ML19319E364
| ML19319E364 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1967 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| References | |
| NUDOCS 8004090495 | |
| Download: ML19319E364 (65) | |
Text
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I e suun SACRAMENTO MUNICIPAL UTILITY DISTRICT M
RANCHO SECO NUCl. EAR GENERATING STATION 4Q UNIT NO.1
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PRELIMINARY SAFETY ANALYSIS REPORT Volume 11
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NOVEMBER 1967 8004090 %
377g
Doc et No. 50-312
_s T('w_/i-LIST OF April 15, 1968 E F F ECTIV E PAGES Amendment No. 2 The active pages in this report are as follpws:
Page or Fig. No.
Issue Page or Fig. No.
Issue Title Page.......... Original' Fig. 2.2-1 thru 2.2-10.
. Original A thru H
....... Amendment 2 2.3-1 thru 2.3-2.
. Amendment 2
- i.............. Original 2.3-3 thru 2.3-4.
. Original 11.
... Amendment 2 2.3-5 thru 2.3-8...
. Amendment 2 111.
..... Original Fig. 2.3-1 thru 2.3-6..
. Original
,~
ivr"........... Amendment 2 2.4-1..
. original v thru vii.
... Amendment 1 2,4-2.
. Amendment 2 viii.
Amendment 2 2.4-3.
. Original ix.
Amendment 1 Fig. 2.4-1 thru 2.4-2.
. Original'
. Original 2.5-1.
. Original x.
xi.......#
Amendment 1 2.6-1.
. Original xii thru xiv.
Amendment 2 2.7-1
. Original 1-1.
. Original 2.8-1 thru 2.8-4.
... Amendment 1 1-11.
Amendment 2 2.9-1.
. Original 1-111.
. original 3-i thru 3-111.
. Ame.ndment 2 1-iv.
Amendment 2 3-iv thru 3-vi.
. Original 1.1-1 thru 1.1-2.
. Original 3.1-1.
. Original
,e~3 4
)
Fig. 1.1-1.
. Original 3.1-2 thru 3.1-4.
. Amendment'2 I\\#
Fig. 1.1-2 thru 1.1-8.. Amendment 2 3.1-5.
. Amendment 2 1.2-1.
. Original 3.1-6
. Amendment 2 1.2-2 thru 1.2-4.
Amendment 2 3.2-1 thru 3.2-2.
. Original 1.3-1 thru 1.3-3..
. Original 3.2-3.
. Amendment 2 1.3-4.
Amendment 1 3.2-4 thru 3.2-10.
. Original 1.3-5.
. Original 3.2-11.
. Amendment 2 1.3-6 thru 1.3-7.
Amendment 2 3.2-12 thru 3.2-69.
. Original 1.3-8.
. Original 3.2-70 thru 3.2-101.
. Amendment 2
,y 1.3-9..
Amendment 2 Fig. 3. 2-1 thru 3. 2-59..
. Original 1.4-1.
. Original Fig. 3.2-59 thru 3.2-61. Amendment 2 1.4-2.
Amendment 2 Fig. 3.2-62 thru 3.2-63.
. Original 1.4-3..
Amendment 1 Fig. 3.1-64..
. Amendment 2 1.4-4 thru 1.4-6.
. Original Fig. 3.2-65.
. Amendment 1 1.4-7 thru 1.4-8..... Amendment 2 Fig. 3.2-66.
. original 1.4-9 thru 1.4-37.... Amendment 2 Fig. 3.2-67.
. Amendment 2 1.5-1 thru 1.5-2.
Amendment 2 Fig. 3.2-68.
. Amendment 1
~~ T 1.6-1.
. Original Fig. 3.2-69.
. Original 1.6-2 thru 1.6-3.
Amendment 2 3.3-1.
. original Fig. 1.6-1 thru 1.6-2.
. original 3.3-2.
. Amendment 2 1.7-1..
. original 3.3-3 thru 3.3-5'.
Original 1.8-1 thru 1.8-2.
. Original 3.3-6 thru 3.3-7.
. Amendment 2 1.9-1.
. original 3.3-8 thru 3.3-10.
Original
.,rm.
2-1 thru 2-i'i.
Amendment 1
- 3. 3-11 thru 3. 3-12.... Amendment 2 g-'y 111.
......... Original 3.4-1 thru 3.4-5...... Original 2
]
2.1-1.
. Original 4-1 thru 4-11.
Original
(}]f 2.2-1 thru 2.2-5....... Original 4.1-1 thru 4.1-15.
Original Amendment 2 C
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2.. _
3 f
.s Docket No. 50-312
.l I
LIST' OF April 15, 1968
- f.
\\=>)
E F F ECTIV E PAdES Amendment No. 2 a
Page or Fig. No.
Issue
'Page or Fig.'No.
Issue t
Fig. 4.1-1....
. Original Fig. 5.711......... Amendment 2 Fig. 4.1-2...,.......' Amendment 2 ' 5.8-1.
. Original ~
.J Fig. 4.1-3 thru 4.1-4.... Original 5.9-1.
. Original 4.2-1.
. Amendment 2 6-1.
. Amendment.1 4.2-2 thru 4.2-6.
. Original 6-11.
. Amendment 2 4.2-7 thru 4.2-8.
. Amendment 2 6.0-1.
. Amendment 2 4.2-9.
. Original Fig. 6. 0-1.
. Amendment 2
~
4.2-10 thru 4.2.11.
. Amendment 2 6.1-1 thru l-7.
. Amendment 2 4.2-12.
. Original 6.1-8.
. Original Fig. 4.2-1.
.... Amendment 2 6.1-9 thru 6.1-10.
. Amendment 2
. Amendment 1 Fig. 4.2-2 thru 4.2-8.
. Original.
6.1-11.
4.3-1.
. Amendment 2 6.1-12 thru 6.1-14.
. Amendment 2 4.3-2 thru 4.3-7....,.... Original 6.1-15........... Amendment 1 4.3-8 thru 4.3-10.. '.
. Amendment 2 6.1-16.
. Amendment 2 4.3-11.
. Amendment 1 Fig. 6.1-1 thru -6.1-2.
. Amendment 2 4.4-1 thru 4.4-3.
. Orfginal Fig. 6.1-3.
. Original 4.4-4.
. Amendment 2 Fig. 6.1-4.
. Amendment 2 4.4-5.
. Original 6.2-1 thru 6.2-8.
. Amendment 2 4.5-1.
. Original Fig. 6.2-1.
. Amendment 2 l
)
5-1 thru 5-111.
. Amendment 1 6.3-1 thru 6.3-2.
. Original 5.1-1........... Amendment 1 6.3-3.
. Amendment 2
(
5.1-2.
. Original 7-1.
. Amendment 2
>!s 5.1-3.
. Amendment 1 7-11.
. Amendment 1 5.1-4 thru 5.1-9.
. Original 7-111.
. Original 5.1-10.
. Amendment 1 7.1-1 thru 7.1-20.
. Amendment 2 5.1-11 thru 5.1-24...... Original Fig. 7.1-1.
. Original 5.1-25 thru 5.1-26.
. Amendment 1 Fig. 7.1-2 thru 7.1-3.
. Amendment 2 5.1-27 thru 5.1-29.
. Original Fig. 7.1-4.
. Original 177-3D........... Amendment 1 7.2-1 thru 7.2-5..
. Original 5?l-)l............ Orig,inal 7.2-6.
. Amendment 2 5.1-32 thru 5.1-33.
. Amendment 1 T.2-7 thru 7.2-8.
. Amendment i Fig. 5.1-1 thru 5.1-3.... Original 7.2-9 thru 7.2-11.
. Original Fig. 5.1-4.
. Amendment 1 Fig. 7.2-1 thru 7.2-4.
. Original Original 7.3-1 thru 7.3-3.
. Original 5.2-1 thru 5.2-5.
5.3-1.
. Original 7.3-4 thru 7.3-7.
. Amendment 2
. Amendment 2 5.4-1.
. Amendment 1 Fig. 7.3-1.
5.4-2 thru 5.4-5.
. Original Fig. 7.3-2.
. Original
'T 5.4-6.
. Amendment 1 Fig. 7.3-3.
. Amendment 1 5.4-7.
. Original Fig. 7. 3-4 thru 7. 3-5.
. Amendment 2 5.4-8 thru 5.4-9..
. Amendment 1 7.4-1.
. Original 5.5-1 thru 5.5-3.
. Original 7.4-2 thru 7.4-5.
. '.. Amendment 1 5.6-1 thru 5.6-2.
. Original Fig. 7.4-1.
. Original 8-1 thru 8-11.
.,. Amendment 2
.\\....Amendmcqt 2 5.6-3 thru 5.6-6.
. Amendment'2 8.1-1.
.:.. Original Fig. 5.6-1.
{
5.7-1.
. Amendment 2 8.2-1 thru 8.2-18 7..'.'. Amendment 2 5.7-2.
. Original Fig. 8.2-1 thru 8.2-3.
. Amendment 2 I $5
<114 D
Amendment 2
Docket No. 50-312 ly LIST OF Apr d. U, 8
)
EFFECTIVE PAGES Amendmcat No. 2 Page or Fig. No.
Issue Page or Fig. No.
Issue e
8.3-1 thru 8.3-3...... Amendment 2 12-1.
. Amendment 1 8.3-4.
. Amendment 1 12.1-1...
. Original 8.4-1.........
. Original ' 12.2-1 thru 12.2-2.
. Amendment 2 9-1.
. Original Fig. 12.2-1....
. Amendment 2 9-11 thru 9-111..
. Amendment 2 12.3-1 thru 12.3-2.
. Amendment 2 9.0-1 thru 9.0-2....... Original 12.3-3 thru 12.3-5.
. Amendment 1 Fig. 9.0-1.......
. Original Fig. 12.3-1.
. A:uendment 2 9.1-f thru 9.1-2.
. Amendment 2 12.4-1.
. Original 9.1-3.
. Original 12.5-1..
. Original 9.1-4 thru 9.1-7.
. Amendment 2 12.6-1.
. Original 9.1-8.
. Original 12.7-1..
. Original Fig. 9.1-1.......... Amendment 2 13-1.
. Original 9.2-1.
. Original
.13.1-1 thru 13.1-2.
. Original 9.2-2 thru 9.2-7....
. Amendment 2 13.1-3.
. Amendment 2 9.2-8....
... Original 13.2-1.
.. Original 9.2-9........... Amendment 2 13.3-1.
. Original Fig. 9.2-1..
. Original 14-1.
. original 9.3-1 thru 9.3-6...
. Amendment 2 14-11 thru 14-111.
. Amendment 1 Fig. 9.3-1 thru'9.3-3.
. Amendment 2 14-iv thru 14-viii.
. Original f((3 9.4-1 thru 9.4-4...
. Original 14.1-1 thru 14.1-7.
. Original
,/
Fig. 9.4-1......... Amendment 1 14.1-8 thru 14.1-10.
. Amendment 2 9.5-1 thru 9.5-3.
. Amendment 2 14.1-11 thru 14.1-15..
. Original
. Original 14.1-16.
. Amendment 2 9.5-4..
Fig. 9.5 *
.tru 9.5-2.
. Amendment 2 14.1-17 thru 14.2-19.
. Original 9.6-1..
. Amendment 2 14.1-20.
. Amendment 2 9.6-2 thru 9.6-7.
. Original Fig. 14.1-1 thru 14.1-21.
. Original
.. Original.14.2-1.
. original
', Fig. 9.6-1.
m 9.7-1 thru 9.7-2...
. Original 14.2-2.
. Amendment 2 Iig.9.7-1.
. Amendment 2 14.2-3...
. Amendment 1 10-1...
. Original
-14.2-4.
. Original 10.1-1.
. Original 14.2-5 thru 14.2-6.
. Amendment 1 10.2-1.
. Amendment 2 14.2-7 thru 14.2-9.
. Original 10.2-2..
. Original 14.2-10.
. Amendment 2 Fig. 10.2-1......
. Original 14.2-11 thru 14.2-23.
... Original 10.3-1 thru 10.3-2.
. Original 14.2-24.'.
..'. Amendment 2 10.4-1............ Original 14.2-25.
. Original 11-1 thru 11-11.. '..
. Amendment i 14.2-26 thru 14.2-30.
. Amendment 2 11.1-1 thru 11.1-4..
. Amendment 1 14.2-31 thru 14.2-32.
. Original 11.1-5....
. Amendment 2 14.2-33.
. Amendment 2 11.1-6 thru 11.1-8..
. Amendment 1 14.2-34.
. Original Fig. 11.1-1, 11.1-2.
.. Original 14.2-35 thru 14.2-37.
. Amendment 1 11.2-1.
. Amqndment 1 14.2-38.
. Amendment 2 11.2-2 thru 1112-5...
.. Original 14.2-39...*,..,,..... original 11.2-6'
. Amendment 1 Fig. 14.2-1 thru 14;2-18.
. Original
(
)
11.2-7 thru 11.2-11.
. Original Fig. 14.2-19 thru 14.2-20. Amendment 2 pj 11.3-1....
. Original Fig. 14.2-21 thru 14.2-28.
. Original Amendment 2 E
=
LIST OF Docket No. 50-312
- g*s 1
P E F F ECTIV E PAGES A ril 15,1968 P
Amendment No. 2 Page or Fig. No Issuo Page or Fig. No.
Issue e
Fig. 14.2-29.
.... Amendment 2 Preliminary Projections to j
Fig. 1s. 2-30 thru 14.2-31... Origins 1 1985-1 thru 4
. Original Fig. 14.2-32.
. Amendment 2 2C* Geology and Seismology-Fig. 14.2-33.
. Original Fig. 14.2-34.
. Amendment 2 Fig. 2C-1 thru 2C-11
. Original Fig. 14.7-35..
. Original Geophysical Report-
-)
Fig. 14.2-36....... Amendment 1 1 thru 6.
. Original Fig. 14.2,37 thru 14.2-47... Original Additional Seismic Exploration-Fig. 14.2-48 thru 14.2-50. Amendment 1 1 thru 2, Plate 1 thru 14.3-1 thru 14.3-2.
. Original Plate 3.
. Original 14.3-3 thris 14.3-8.
. Amendu.ent 1 Geological Log of Drill 14.3-9 thru 14.3-10..
. Amendment 2 Holes-91 Sheets.
. Original
' 14.3 -11 thru 14. 3-13. s... Amendment 1 2D Seismic Report 14.3-14.
. Amendment 2 Seismic Hazard at the Fig. 14.3-1 thru 14.3-3.
. Original Clay Site, I thru 14.
. Original Fig. 14.3-4 thru 14.3-5.. Amendinent 1 Addendum to Seismic Hazard Fig. 14.3-6.
. Original at the Clay Site-1 sheet. Original Fig. 14.3-7.
. Amendment 1 Seismic Hazard at the 14.4-1 thru 14.4-2.
.Criginal Sierran Sites Area 15-1 thru 15-5.
...Origina!
I thru 10.
. At.endment 1
,20 q
) Appendix 1 Table of 2E Soil and Foundations th Contents.
Amendment 2 Investigation Report 1A Answers to Questions.
Amendment 1 2E-1 thru 2E-ll, IA-1 thru 1A-14.
Amendment 1 Fig. C-119-E thru Fig. LA.2-1.
Amendment 1 C122-E,
. Original 1A-15 thru LA-16.
. Amendment 1 Report of Laboratory 1 A-17 thru 1A-23.
Amendment 2 Testing-1 thru 9, 3 Tables, 1B,Qpality Assurance Fig. I thru 2 and 9, curves
"*Dpe,ra tions.
... Amendment 1 1 thru 7.
Original IB-1Y...........
Original 2F Meteorological Station 18-2 thru 1B-4.
. Amendment i 1B-5.
Amendment 2 2G Storage Reservoir Criteria Fig. 13-1.
. Amendment 1 1C Rancho Seco Project 2GA-1.
. Amendment i e
Engineering Staff.
Amendment 2 2H Answers to Questions
~
IC-1 thru LC-4.
.Am'endment 2 Fig. IC-1.
Amendment 2 Letter pg. I and 2.
. Amendment 2 cp d'
Appendix 2 Table of Fig. 2H.2-1 thru 2H.2-2. Amendment 2 Contents.
Amendment 2 Appendix 3 Table of 2A Final Report Contents.
. Amendment 1 i thru vi.
Original 3A Answers to Questions ~
1 thru 63....
. Amendment 1
-Supplement Fig. 3A.2-1 thru,.3A.2-3.,. Amendment 1 1 thru 18.
.i.
Amendment 2 Fig. 3A.4-1.
... Amendment 1 2B Southeast Area Plan 3A-15 thru 3A-23.'..'.. Amendment 2 (Bound)-14 pages.
Original Fig. 3A.14-1.
.Amendm nt 2
{
%)
s-F Amendment 2 i
e C.
Docket No. 50-312 LIST OF April 15,1968 i
E F F ECTIV E PAGES Amendment No. 2 L
i Page or Fig. No.
Issue Page,or Fig. No.
... Amendment 2 Figure SG-1.
Original Appendix 4 Table of 5H Quality Control Procedure Contents........ Amendment i for Field Welding 4A Answers to Questions
Original 4A-1 thru 4A-6 SI Containment Structure Fig. 4A.1-1 thru 4A.1-15. Amendment 1 Instrumentation Appendix A SI-1.
Original The Properties and Micro-5J Answers to Questions structure of Spray-Quenched SJ-1 thru SJ-2.
. Amendment 1 Thick-Section Steels Fig. 5J2-1.
. Amendment i 15 pages.*.
Original 5J-3 thru SJ-14.
. Amendment 2 Original Appendix 6 Table of 4A-1.
. Amendment 1 Contents.
. Amendment 1 Appendix 8, B & W Data 6A Answers-to Questions.
(2 pgs.)
. Amendment 1 6A-1.
. Amendment 1 6A-2
. Amendment 2 4A-13 thru 4A-18.
. Amendment 1 Appendix 5 Table of 6A-6 thru 6A-7
. Amendment 2
)
Contents.
. Amendment i Fig. 6A.8-1.
. Amendment 2 (x_)
SA Structural Design Bases 6A-8 thru 6A-18.
. Amendment 2 SA-1 thru SA-5....
. Amendment 2 Fig. 6A.16-1 thru Fig. 5A-6 thru SA-7.
. Amendment 1 6A.16-2
. Amendment 2 Fig. SA-1 thru 5A-2.
. Amendment 1 Appendix 7 Table of SB Justification of Contents.
. Amendment 1 Structural Proof Test -
7A Answers to Questions Pressures 7A-1.
. Amendment 1
,,,.TS-1 thru 5B-3.
. Amendment 2
, 5g Specification for Splicing Appendix 8 Table of Reinforcing Bar Using the Contents.
