ML19329D998
| ML19329D998 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1967 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| References | |
| NUDOCS 8004090532 | |
| Download: ML19329D998 (24) | |
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. 4g PRELIMINARY SAFETY ANALYSIS REPORT 000 034?
Volume IV O
REGU3 TORY DOCl'1E ru u0<
8 o U 4 0 9 0 '[]2 NOVEMBER 1967 377f.
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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 fr s has been inserted, Appendix 3 should be the first amendment in the new f')
Volume V.
The 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.
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SACRAMENTO MUNICIPAL UTILITY DISTRICT l
RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 j
Docket No. 50-289 LIST OF EFFECTIVE PAGES February 2, 1968 Amendment No. 1 The active pages in this report are as follows:
Page or Fig. No.
Issue Page or Fig. No.
Issue Title...................... 0riginal 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.......................... 0riginal 5.1-3..................... Amendment 1 xi thru xiv.............. Amendment 1 5.1-4 thru 5.1-9............. original 1-1.......................... 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 1.2-1 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 6-ii............. Amendment 1 2.3-6 thru 2.3-8.......... Amendment 1 6.0-1........................Original 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................ Ame nd me n t 1 2.8-1 thru 2.8-4.......... Amendment 1 6.2-1..................... Amendment 1 2.9-1........................ Original 6.2-2 thru 6.2-8............. Original 3-1 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 -6 5............... Amendment 1 7-1.......................... Original F ig. 3. 2 -68............... Amendment 1 7 - ii...................... Ame nd me n t 1 sst B
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(V SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. I s
LIST OF Docket No. 50-289 E F F E CTIV 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 11.3-1............ Original 7.1-10 thru 7.1-11........ Amendment 1 12-i.......................... 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................ Amendment 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 13-1....
.................... 0riginal 7.2-9 thru 7.2-11............ Original 13.1-1 thru 13.3-1............ Original m
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7.3-1 thru 7. 3-7............. original 1........................... 0riginal k2 Fig.
7.3-3................ Amendment 1 11 thru 111............... Amendment 1 7. 4 - 1........................ or i g ina l 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............. Original 14.1-20.................... Amendment 1 8.2-2 thru 8.2-17......... Amendment 1 14.2-1 tl.ru 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 14.2-33.................... Amendment 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 1 9.5-3..................... Amendment 1 14.3-1 thru 14.3-2............ Original 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 ll-ii..................... Amendment 1 Table of Contents....... Amendment 1 11.1-1 thru 11.2-1........ Amendment 1 1A 1A-1 thru 1A-17....... Amendment 1 i n/
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SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 Docket No. 50-289 LIST OF February 2, 1968 E F F E CTIV E PAGES Amendment No. 1 Page or Fig. No.
Issue Page or Fig. No.
Issue 1B 1B-1 thru 1B-5........ Amendment 1 2GA-1................... Amendment 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 4A.2-1 thru 2........... Amendment 1 2C Geology and Seismology-Questions-4A-9 thru 2C-1 thru 2C-13, Fig 2C-1 4A-12................... Amendment 1 thru 2C-ll................. Original Appendix 5 Geophysical Report-Table of Contents....... Amendment 1 1 thru 6................... Original 5A 5A-1..................... Original Additional Seismic Exploration-SA-2 thru 5A-6.......... Amendment 1 1 thru 2, Plate 1 thru Fig SA-1 thru SA-2...... Amendment 1 Plate 3.................... Original SB SB-1 thru 5B-3........... original Geological Log of Drill SC SC-1 thru SC-3........... Original Holes-91 Sheets............ Original 5D 1 thru 10................ Original 2D Seismic Hazard at the SE 5E-1 thru 5E-2........... Original Clay Site 1 thru 14........ Original SF 5F-1 thru 5F-2........... Original Addendum to Seismic Hazard SG SG-1 thru 5G-2........... Original at the Clay Site-1 sheet... Original 5H 5H-1 thru 5H-5........... Original Seismic Hazard at the SI SI-1..................... Original Sierran Sites Area 5J 5J-1 thru 5J-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 Testing-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 D
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SACRAMENTO MUNICIPAL UTILITY DISTRICT
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RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 LIST OF Docket No. 50-289 E F F ECTIV E PAGES February 2, 1968 Amendment No. 1
)
Page or Fig. No.
