two Loops Operating, No Loop Stop Valves Closed, Two Pumps Loss of Flow Incident, Sheet 2
,REVISION 8 8/89 SPS UFSAR 14.2-5
- 14.2.1.2 Results Figure 14.2-2 shows the effect of the initial power level on peak heat
_ flux for various *reactivity insertion rates from 20 to 60 pcm/sec. It shows that peak heat flux initially decreases with increasing initial power level and then, depending on the rate, it increases again and approaches 35% of full power (reactor trip is assumed to be initiated at this value). It can also be seen that for the faster insertion rates, which result in the greatest energy addition, the flux peak is greatest for the lowest initial power level. Figures 14.2-1, 14.2-3, and 14.2-4 show the transient behavior for a reactivity insertion rate of 75 pcm/sec with the incident terminated by reactor trip at 35% power. This insertion rate is greater than that for the two highest worth banks, both assumed to be in their highest incremental worth region. Figure 14.2-1 shows the nuclear power increase. The nuclear power is seen to increase to the trip point in 10.47 sec. The nuclear power overshoots to approximately 386% but this occurs for only a very short time period. Hence, the energy release and the fuel temperature increases are small. The thermal flux response, of interest for departure-from-nucleate-boiling (DNB) considerations, is shown on Figure 14. 2-3. The beneficial effect of the inherent thermal lag of the fuel is evidenced by a peak heat flux of only 38% of the nominal value. There is a large margin to DNB during [ the transient since the rod surface heat flux remains below the design value, and there is a high degree of subcooling at all times in the core. Figure
- 14. 2-4 shows the response of the average fuel, cladding, and coolant temperature. The average fuel temperature increases to 763°F, which is lower f than the nominal full-power value of about 14 7 5°F. The average coolant temperature increases to only 585°F. I 14.2.1.3 Conclusion Taking into account the conservative assumptions used in the 75-pcm/sec reanalysis, it is concluded that, in the unlikely event of a control-rod
REVISION 8 8/89 SPS UFSAR 14.2-6 assembly withdrawal incident, the core and reactor coolant system are n o t . adversely affected, since the thermal power reached is only approximately 38% - , of the nominal value and the core water temperature increases to only 585°F. I This combination of thermal power and core water temperature results in a DNBR weli' above the limiting value of 1.30. The peak average clad temperature is less than the nominal full-power value of 633°F and thus there is no clad damage. 14.2.2 UNCONTROLLED CONTROL-ROD ASSEMBLY WITHDRAWAL AT POWER An uncontrolled control-rod assembly withdrawal at power results in an increase in core heat flux. Since the heat extraction from the steam generator remains constant until the steam-generator pressure reaches the relief or safety valve setpoint, there is a net increase in reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result_ in DNB. Therefore, to prevent the possibility of damage to the cladding, _,the reactor protection system is designed to terminate any such transient befor the DNBR falls below 1.3. The automatic features of the reactor protection system that prevent core damage in a control-rod assembly withdrawal incident at power include the following:
- 1. Nuclear power range instrumentation actuates a reactor trip if two out of the four channels exceed an overpower setpoint.
- 2. Reactor-trip is actuated if any two out of three delta T channels exceed an overtemperature delta T setpoint. This setpoint is automatically varied with axial power distribut_ion to ensure that the allowable fuel power rating is not exceeded ..
- 3. Reactor trip is actuated if any two out of three delta T channels exceed an overpower delta T setpoint. This setpoint is automatically varied with axial power distribution to ensure t h a .
the allowable fuel power rating is not exceeded.
SPS UFSAR 14.2-7
- 4. A high-pressure reactor trip, actuated from any two out of three pressure channels, is set at a fixed point. This set pressure is less than the set pressure for the pressurizer safety valves.
- 5. A high pressurizer water level reactor trip, actuated from any two out of three level channels, is actuated at a setpoint. This affords additional protection for control-rod assembly withdrawal incidents.
The Technical Specifications require that the reactor be maintained subcritical by some minimum amount until normal water level is established in the pressurizer.
- 6. In addition to the above.;.listed reactor trips, there are the following control-rod assembly withdrawal blocks:
- a. High nuclear power (one out of four).
- b. High overpower delta T (two out of three).
- c. High overtemperature delta T (two out of three) *
- The manner in which the combination of overpower and overtemperature delta T trips provide protection over the full range of reactor coolant system conditions is illustrated in Chapter 7. Figure 7.2-8 represents the allowable conditions of reactor vessel average tem~erature and delta T with the design power distribution in a two-dimensional. plot. . The boundaries of operation defined by the overpower delta T trip and the overtemperature delta T trip are represented as protection lines on this diagram. The protection lines are drawn to include all adverse instrumentation and setpoint errors, so that under nominal conditions trip would occur well within the area bounded by these lines.