. Amendment 2 Coldwell Process 8A Answers to Questions 5C-1 thru 5C-3...
. Amendment 2 SD Turbine Generator Missiles Appendix 9 Table of 1 thru 10.
Original Contents.
. Amendment 2 4 sheets of Parts Drawings 9A Answers to Qucstions-SE Justification for Load 9A-1.
. '.. Amendment 2 Factors 9A-2.
. Amendment 1
. Amendment 2 SF Justification for Yield Appendix 10 Contains nothing Reduction Factors.
Appendix 11 Table of 5F-1 thru SF-2'.
Original Contents
. Amendment 1 3G Description of the Finite 11A Answers to Questions Element Techn%que Used in#' 'n 11A-1 thru'llA-2.:.
. Amendment 1 Containment St'ructural Fig. 11A.1-1.
. Amendment 1
(
i Analysis Fig. IlA.1-2..
. Amendment 2
{ ~7[
3G-1 thru SG-2 Original 11A-3 thru llA-6.
. Amendment 1 Amendment 2 G
4
' m-Docket No. 50-312
)
LIST OF I x- /
~
E F F ECTIV E PAGES April 15, 1968 Y"'-
Amendment No. 2 Page or Fig. No.
Issue Page or Fig. No.
Issue
)
Appendix 12 Table of Appendix 15 Table of Contents..
. Amendment 1 Contents.
. Amendment 2 12A Answers to Questions 15'A Answers to Questions 12A-1 thru 12A-4
. Amendment i 15A-1 thru 15A-2.
. Amendment 2 12A-5 thru 12A-10.
. Amendment 2 Fig. 12A.5-1.
. Amendment 2 12A-11.
. Amendment 2 Fig. 12A.6-1 thru 12A.6-3......... Amendment 2 12A-12 thru 12A-17.
. Amendment 2 Appendix 13 Table of Contents..'.
. Amendment 2 13A Answers to Ques'tions L3A-1 thru 13A-3.... Amendment 2 Appendix 14 Table of Contents.
. Amendment i 14A Answers to Questions 14A-1.
. Amendment 1 14A-2.
. Amendment 2
. Amendment 1 1
7s pS*TN 14A-7 thru 14A-9.... Amendment 2 133f x,
Fig. 14A.6-1 thru 14A.6-3......... Amendment 1 Fig. 14A.6-4 thru 14A.6-5......... Amendment 2 14A-11 thru 14A-13.
. Amendment 1 14A-14.
. Amendment 2 F4 g. 14A.8-1.
. Amendment 2 14A-15 thru 14A-20.
. Amendment 1
' F)$. 14A.11-1 thru 14A.11-2
. Amendment 1 14A-21 thru 14A-22
. Amendment 1 14A-23 thru 14A-29... Amendment 2 J
Fig. 14A.18-1.
. Amendment 2 c,
14A-30. ~........ Amendment 2
~
Fig. 14A.19-1.
. Amendment 2 14A-31 thru 14A-32
. Amendment 2 Fig. 14A.21-1 thru 14A.21-4.
. Amendment 2 14A-33 thru 14A-34
. Amendment 2 Fig. 14A.22-1.
. Amendment 2 14A-35 thru 14A-36... Amendment 2 Fig. 14A.25-1.
.s.
. Amendment 2 Fig. 14A.26-1.. \\.... Amendment 2
.l l
. Amendment 2 l.'
O. \\l r
I H
Amendment 2
A
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Docket No. 50-312 February 2, 1968 AMENDMENT NO 1
- SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1
~
Amendment No.1 to the Sacramento Municipal' Utility District's Prelimi-nary Safety Analysis Report includes both replacement.pages and new
. pages and-tabs. All pages to be inserted are identified as Amendment 1.
Any technical text material changed by_this amendment is coded in the outside. margin-by a black-bar and the numeral one.
Before inserting the Amendment 1 material (contained in this new Volume V) in the different volumes, it is suggested that the Appendix 5 material be removed from Volume IV to provide space. After the Amendment 1 material has _ been inserted, Appendix 3 should be the first amendment in the new
.(('"%).
Volume V.- 1te List of Effective Pages should be checked to verify the completeness of Volumes I-thru V.
It should be noted that-License Application page 4 is replaced with a-
~ new page 4 plus two new additional pages, 8 and 9.
d r
77W
'A
o SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 Docket No. 50-285 [
LIST OF gg, E F F E CTIV E PAGES February 2, 1968 i
Amendment No. 1 The active pages in this report are as follows:
Page or Fig. No.
Issue Page or Fig. No.
Issue Ti,le.......................Criginal 3.3-1 thru 3.4-5............. Original A......................... Amendment 1 4-1 thru 4-iv................ original B th ru E.................. Amendmen t 1 4.1-1 thru 4.3-10............ Original i........................... Original 4.3-11.................... Amendment 1 ii........................ Amendment 1 4.4-1 thru 4.5-1............. original iii......................... 0riginal 5.1-1..................... Amendment 1 iv thru ix............... Amendment 1 5.1-2........................ Original x........................... Original 5.1-3..................... Amendment 1 xi thru xiv.............. Amendment 1 5.1-4 thru 5.1-9............. Original 1-i..........................Original 5.1-10.................... Amendment 1 1-ii...................... Amendment 1 5.1-11 thru 5.1-24........... Original 1-111 thru iv...............
0riginal 5.1-25 thru 5.1-26........ Amendment 1 1-v.......................... Original 5.1-27 thru 5.1-29........... Original 1.1-1 thru 1.1-2............. Original 5.1-30.................... Amendment 1 Fig. 1.1-4 thru 1.1-8..... Amendment 1 5.1-31....................... original
- 1. 2 thru 1. 2 -4............. original 5.1-32 thru 5.1-33........ Amendment 1 1.3-1 thru 1.3-3............. original Fig.
5.1-4................ Amendment 1 1.3-4..................... Amendment 1 5.2-1 thru 5.3-1............. Original
- 1. 3
.5 thru 1.3-9............. Original 5.4-1..................... Amendment 1 1.4-1........................Original 5.4-2 thru 5.4-5............. Original 1.4-2 thru'1.4-3.......... Amendment 1 5.4-6..................... Amendment 1 1.4-4.thru 1.9-1............. Original 5.4-7........................ Original 2-1 thru 2-ii............. Amendment 1 5.4-8 thru 5.4-9.......... Amendment 1 2-iii........................ Original 5.5-1 thru 5.9-1............. original 2.1-1 thru 2.3-5............. original 6-1 thru f -ii............. Amendment 1 2.3-6 thru 2.3-8.......... Amendment 1 6.0-1........................Originct 2,4-1 thru 2.5-1............. Original 6.1-1.....
.................. Original 2.6-1..................... Amendment 1 6.1-2 thru 6.1-16......... Amendment 1 2.7-1........................ Original Fig. 6.1-1................ Amendment 1 2.8-1 thru 2.8-4.......... Amendment 1 6.2-1..................... Amendment 1 2. 9 - 1........................ or igina l 6.2-2 thru 6.2-8............. Original 5-i.thru 3-vi................ Original Fig. 6.2-1................ Amendment 1 3.1-1 thru 3.1-6............. Original 6.3-1 thru 6.3-3............. Original Fig. 3.2-65............... Amendment 1-7-i.......................... Original Fig. 3.2-68............... Amendment 1 7 -ii...................... Amend me n t 1 B
j'm N-SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 L I S 'T OF Docket No. 50-289 E F F ECTIV E PAGES February 2, 1968 Amendment No. 1 Page or Fig. No.
Issue Page or Fig. No, Issue 7.1-1 thru 7.1-5............. Original 11.2-2 thru 11.2-5........... Original 7.1-6 thru 7.1-7.......... Amendment 1 11.2-6.................... Amendment 1 7.1-9........................ Original
- 11. 2-7 thru 11. 2-1............ Original 7.1-10 thru 7.1-11........ Amendment 1 12-1.......................... Original 7.1-12 thru 7.1-13........... Original 12.1-1........................ Original 7.1-14 thru 7.1-15........ Amendment 1 12.2-1 thru 12.2-2......... Amendment 1 7.1-16 thru 7.1-19........... Original Fig.
12.2-1................ Amendment 1 Fig. 7.1-2................knendment 1 12.3-1 thru 12.3-5......... Amendment 1 7.2-1 thru 7.2-6............. Original 12.4-1 thru 12.7-1............ Original 7.2-7 thru 7.2-8.......... Amendment 1 13-i.......................... Original 7.2-9 thru 7.2-11............ Original 13.1-1 thru 13.3-1............ original
-~ s l
)
7.3-1 thru 7.3-7............. Original i........................... 0riginal
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F i g. 7. 3 -3................ Ame nd men t 1 11 thru lii............... Amendment 1 7.4-1........................Original 14.1-1 thru 14.1-8............ Original 7.4-2'thru 7.4-5.......... Amendment 1 14.1-9..................... Amendment 1 8-1 thru 8-ii............. Amendment 1 14.1-10 thru 14.1-19.......... Original 8.1-1 thru 8. 2-1............ 0t iginal 14.1-20.................... Amendment 1 8.2-2 thru 8.2-17......... Amendment 1 14.2-1 thru 14.2-2............ Original Fig. 8.2-1 thru 8.2-3..... Amendment 1 14.2-3..................... Amendment 1 8.3-1........................ Original 14.2-4........................ Original 8.3-2 thru 8.3-4.......... Amendment 1 14.2-5 thru 14.2-6......... Amendment 1 8.4-1........................ Original 14.2-7 thru 14.2-32........... Original 9-1 thru iii................. original 14. 4-3 3.................... Amendmen t 1 9.0-1.thru 9.2-9............. Original 14.2-34....................... Original 9.3-1..................... Amendment 1 14.2-35 thru 14.2-37....... Amendment 1 9.3-2 thru 9.4-4............. Original 14.2-38 thru 14.2-39.......... Original Fig.
9.4-1................ Amendment 1 Fig. 14.2-36............... Amendment 1 9.5-1 thru 9.5-2.....
..... Original Fig. 14.2-48 thru 14.2-50.. Amendment i 9. 5 -3................
... Amendment 1 14.3-1 thru 14.3-2............ Original
'I 9.5-4........................ original 14.3-3 thru 14.3-14........ Amendment 1 Fig.
9.5-1................ Amendment 1 Fig. 14.3-4 thru 14.3-5.... Amendment 1 9.6-1 thru 9.6-7............. Original Fig. 14.3-7................ Amendment 1 10-1......................... Original 14.4-1 thru 14.4-2............ Original 10.1-1.thru 10.4-1........... Original 15-1 thru 15-5................ Original 11-1......................... Original Appendix 1 11-ii..................... Amendment 1-Table of Contents....... Amendment 1
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,11.1-1 thru 11.2-1........ Amendment 1 1A 1A-1 thru 1A-17....... Amendment 1 iQ)
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SACRAMENTO MUNICIPAL' UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 Docket No. 50-289 LIST O F Uk.
- E F F E CTI V E PAGES February 2, 1968 Amendment'No. 1 Page or Fig. No..
Issue Page or Fig. No.
Issue 1B 1B-1 thru 1B-5........ Amendment 1 2G A-1................... Am en dmen t 1 Appendix 2 Appendix 3 Table of Contents....... Amendment 3 Table of Contents....... Amendment 1 2A Title Page................ Original 3A 3A-1 thru 3A-14....... Amendment 1 i thru'vi.................
0riginal 4A Questions-4A-1 1.thru 63.................. original thru 4A-8............... Amendment 1 2B Southeast Area Plan Appendix A to Question (Bound)-17 pages........... Original 4A.2-1 thru 15.......... Amendment 1 Preliminary Projections...
Appendix B to Question to 1985-1 thru 4........... Original 4 4. 2-1 thru 2........... Amendment 1 2C Geology and Seismology-Questions-4A-9 thru 2C-1 thru 2C-13, Fig 2C-1 4A-12................... Am en dmen t 1 thru 2C-11................. Original Appendix 5 Geophysical Report-Table of Contents....... Amendment 1 1
thru'6................... original 5A SA-1..................... Original Additional Seismic Exploration-SA-2 thru 5A-6.......... Amendment I l thru 2, Plate 1 thru Fig SA-1 thru SA-2...... Amendment 1 Plate 3.................... Original 5B SB-1 thru 5B-3........... Original Geological Log of Drill 5C 5C-1 thru SC-3........... Original Holes-91 Sheets............ Original SD 1 thru 10................ Original 2D Seismic Hazard at the SE SE-1 thru 5E-2........... Original Clay Site-1 thru 14........ Original 5F SF-1 thru SF-2........... Original Addendum to Seismic Hazard 5G SG-1 thru 5G-2........... Original at the Clay Site-1 sheet... Original SH 5H-1 thru 5H-5........... original 1
Seismic Hazard at the 51 SI-1..................... original Sierran Sites Area 5J 5J-l thru SJ-2........ Amendment 1 1 thru 10............... Amendment 1 Appendix 6 2E Soil and Foundations Table of Contents....... Amendment 1 Investigation Report 6A-1 thru 6A-6.......... Amendment 1 2E-1 thru 2E-ll, Fig C-119-E Appendix 7 thru C122-E................ Original Table of Contents....... Amendment 1 Report of Laboratory 7A-1.................... Amendment 1 Tes' ting-1 thru 9,.3 Tables, Appendix 9
' Fig 1 thru 2 and 9, curves Table of Contents....... Amendment 1 1 thru 7................... Original 9A-1 thru 9A-2.......
.. Amendment 1 2F ~2F-1 thru 2F-2........ Amendment 1 Appendix 11 2G 2G-1 thru 2G-3........ Amendment 1 Table of Contents....... Amendment 1 f
D e
l SACRAMENTO MUNICIPAL UTILITY DISTRICT i
l l
RANCHO SECO NUCLEAR GENERATING SIATION l
l ENIT NO. 1 t
Docket No. 50-289 LIST OF 4g E F F E CTIV E. P A G ES February 2, 1968 Amendment No. 1 Page or Fig. No.
Issue m.
11A-1 thru llA-6........ Amendment 1 Appendix 12 Table o f Contents....... Amendment 1 12A-1 thru 12A-4........ Amendment 1 Appendix 14 Table of Contents....... Amendment 1 14A-1 thru 14A-22....... Amendment 1 l
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TABLE 07 CONTENTS 7ykj VOLLME I
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1.
INTRODUCTION AND
SUMMARY
~Section Page q
1.1 I_NTRODUCTION.
1.l-1 1.2 DESIGN HIGHLIGHTS' 1.2-1
'1.2.1 SITE CHARACTERISTICS-1.2-1
-1.2.2
_-POWER LEVEL.
1.2-1
.1.2.3 PEAK SPECIFIC POWER LEVEL
- 1.2-1
-1.2.4.
REACTOR BUILDING 1.2-1
. -J 1.2.5 ENGINEERED SAFEGUARDS 1.2-2 1.2.6 ELECTRICAL ~ SYSTEMS AND-EMERGENCY POWER 1.2-3 1.2.7 ONCE-THROUGH STEAM GENERATORS 1.2-4 1.3 TABULAR CHARACTERISTICS 1.3-1 1.3.1 ITEM l - HYDRAULIC AND THERMAL DE3IGN PARAMETERS 1.3-1 1.3.2
-ITEM 2 - CORE-MECHANICAL DESIGN PARAMETERS 1.3-1 1.3.3 ITEM 3-- PRELIMINARY NUCLEAR DESIGN DATA 1.3-8 1.3.4
. ITEM 4 - PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT SYSTEM :
1.3-8 1.3.5 ITEM 5 - REACTOR C00IANT SYSTEM - CODE REQUIREMENTS 1.3-8
[(h l'. 3. 6 ITEM 6 - PRINCIPAL DESIGN PARAMETERS OF THE REACTOR j
,/
VESSEL 1.3-9
'1.3.7 ITEM 7 - PRINCIPAL DESIGN FEATURES OF THE STEAM GENERATORS 1.3-9
.1.3.8' ITEM 8 - PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PUMPS 1.3-9
- 1. 3. 9 - ITEM 9 - PRINCIPAL DESIGN PARAMETERS OF THE REACTOR C001 ANT PIPING 1.3-9 1.3.10 ITEM 10 - REACTOR BUILDING PARAMETERS 1.3-9 1.3.11 _ ITEM 11 - ENGINEERED SAFEGUARDS '
l.3-9 1.4 PRINCIPAL DESIGN CRITERIA 1.4-1
- 1. 4'.1 -
CRITERION 1 - QUALITY STANDARDS.(CATEGORY A) 1.4-1 1.4.2 CRITERION 2 - PERFORMANCE STANDARDS (CATEGORY A) 1.4-2 1.4.3 CRITERION 3 - FIRE PROTECTION (CATEGORY A) 1.4-3 1.4.4
' CRITERION 4 - SHARING-OF SYSTEMS;(CATEGORY A) 1.4-5 1.4.5 CRITERION 5 - RECORDS-REQUIREMENTS (CATEGORY A) 1.4-5 l.4.6-
' CRITERION 6 - REACTOR CORE DESIGN (CATEGORY A) 1.4-5
,E
-1.4.7 CRITERION ~7 -_ SUPPRESSION OF-POWER OSCILIATIONS (CATEGORY B) 1.4-6 1.4.8
- CRITERION 8 ~ - OVERALL POWER COEFFICIENT -(CATEGORY B) 1.4-7 1.4.9 CRITERION 9 - ~ REACTOR C001 ANT PRESSURE BOUNDARY (CATEGORY A) 1.4-7
- 1. 4. CRITERION 10 - CONTAINMENT (CATEGORY A) 1.4-8
~
j
- 1.4.11 CRITERION 11 - CONTROL ROOM '(CATEGORY B) 1.4-8 1.b) 1
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Amendment-3 i
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Section-Page 1.4.12 ' CRITERION 12 - INSTRUMENTATION AND CONTROL SYSTDIS (CATEGORY B).
1.4-10 1.4.13. CRITERION 13 - FISSION PROCESS MONITORS AND CONTROLS (CATEGORY B) 1.4-11 1.4.14 CRITERION 14 - CORE PROTECTION SYSTEMS (CATEGORY B) 1.4-12 1.4.15 CRITERION 15 - ENGINEERED SAFETY FEATURES PROTECTION SYSTEMS (CATEGORY B) 1.4-12 G-1.4.16 CRITERION 16 - MONITORING REACTOR COOLANT PRESSURE BOUNDARY (CATEGORY B) 1.4-12 1.4.17 CRITERION 17 - MONITORING RADIOACTIVITY RELEASES L
(CATEGORY B) 1.4-13 1.4.18 CRITERION 18 - MONITORING FUEL AND WASTE STORAGE (CATEGORY B) 1.4-15 1.4.19 CRITERION 19 - PROTECTION SYSTEMS RELIABILITY (CATEGORY B) 1.4-15 1.4.20 CRITERION 20 - PROTECTION SYSTDiS REDUNDANCY AND INDEPENDENCE (CATEGORY B) 1.4-16 1.4.21 CRITERION 21 - SINGLE FAILURE DEFINITION (CATEGORY B) 1.4-16 1.4.22 CRITERION 22 - SEPARATION OF PROTECTION AND CONTROL INSTRUMENTATION SYSTDIS (CATEGORY B) 1.4-16 1.4.23 CRITERION 23 - PROTECTION AGAINST MULTIPLE DISABILITY FOR PROTECTION SYSTEMS (CATEGORY B) 1.4-17 1.4.24 CRITERION 24 - EMERGENCY POWER FOR PROTECTION SYSTEMS (CATEGORY B) l'.4-17 1.4.25 CRITERION 25 - DDIONSTRATION OF FUNCTIONAL OPERABILITY 0F PROTECTION SYSTDIS (CATEGORY B) 1.4-17 1.4.26 CRITERION 26 - PROTECTION SYSTEMS FAIL-SAFE DESIGN (CATEGORY B) 1,4-18 1.4.27 CRITERION 27 - REDUNDANCY OF REACTIVITY CONTROL
.(CATEGORY A) 1.4-19
~
1.4.28 CRITERION 28 - REACTIVITY HOT, SHUTDOWN CAPABILITY (CATEGORY A) 1.4-19 1.4.29 CRITERION 29 - REACTIVITY SHUTDOWN CAPABILITY (CATEGORY A) 1.4-19 1.4.30 CRITERION 30 - REACTIVITY HOLDDOWN CAPABILITY (CATEGORY B) 1.4-20 1
1.4.31 CRITERION 31 - REACTIVITY CONTROL SYSTEMS MALFUNCTION (CATEGORY B) 1,4-20
~
1.4.32 CRITERION 32 - MAXIMUM REACTIVITY WORTH OF CONTROL RODS'(CATEGORY A) 1.4-20 1.4.33 CRITERION 33 - REACTOR COOIANT PRESSURE BOUhTARY CAPABILITY (CATEGORY A) 1.4-21 1.4.34 ~ CRITERION 34 - REACTOR COOLANT PRESSURE BOUNDARY RAPID PROPAGATION FAILURE PREVENTION (CATEGORY A) 1.4-21 Y
1.4.35 CRITERION 35
. REACTOR COOLANT PRESSURE BOUNDARY BRITTLE FRACTURE PREVENTION (CATEGORY A) 1.4-22 1.4.36 CRITERION 36 - REACTOR C00IANT PRESSLEE BOUNDARY SURVEILLANCE (CATEGORY A) 1.4-22 11 Amendment 3
.A f('
Section Page Q
-l~.4.37' CRITERION 37 - ENGINEERED SAFETY' FEATURES BASIS FOR
~
DESIGN (CATEGORY A) 1.4-22 1.4.38 ' CRITERION 38 - RELIABILITY AND TESTABILITY OF ENGINEERED SAFETY FEATURES (CATEGORY A) 1.4-23 1.4.39 i CRITERION 39
. EMERGENCY POWER FOR ENGINEERED SAFETY FEATURES.(CATEGORY A) 1.4-24 1.4.40 CRITERION 40. -- MISSILE PROTECTION (CATEGORY A) 1.4-24 1.4.41 CRITERION 41 - ENGINEERED SAFETY FEATURES PERFORMANCE CAPABILITY (CATEGORY A) 1.4-25 l'.4.42 CRITERION 42 -. ENGINEERED SAFETY. FEATURES COMPONENTS
' CAPABILITY - (CATEGORY A) 1.4-25 1.4.43 CRITERION 43 - ACCIDENT AGGRAVATION PREVENTION
.~
. (CATEGORY.A).
~
1.4-26
' l.'4.44' CRITERION 44 - EMERGENCY CORE COOLING SYSTEMS CAPABILITY -(CATEGORY A) -
- 1. 4 -1.6
-1.4.45 -CRITERION 45 - INSPECTION OF EMERGENCY CORE COOLING SYSTDIS (CATEGORY A) 1.4-27
^
1.4.46: CRITERION 46 - TESTING OF EMERGENCY CORE COOLING
. SYSTEMS COMPONENTS (CATEGORY A) 1.4-27
.l.4.47 ' CRITERION 47 - TESTING OF EMERGENCY CORE COOLING SYSTDIS (CATEGORY A) 1.4-28 1.4'.48 CRITERION 48 - TESTING OF OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS (CATEGORY A) 1.4-28
- f 1.4.49 CRITERION 49 - CONTAINMENT DESIGN BASIS (CATEGORY A) 1.4-28
~ -(,
1.4 50 - CRITERION.50 - NDT REQUIREMENT FOR CONTAINMENT MATERIAL (CATEGORY A) 1.4-29 1.4.51 CRITERION 51
. REACTOR C00IANT PRESSURE BOUNDARY 0UTSIDE CONTAINMENT -(CATEGORY A) 1.4-29
.1.4.52 CRITERION 52 - CONTAINMENT HEAT REMOVAL SYSTEMS (CATEGORY A) 1.4-30
=1.4.53 CRITERION 53 - CONTAINMENT ISOIATION VALVES (CATEGORY A) 1.4-31 1.4.54 CRITERION 54
-CONTAINMENT LEAKAGE RATE TESTING (CATEGORY A) 1,4-31 1.4.55. CRITERION 55 - CONTAINMENT PERIODIC LEAKAGE RATE TESTING (CATEGORY A) 1.4-32 1.4.561: CRITERION 56 - PROVISIONS FOR TESTING OF PENETRATIONS (CATEGORY A)'
l.4-32 1.4.57 CRITERION 57 :- PROVISIONS'FOR TESTING OF ISOIATIONS VALVES. (CATEGORY ~ A) 1.4-32 1.4.58' ' CRITERION 58 - INSPECTION OF CONTAINMENT PRESSURE-REDUCING ~ SYSTDIS (CATEGORY A) 1.4-33 T
1.4.59 CRITERION 59 - TESTING OF CONTAINMENT PRESSURE-REDUCING SYSTDIS COMPONENTS (CATEGORY A) 1.4-33
'l.4.60 CRITERION 60 - TESTING OF CONTAINMENT SPRAY SYSTEMS
..(CATEGORY A) 1.4-34
=
1.4 61.. CRITERION 61
. TESTING OF OPERATIONAL SEQUENCE OF CONTAINMENT PRESSURE-REDUCING SYSTDIS (CATEGORY A) 1.4-34 O.a AnendmentL3.
iii
Section Page
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1.4.62 CRITERION 62 - INSPECTION OF AIR CLEANUP SYSTDiS (CATEGORY A) 1.4-35 1.4.63 CRITERION 63 - TESTING OF AIR CLEANUP SYSTEMS COMPONENTS (CATEGORY A) 1.4-35 1.4.64 ' CRITERION 64'- TESTING OF AIR CLEANUP SYSTEMS (CATEGORY A) 1.4-35 1.4.65 ' CRITERION 65 - TESTING OF OPERATIONAL SEQUENCE OF AIR CLEANUP SYSTEMS (CATEGORY A) 1.4-36 Q-1.4.66 -CRITERION 66 - PREVENTION OF FUEL STORAGE CRITICALITY (CATEGORY B) 1.4-36 1.4.67 CRITERION 67 - FUEL AND WASTE STORAGE DECAY HEAT (CATEGORY B) 1.4-36 1.4.68 CRITERION 68 - FUEL'AND WASTE STORAGE RADIATION SHIELDING (CATEGORY B)'
1.4-37 1.4.69 CRITERION PROTECTION AGAINST RADIOACTIVITY RELEASE FROM SPEET FUEL AND WASTE STORAGE (CATEGORY B) 1.4-37 1.4.70 CRITERION 70 - CONTROL OF RELEASES OF RADIOACTIVITY TO THE EWIRONMENT (CATEGORY B) 1.4-38 1.5 RESEARCH AND DEVELOPMENT REQUIRDIENTS 1.5-1 1.5.1 XENON OSCILLATIONS 1.5-1'
- 1. 5. 2 THERMAL AND HYDRAULIC PROGRAMS 1.5-1 1.5.3 FUEL ROD CLAD FAILURE 1.5-2
- 1. 5. 4 HIGH BURNUP FUEL TESTS 1.5-3 1.5.5 INTERNALS VENT VALVES 1.5-3 1
1.5.6 CONTROL ROD DRIVE TEST 1.5-4
.)
1.5.7 ONCE-THROUGH STEAM GENERATOR TEST 1.5-4 1.5.8 SELF POWERED DETECTOR TESTS 1.5-5 1.5.9 BLOWDOWN FORCES ON INTERNALS 1.5-5 1.5.10 RADIO IODINE SPRAY REMOVAL SYSTEM 1.5-6 1.6 SMUD'S COMPETENCE TO BUILD AND OPERATE NUCLEAR PLANT 1.6-1 1.7 IDENTIFICATION OF CONTRACTORS AND AGENTS 1.7-1
1.8 CONCLUSION
S 1.8-1
'~
1.9 REFERENCES
1.9-1 2.
SITE AND ENVIRONMENT 2.1
SUMMARY
2.1-1 2.2 SITE AND ADJACENT AREAS 2.2-1 k
2.2.1 SITE LOCATION 2.2-1 2.2.2 SITE OWNERSHIP 2.2-1 2.2.3.
' SITE ACTIVITIES 2.2-1
- 2. 2.4 '
POPULATION.
2.2-1 2.2.5 LAND USE 2.2-3 2.2.6~
ACCESS AND EGRESS 2.2 3 2.2.7 MAKE-UP WATER SUPPLY 2.2-5 iv Amendment.4
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Page
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-Section.
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2'.3
? METEOROLOGY 2.3-1
- 2. 3. l'-
INTRODUCTION 2.3-1 2.3.2 DESCRIPTIVE METEOROLOGY 2.3-1 2.3.3-METEOROLOGICAL DATA 2.3-2 2.3.4 -
PROGRAM OF METEOROLOGICAL INVESTIGATION 2.3-5
.2.3.5:-
PRELIMINARY ESTIMATES OF DIFFUSION 2.3-5
- 2.'4 HYDROLOGY 2.4-1 2.4.1: CHARACTERISTICS OF STREAMS AND LAKES IN VICINITY 2.4-1
- 2. 4 '. 2 TOPOGRAPHY 2.4-1 2.4.3.
. TERMINAL DISPOSAL OF. STORM RUNOFF
-2.4-1
.2.4.4
. HISTORICAL FLOODING-2.4-1 2.4.5 PREDICTION OF IAND URBANIZATION 2.4-1 2.4.6 GROUNDWATER 2.4-3 2.5 -
GEOLOGY 2.5-1 2.6 SEISMOLOGY 2.6-1 2.7 SOILS 2.7-1 2.8 SITE ENVIRONMENTAL RADIOACTIVITY PROGRAM 2.8-1 f ]/
2.8.1 GENERAL 2.8-1
'Q, 2.8.2 LAND ENVIRONMENT 2.8-1 2.~ 8. 3 WATER ENVIRONMENT 2.8-2 2.8.4 SAMPLING 2.8-2
2.9 REFERENCES
2.9-1 3.
REACTOR 3.1 DESIGN BASES ~
3.1-1 3.1.1 PERFORMANCE OBJECTIVES 3.1-1 3.1.2 LIMITS
'3.1-1 3.2-
- EACTOR DESIGN 3.2-1
-3~2.1 GENERAL
SUMMARY
3.2-1 3.2.2 NUCIEAR DESIGN AND EVALUATION 3.2-2
- 3.2.3 THERMAL AND HYDRAULIC DESIGN AND EVALUATION 3.2-29
'T; T3.2.4
-3.3-1
.3.3.1 NUCLEAR. TESTS AND INSPECTION 3.3-5
~3.3.2 THERMAL AND HYDRAULIC TESTS. AND --INSPECTION 3.3-2 3.3.3
-FUEL AGSEMBLY, CONTROL' ROD ASSEMBLY, AND CONTROL ROD
. DRIVE MECHANICAL TESTS AND INSPECTION 3.3-5
' f] -
3.3.4 LINTERNALS TESTS AND INSPECTIONS 3.3-10
^
3.4~
REFERENCES 3.4-1 Amendment.3 v
VOLUME II 6ection Page 4.
REACTOR COOLANT SYSTEM 4.1 DESIGN ' BASES 4.1-1 4.1.1 PERFORMANCE OBJECTIVES 4.1-1 4.1. 2 DESIGN CHARACTERISTICS 4.1-2 3-4.1.3 EXPECTED OPERATING CONDITIONS 4.1-7 4.1.4 SER" ICE LIFE 4.1-8
- 4.1. 5 '
CODES AND CLASSIFICATIONS 4.1-15 4.2 SYSTEM DESCRIPTION AND OPERATION 4.2-1 4.2.1 GENERAL DESCRIPTION 4.2-1 4.2.2 MAJOR COMPONENTS 4.2-1 4.2.3 PRESSURE-RELIEVING DEVICES 4.2-7 4.2.4 ENVIRONMENTAL PROTECTION 4.2-7 4.2.5 MATERIALS OF CONSTRUCTION 4.2-7 4.2.6 MAXIMUM HEATING AND COOLING RATES 4.2-11 4.2.7 LEAK DETECTION 4.2-11 4.3 SYSTEM DESIGN EVALUATION 4.3-1 4.3.1 SAFETY FACTORS 4.3-1
.4.3.2 RELIANCE ON INTERCONNECTED SYSTEMS 4.3-8 14.3.3 SYSTEM INTEGRITY 4.3-9
)
4.3.4 PRESSURE RELIEF 4.3-9
/
4.3.5 REDUNDANCY 4.3-10 4.3.6 SAFETY ANALYSIS 4.3-10 4.3.7 OPERATIONAL LIMITS 4.3-10 4.4 TESTS AND INSPECTIONS 4.4-1 4.4.1 COMPONENT IN-SERVICE INSPECTION 4.4-1 4.4.2 REACTOR'C00LANT SYSTEM TESTS AND INSPECTIONS 4.4-3
~
4.4.3 MATERIAL IRRADIATION SURVEILLANCE 4.4-4
4.5 REFERENCES
4.5-1 5.
CONTAINMENT SYSTEM 5.1 STRUCTURAL DESIGN 5.1-1 5.1.1 GENERAL DESCRIPTION OF CONTAINMENT STRUCTURE 5.1-1 5.1. 2 BASIS FOR DESIGN LOADS 5.1-1 5.1.3 CONSTRUCTION MATERIALS 5.1-4 5.1.4 CONTAINMENT STRUCTURE DESIGN CRITERIA 5.1-11 5.1.5 STRUCTURAL DESIGN ANALYSIS 5.1-27 5.2 DESIGN,-CONSTRUCTION, AND TESTING OF PENETRATIONS 5.2-1 5.2.1 TYPES OF PENETRATIONS 5.2-1 5.2.2 DESIGN OF PENETRATIONS 5.2-3 w
5.2.3-INSTALLATION OF PENETRATIONS 5.2-5
)
'5.2.4 TESTABILITY OF PENETRATIONS AND WELD SEAMS 5.2-5 vi
, Amendment 4
h
'. See tion Par 5.3
. CONTAINMENT ACCESSIBILITY CRITERIA 5.3-1 l
5.4 CONSiRUCTION -PRACTICES AND QUALITY ASSURANCE 5.4-1 5.4.1 ORGANIZATION OF QUALITY ASSURANCE PROGRAM 5.4-1 5.4.2' APPLICABLE CONSTRUCTION CODES 5.4-1 1
5.4.3 CONSTRUCTION MATERIALS INSPECTION AND INSTALLATION 5.4-2 5.4.4 SPECIFIC CONSTRUCTION TCPICS 5.4-7 5.5 CONTAINMENT SYSTEM INSPECTION, TESTING, AND SURVEILLANCE 5.5-1 5.5.1 TESTS TO ENSURE LINER INTEGRITY 5.5-1 5.5.2 STRENGTH TEST 5.5-3 5.6
- ISOLATION SYSTEM 5.6-1
- 5. 6.1 DESIGN BASES 5.6-1 5.6.2 SYSTEM DESIGN 5.6-1 5.7 VENTILATION SYSTEM 5.7-1 5.7.1.
DESIGN BASES 5.7-1 5.7.2 SYSTEM DESIGN 5.7-1
~
5.8 LEAKAGE MONITORING SY' STEM' 5.8-1 j
5.9 SYSTEM DESIGN EVALUATION 5.9-1 6.
ENGINEERED SAFEGUARDS 6.1 EMERGENCY INJECTION 6.1-1 6.1.1.
DESIGN BASES-6.1-1 6.1. 2 DESCRIPIION.
6.1-1 6.1.3 DESIGN EVALUATION 6.1-5 6.1.4 TEST AND INSPECTIONS 6.1-15 6.2 REACTOR BUILDING ATMOSPHERE COOLING AND WASHING 6.2-1 6.2.1 DESIGN BASES 6.2-1 6.
2.2 DESCRIPTION
6.2-1 6.2.3 DESIGN EVALUATION 6.2-2 6.2.4 TESTS AND INSPECTIONS 6.2-7 6.3
. ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS
- 6.3-1
,g 6.
3.1 INTRODUCTION
6.3-1 6.3.2
SUMMARY
OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATION 6.3-1 6.3.3 LEAKAGE ASSUMPTIONS 6.3-2 6.3.4 '. DESIGN BASIS LEAKAGE 6.3-2 6.3.5 LEAKAGE ANALYSIS CONCLUSIONS 6.3-2 i
)
'L' Amendment 3 vii b
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Section 7.
INSTRUMENTATION AND CONTROL 7.1 '
PROTECTION SYSTDiS 7.1-l
' 7.1.1 DESIGN BASES 7.1-1
'7.1.2 SYSTD1 DESIGN 7.1-6 7.1.3 SYSTEMS EVALUATION 7.1-17 b
7.2 REGUIATING SYSTEMS 7.2-1 d.
7.2.1 DESIGN BASES 7.2-1 7.2.2 SYSTD1 DESIGN 7.2-3 7.2.3 SYSTDi EVALUATION 7.2-9 7.3 INSTRUMENTATION 7.3-1 7.3.1 NUCLEAR INSTRUMENTATION 7.3-1 7.3.2 NONNUCLEAR PROCESS INSTRUMENTATION 7.3-3
~
7.3.3 INCORE MONITORING SYSTai 7.3-5 7.4 OPERATING CONTROL STATIONS 7.4-1 7.4.1 GENERAL LAYOUT 7.4-1 7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION 7.4-1 7.4.3 SUmiARY OF AIARMS 7.4-2 7.4.4 COMMUNICATION 7.4-2 7.4.5 OCCUPANCY 7.4-2 7.4.6 AUXILIARY CONTROL STATIONS 7.4-3 7.4.7 SAFETY FEATURES 7.4-4 7.4.8 SYSTEM EVALUATION 7.4-4 8.