The utility of the diagram just described is in the fact that the operating limit imposed by any given DNBR can be represented as a line on this coordinate system. The DNB lines represent the locus of conditions for which the DNBR equals 1.3. All points below and to the left of the line for a given pressure have DNBRs greater than 1.3. The diagram shows that DNB is prevented for all cases if the area enclosed within the maximum protection lines is not traversed by the applicable DNBR line at any point.
SPS UFSAR 14. 2:--8 The region of permissible operation (power, pressure, and temperature) is completely bounded by the following reactor trips: nuclear overpower (fixed. setpoint), high pressure (fixed setpoint), low pressure (fixed setpoint), and overpower and overtemperature delta T (variable setpoints). These trips are designed to prevent a DNBR of less than 1.30. 14.2.2.1 Method of Analysis The purpose of this analysis is to demonstrate the manner in which the above protective systems function for various reactivity insertion rates from different initial conditions. Reactivity insertion rates and initial conditions govern which protective function occurs first. Analysis is performed using several digital computer codes. First, the actual core limits are determined from the W-3 DNB correlation described in Section 3. Protection lines, illustrated in Figure 7.2-1 are then selected and incorporated into a transient analysis by a detailed digital simulation of the unit. In order to obtain conservatively low DNBRs, the following assumptions are made:
- 1. Initial conditions assume maximum power and reactor coolant temperatures and m~nimum pressure; that is, the power is assumed 2%
high, the average temperature is assumed 4°F high, and the pressure is assumed 30 psi low. This gives the minimum initial margin to DNB.
- 2. For the initial FSAR analysis, a zero moderator coefficient of reactivity was assumed corresponding to the beginning of core life.
A conservatively small (in absolute magnitude) Doppler power reactivity coefficient was used (see Figure 14.2-5). The assumed reactivity coefficients result in a minimum of negative feedback reactivity and, therefore, higher peak powers and temperatures.
- 3. The reactor trip on high nuclear power is assumed to be actuated at a conservative value of 118% of nominal full power. The delta T trips
SPS UFSAR 14.2-9 include all adverse instrumentation and setpoint errors. The delays for trip signal actuation are assumed at their maximum values, that is, 0.5 sec for the high nuclear power trip and 3.5 sec for the delta T trip.
- 4. The rate of negative reactivity insertion corresponding to the tri~
of the control-rod assemblies is based on the assumption that the. highest worth control-rod assembly is stuck in its fully withdrawn position.
- 5. The analysis assumes reactor coolant system flow because the Technical Specifications preclude critical operation under natural..
circulation conditions. When the reactor is critical, except for special low-power tests during startup testing after reloads, at least one *reactor coolant pump must be in operation. The special low-power tests are conducted under careful supervision using approved procedures. Because of the aforementioned Technical
- Specification and procedural restrictions, it is not necessary to specifically analyze a startup accident without reactor coolant system flow.
The effect of the control-rod assembly movement on axial core power distribution is accounted for by a decrease in overtemperature delta T and overpower delta T trip setpoints proportionate to the decrease in the DNB margin. 1 A subsequent reanalysis was performed to verify the acceptability of_ operation with high levels of steam-generator tube plugging. This analysis used the following assumptions, in addition to those in the initial FSAR analysis:
- 1. The reanalysis assumes that 40% of the steam-generator tubes are,,
plugged.
- 2. A thermal design flow of 79,650 gpm per loop is assumed. This value is 90% of the initial thermal design flow. It is based on plant flow
SPS UFSAR 14. 2-10 measurements and a conservative prediction of flow with 40% steam-generator tube plugging.
- 3. The overtemperature delta T reactor trip function was modified to account for increased coolant loop transit time and to be conser-vative with respect to core thermal limit lines recalculated based on the lower thermal design flow. A trip delay time of 2.0 sec was used.
- 4. A positive moderator coefficient of reactivity was assumed. The value was conservative with respect to Technical Specification limits for each case analyzed.
14.2.2.2 Results Figures 14.2-6 and 14.2-7 show the response of nuclear power, average coolant temperature, pressure, *and the DNBR to a rapid control-rod assembly 4 withdrawal (6.0 x 10- delta k/sec) incident starting from full power, assuming a zero moderator temperature coefficient (original FSAR analysis). The reactivity insertion rate is greater than that for the two highest worth banks, both assumed to be in their highest incremental worth region. Reactor trip on high nuclear power occurs approximately 2.4 sec after the start of the accident. Since this is rapid with respect to the thermal time constants of the plant, small changes in T and pressure result. A larger DNB margin is avg maintained, the minimum DNBR being 1.53. The response of nuclear power, average coolant temperature, pressure, and the DNBR to a slow control-rod withdrawal (2.0 pcm/sec) from full power*is shown in Figures 14.2-8 and 14.2-9 (original FSAR analysis). Reactor trip on overtemperature delta T occurs after approximately 48 sec. The rise in temperature and pressure is larger than that for the rapid control rod assembly withdrawal. The minimum DNBR reached during the transient is 1.36. Figures 14.2-10 and 14.2-11 show the minimum DNBR as a function of reactivity insertion rate for control-rod assembly withdrawal incidents
SPS UFSAR 14.2-11 starting at 60% and 10% power, respectively. The results are very similar to the 100%-power case (Figure 14.2-12), except that as the initial power is decreased, the range over which the overtemperature delta T trip operates is increased. Figures 14.2-13 through 14.2~15 show the minimum DNBR as a function of 1 reactivity insertion rate for the 40% steam-generator tube plugging analysis. The limiting case for DNB margin is a reactivity insertion rate of 6.2 x 10-S delta k/sec from full power initial conditions, which results in a minimum DNBR of 1. 31. 14.2.2.3 Conclusions In the unlikely event of a control-rod assembly withdrawal incident during power operation, the core and reactor coolant system are not adversely affected since the minimum value of the DNBR reached is in excess of 1.3 for all control-rod assembly reactivity rates. Protection is provided by the
- nuclear flux overpower and overtemperature delta T trips. The preceding sections have described the effectiveness of these protection channels.