ELECTRICAL SYSTEMS 8.1 DESIGN BASIS 8.1-1 8.2 ELECTRICAL SYSTDi DESIGN 8.2-1 8.2.1 ELECTRICAL SYSTEM DESIGN NEIWORK INTERCONNECTIONS 8.2-1 8.2.2 STATION DISTRIBUTION SYSTEM 8.2-2 8.2.3 EMERGENCY POWER SYSTDi 8.2-9 8.3 DESIGN EVALUATION 8.3-1 8.3.1 EVALUATION'0F THE PHYSICAL IAYOUT 8.3-1 8.3-2 ACCIDENTAL PHASE REVERSAL 8.3-2 8.4.
TESTS AND INSPECTIONS 8.4-1 9.
AUXILIARY AND DIERGENCY SYSTEMS R
9.1 MAKEUP AND PURIFICATION SYSTDI 9.1-1 9.1.1-DESIGN BASES 9.1-1 9.1.2
- SYSTD1 DESCRIPTION AND EVALUATION 9.1-2 9.2 CHEMICAL ADDITION AND SAMPLING SYSTEM 9.2-1
- 9.2.1 4 DESIGN BASES 9.2-1
.9.2.2
. SYSTEM DESCRIPTION AND' EVALUATION 9.2-1 viii Amendment 3
i
-Section' Page k,./
- 9'. 3 COOLING WATER SYSTEMS-9.3-1 9.3.1 DESIGN BASES 9.3-1 9.'3.2.-
SYSTD1 DESCRIPTION AND EVALUATION 9.3-1 9.4 SPENT = FUEL COOLING SYSTDi 9.4-1 9.4.1 DESIGN BASES.
9.4 9.4.2 SYSTEM DESCRIPIION AND EVALUATION 9.4-1 9.5 DECAY HEAT REMOVAL SYSTDI 9.5-1 9.5.1 DESIGN BASES 9.5-1 9.5.2
' SYSTDI DESCRIPIION AND EVALUATION 9.5-1 9.6
. FUEL HANDLING SYSTEM 9.6-1
--T
- 9. 6.1 -. DESIGN BASES 9.6-1 9.6.2 SYSTEM DESCRIPIION AND EVALUATION.
9.6-2 9.7
. STATION VENTILATIO" SYSTEMS 9.7-1 9.7.1 DESIGN BASES 9.7-1
~
9.7.2 SYSTEM DESCRIPTION AND EVALUATION 9.7-1 VOLUME III k(
)
U 10.
STEAM AND POWER CONVERSION SYSTEM 10.1 DESIGN BASES 10.1-1 10.1.1 OPERATING.AND PERFORMANCE REQUIREMENTS 10.1-1 10.1.2 ELECTRICAL SYSTDI CHARACTER 1STICS 10.1-1 10.1.3 FUNCTIONAL LIMITATIONS 10.1-1 10, 1.4 SECONDARY FUNCTIONS 10.1-1 10.2 SYSTDi DESIGN AND CPERATION 10.2-1 10.2.1 SCHEMATIC FLOW DIAGRAM 10.2-1
'10.2.2 CODES.AND STANDARDS 10.2-1 10.2.3 DESIGN FEATURES ~
10.2-2
.10.2.4 SHIELDING 10.2-2 10.2.5 CORROSION PROTECTION-10.2-2
.10 2.6 IMPURITIES CONTROL 10.2-2
.10.2.7 RADIOACTIVITY 10.2-2
,g 10.3 SYSTDI ANALYSIS 10.3-1 10.3.1. TRIPS, AUTOMATIC CONTROL - ACTIONS, AND ALARMS 10.3-1 10.3.2 TRANSIENT CONDITIONS 10.3-2
-10.3.3-MALFUNCTIONS 10.3-2
-10i3.4 OVERPRE SSURE ' PROTECTION 10.3-2 10'.3.5 INTERACTIONS 10.3-2 10.3.6 O
' OPERATIONAL LIMITS 10.3-2 O _)
- 10.4 - TESTS AND INSPECTIONS 10.4-1 Amendment 03.
ix-
--e
Section-Page 11.
RADIOACTIVE WASTES AND RADIATION PROTECTION 11.1 RADI0 ACTIVE WASTE HANDLING 11.1-1 11.1.1 DESIGN BASES 11.1-1 11.1.2 SYSTEM DESIGN AND EVALUATION 11.1-6 11.1.3 TESTS AND INSPECTIONS 11.1-8 11.1.4 TRITIUM MANAGEMENT FOR NORMAL OPERATION 11.1-8 11.2 RADIATION SHIELDING 11.2-1 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY 11.2-1
~
SHIELDING 11.2.2 AREA RADIATION MONITORING SYSTEM 11.2-6 11.2.3
' HEALTH PHYSICS 11.2-8
11.3 REFERENCES
11.3-1 12.
CONDUCT OF OPERATIONS
12.1 INTRODUCTION
12.1-1 12.2 ORGANIZATION AND RESPONSIBILITY 12.2-1 12.3 PERSONNEL TRAINING 12.3-1 12.3.1 TRAINING INITIAL PLANT STAFF 12.3-1
.')
12.3.2 REPLACEMENT AND REFRESHER TRAINING 12.3-4 12.3.3 EMERGENCY DRILLS 12.3-5 12.4 WRITTEN PROCEDURE 12.4-1 12.5 RECORD 12.5-1 12.6 ADMINISTRATIVE CONTROLS 12.6-1 12.7 INDEPENDENT AUDIT OF PLANT OPERATIONS 12.7-1 13.
INITIAL TESTS AND OPERATION 13.1-1 13.1 TESTS - PRIOR TO REACTOR FUELING 13.2-1 13.2 INITIAL CRITICALITY 13.3-1 13.3 POSTCRITICALITY TESTS 93
~
~
Amendment 4 x
)-
'Section.
Page
- 14..
SAFETY ANALYSIS 14.1 ~ ' CORE AND C001 ANT ' BOUNDARY PROTECTION ANALYSIS 14.1-1 1
14.1.1 ABNORMALITIES.
14.1-1 14.1.2 ANALYSIS OF EFFECTS AND CONSEQUENCES 14.1-2 14.2 STANDBY SAFEGUARDS ANALYSIS 14.2 14.2.1-; SITUATIONS Ar.ALYZED_AND CAUSES 14.2-1
~
14.2.2 ACCIDENT ANALYSES 14.2-1 14.3 ENVIRONMENTAL CONSEQUENCES OF HYPOTHETICAL ACCIDENTS 14.3-1 14.3.1 GENERAL APPROACH 14.3-1 14.3.2 STEAM GENERATOR TUBE FAILURE 14.3-1 14.3.3 LOSS OF ELECTRIC POWER.
14.3-1 14.3.4 STEAM LINE~ FAILURE 14.3-3
=14.3.5 FUEL HANDLING ACCIDENT 14.3-4 14.3.6 ROD EJECTION ACCIDENT-14.3-4 14.3.7 WASTE GAS TANK RUFIURE 14.3-4 14.3.8 LOSS-OF-COOIANT ACCIDENT 14.3-5 14.3.9-MAXIMUM HYPOTHETICAL ACCIDENT 14.3-6 14.3.10 IODINE REMOVAL SENSITIVITY' ANALYSIS 14.3-10 14.3.11 POPUIATION DENSITY CONSIDERATIONS 14.3-12
14.4 REFERENCES
14.4-1 15.
TECHNICAL SPECIFICATIONS t
1 1
)
.y
'I Amendment:3'-
xi
l g7 v0LUME IV
)
APPENDIX 1 1A ANSWERS TO QUESTIONS 1B QUALITY ASSURANCE OPERATIONS
~
1C RANCHO SECO PROJECT ENGINEERING STAFF APPENDIX 2 2A METEOROLOGY 1
2B LAND USAGE AND POPULATION 2C GEOLOGY AND SEISMOLOGY
~-
2D SEISMIC REPORT 2E SOIL AND FOUNDATIONS INVESTIGATION REPORT 2F METEOROLOGICAL STATION 2G STORAGE RESERVOIR 2H
-ANSWERS TO QUESTIONS APPENDIX 3 3A ANSWERS TO QUESTIONS APPENDIX 4 4A ANSWERS TO QUESTIONS h
xii Amendment 3
i i
(
APPENDIX 5' 5A'. STRUCTURAL DESIGN BASES-i SB JUSTIFICATION OF STRUCTURAL PROOF TEST.- PRESSURES SC SPECIFICATION FOR. SPLICING REINFORCING'BAR USING THE CADWELD PROCESS 5D TURBINE GENERATOR MISSILES
~
SE~ = JUSTIFICATION FOR' LOAD FACTORS
~ -
5F' JUSTIFICATION FOR YIELD REDUCTION FACTORS 5G
- DESCRIPTION OF THE FINITE ELEMENT TECHNIQUE USED IN CONTAINMENT STRUCTURAL ANALYSIS 5H QUALITY CONTROL PROCEDURE FOR FIELD WELDING SI -CONTAINMENT STRUCTURE INSTRUMENTATION 5J-ANSWERS TO QUESTIONS APPENDIX 6 6A~
ANSWERS TO QUESTIONS APPENDIX 7 4
e 7A' ANSWERS TO QUESTIONS
]
.y APPENDIX 8 28A -
ANSWERS TO QUESTIONS i
O i %.. }.:
1 Amendment 3.
- xiii,
- a..
i APPENDIX 9 9A ANSWERS TO QUESTIONS g,
APPENDIX 11 llA ANSWERS TO QUESTIONS APPENDIX 12 12A ANSWERS TO QUESTIONS APPENDIX 13 13A ANSWERS TO QUESTIONS APPENDIX 14 14A ANSWERS TO QUESTIONS 4
APPENDIX 15 15A.
ANSTiERS TO QUESTIONS I
&}
xiv Amendment 3
TABLE OF CONTENTS
' '_ Q-.
i 4.
REACTOR COOLANT SYSTEM Section-Page 4.1 DESIGN' BASES-4.1-1 4.1.1 PERFORMANCE OBJECTIVES 4.1-1 4.1.2 DESIGN CHARACTERISTICS 4.1-2
~4.1.2.1 Design Pressure 4.1-2 4.1.2.2
' Design Temperature 4.1-2 4.1.2.3 Reaction Loads 4.1-6 4.1.2.4 Seismic Loads 4.1-6 4.1.2.5 Cyclic Loads-4.1-6 4.1.2.6 Water Chemistry 4.1-7 4.1.3 EXPECTED OPERATING CONDITIONS 4.1-7 4.1.4
-SERVICE ~ LIFE 4.1-8 4.1.4.1 Nbterial Radiation Damage 4.1-8 4.1.4.2-Nuclear Unit Operational Thermal Cycles 4.1-11 4.1.4.3-Operating Procedures 4.1-13 4.1.4.4 Quality Manufacture 4.1-13 4.1.5 CODES AND CLASSIFICATIONS 4.1-15 4.2 SYSTEM DESCRIPTION AND OPERATION 4.2-1 4.2.1 GENERAL DESCRIPTION 4.2-1 4.2.2
~ MAJOR COMPONENTS 4.2-1
([-, -
4.2.2.1 Reactor Vessel 4.2-1 4.2.2.2 Pressurizer 4.2-2 4.2.2.3
' Steam Generator 4.2-3 1
4.2.2.4
' Reactor Coolant Pumps 4.2-6 4.2.2.5 Reactor Coolant Piping 4.2-7 4.2.3 PRESSURE-RELIEVING DEVICES 4.2-7 4.2.4 ENVIRONMENTAL PROIECTION 4.2-7 4.2.5' MATERIALS OF CONSTRUCTION 4.2-7 4.2.6 MAXIMUM HEATING ~AND COOLING RATES 4.2-11 4.2.7 LEAK DETECTION 4.2-11 4.3 SYSTEM DESIGN EVALUATION 4.3-1 4.3.1 SAFETY FACTORS 4.3-1 4.3.1.1 Pressure Vessel Safety 4.3-1 4.3.1.1.1 Design and Stress Analysis 4.3-1 4.3.1.1.2 Quality Control 4.3-2 4.3-5 4.3.1.1.3 Operation
,gj 4.3-6
- i 4.3.1.1.4.
Additional Pressure Vessel Safety Factors 4.3.1.2 Piping 4.3-7 4.3.1.3 Steam Generator 4.3-7 4.3.2' ' RELIANCE ON INTERCONNECTED SYSTEMS 4.3-8 4.3.3 SYSTEM INTEGRITY 4.3-9 4.3.4 PRESSURE RELIEF 4.3-9 4.3.5
-REDUNDANCY 4.3-10 4.3.6 SAFETY ANALYSIS-4.3-10
.f'}1 i
4.3.7 OPERATIONAL LIMITS-4.3-10 xi -
Amendment 3
'4-i
]
Section Page 4.4 TESTS AND INSPECTIONS 4.4-1 4.4.1 COMPONENT IN-SERVICE INSPECTION 4.4-1 l
4.4.1.1 Reactor Vessel 4.4-1 1
4.4.1.2 Pressurizer 4.4-1 4.4.1.3 Steam Generator 4.4-2
- 4. 4.1 ~. 4 Reactor Coolant Pumps 4.4-2 4.4.1.5 Piping 4.4-2 l
4.4.1.6 Dissimilar thtal and Representative Welds 4.4-2 '
4.4.1.7 Inspection Schedule 4.4-3
' 4.4.2 REACTOR COOLANT SYSTEM TESTS AND INSPECTIONS 4.4-3 4.4.2.1 Reactor Coolant System Precritical and Hot Leak Test 4.4-3 4.4.2.2 Pressurizing System Precritical Operational Test 4.4-3 4.4.2.3 Pressurizer Surge Piping Temperature Gradient Test 4.4-3 4.4.2.4 Relief System Test 4.4-3 4.4.2.5 Plant Power Startup Test 4.4-3 4.4.2.6 Plant Power Heat Balance 4.4-3 4.4.2.7 Plant Power Shutdown Test 4.4-4 4.4.3
. MATERIAL IRRADIATION SURVEILLANCE 4.4-4
4.5 REFERENCES
4.5-1 t
'4-11 Amendment 3
g I
' LIST OF TABLES y;
-Table i psg Title Page
'4.1-1
~ Tabulation of Reactor Coolant System Pressure Settings 4.1-2 p
- '~
,4.1-2 Reactor Vessel. Design Data
.4.1-3 e
4.1-3 Pressurizer Design Data 4.1-4
.4.1-4 Steam Generator Design Data 4.1-4
.m 4.1-5 Reactor Coolant Pump Design Data-4.1-5 14.1-6'
~
Reactor Coolant Piping Design Data 4.1-6
-4.1 Transient Cycles 4.1 7 4.1-8 Reference-for Figure 4.1-4 -' Increase in Transition 4.1-9 Terperature Due to Irradiation Effects for A302B' e
. Steel 4.1-10
'4.1-9 Reactor Coolant System Codes and Classifications 4.1-11 4.1-10 Design Transient Cycles 4.1-12 4.2-1~
Materials of Construction 4.2-8 l
~
' u',,
Amendment-31 4-111
LIST OF FIGURES Figure Number Title 4.1-1 Reactor Coolant System 4.1-2
' Reactor Coolant System Arrangement - Elevation t5a 4.1-3 Reactor Coolant System Arrangement - Plan 4.1-4 Nil-Ductility Transition Temperature Increase Versus Integrated Neutron Exposure for A302B Steel 4.2-1 Reactor Vesse' 4.2-2 Pressurizer 4.2-3 Steam Generator 4.2 Steam Generator. Heating Regions 4.2-5 Steam Generator Heating Surface and Downconer Level Versus Power 4.2-6 Steam Generator Temperatures
{)
4.2-7 Reactor Coolant Pump 4.2-8 Predicted NDTT Shif t Versus Reactor Vessel Irradiation 2
4-iv Amendment 3
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4~.1
- DESIGN BASES
<g The' reactor. coolant system consists of the reactor vessel, coolant
-a
- pumps,= steam generators, p;essurizer,'and interconnecting piping.
4
.The-functional relationship between coolant system components is shown in Figure 4.1-1. -The coolant system physical arrangement is shown in Figures 4.1-2 and 4.1-3..
..The. reactor coolant-system is designed in accordance~with the following codes:
- Piping and Valves - USASIB31.1-1955 (pressure piping) including nuclear-cases.
Pump Casing - ASHE Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
' Steam Generators - ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
, Pressurizer - ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
.2-t
)
~ikd-
- Reactor Vessel - ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
Welding Qualifications - ASNE Boiler and ' Pressure Vessel Code,Section IX.
To assist in' the review of the system drawings, a standard set of
-symbols and abbreviations has been used and is summarized in Figure 9.0-1.
4.1.1 PERFORMANCE OBJECTIVES The reactor coolant system is designed to contain and circulate reactor coolant at pressures and flows necessary tio transfer the '
)
' heat. generated in the reactor core to the~ secondary fluid in the g
,g'
' steam generators.' In addition to serving as a heat transport medium, the coolant also serves'as a' neutron moderator and reflector, and as a solvent for.the soluble boron utilized in chemical shim reactivity
' Control.
As the1 coolant energy and radioactive material container, the reactor coolantisystem is designed to maintain its integr'ity under all oper-(*'g:
ating conditions'.
While performing this' function, the system serves S
i b $> A-
-4.1-1 9
.a g.
e ty ?
Y4'%"NN9
Design Bases the safeguard objective of preventing the release to the reactor building of any fission products that escape the primary barrier, the core cladding.
4.1.2 DESIGN CHARACTERISTICS 4.1.2.1 Design Pressure 0*
The reactor coolant system design, operating, and control set point pressures are listed in Table 4.1-1.
The design pressure allows for operating transient pressure changes. The selected design margin f
considers core thermal lag, coolant transport times and pressure drops, instrumentation and control response characteristics, and system relief valve characteristics.
The design pressures and data for the respective system components are listed in Tables 4.1-2 through 4.1-6.
4.1.2.2 Design Temnerature The design temperature for each component is selected above the maximum anticipated coolant temperature in that component under all normal and transient load conditions.
The design and operating temperatures of the respective system components are listed in Tables 4.1-2 through 4.1-6.
TABLE 4.1-1 TABULATION OF REACTOR COOLANT SYSTEM PRESSURE SETTINGS Item Pressure, psig Design Pressure 2,500 Operating Pressure 2,185 Code Relief Valves 2,500
~
High Pressure Trip 2,350 High Pressure Alarm 2,300 Low Pressure Alarm 2,150 Low Pressure Trip 2,050 f
)
4.1-2
1 Deo1gn Bases k
x
)
TABLE 4.1-2.
REACTOR' VESSEL DESIGN DATA Item Data Design / Operating Pressure, psig 2,500/2,,185
-I Hydrotest Pressure (cold),psig 3,125 Design / Operating Temperature, F 650/603 1
Overall Height of Vessel and closure
' Head, ft-in.
37-4 Straight Shell Thickness, in.
8-7/16 3
Water Volume, ft.
4,150 Thickness of Insulation, in.
3
!( 2 Number of Reactor Closure Head Studs 60 s
Flange ID, in.
165 Shell ID, in.
171
~
Inlet Nozzle ID, in.
28 4
Outlet ' Nozzle ID, in.
36 Core _ Flooding Water Nozzle ID, in.
11-1/2 L
- {i l
i b
aw
i
)
4.1 J
\\
, ~......,.. _. -.,....
.i
Design Bases
-~
TABLE 4.1-3 PRESSURIZER DESIGN DATA Item Data Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 5
Design / Operating Temperature, F 570/650 Normal Water Volume, ft3 800 Normal Steam Volume, ft3 700 Surge Line Nozzle Diameter, in.
10 Overall Height, ft-in.
44-0 TABLE 4.1-4 STEAM GENERATOR DESIGN DATA Item Data per Unit
.~.
Design Pressure, Reactor Coolant / Steam, psig 2,500/1,050 Hydrotest~ Pressure (tube side-cold, reactor coolant), psig 3,125 Design Temperature, Reactor Coolant / Steam, F 650/603 Reactor Coolant Flow, lb/hr-65.66 x 106 Heat Transferred, Btu /hr 4.21 x 109 Steam Conditions at Rated Load, Outlet Nozzles:
6 Steam Flow, lb/hr 5.30 x 10
~
Steam. Temperature, F 570 (35 F superheat)
Steam Pressure, psig 910 Feedwater Temperature, F 455 Overall Height, ft-in.
73 1/2 Stiell OD, in.
147-1/4 3
Reactor Coolant Water Volume,_ft 2,030 4.1-4
.x Design Bases
- gO[
- i. U }'
TABLE 4.1-5
- REACTOR C001 ANT PUMP DESIGN DATA l Item Data per Unit Number of_ Pumps-
'4 d
~
i Design-Pressure, psig 2,500.
'l
~l
- Hydrotest Prescure (cold), psig 3,125 j
Design: Temperature, F 650 T
Operating Speed ~(nominal), rpm 1,180 Pumped. Fluid Temperature, F~
60 to 580 Developed. Head, ft 370 Capacity, gpm 88,000 Hydraulic Efficiency, %
86
~
-Seal Water Inject' 9, gpm 60 f
Seal-Water Return smax), gpm 58 Pump Nozzle TD, in.-
28
. Overall Unit. Height, ft 24 Water Volume,-ft3 95 Motor Stator Frame Diameter, ft 8
l Purp-Motor Moment of Inertia, Ib-ft2 70,000 Motor Data:
. Type Squirrel-Cage Induction, Single Speed Voltage-6,600
.d Phase 3
Frequency, cps 60 Starting-Across-the-Line
_Input -(hot reactor coolant), kw 5,600 Input -(cold reactor ' coolant), kw 7,400 J
%,.f :
4.1-5
Design B3s:s TABLE 4.1-6
)
REACTOR C001 ANT PIPING DESIGN DATA Item Data Reactor Inlet Piping ID, in.
28 Reactor Outlet Piping ID, in.
36
(*j, Pressurizer Surge Piping, in.
10 Sch. 140 Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 Design / Operating Temperature, F 650/603 Design / Operating Temperature (pressurizer surge line), F 670/650 3
Water Volume, ft 1,910 4.1.2.3.
Reaction Loads All components in the reactor coolant system are supported and in.er-T connected so that piping reaction forces result in combined mecht nical
../
and thermal stresses in equipment nozzles and struccural walls w; thin established code limits. Equipment and pipe scapports are designed to absorb piping rupture reaction loads for elimt.2 tion of secondary a
ident effects such as pipe motion and equipment foundation sb.f ting.
4.1.2.4 seismic Loads Reactor coolant system components are designated as Class I equ.p-ment, and are designed to maintain their functional integrity during earthquake, The basic design guide for the seismic analysis is the AEC publication TID-7024, " Nuclear Reactors and Earthquake".
Structures and equipment will be designed in accordance with Appendix SA.
4.1.2.5 Cyclic Loads
(
All components in the reactor coolant system are designed to with-
{
stand the effects of cyclic loads due to reactor system temperature and pressure changes.
These cyclic loads are introduced by normal unit load transients, reactor trip, and startup and shutdown operation.
Design cycles are shown in Table 4.1-7.
During unit startup and shutdown, the rates af temperature and pressure changes
- s are limited.
)
4.1-6 L
I Design Bases
- 7) '
TRANSIENT CYCLES *
(
TABLE 4.1-7
\\'
(Rated Steam Load Basis)
Estimated Transient Description Design Cycles Actual Cycles' 1.
Heatup, 70 to 579 F, ar.d Cooldown, 579 to 70 F 480 80 2.
Heatup,.540 to 579 F, and Cooldown, 579 to 540 F 1,440 770 3.
Ramp Loading and Ramp Unloading (15-100-15%)
12,000 9,000
'4.
Step Loading. Increase. (10%)
2,000 1,500 5.
Step Unloading-Decrease (10%)
2,000 1,500 6.
Step-Load Reduction to Auxiliary Load l(100-5%)
160 120
'7.
Reactor Trip from Rated Power 400 300 O
f; ( j 8.
Miscellaneous Transients 10 5
- The cycles above are based on 40 year design life 4.1.2.6 Water' Chemistry The water chemistry is selected to provide the necessary boron content for reactivity control and to minimize corrosion of reactor
-coolant system surfaces.- The reactor coolant chemistry is discussed
.in further detail in 9.2..
4.1.3 EXPECTED OPERATING CONDIT10NS Throughout the load range from 15 to 100 percent power, the reactor
,gg coolant system is operated at a constant; average temperature.
Reactor coolant system pressure is controlled to provide sufficient overpressure to. maintain adequate core subcooling.
~
The minimum operating pressure is established from core thermal analysis. This analysis is based upon the maximum expected inlet and. outlet temperatures, the maximum reactor power, the minimum DNBR
. required (including instrumentation errors and the reactor control system
(
-deadband), and'a core-flow distrib'ition factor.
t 4 1-7'
Design Bases The maximum operating pressure is established on the basis of ASME
]
Code relief valve characteristics and the margins required for normal pressure variations in the system.
Pressure control between the preset maximum and minimum limits is obtained directly by pressurizer. spray action to suppress high pressure and pressurizer heater action to compensate for low pressure.
Normal operational lifetime transient cycles are discussed in detail in 4.1.4.
4.1.4 SERVICE LIFE
';p The service life of reactor coolant system pressure components depends upon the end-of-life material radiation damage, nuclear unit operational thermal cycles, quality manufacturing standards, environmental protection, and adherence to established operating procedures.
In the following discussion each of these life-dependent factors will be discussed with regard to the affected components.
4.1.4.1 Material Radiation Damage The reactor vessel is the only reactor coolant system component exposed to a significant level of neutron irradiction and is therefore the only component subject to material radiation damage.
~
To assess the potential radiation damage at the end-of-reactor service life, the maximum exposure from fast neutrons (E > 1.0 Mev) 80 percent load factor.
Reactor vessel irradiation exposure calcu-
~)
has been computed to be 3.0 x 1019 n/cm2 over a 40 year life with an 1ations are described in 3.2.2.1.7.
For this neutron exposure, the predicted nil-ductility transition temperature (NDTT) shift is 250 F based on the curve shown in Figure 4.1-4.
Based on an initial NDTT of 10 F, this shift would result in a predicted NDTT of 260 F.
~
The " Trend curve for 350 F Data", as shown in Figure 4.1-4, repre-sents irradiated material test results and was compiled from the reference documents listed in Table 4.1-8.
L '
To evaluate the NDTT shif t of welds, heat-affected zones, and base material for the material used in the vessel, test coupons of these three material types have been included in the reactor vessel sur-veillance program as described in 4.4.3.
t 4.1-8
Design Bases f5 (h
TABLE 4.1-8
. REFERENCES FOR FIGURE 4.1-4 INCREASE IN TRANSITION TEMPERATURE DUE TO IRRADIATION EFFECTS FOR A302B STEEL Neutron Ref
- Temp, Exposure, NDTT No Reference Material Type
.F n/cm2 ( > 1 Mev)
F 1
ASME Paper All Steels 550 Maximum Curve **
Data No. 63-WA-100.
2 ASTM-STP 380,.
A302B Plate 550 Trend Curve **
Data
-)
p.295
-J
'3 NRL Report 6160 A302B Plate 550 5 x 1018 65 p 12 4
ASTM-STP 341, A302B Plate 550 8 x 1018 85*
.{
p 226 5
ASTM-STP 341, A302B Plate 550 8 x 1018 100 p 226 6
ASTM-STP 341, A302B Plate 550 1.5 x 1019 p 226 130*
7 ASTM-STP 341, A302B Plate 550 1.5 x 1019 p 226' 140
'8 Quarterly Report A302B Plate 550 3 x 1019
~,
120 of Progress,
" Irradiation Ef-fects on Reactor Structural' Mate-rials", 11-1-64/
1-31-65 l
9~ Quarterly Report A302B Plate 550 3 x 1019 of Progress, 135
" Irradiation Ef-fects on Reactor Structural Mate-rials", 11-1-64/-
-O 1-31-65
^ * ! ansverse specimens.
- See. Figure 4.1-4.
u) 4.1-9
- 1
y Design Bases TABLE 4.1-8 Continued 1
Neutron i
Ref
- Temp, Exposure NDTT No Reference Material Type F
n/cm2(> 1 Mev)
F 10 Quarterly Report A302B Plate 550 3 x 1019 140 of Progress,
" Irradiation Ef-fects on Reactor Structural Mate-rials", 11-1-64/
1-31-65 11 Quarterly Report
'A302B Plate 550 3 x 1019 170 of Progress,
" Irradiation Ef-fects on Reactor Structural Mate-rials", 11-1-64/
1-31-65 12 Quarterly Report A302B Plate 550 3 x 1019 205 of Progress,
.)
" Irradiation Ef-fects on Rea. tor Structural Mate-rials", 31-1-64/
1-31-65 18 13 Welding Research A302B Weld 500 5 x 10 70 Suppleinent, Vol.
to 27, No. 12, Oct.
575 1962, p 465-S 14 Welding Research A302B Weld 500 5 x 1018 50 Supple =cnt, Vol.
to 27, No. 10, Oct.
575
~
1962, p 465-S 18 15 Welding Research A302B Weld 500 5 x 10 37 Supplement, Vol.
to 27, No. 1, Oct.
575 t.
1962, p 465-S s
16 Welding Research A302B Weld 500 5 x 1018 25 Supplement, Vol.
to 27, No. 10, Oct.
575 1962, p 465-S
-4.1-10
_y 1
Design Bases (f
.4.1.4.2~
Nuclear Unit Operational Thermal Cycles To cstablish the service life of the reactor coolant system components
~
as required by the.ASME III for Class "A": vessels, the nuclear unit-
~
operating.co.ditions that involve the cyclic application of. loads and thermal conditions have been-established.for the 40-year design life.
i The number:of thermal'and-loading cycles to be uses for design I
purposes are listed in Table 4.1-7,' " Transient Cycles." The estimated-actual cycles based on.r_eview of existing nuclear stations operations
- are also provided -in Table 4.1-7.
Table 4.1-9. lists those components designed to ASME IIIL-Class "A".
The effect of individual transiente,
- and the sum of.these transients are evaluated to determine the fatigue usage factor.during the detail design and stress analysis effort.
- As sp'ecified in ASME III Paragraph.415.2 (d) (6), the cumulative fatigue. usage factor will be'less than 1.0 for the design cycles
- listed-in Table 4.1-7.
~
TABLE 4.1-9 REACTOR COOLANT SYSTEM CODES AND CLASSIFICATIONS Component Code Classification
. Reactor Vessel ASME.a III Class A
[(j~N
\\\\
Steam Generator ASME a III Class A Pressurizer
'ASME a III Class A Reactor Coolant Pump Casing ASME a III Class A d
Motor IEEE,b NEMA, e and USASI Piping and Valves USASI e B31.1-1955 and Asso-ciated Nuclear Code Cases r
'Ameridan Society of Mechanical Engineers, Boiler and Pressure 8
- Vessel Code.
Section III covers ' Nuclear Vessels.
b Institute of Electrical and Electronics Engineers.
.g
~
National Electrical Manufacturers Association
.dL United States of America Standards Institu*.e Nos. C50-2-1955 and C50.20-1954.
United" States.of.-America Standards Institute No. B31.1.
f~N V;
y 4.1-11' a
Design BSg s 7
The transient cycles listed in Table 4.1-7 are conservative and
/
complete in that they include all significant modes of normal and emergency operation.
The estimated frequency bases for the design transient' cycles are listed in Table 4.1-10.
A large number of cycles of smaller magnitudes than those described can be tolerated.
A heatup and cooldown rate of 100 F/ hour is used in the analysis of Transients 1 and 2 in Table 4.1-7.
jy A ramp loading and ramp unloading transient is defined as a change in power level from 15 to 100 to 15 percent of rated power at a rate of change of 10%/ min. A step loading transient is an instantaneous power increase or decrease of 10 percent of rated power. A step load reduction to auxiliary load is an instantaneous reduction in elec-trical load from 100 to 5 percent of rated load.
The miscellaneous transients (Item 8) listed in Table 4.1-7 include the initial hydrotests, plus an allowance for future hydrotests in the event that reactor coolant system modifications or repairs may be re-quired.
Subsequent to a normal refueling operation only the reactor vessel closure seals are hydrocested for pressure integrity; there-fore, reactor coolant system hydrotests before startup are not included.
TABLE 4.1-10 DESIGN TRANSIENT CYCLES
-}'
(Rated Steam Load Basis)
Transient No. (See Table 4-7)
Frequency 1.
Heatup, 70 to 579 F, and Cooldown, 579 to 70 F 12 per Year 2.
Heatup, 540 to 579 F, and Cooldown, 579 to 540 F 36 per Year 3.
Ramp Loading and Ramp Unioading (15-100-15%)
6 per Week 4.
Step Loading Increase (10%)
1 per Week 5.
Step Unioading Decrease (10%)
1 per Week 6.
Step Loading Reduction to Auxiliary Load (100-5%)
4 per Year 7.
Reactor Trip From Rated Power 10 per Year
~
)
4.1-12
Design Bases q
4.1.4.3 Operating Procedures The reactor coolant' system pressure vessel components are designed using. the ' transition. temperature method of minimizing the possibility
.of brittle fracture of the vessel materials. The various combinations sof stresses are evaluated and employed to determine the system opera-
)
ting procedures.
The basic determination of vessel operation from cold startup and shut-down to full pressure and temperature operation is performed in accord-ance with a " Fracture Analysis Diagram" as published by Pellini and Puzak.2 At temperatures below the nil-ductility transition temperature (NDTT) and the design transition temperature (DTT), which is equal to NDTT + 60 F, the pressure vessels will be operated so that the stress levels will be restricted to a value that will peevent brittle failure.
These levels are a.
Below the temperature of DTT minus 200 F, a maximum stress of 10 percent yield strength.
b.
From the: temperature of DTT minus 200 F to DTT, a maximum stress which will increase from 10 to 20 percent yield strength.
Ok,)
c.
At the temperature of DTT, a maximum stress of 20 percent yield strength.
If the nominal stresses are held within the referenced stress limits (a. through c. above), brittle fracture will not occur.
This state-ment 'is based on data reported by Robertson3 and Kihara and Masubichi4 in published. literature.
It can be shown that stress limits can be iantro11ed by imposing operating procedures that control pressure and temperature during heatup and cocidown (See Reference 3).
This pro-cedure will ensure that the' nominal stress levels do not exceed those specified in a. t!1 rough c. above.
4.1.4.4.
~ Quality Manufacture d
Material selection is discussed in detail in 4.2.5.
'Th Af ter receipt of the material, a program of qualification of all
.)
-welding and heat treating processes that could affect mechanical or metallurgical properties of the material during fabr: cation is under-
.taken. :This' program will establish-the properties.of the material, as. received, and certify that the mechanical properties of the mate-rials in the finished vessels are consistent with.those used in the design analysis. This program consists of:
p)l
- a.. Weld qualification test plates using production proce-
\\'
dures and' subjecting. test plates to the heat treatments to be used.in fabricating the. vessels.
4.1-13
'1
Design Bases b.
Subjecting qualification test plates to all non-O!
destructive tests to be employed in production, such as x-ray, dye-penetrant, magnetic particle, and ultra-sonic. Acceptance standards are the same as used for production.
Subjecting qualification test plates to destructive c.
tests to establish 5'
(1) Tensile strength (2) Ductility (3) Resistance to brittle fracture of the weld metal, base metal, and heat-affected zone (HAZ) metal.
Af ter completion of the qualification test program, production welding and inspection procedures are prepared.
All plate or other materials are permanently identified, and the identity is maintained throughout manufacture so that each piece can be located in the finished vessels.
In-process and final dimen-sional inspections are made to ensure that parts and assemblies meet the drawing requirements, and an "as-built" record of these dimensions is kept for future reference.
}
All welders are qualified or requalified as necessary in accordance with The Babcock & Wilcox Company and ASNE IX requirements. Each lot of welding electrodes and fluxes is tested and qualified before release to ensure that required mechanical properties and as-deposited chemical properties can be met.
Electrodes are identified and issued only on an approved request to ensure that the correct materials are used in each weld. All welding electrodes and fluxes are maintained dry and free from contamination before use.
Records are maintained and reviewed'by welding engineers to ensure that approved procedures and materials are being used. Records are maintained for each weld joint and include the welder's name, essential weld parameters, and electrode heat or lot number.
The several types of nondestructive tests performed during vessel fabrication are as follows:
a.
Radiography, including x-ray, high voltage linear accel-erator, or radioactive sources, will be used as applicable to determine the acceptability of pressure integrity welds and other welds as specifications require.
b.