14.2.3 MALPOSITIONING OF THE PART-LENGTH CONTROL-ROD ASSEMBLIES The part-length control-rod assemblies have been removed from the core. 14.2.4 CONTROL-ROD ASSEMBLY DROP/MISALIGNMENT Control-rod misalignment accidents include (1) dropped full-length assemblies, (2) dropped full-length assembly groups, and (3) statically misaligned assemblies. Each control-rod assembly has a rod position indicator channel that. displays the position of the assembly. The displays of assembly position are grouped for the operator's convenience. Fully inserted assemblies are further indicated by rod bottom lights. Bank (demand) position is also indicated. The assemblies are always moved in preselected banks and the banks are always moved in the same preselected sequence.
SPS UFSAR 14.2-12 The dropping of a control-rod assembly could occur only when the drive mechanism is deenergized. This would result in a power reduction and an increase in the hot-channel factor. If no protective action occurred, the reactor control system would restore the power to the level that existed before the incident. This would lead to a reduced safety margin or possibly DNB, depending on the magnitude of the hot-channel factor. Dropped assemblies or banks are detected by
- 1. A sudden drop in the core power level.
. power d.istri.b ution
. 2 2* Asymmetric as seen on ex-core neutron d etectors or core exit thermocouples.
- 3. Rod bottom light(s).
- 4. A rod deviation alarm.
The rod bottom signal device provides an individual position indication signal for each control-rod assembly.
- The initiation of this signal is independent of lattice location, reactivity worth, or power distribution changes inherent with the dropped control-rod assembly. The other independent indic~tion of a control-rod assembly drop
- is obtained by using the ex-core power range channel signals.
This rod drop detection circuit is actuated upon the sensing of a rapid decrease in local flux such as could occur from the depression of flux in one region by a dropped control-rod assembly. This detection circui~ is designed such that normal load variations do not cause it to be actuated. A rod drop signal from any control-rod assembly position indication channel, or from one or more of the four power range channels, initiates . protective action by reducing turbine load by a preset adjustable amount and blocking further automatic control-rod assembly withdrawal. The turbine runback is redundantly obtained by acting on the turbine load limit and on the turbine load reference. The control-rod assembly stop is also redundantly actuated.
SPS UFSAR Figure 14.2-10 a:
<I:
w (,)
-a:
.... a.
z I-
- c: X
-------- ---=---
C, :::,
-...1 u.
--------- 7 I '<I' 0
I cJ w
,S!?
- ii::
<J w*
I-a: z 0
~
a: w en z
- LI)
~
I
~
(,)
<I:
w a: w-a:_ I-a: w a.-
~
w a. - I- - a: a: w I-
> I-0 <J IO r I I I I I b
- 0) co ll)
":J3S EFFECT OF REACTIVITY INSERTION RATE ON MINIMUM DNBR FOR A ROD WITHDRAWAL ACCIDENT FROM 10 PERCENT OF FULL POWER
SPS UFSAR Figure 14.2-14 l'I 0 I
~
"" ~
/
/
/
"'0 I
e,j w
~
w* I-
/ c(
a:
,/ in z 0
/ j::
/ a:
w
/ en z
I
- I I
I II) 0 I 2: I-u c( w a:
......... a,
'- CD
'- \.
\.
\.
co in
~
cw, N I I I I b... 0 0
~
0
~
0
..:I' 0
..."?
HBNC wnWINIW ROD WITHDRAWAL AT 60% POWER
I - SPS UFSAR Figure 14.2-15
"'0I I
\
\
\ in
\
\
\
\
\
\
\
\ \.
\.
0 I d w se
~
\. w I-
\. - c(
cc I I in z 0 I ;:: I cc w I en z I > I !:: 2: I I-CJ c( I ID w cc 0 I a, m I co in "it C"J N l I I 0 I b... 0 0 0 0
~ ~
i.q HBNO llllnlllllNll/11
~ C"J ROD WITHDRAWAL AT 100% POWER}}