Ultrasonics is used to examine all pressure-integrity raw material and the bond between corrosion-resistant cladding 3
to base material.
In addition, pressure-containing welds t
where applicable are inspected by ultrasonics.
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1.
All data is for 30 f t-lb "Fix".
2.
Numbers on Curves Indicate References in Table 4-11.
FIGURE 4.1-4 NIL-DUCTILITY TRANSITION TEMPERATURE INCREASE VERSUS INTEGRATED NEUTRON EXPOSURE FOR A302B STEEL
$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT
m 1( N; 4.2 SYSTEM DESCRIPTION AND OPERATION 4.2.1 GENERAL ^ DESCRIPTION.
The' reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed coolant
-circulating pumps, an electrically-heated pressurizer, and inter-connecting piping. The s'ystem-is arranged as two heat transport loops, each.with two circulating pumps and one steam generator.
Reactor system design data are. listed in Tables 4.1-2 through
~
4.1-6, and a system schematic diagram is shown in Figure 4.1-1.
Elevation and plan views of the arrangement of the major components are shown in Figures 4.1-2 and 4.1-3.
~
4.2.2 MAJOR COMPONENTS 4.2.2.1 Reactor Vessel The reactor. vessel consists of a cylindrical shell, a cylindrical
. support skirt, a spherically dished bottom head, and a ring flange to which a removable reactor closure head is bolted.
The reactor closure head is a spherically-dished head welded to a ring flange.
The vessel has six major nozzles for reactor coolant flow, 69 control rod drive nozzles mounted on the reactor closure head, and two core flooding nozzles--all located above the core. The reactor vessel will be vented through the control rod drives. The vessel closure seal is formed by two concentric 0-rings with pro-visions between them for leakage collection.
The reactor vessel, l2 nozzle design, and seals incorporate the extensive design and fab-rication experience accumulated by B&W. Fifty-two in-core instru-l2 mentation nozzles are located on the lower head.
The reactor closure head and the reactor vessel flange are joined by sixty 1/2 in. diameter studs. Two metallic 0-rings seal the reactor vessel when the reactor closure head is bolted in place.
Leakoff and test taps are provided in the annulus between the two e
M g::: c.,f leakage and to hydrote::t the vessel closure 2
0 *-ins M seal'after refueling
..y The vessel -is insulated with tretallic reflective-type insulation.
Insulation' panels are provided for che reactor closure head.
The' reactor vessel internals are designed to direct the coolant
. flow, support the, reactor core, and guide the control rods in the withdrawn position.
The reactor vessel contains the core support assembly,. upper plenum assembly, fuel assemblies, control rod assemblies, surveillance specimens, and incore instrumentation.
, f' p t
j.
Ms-
- Amendment 2.
4.2-1.
y System Description and Operation i
The reactor vessel shell material is protected against fast neutron flux and gamma heating effects by a series of water annuli and the thermal shield located between the core and vessel wall.
This protection is further described in 3.2.4.1.2, 4.1.4, and 4.3.1.
Stop blocks welded to the reactor vessel inside wall limit reactor internals and core vertical drop to 1/2 in, or less and prevent rotation about the vertical axis in the unlikely event of a major b
internals component failure.
Surveillance specimens made from reactor steel are located between the reactor vessel wall and the thermal shield. These specimens will be examined at selected intervals to evaluate reactor vessel material NDTT changes as described in 4.4.3.
-The reactor vessel general arrangement is shown in Figure 4.2-1, and the general arrangement of the reactor vessel and internals is shown in Figures 3.2-59 and 3.2-60.
Reactor vessel design data are listed in Table 4.1-2.
4.2.2.2 Pressurizer The general arrangement of the reactor coolant system pressurizer is shown in Figure 4.2-2, and the design characteristics are tab-
}
ulated in Table 4.1-3.
The electrically-heated pressurizer estab-
./
lishes and maintains the reactor coolant pressure within prescribed limits and provides a surge chamber and a water reserve to accommo-date reactor coolant volume changes during operation.
The pressurizer is a. vertical cylindrical vessel connected to the reactor outlet piping by the surge piping. The pressurizer vessel is protected from thermal effects by a thermal sleeve on the surge line and by a distribution baffle located above the surge pipe entrance to the vessel.
Relief valves are mounted on the top of the pressurizer and function u
- to relieve any system overpressure. Each valve has one-half the required relieving capacity. The capacity of these valves is dis-cussed in.4.3.4.
The relief valves discharge to a pressurizer relief tank located within the reactor building. The pressurizer relief tank has a stored water supply and cooling coils to condense the steam. A relief valve protects the tank against overpressure should a pressurizer valve fail to reseat.
.The pressurizer contains replaceable electric heaters in its lower section and a water spray nozzle in its upper section to maintain the steam and water at the saturation temperature corresponding to the desired reactor coolant system pressure.
During outsurges, as
(,
the pressure in the reactor-decreases, some of the water in the pressurizer flashes to steam to maintain pressure.
Electric heaters 4.2-2
P u.
s 1Syst'em Description and Operation -
Lare actuated to~ restore the. normal operating pressure. During;in-Tys /g osurges, as: pressure in the; reactor system increases, steam is con-densed by aLwater spray from:the reactor inlet lines, thus reducing
. p re's sure. -Spray' flow and heaters are controlled by the pressurizer
- pressure. controller.
InstrumentationL for th'e pressurizer is ' discussed in 7.3.2.
l 4. 2 '. 2. 3
~Ste'am Generator
~
The general'arrangemeat of the steam generators is shown in Figure
~
'4.2-3, and design data are tabulated in Table 4.1-4.
The1 steam generator is a vertical,~ straight-tube-an'd-shell heat-exchanger and' produces superheated steam at constant pressure over the power range.
Reactor 1 coolant--flows downward through the tubes,-
and steam.is; generated;on'the shell' side.
The high pressure parts of'the unit are'the hemispherical. heads,.the tubesheets, and the straight Inconel*' tubes between.the tubesheets.
Tube supports
. hold'the tubes in a uniform pattern along their length.
The shell, the outside of the tubes, and the tubesheets form the boundaries of the steam-producing section of the vessel. Within the shell, the tube bundle is surrounded by a baffle, which is
['{l divided intoL two sections. -The upper part of the annulus between the shell and' baffle is the super-heater. outlet, and the lower part L
is the feedwater. inlet-heating: zone.
Vents, drains, instrumentation nozzles, and inspection openings are provided:on the shell side of
- the: unit. The reactor coolant side-has instrumentation connections
'on theitop and bottom heads, manways on both heads, and a drain nozzle forlthe bottom head.
Venting of the reactor coolant side-of the; unit is accomplished by a vent connection on the reactor coolant inlet pipe to 'each unit.- The unit is supported by a skirt attached.to the~ bottom head.
Reactor coolant. water (enters the steam generator at the upper plenum, flows'down the-Inconel tubes while transferring' heat to the
-secondary.'shell-side fluid,.and exits' through the lower plenum.
~/
Figure.4.2-4 shows.the flow paths and. steam generator. heating regions.'
..g
~
- I onel isfa~ trade name of an alloy:manufs tured:by the InternationaliNickel Company.. It also has substantial common
- usagenas-a generic-description of a Ni-Fe-Cr alloy conforming.
Lto; ASTM Specification SB-163.
It is in the latter context' k[r
.that. reference'is:made here.
2 ~4.2.
e 1 3
System Description and Operation Four heat transfer regions exist in the steam generator as feed-water is converted to superheated steam.
Starting with the feed-water inlet these are:
a.
Feedwater Heating Feedwater is heated to saturation temperature by direct contact heat exchange. The feedwater enter-ing the unit is sprayed into a feed heating annulus
.3' (downcomer) formed by the shell and the baf fle around the tube bundle. The steam that heats the feedwater to saturation is drawn into the downcomer by condensing action of the relatively cold feedwater.
The saturated water in the downcomer forms a static head to balance the static head in the nucleate boiling section. This provides the head to over-come pressure drop in the circuit formed by the downcomer, the boiling sections, and the bypass steam flow to the feedwater heating region. With low (less than 1 f t/sec) saturated water velocities entering the generacing section, the secondary side pressure drops in the boiling section are negligible.
The majority of the pressure drop is due to the static head of the mixture. Consequently, the downcomer phase boiling mixture in the nucleate boiling region.
-)
1cvel of water balances the mean density of the two-b.
Nucleate Boiling The saturated water enters the tube bundle, and the steam-water mixture flows upward on the outside of the Inconel tubes counter-current to the reactor coolant flow.
The vapor content of the mixture increases almost uniformly until DNB, i.e., departure from nucleate boiling, is reached, and then film boiling and super-heating occurs.
The quality at which transi-tion from nucleate boiling to film boiling occurs is a function of pressure, heat flux, and mass velocity.
c.
. Film Boiling Dry saturated steam is produced in the film boiling region at the upper end of the tube bundle.
6.
Superheated Steam Saturated steam is raised to final temperature in the super-heater region.
)
4.2-4
g
~
Systam Description end Operation a
e j
- Shown' en Figure 4.2-5 is a plot.of heating surface and downcomer
- V) +
-level versus load.. As shown, the downcomer water level is propor-tional to steam flow from 15 to 100 percent load.
A constant minimum level _is held below'15 percent' load. The amount of surface (or length) of the nucleate boiling section and the film boiling section-is proportional'to load. The surface available for super-heating varies inversely with load, i.e., as load decreases the superheat section' gains from the nucleate and film boiling regions.
~ Mass inventory in the steam generator increases with load as the
-length of the heat transfer regions varies.
The simple concept with ideal counterflow conditions results in highly stable flow characteristics on both the reactor coolant
- and secondary. sides.-. The hot reactor coolant fluid is cooled uniformly as it flows downward. The secondary side mass flow is low, and the majority of.the pressure: drop is due to the static effect of the mixture. The boiling-in the steam generator is
- somewha't.similar to " pool boiling", except that there is motion upward,that permits some parallel flow of water and steam.
. A plot of reactor coolant and steam temperatures versus reactor power
- is shown.in Figure 7.2-1.
As shown, both steam pressure and average reactor coolant temperature are held constant over the load range'from fp 15 to 100 percent rated power.
Constant steam pressure is obtained by Q
a variable two-phase boiling length (see Figure 4.2-4).and-by the regulation of feed flow to obtain proper steam generator secondary mass inve ntory'.'
In addition to average reactor coolant temperature, reactor coolant flow is also held constant.
The difference between reactor ecolant inlet and outlet temperaturesiincreases proportionately as load is increased.
Saturation pressure and temperature are constant, resulting in a variable superheater outlet temperature.
Figure;4.2-6, a plot of temperature versus tube length, shows the temperature differences between shell and tube throughout the steam generator at full load. 'The excellent heat transfer coefficients permit.the use of a secondary operating pressure and temperature sufficiently close to the reactor coolant average temperature so that a straight-tube design can be used.
~
The shell temperature is controlled by the use of direct contact g
. steam that heats the feedwater to saturation, and the shell is bathed with saturated water from feedwater inlet.to the lower tubesheet.
In the superh' eater'section, the-tube wall temperature approaches
~
' the' reactor coolant-fluid ~ temperature since the steam film heat transfer. coefficient:is considerably lower than the reactor cool-ant heat; transfer coefficient.
By baffle arrangement in the super-u( }-
heater section, the shell'section is bathed with superheated steam 4.2-5 n
System Description and Operation above the steam outlet nozzle, further reducing temperature dif-ferentials between tubes and shell.
The steam generator design and' stress analysis will be performed in accordance with the requirements of the ASME III as described in 4.3.1.1.
4.2.2.4 Reactor Coolant Pumps The general arrangement of a reactor coolant pump is shown in Figure 4.2-7, and the pump design data are tabulated in Table 4.1-5.
The reactor coolant pumps are vertical, single-speed, shaft-sea}ed units having bottom suction and horizontal discharge.
Eac.h ptonp has a separate, single-speed, top-mounted motor, which is connected to the pump by a shaft coupling.
Shaft sealing is accomplished in the upper part of the pump housing using a throttle bushing, a seal chamber, a mechanical seal, and a drain chamber in series.
Seal water is injected ahead of the throttle bushing at a pressure approximately 50 psi above reactor system pressure.
Part of the seal flow passes into the pump volute through the radial pump bearing. The remainder flows out along the throttle bushing, where its pressure is reduced, to the seal chamber and is returned to the seal water supply system. The out-board mechanical seal normally operates at a pressure and tempera-
)
ture of approximately 50 psig and 125 F.
However, it is designed for full reactor coolant system pressure and, if seal chamber cooling were maintained, would continue to operate satisfactorily without seal water injection for several weeks. The outboard drain chamber would further prevent leakage to the reactor building if deterioration of the mechanical seal performance should occur.
A water-lubricated, self-aligning, radial bearing is located in the pump housing. An oil-lubricated radial bearing and a Kingsbury, type, double-acting, oil-lubricated thrust bearing are located in the pump motor. The thrust bearing is designed so that reverse rotation of the shaft will not lead to pump or motor damage.
Lube oil cooling is accomplished by cooling coils in the motor oil reservoir.
Oil pressure required for bearing lubrication is main-tained by internal pumping provisions in the motor, or by and external system if required for " hydraulic-jacking" of the bearing surfaces for startup.
An antirotation device will be furnished with each pump motor to 2
prohibit reverse rotation of the pump.
Factory thrust, vibration, and seal performance tests will be made in a closed loop on the first pump at rated speed with the pump end at rated temperature and pressure.
Sufficient testing will be done on subsequent units to substantiate that they conform to the initial test pump characteristics.
4.2-6
-=
ym System Description and Operation (7Q
-, j t
'i,_) '
4. 2. 2.'5 ;
Reactor Coolant Piping
' The-general arrangement of the. reactor coolant system piping is shown in Figures 4.1-2 and 4.1-3.
Piping design data are presented in_ Table.4.1-6.
In addition to the pressurizer surge piping con-nection,-the piping is equipped with welded connections for pressure taps, temperature elements, vents, drains, decay heat removal, and emergency core cooling high pressure injection water. Thermal sleeves are provided in the pressurizer surge piping, the emergency 2
.high pressure injection, and the core flooding connections.
4.2.3-
- PRESSURE-RELIEVING DEVICES The. reactor coolant system is protected against overpressure by control and protective circuits such as the high pressure trip and code relief valves located on the top head of the pressurizer.
JThe relief' valves discharge into the pressurizer relief tank which condenses and collects the effluent. The schematic arrangement of the relief devices is shown in Figure 4.1-1.
Since all sources of heat in the system, i.e., core, pressurizer heaters, and reactor coolant pumps, are interconne:ted by the reactor coolant piping with no intervening isolation valves, all relief protection can conveniently be located on the pressurizer.
]
'4.2.4 ENVIRONMENTAL PROTECTION The reactor coolant system is surrounded by con'
-~o. shield walls.
These-walls provide shielding to permit access c ;a the reactor building forfinspection and maintenance of miscellaneous rotating equipment during rated power operation and for periodic calibration of the incore monitoring system.. These shielding walls act as missile protection for the reactor building liner plate.
Lateral bracing will be provided near the steam generator upper tube-shect elevation to. resist lateral loads, including those resulting from seismic forces, pipe rupture, thermal expansion, etc. Additional bracing-is provided at a lower elevation to-restrain the 36-in. ID vertical pipe leg fr9m whipping.
-d
' 4 ~. 2. 5 - MATERIALS OF CONSTRUCTION
~~
Each'of the materials used in the reactor coolant system has been selected for the expected environment and service conditions. The
~
major component materials are listed in Table 4.2-1.
fS; b2-l-
lAmendment 2 4.2-7
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System Description and Operation TABLE 4.2-1 MATERIALS OF CONSTRUCTION Component Section Materials Reactor Vessel Pressure Plate Sa-533, Grade B, Class 1*
2 Pressure Forgings A-508-64, Classes 1 & 2 (Code Case 1332-3)
Cladding 18-8 Stainless Steel Thermal Shield and Internals SA-240, Type 304, and Inconel-X Steam Generator Pressure Plate SA-516, Grade 70 SA-533, Grade B, Class 1 Pressure Forgings A-508-64, Class 1 (Code Case 1332-3)
Cladding for Heads 18-8 Stainless Steel 2
Cladding for Tubesheets Ni-Cr-Fe Tubes SB-163 9/
1 Pressurizer Shell, Heads, and External Plate SA-516, Grade 70 Forgings A-508-64, Class 1 (Code Case 1332-3)
Cladding 18-8 Stainless Steel Internal Plate SA-240, Type 304 Internal Piping SA-312, Type 304 Reactor Coolant 28 in, and 36 in.
SA-516, Grade 70 (Elbows)
Piping A-106, Grade 6 (Straights)
Ciadding 18-8 Stainless Steel 2
10 in.
A-403, Grade WP 316 (Elbows)
A-376, Type 316 (Straights)
- This mat'erial is metallurgically identical to SA-302, Grade B, as modified by Code Case 1339.
O>
4.2-8 Amendment 2
~
J
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System Desctiption ano Operation
,; Q "All_ reactor' coolant system materials exposed to the coolant'are
. corrosion-resistant. materials consisting of 304 or 316 SS, weld ideposic 304 :SS cladding,- Inconel -(Ni-Cr-Fe), and 17-4 PH (H1100).
.These' materials were. chosen forLspecific purposes at various
- locations within'the system because of their superior compatibility with the reactor coolant.
.j Periodic. analyses 'of the coolant chemical composition will be
-performed _to monitor the adherence of the system to the reactor coolant' water quality listed-in Table 9.2-2.
Maintenance of the
-water quality to minimize corrosion is performed by the chemical
~'
-addition _and sampling 1 system which-is described in detail in 9.2.
The[feedwaterqualityenteringthesteamgeneratorw'llbeheld
^
~ i
.within the-limits listed in Table 9.2-1 to prevent deposits and' corrosion.inside the steam generators. This required feedwater
~
quality has been sucessfully.used in comparable.once-through,.
non-nuclear--steam generators. The phenomena of stress-corrosion cracking and corrosion fatigue are not generally encountered unless a combination'of elements in varying degrees is present. The Jnecessary. conditions are a~ susceptible alloy, an aggressive environ-ment, a' stress, and time.
All external insulation'of' reactor coolant system components will be compatible with the component materials. The reactor vessel is 1-insulated with metallic' reflective insulation on the cylindrical shell exterior. The closure flanges and the top and bottom heads
.in the area of corrosion-resistant penetrations will be insulated
.with; low-halide-content' insulating material. All other external corrosion-resistant surfaces in the reactor coolant system will-be insulated with low or halide-free insulating material as required.
The reactor vessel plate' material opposite the. core is purchased to a.specified Charpy V-notch test result of 30 ft-lb or greater at a corresponding nil-ductility transition' temperature - (NDTT) of 10 F or less,'and the material will.be tested to verify conformity to specifiedl requirements and to determine the actual NDTT v11ue.
In addition,1thiszplatelwill be 100' percent-volumetrically inspected by ultrasonic l test using both' normal and: shear wave.
-q The-reactor vesse1. material is_ heat-treated specifically,to obtain
' good notch-ductility which will ensure a. low NDTT and thereby give
,g assurance that the. finished vessel can be initially hydrostatically tested and. operated at room temperature _without restrictions. The stress ' limits _ e'stablished for the reactor vessel are _ dependent upon
' theitemperature at which'the stresses are applied. Asfa result of 4
- ifest~ neutron absorption in the region of the core', the material Jductility wiu change. The effect.is an increase in the NDTT.
The a
Jpredicted end-of-life NDTT'value of the reactor vessel' opposite the
~
- coreis'260 F:or 1 css.. The: predicted neutron exposure and NDTT ishiftiare discussed? n 4.1.4'and shown on Figure 4.2-8.
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System Description and Operation
/
The unirradiated or initial NDTT of pressure vessel base plate material is presently measured by two methods:
the drop weight test given in ASTM E208, and the Charpy V-notch impact test (Type A) given in ASTM E23. The NDTT is defined in ASTM E208 as "the tem-perature at which a specimen is broken in a series of tests in which duplicate no-break performance occurs at a 10 F higher temperature".
Using the Charpy V-notch test, the NDTT is defined as the tempera-ture at which the energy required to break the specimen is a certain E.
" fixed" value. For SA 302B steel the ASME III Table N-332 specifies an energy value of 30 ft-lb.
This value is based on a correlation with the drop weight test and will be 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 per-formed 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 degrees between curves plotted through the minimum and average values respectively. The determination of NDTT from the average curve is considered representative of the material and is consistent with procedures specified in ASTM E23.
In assessing the NDTT shift due to irradiation, the translation of the average curves is used.
The material for these tests will be treated by the methods outlined
)
in ASME III Paragraph N-313.
The test coupons will be taken at a
/
distance of T/4 (1/4 of the plate thickness) from the quenched surfaces and at a distance of T from the quenched edges. These tests are performed by the material supplier to certify the material as delivered to BW.
The exact test coupon locations are reviewed and approved by BW to ensure compliance with the applicable ASME Code and specifications.
In accordance with ASME III Paragraphs N-712 2
and N-330, BW performs Charpy V-notch impact tests on heat-affected zone (HAZ), base metal, and weld metal on all pressure vessel test plates.
Differences of 20 to 40 F in NDTT have been observed between T/4 and 1 -
the surface in heavy plates.
The T/4 location for Charpy V-notch impact specimens is conservative since the NDTT of the surface mate-
~
rial is lower than that of the internal material.
The reactor vessel design includes surveillance specimens which will permit an evaluation of the neutron exposure-induced shift on the material nil-ductility transition temperature properties.
The remaining material in the reactor vessel and the other reactor coolant system components are purchased to the appropriate design code requirements and specific component function.
The material irradiation surveillance program is described in 4.4.3.
4.2-10 Amendment 2
System Description and Operation
['()
4.2.6 MAXIMUM HEATING AND COOLING RATES The normal reactor coolant system operating cycles are given in Tables 4.1-7 and 4.1-10 and described in 4.1.4.
The normal system heating and cooling rate is 100 F/hr. The exact final rates are determined during the detail design and stress analysis of the
)
vessel.
The fastest cooldown rates resulting from the break of a main steam line are discussed in 14.1.2.9.
4.2.7:
LEAK DETECTION To minimize leakage from the reactor coolant system all components are interconnected by an all-welded piping system.
Some of.the components have access openings of a flanged-gasketed design. The largest of these is the reactor closure head, which has a double metal 0-ring seal with provisions for disposing of leakage between the l2 0-rings.
With regard to the reactor vessel, the probability of a leak occur-ring is considered to be remote on the basis of reactor vessel design, fabrication, test, inspection, and operation at temperatures
((
above the material NDTT as described in 4.3.1.
Reactor closure head t( g Icakage will be zero from the annulus between the metallic 0-ring seals during vessel steady-state and virtually all transient operat-ing conditions. Only in the event of a rapid transient operation, such as an emergency cooldown, would there be some leakage past the innermost 0-ring seal. A stress analysis on a similar vessel design indicates this leak rate would be approximately 10 cc/ min through the seal monitoring taps to a drain, and no leakage will occur past the outer 0-ring seal. The exact nature of this transient condition and the resulting small leak rate will be determined by a detailed stress analysis.
In'the unlikely event that an extensive '.e.ak should occur from the system into the reactor building.during reactor operation, the leakage will be_ detected by nna or. more of the following methods:
a.-
Instrumentation in the. control room will indicate the addition rate of makeup water required to maintain normal
,j water level in the pressurizer and in the makeup tank.
Deviation from normal makeup and letdown to the reactor coolant system will provide an indication of the mag-nitude of the: leak.
D C
Amendment 2 4.2-11~
System Description and Operation b.
Control room instrumentation vill indicate additional reactor building atmosphere particulate or radicactive gas activity.
c.
Control room instrumentation will indicate the existance of a change in the water level in the reactor building sump.
If any one of the methods above indicates an excessive reactor 3'
coolant leakage rate during operation, the reactor will be taken to a cold shutdown, and the cause of the problem sill be determined.
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4.3 SYSTEM DESIGN EVALUATION 4.3.1 SAFETY FACTORS The reactor coolant' system is designed, fabricated, and erected in 1
accordance with proven and recognized design codes and quality standards applicable for the specific component function or class-ification. These components are designed for a pressure of 2,500 psig at a nominal temperature of 650 F.
The corresponding nominal operating pressure of 2,185 psig allows an adequate margin for normal load changes and operating transients. The reactor system components are designed to meet the cades listed in Table 4.1-9.
Aside from ~the -safety factors introduced by code requirements and quality control programs, as described in the following paragraphs, the reactor coolant system functional safety factors are discussed in Sections 3 and 14.
4.3.1.1 Pressure Vessel Safety The safety of the nuclear reactor vessel and all other reactor r~
coolant system pressure vessels is dependent upon four major factors:
(a) design and stress analysis, (b) quality control, (c) proper operation, and (d) additional safety factors.
The special care l2 and detail used in implementing these factors in pressure vessel manufacture are briefly described as follows:
4.3.1'.1.1 Design and Stress Analysis These pressure vessels are designed to the requirements of the ASFE III code. This code is a result of ten years of effort by representatives from industry and government who are skilled in the design and fabrication of pressure vessels.
It is a comprehensive code based on the most applicable stress theory.
It requires a stress analysis of the entire vessel under both steady-state and transient operations. The result is a complete evaluation of both primary and secondary stresses, and the fatigue life of the entire vessel. This is a contrast with previous codes which basically established a vessel thickness during steady-state operations only.
-gr In establishing the fatigue life of these pressure vessels, using the design cycles from Table 4.1-7, the fatigue evaluation curves of ASME III are. employed.
Since ASME III requires a complete stress analysis, the designer 3
must have at his disposal the necessary analytical tools to accomplish this. 'These tools are.the solutions to the basic mathematical theory of elasticity equations.
In recent years the capability and use of s_
Amendment 2 4.3-1 L
System Design Evaluation computers have played a major part in refining these analytical solutions. The Babcock & Wilcox Company has confirmed the theory of plates and shn11s by measuring strains and rotations on the large flanges of. actual pressure vessels and finding them to be in agree-ment with those predicted by the. theory.
B&R has also conducted
' laboratory deflection studies of thick shell and ring combinations to define the accuracy of the theory, and is using computer programs developed on the basis of.this test data.
.9 The analytical procedure considers all process operation conditions.
A detailed design and analysis of every part of the vessel is pre-pared as follows:
a.
'*Ne vessel size and configuration are set to meet the process requirements, the rhickness requirements due to pressure and other structural dead and live loads, and the special fillet contour and transition taper requirements at nozzles, etc., required by ASME III.
b.
The vessel pressure and temperature design tran-
~
sients given in Table 4.1-7 are employed in the determination of the pressure loading and tempera-ture gradient and their variations with time throughout the vessel. The resulting combinations
~T of pressure loading and thermal stresses are cal-J culated.
Computer programs are used in this devel-opment.
c.
The stresses through the vessel are evaluated using as criteria the allowable stresses per ASNE III.
This code gives safe stress level limits for all the types of applied stress. These are membrane stress (to ensure adequate tensile strength of the vessel), membrane plus primary bending stress (to ensure a distortion-free vessel), secondary stress (to ensure a vessel that will not progressively deform under cyclic loading), and peak stresses (to ensure a vessel of maximum fatigue life).
A design report is prepared and submitted-to the jurisdictional authorities and regulatory agencies, i.e.,
state, insurance, etc.
This report defines in sufficient detail the design basis, loading conditions, etc., and will summarize the conclusions to permit in-dependent checking by interested parties.
k 4.3.1.1.2 Quality Control In-process and final dimensional inspections are made to ensure that parts and assemblies meet the drawing requirements, and an "as-built" l
4.3-2
'h System Design Evaluation
' M,f crecord of these dimensions is kept.for reference. A temperature-Econtrolled gage-room is maintained ~to. keep all measuring equipment
'in1 proper, calibration, and personnel' supervising this work are
. trained in formal programs sponsored by gage equipment manufacturers.
i The practice of applied radiography is being continually improved to enhance flaw' detection. Present procedures are:
h' la.. All welds r.re praperly prepared by chipping and grinding valleys between stringer beads so that
. radiographs can be properly interpreted.
- b. ' All radiographs are ' reviewed by two people know-
~,
ledgeable and skilled in'their interpretation.
c.
An 0.080 in. lead filter is used at the film to absorb " broad-bear cestter" when using high voltage
~
equipment (above 1 Mev).
d.
Fine grain or extra fine grain film is used for all exposures.
e.
Densities of radiographs are controlled by densi-tometers.
f.
Double film technique is used on all gamma-ray exposures as well as high voltage exposures.
. g.
Films are processed through an automatic processor which has a controlled replenishment, temperature, and' process cycle, all contributing to better quality.
h.
Energies are controlled so as to be in the optimum range.
- Ultrasonics, one-of the most useful of inspection tools, is being
- used as.follows:-
- a.
l&t addition to radiography, pressure-containing J
welds, where applicable, are inspected.by ultrasonics.
l3 b.
In order to detect laminations which are normally parallel to the surface, plates are also inspected
. by a shear wave.
~
c.
xThe bond between cladding and base material is inspected bylultrasonics.
d.
- . All. plate is 100 percent volumetrically inspected by
. ultrasonics using both normal and shear wave.
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Personnel conducting ultrasonic inspections are given extensive training.
The magnetic particle examination is used to aid in detecting sur-face and near surface defects and is employed on both parts and finished ve m is as follows:
a.
Welds are inspected with the magnetic particle
,ji.
method'after removal of backup strips.
'o. Weld preparations are inspected by the magnetic particle method, c.
The external surface of the entire vessel, including weld seams, is inspected-after all heat treatment.
d.
Personnel using this method are trained by B&W and by the equipment manufacturers who offer fomal training programs.
The liquid penetrant examination helps to detect surface defects and is particularly adaptable to the nonmagnetic materials such as stainless steel.
It is presently being used as follows:
a.
Inspection of weld-deposited cladding.
b.
Inspection of reactor vessel studs.
c.-
Personnal using this method of examination are trained by B&W and by the equipment manufacturers.
The primary purpose of these quality control procedures and methods is to locate, define, and determine the size of material defects to allow an evaluation of defect acceptance, rejection, or repair.
The size of defect that can conceivably contribute to the rupture of a vessel depends not only on the size effect but on the orient-ation of the defec.t, the magnitude of the stress field, and tempera-ture. 'These major parameters have been correlated by Pellini and c
Puzak2 who have prepared a " Fracture Analysis Diagram" which is the basis of vessel operation from cold startup and shutdown to full pressure and temperature operation.
The diagram predicts that, for a given level of stress, larger flaw sizes will be required for fracture initiation above the NDTT tempera-f ture. For example, at stresses in the order of 3/4 yield strength, a flaw in the order of 8 to 10 inches may be sufficient to initiate fracture at temperatures below the NDTT temperature.
However, at NDTT + 30 F, a flaw of 1-1/2 times this size may be required for i
initiation of fracture. While at a temperature of NDTT + 60 F,
~
brittle fracture is not possible under elastic stresses because
-4.3-4
4 System Design Evaluation s-
' h(
)
- brittle ~ fracture propagation does not' take place lat this temperature.
JFractures abovei his temperature are of the predominantly ductile t
type and are dependent upon the-member net section area and section modulus as'they establish.the applied stress.
Stud forgings will be-inspected for flaws by two ultrasonic inspec-j tions.-.An~ axial longitudinal beam inspection will be performed.
The rejection standard will be loss-of-back-reflection greater than that ifrom a '1/2-in. diameter flat bottom hole. A radial inspection
. will be made using the' longitudinal beam technique. This inspection will. carry-the same rejection standards.as_the axial inspection.
In addition to the' ultrasonic tests,' liquid-penetrant inspection will be performed ~on the~ finished studs.-
- The stress analysis of the studs will include e fatigue evaluation.
It 'is not expected that fatigue ' evaluation will yield a significantly high usage factor for the 40-year design life..Therefore, there-will betno planned frequency for stud replacement.
If an indication
.i;s found when the studs are inspected during refueling, as described below, the stud will be replaced.
One-third'of the studs'will be visually examined and dye-penetrant
~
inspected at every refueling.
Any positive indications found will be cause for rejection.
I \\
~The reactor closure head is attached to the reactor vessel with
' '\\
sixty 6-1/2-in. diameter studs. The stud material is A-540, Grade B23 - (ASME;III, Case 1335) which has 'a minimum yield strength of 130,000 psi. The studs, when tightened for operating conditions, will have a tensile stress of approximately 30,000 psi.
Thus, at operating' conditions (2,185 psig):
a.
10 adjacent studs can fail before a leak occurs.
b.
25 adjacent studs can fail before the remaining studs reach yield strength.
.c.. 26 adjacent stud: can fail before the remaining studs reach the ultimate tensile. strength.
d.
43 symmetrically located studs.can fail before the remaining studs reach yield strength.
-]
i4.3.1.1.3 operation As previously mentioned in 4.1.4, pressure vessel service life is
. dependent on adherence to established operating procedures.
Pres-sure_vesselfsafety is also dependent on proper vessel operation.
- j,s Therefore, particular attention is given to fatigue evaluation of
< -t Y
.the; pressure vessels and to the factors that affect fatigue life.
&J 4.3-5 L
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System Design Evaluation The fatigue criteria of ASME III are the bases of designing for fatigue..They are based on fatigue tests of pressure vessels spon-sored by the AEC and the Pressure Vessel Research Committee.
The stress limits established for the pressure vessels are dependent upon the temperature at which the stresses are applied.
As a result of fast neutron absorption in the region of the core, the reactor vessel material ductility will change. The effect is an : increase in the nil-ductility transition temperature (NDTT).
5 The determination of the predicted NDTT shift is described in 4.1.4.1.
This NDTT shift is factored into the plant startup and shutdown procedures so that full operating pressure is not attained until the reactor vessel temperature is above the design transition temperature (DTT). Below the DTT the total stress in the vessel wall due to both pressure and the associated heatup and cooldown transient is restricted.t'o 5,000 - 10,000 psi, which is below the threshold of concern for safe cperation. These stress levels define an operating coolant pressure temperature path or envelope for a stated heatup or cooldown rate that must be followed. Additional information on the determination of the operating procedures is provided in 4.1.4.1, 4.1.4.2, and 4.1.4.3.
4.3.1.1.4 Additional Pressure Vessel Safety Factors Additional methods and procedures used in pressure vessel design,
)'
not previously mentioned in 4.3.1.1 above but which are considered conservative and provide an additional margin of safety, are as follows:
a.
Use of a stress concentration factor of 4 on assumed flaws in calculating stresses.
b.
Use of minimum specified yield strength of the mate-rial instead of the actual values.
Neglecting the increase in yield strength resulting c.
from irradiation effects.
d.
The design shift in NDTT as given in 4.1.4.1 is based on maximum predicted flux levels at the reactor ves-sel inside wall surface, whereas the bulk of the re-actor vessel material will experience a lesser expo-sure of radiation and consequently a 1cwer change in NDTT over the life of the vessel.
e.
Because the irradiation dosage is higher at the in-side surface of the reactor pressure vessel wall, the surveillance specimens will be subjected to a greater degree of irradiation and therefore to a larger shift in NDTT value than will be experienced by the vessel.
c 4.3-6
System Design Evaluation
- fm :)
'b M The specimens' lead the vessel with respect to irra-1diation effects and -impart a degree of conservatism
-in the evaluation of the capsule specimens. The material. irradiation surveillance program is des-cribed in 4.4.3.
'f.'
Lesults from the method of neutron flux calculations, as described in 3.2.2.1.7, 'have. increased.the flux calculations by.a factor of 2 in predicting the nyt-in the reactor. vessel wall. The conservative assump-tions, uncertainties, and comparisons of calculational codes used in determining this factor are discussed in detail in 3.2.2.1.7.
--The foregoing discussion. presents a detailed ' description of quality design, fabrication, inspection, and operating procedures used to ensure confidence in the. integrity of pressure vessels. Experience
~
reported by Reference.5, BG, and the satisfactory experience of
'BW customers support the conclusion that pressure vessel rupture is incredible..
4.3.1.2 Piping
'O Total stresses resulting from thermal expansion and pressure and mechanical and seismic loadings are all considered in the design
~ '
- of the reactor coolant piping. The total stresses that can be expected in the piping are within-the maximum code allowables. The pressurizer surge line connection and the high pressure injection connections are equipped with thermal sleeves to limit stresses from
~
thermalLshock to acceptable values. All materials and fabrication procedures will meet the requirements of the'specified code. All material will be ultrasonically inspected. All welds will be radio-graphically-inspected.. All interior surfaces'of the interconnecting piping are clad with stainless steel co eliminate corrosion problems and to reduce coolant' contamination.
4.3.1.3-Steam Generator Because the basic concept of'the once-through steam generator would indicate the-possible existance of differential thermal expansion-
.p induced. stresses in either the-tubes or shell, the thermal loadings have been evaluated using the most-. severe design transients from Table.4.7-7.
- The' basic structural premise of the steam generator is that the
'tubesheets:themselves are. designed;to take the full design pressure
-on either-sidenof the tubesheet with zero pressure on the other side.
~
That is, the tubes-are not counted upon for any structural aid or
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support.
I
.4.3-7
?'
System Design Evaluation Tne steam line failure antlyzed in 14.1.2.9 closely simulates this
. design premise in a transient manner.. Secondary temperature varia-tions during the accident are essentially transient skin effects with the controlling temperature for the tubesheets and tubes being that of the reactor coolant. Thermal stresses for this case will be below ASME allowable values.
Some tube deformation may occur but will be restrained by the tube supports.
During normal power operation the tubes are hotter than the shell of
+
3 the steam generator by 10 to 20 F depending on load. -The effect is.
to put the tubes in a slight' compression of 3,000 psi at the 20 F maximum temperature difference.
During startup and shutdown operations the tubes are hotter than the shell of -the steam generator by 40 F.
This places the tubes in a
' compressive stress of 6,000 psi. Thus, the stress levels developed-during normal startup or shutdown operations cause no adverse effect on the tubes since these stresses are well below the allowable stress 2l of 23,300 psi for SB-163 material. Buckling of the tubes does not occur.since they are supported laterally at 40-in. intervals along their length. To demonstrate the. structural adecuacy of the steam generator at this condition, a laboratory unit was constructed of the same tube size, length, and material as the steam generator, but of seven tubes in number.
It was structurally tested 'with a thermal difference of shell and tube of 80 F for 2,000 cycles.
This severe difference twice as great as the maximum expected during startup
- )
thermal. cycle test was performed with a tube-to-shell temperature and shutdown (Transients 1 and 2, Table 4.1-7).
Destructive exam-ination of the unit after this test indicated no adverse eifects from fatigue, stress, buckling, or tube-to-tubesheet joint leakage.
-4.3.2 RELIANCE ON INTERCONNECTED SYSTEMS
'The principal. heat removal systems which are interconnected with the reactor coolant system are the steam and feedwater systems and the decay heat removal system.
The reactor coolant system is dependent upon the steam generators, and the steam, feedwater, and condensate systems for decay heat removal from normal operating conditions to a reactor _ coolant temperature of approximately 250 F.
All vital
. active. components 'in these systems are duplicated for reliability purposes.
The engineering flow diagram of the steam an' feedwater systems is shown in Figure 10-2-1.
In the event that the conitasers are not available to receive the: steam generated by decay heat, the water stored in the feed-water system may be-pumped into the steam generators and the resultant
' steam vented to the atmosphere. Either the turbine-driven, or motor-driven auxiliary boiler. feed pumps, both rated at five percent capacity,
- will aupply water to the steam generators.
4.3-8 Amendment 2
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. thermal;drivingihea'd'off thelreactor: coolant system is no longer' adequate StoTgenerate: steam.D This: system is comple'tely-described in 9.5.
The-heat
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D sater system which also.contains sufficient redundancy t'o guarantec j
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= proper operation. ; A schematic diagram: of;the component' cooling water systemispresentedfin: Figure 9.L3-2.-
[4)3.3 SSYSTEM INTEGRITY-
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.The$ integrity-of the~ reactor' coolant system is ensured by proper.
cmaterials selection,afabrication quality control,; design, and O
Loperation. JAll components in.the?reactorfcoolant system are fabri-catedifrom; materials %initi' ally having-a low nil-ductility transition d
1temperaturet(NDTT)1to eliminate the; possibility -of propagating-type ifailures. 'Where material properties are subject to change through-out unit: lifetime,.such'as'the case with the-reactor vessel, provi-l
?
-sions are'inclu'ded'for: materials. surveillance specimens. These will
.be periodically examined,'and.any required temperature-pressure
' restrictions will be' incorporated into reactor operation to-ensurc
' operation above NDTT.
The_ reactor coolant:sys_ tem'is designed in accordance-with ASME h=-
pressure vessel and USASI power piping. codes as covered in 4.1.
l2
_ Relief _ valves on_the pressurizer are sized to prevent system pressure
~
fromLexceeding the. design. point by more than-10 percent.
~
'As a further assurance.of system integrity, all_ components in the.
3 system willsbe.hydrotested at 3,125 psig before initial operation.
'Theelargest an'd most frequently.used opening-in the reactor coolant
- sy. stem,~the reactor.
- closure. head,-contains provisions for separate
- hydrostatiefpressurization-betweenthe'0-ringtypegaskets.
,4.3.41
' PRESSURE RELTEF
~4
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. The reactor ' coolant system _ is protected against overpressure.by lu
.' safety valvesflocated'on.the top;of the pressurizer.
[Theichpacityofthepressurizer.safetyvalvesisdeterminedfrom N
t
. considerations of'U(1) ~the _ reactor protection system,- '(2) pipe pressurc
. dropt (static and dynamic) between the point lof highest pressure in
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- the rreact'or. coolant system fand the pressurizer,. (3) the pressure drop
~
Kin.the safety valve discharge piping, 'and -(4) _ accident or transient V
Leonditions that' may 'potentially cause ' overpre'ssure.
4 4
Preliminary analysis Lin'dicates that the hypothetical case of with-
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. trip ~ with resulting turbine trip. This accident condition produces a power mismatch between the reactor coolant system and steam system larber than that caused by. a turbine trip without immediate reactor trip, or by a partial load rejection from full load.
The ASME Section III required safety valve capacity as determined on the basis of the accident described above is 600,000 lb/hr. Two, 4
300,000 lb/hr valves are installed. An additional pilot-operated b.
safety valve, capable of 100,000 lb/hr, is provided to limit the lifting frequency of the code safety valves.
4.3.5 REDUNDANCY.
The reactor coolant system contains two steam generators and four reactor coolant pumps.
Operation at reduced reactor power is.possible with one or more pumps out of service. For added reliablility, power to the pumps is available from either of two electrically separated 1 sources as shown in Figure 8.2-1.
Separate core flooding nozzles are provided on opposite sides of the reactor vessel to ensure core reflooding water in the event of a single nozzle failure. Reflooding water is available from airher the core flooding tanks or the decay heat removal pumps which provide
- y engineered' safeguard. low pressure injection. The high pressure injection
-)
2l pipes are connected to the reactor coolant system on each of the four l
coolant inlet pipes.
I l
4.3.6 SAFETY ANALYSIS
(.
The components of the reactor coolant system are interconnected by an all-welded piping system.
Since the reactor inlet and outlet nozzfes are located above the core, there is never any danger of the reactor coolant uncovering the core when any other system component is drained for inspection or repair.
4.3.7 OPERATIONAL LIMITS Reactor coolant system heatup and cooldown rates are described in detail in 4.1.4 and 4.2.6.
The component stress limitations dictated by material NDTT considerations are described in 4.1.4 and 4.3.1.
N The reactor coolant system is designed for 2,500 psig at 650 F.
The normal operating conditions will be 2,185 psig at an average system temperature of 579 F at rated power.
In this mode of operation, the reactor vessel outlet-temperature is 603 F.
Additional temperature variations at various~ power levels are shown on Figure 7.2-1.
- 4.3 Amendment 4.
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Reactor: trip' signals wi:1 Exi fed to the reactor protection system Jas airesult.ot.high' reactor coolant temperature,.high' pressure,-le c
f pressure,, and---low flow, i.e.,
flux-flow' comparison and pump monitor i
. ~
By 'elatingilow flow-to the reactor power, operation at signals.
r
. partial-power is ; feasible withjless' than-four reactor coolant pumps operating.
The reactor maximum-calculat d operating limits are as
- follows:
' Performance Vs Pumps In-Service
-Reactor Coolanti 2 9 umps Pumps Operating
_ 4 Pumps.
3 Pumps (2 Loops)
^
Maximum Reactor Power,'
.% of' Rated-1001 86 60 Reactor Coolant Flow,-
% of' Rated 100 74 38 Reactor operating limits under natural circulation conditions are-
-discussed ~1n 14.1.2.6.3.
The. bases for the selection of operational ibnits are discussed further. in 7.1.2.4.
The reactor. coolant system is designed for continued operation uith
-s f(
j l' percent of the fuel rods in the failed condition. The tolerable i '-<
radioactivity content of the coolant is based on long-term saturation activities with 1 percent failed fuel '(Table 11.1-3).
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- TESTS AND-INSPECTIONS 4'.4.1
' COMPONENT IN-SERVICE INSPECTION' Consideration has been given to.the inspectability of -the reactor 1 coolant = system in the design of components,- in the equipment layout, and in the support structures tx) permit access for.the purposes of inspection. Access.for inspection is. defined to be access for visual examination by ' direct or remote meansL and/or by contacting vessel surfaces.during nuclear-unit ' shutdown.
.a
' 4.4.1.1: : Reactor Vessel'
-Access for. inspection of the reactor vessel will be as follows:
a..
Closuie-studs, spherical washers, and nuts can be
- inspected visually or by surface contact methods, b.
External: surfaces and welds on the closure head can be inspected visually or by contact following removal
-of the insulation.
Internal surfaces of the closure head can be examined visually by remote means, r~
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- \\_,/
c.
Inner surfaces of the vessel outlet nozzles can be
- inspected visually by remote means during refueling
. periods..The complete internal surface can be inspected by remote visual means following removal of the reactor core and vess.1 internals, d.
External surfaces of the vessel nozzle to piping welds can be inspected by remote visual means following removal of annulus shield plugs and vessel insulation.
Insulation around the welds will be made in removable sections.
The external surface of the reactor vessel can be inspec-
-e.
~
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.ted during the reactcr lifetime if it' ~should become necessary. - An annulus has been provided between the
. reactor. vessel and-rhe primary shield to accommodate j
inspection equipment. Access to thi.e cavity can be
"$5 gained by disassembly of the shielcing between the reactor vessel and primary shield-at the top of the annulus, which'is designed to be removable.
4.4.1.2 LPressurizer The-external surface will be accessible for surface and volumetric
" i / (
inspection... The internal surface can be inspected by-remote visual
.- (
.means.
w 4.4-1
Tests and Inspections 4.4.1.3-Steam Generator m
i The external surfaces of the steam gene.ator are accessible for
-surface and volumetric inspection. The reactor coolant side of the steam generator can be inspected internally by remote visual means by removing the manway covers in the steam generator heads.
Po rtiot. 4 of the internal surface of the shell and feedwater nozzles can be inspected by removal of the feedwater ring, handhole covers, and manway covers.
D' 4.4.1.4 Reactor Coolant Pumps r
-The external surfaces of the pump casings are accessible for in-spection. The internal surface of the pump inlet is available for inspection by removing the pump internals.
4.4.1.5 Piping The reactor coolant piping, fittings and attachments to the piping external to the primary shield will be accessible for external surface and volumetric inspection.
Reactor coolant piping welds will not be located in the primary shielding.
4.4.1.6 Dissimilar Metal and Representative Welds All dissimilar metal welds will be made in the manufacturer's shops and will be accessible for inspection during the service 1 fe of the nuclear unit.
Dissimilar metal welds on the-reactor vessel include only the core flooding lines, in-core instrumentation guide tubes, and control rod drive housings.
Dissimilar metal welds in the piping include only attachments and the coolant pump inlets and outlets.
reacto.
~
Dissimilar metal welds in the pressurizer include the surge line, the relief valve header, and the spray line connections.
Dissimilar metal welds on the steam generator occur only at the small drain lines and instrument attachments.
Representative longitudinal and circumferential welds on the piping, steam generator, pressurizer, and pump casing will be inspectable as described above.
Representative welds on the reactor vessel closure head wil1~be.inspectable.
Longitudinal and circumferential weld areas on the reactor vessel interior surfaces will be inspectable.
~
4.4-2
Tests and Inspections f 'j ;
4.4.1.7 LInspection' Schedule
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The schedule for the type and frequency of inspection in each of the
. areas mentioned above_will be established during the detailed design.
4'.4.2 REACTOR COOLANT SYSTEM TESTS AND INSPECTIONS The assembled reactor coolant system will be treated and inspected
'during final ^ nuclear unit construction'and initial startup phases, as follows:
.'4.4.2.1
- Reactor Coolant' System Precritical and Hot Leak Test This. test demonatrates satisfactory. preliminary operation of the-entire system and its individual components, checks and evaluates operating procedures, and determines reactor coolant system integ-rity at normal operating temperature and pressure.
- 4. 4.2.2 Pressurizing System Precritical Operational Test This-test demonstrates satisfactory praliminary operation of the pressurizer and its individual components.
Spray valve adjustments
-}>"5 and heater control adjustments are tested.
G' 4.4.2.3 Pressurizer Surge Piping Temperature Gradient Test' The temperature at the midpoint of the pressurizer surge line is determined after a period of steady state operation to check temper-ature gradients.
,4.4.2.4 Relief System Tes't In this test all relief valves are set and adjusted, and operating procedures are evaluated.
4.4.2.5 Plant Power Startup Test Ej
-This test determines performance characteristics of the entire plant in'short_ periods of operation at steady-state power levels.
4.4.2.6 Plant Power Heat Balanqg
. Thia-< test determines the actual reactor heat balance at various powerJ 1evels - to provide the necessary -data for calorimetric calibra-
, (-f ')
tioiiof the nucleari nstrumentation and reactor coolant system flow i
C rat'e.
4.4.
m Tests cnd In pections
)
4.4.2.7l Plant Power Shutdown Test
~This test checks and evaluates the operating procedures used in shutting down the plant and~ determines the overall plant operating characteristics during shutdown operations. These tests are in addition to the tests in compliance with code requirements.
4.4.3 MATERIAL IRRADIATION SURVEILLANCE
$P Surveillance specimens of the reactor vessel shell section material are installed between the core and inside wall of the vessel shell to monitor the NDTT of the vessel mat 3 rial during operating lifetime.
The type of specimens _ included in the surveillance program will be Charpy V-notch (Type A) and tensile specimens for measuring the changes in material properties, resulting from irradiation. This is in accordance with ASTM E185-66, " Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors."
The reactor vessel material surveillance program vill utilize a total of four surveillance specimen holder tubes located close to the inside reactor vessel wall. The locations of the holder tubes are shown in 3
Figures 3.2-59 and 3.2-60.
In these positions, the specimens will receive approximately three times as much radiation as the reactor vessel receives.
As recommended in ASTM E185-66, paragraph 4.6, and indicated in the table below, specimens will be withdrawn at three or more separate times, and one of the data points obtained shall correspond to the neutron exposure of the vessel wall near the end of its design life.
As can be seen from Figure 4.2-8, the planned specimen removal program will provide sufficient data points (indicated on curve) to allow con-struction of the curve of NDTT shift versus integrated neutron exposure for the Rancho Seco reactor pressure vessel.
The material from the reacto,r vessel will have its initial NDTT deter-mined by the Charpy V-notch impact correlation with drop weight tests.
3 The predicted shift or change in the NDTT of the vessel material resulting from irradiation is discussed in detail in 4.1.4.1.
-The influence of neutron irradiation on the reactor vessel material properties will be evaluated periodically during plant shutdowns for
-refueling as tabulated below. Adequate specimen holders are provided to permit evaluation on approximately the schedule shown below.
Schedule for Capsule Removal 3
Equivalent Vessel Material Holder Tubes Exposure Time (vears) 1-10 2
20 3
30
)
4 45 4.4-4 Amendment 3
W Tests and~ Inspections e,: x.
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lThis evaluation will-be accomplished by testing samples of the
~ ' ~ ',
-material from~the; reactor-vessel which are contained in the surveillance' specimen capsules.
These capsules contain steel-coupons from plate,' weld, Land heat-affected zone material used in fabricating the reactor vessel. -Dosimeters are placed with the
. Charpy1V-notch impact specimens and tensile specimens. The dosimeters j
- will permit evaluation of flux as seen by the specimens and vessel.
. wall.. To prevent corrosion the specimens are enclosed in stainless steel sheaths.
The irradiated samples areLtested to determine.the material proper-ties, such as. tensile, impact, etc., and;the irradiated NDTT.which
' may be measured in a manner similar to the initial-NDTT. 'These-
-test results can-be compared with.the then-existing data on the
~,
effects of neutron flux and spectrum on engineering materials.
The measured neutron flux and NDTT may then be compared with the
~
initial NDTT and the predicted:NDTT shift to monitor the progress of radiation-induced changes in the vessel materials. As the end
- of. reactor design. life nears, a significant increase in measured NDTT.in excess ofi the predicted NDTT shif t could be investigated by reviewing the vessel. stress. analysis and operating records.
If necessary or required in accordance with the advanced knowledge available at that time, the vessel transient limitations on pressure
~
- - (
and temperature may be altered so-that vessel stress limits, ss
(
stated in 4.1.4.3 for heatup and cooldown, are not exceed-A.
9
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Ly.
AMENDMENT NO. 2 i
SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 Amendment No. 2 to the Sacramento Municipal Utility District's Preliminary Safety Analysis Report includes both replacement pages and new pages and tabs.
All pages to be inserted are identified as Amendment 2, except the reprinted appendices.
Any technical text material changed by this amend-ment is' coded in the outside margin by a black bar and the numeral two.
Before inserting dhe Amendment 2 material in the different volumes, it is suggested that Appendicas 2A, 2C, 2D and 2E be removed from Volume IV, discarded and replaced with,the new reprinted appendices 2A, 2C, 2D, and 2E.
Additionally, remove Appendices 3 and 4 (including tabs) from Volume V and place at the back of Volume IV.
The list of Effective Pages should be checked to verify the completeness of Volumes I thru V.
"'}
It should be noted that three new additional pages, 10, 11 and 12 are to
{g(( /
be added to the License Application.
The response to letter from Peter A. Morris, Director, Division of Reactor Licensing to E. K. Davis, General Counsel, Sacramento Municipal Utility District, dated Narch 21, 1968, is arranged in the question order of the above letter.
For convenience a cross reference of the AEC DRL question number and SMUD response number is presented below.
Response to questions
.,j'are to be inserted into the volumes according to the assigned SMUD number.
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-4.5' REFERENCES) 1-~'Porse, L.,. Reactor-Vessel Design'Considering Radiation Effects,
.ASME Paper-No.:63-WA-100.
2 )Pellini,' W. S.:and Puzak, P. P.,' Fracture Analysis Diagram
- Procedures. for' the Fracture-Safe' Engineering Design 'of _ Stal
_ Structures,' Welding Research Council Bulletin 88, May 1963.
3 Robertson, T..S.,-Propagation of Brittle Fracture:in Steel, Journal of Iron and Steel Institute. Volume 175. December 1953.
4 Kihara, H. and Masubichi, K., Effects of Res2. dual Stress on Brittle Fracture, Welding Journal. Volume 38_, April 1959.
~
5' Miller, E.
C., The Integrity'of Reactor Pressure Vessels, ORNL-NSIC-15, May 1966.
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