Issue llA-1 thru 11A-6........ Amendment 1 Appendix 12 Table of Contents....... Amendment 1 12A-1 thru 12A-4........ Amendment 1 Appendix 14 Table of Contents....... Amendment 1 14A-1 thru 141-22....... Amendment 1 O
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l Docket No. 50-312 m
April 15, 1968 AMENDMENT NO. 2 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-is coded in the outside margin by a black bar and the numeral two.
ment Before inserting the Amendment 2 material in the different volumes, it is suggested that Appendices 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.
a It should be noted that three new additional pages, 10, 11 and 12 are to 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 March 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
%.*are to be inserted into the volumes according to the assigned SMUD number.
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QUESTION RESPONSE QUESTION RESPONSE QUESTION RESPONSE NO.
NO.
NO.
NO.
NO.
NO.
r 1.1 1A.4 6.1 6A.7 12.1 12A.2 1.2 1A.5 6.2 6A.8 12.2 12 A. 3 1.3 1A.6 6.3 6A.9
- 12. 3 12A.4 1.4 1A.7 6.4 6A.10 12.4 12A.5 1.5 1A.8 6.5 6A.11 12.5 12A.6 1.6 1A. 9 -
6.6 6A.12 12.6 12A.7 1.7 1A.10 6.7 6A.13 6.8 6A.14 13.1 13A.1 2.1
-14A.14 6.9 6A.15 13.2 13A.2 2.2 14A.15 6.10 6A.16 13.3 13A.3 2.3 14A.16 13.4 12 A. 8 2.4
'14A.17 7.1 7A.2 2.5 14A.18 s.
7.2 7A.3 14.1 14A.20 2.6 14A.19 7.3 7A.4 14.2 14A. 21 2.7 2H.1 7.4 7A.5 14.3 14A.22 2.8 2H.2
'7. 5 7A.6 14.4 14A.23 7.6 7A.7 14.5 14A.24 3.1 3A.6 7.7 7A.8 14.6 14A.25 1;;3 3.2 3A.7 7.8 7A.9 14.7 14A.26
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3.3 3A.8 7.9 7A.10 14.8 14A.27 l
3.4 3A.9 14.9 14A.28
}[7g) f; 3.5 3A.10 8.1 8A.1
-3.6 3A.11 8.2 8A.2 15.1 15A.1 3.7 3A.12 8.3 8A.3 3.8 3A.13 8.4 8A.4 16.1 7A.11 3.9 3A.14 8.5 8A.5 16.2 14A.29 3.10 3A.15 16.3 14A.30 9.1 9A.2 16.4 3A.16 I
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- 4.1 g.
4A.12 9.2 9A.3 16.5 SJ.5
- 4. 2 5J.4 9.3 9A.4 16.6 1A.11 4.3 4A.13 9.4 9A.5 4.4 4A.14 9.5 9A.6 4.5 4A.15 9.6 9A.7
.l 9.7 9A.8
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Amendment 2
1 es Docket No. 50-312 LIST OF s
\\-4 E F F E C T I V E P A G E S April 15, 1968 A
~~--
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 iii.
. Original Fig. 2.3-1 thru 2.3-6.... original ivr'........... Amendment 2 2.4-1............ Orig ina l v thru vii.
Amendment 1 2,4-2.
. Amendment 2 viii.
Amendment 2 2.4-3.
. Original ix............ Amendment 1 Fig. 2.4-1 thr'u 2.4-2.
. Original x.
. Original 2.5-1.
. Original xi.
.#1 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 3-1 thru 3-iii.
Amendment 2 1-iv.
Amendment 2 3-iv thru 3-vi.
. original 1.1-1 thru 1.1-2.
. original 3.1-1.
. original h
Fig. 1.1-1.
. Original 3.1-2 thru 3.1-4.
Amendment 2 l
t<[;ph/
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. 3.2-64.
. 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.
.:w.
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
^
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.
.,s._. Original 3.3-8 thru 3.3-10~.
Original 2-1 thru 2-1,1.
Amendment 1
- 3. 3-11 thre 3. 3-12.... Amendment 2 7N 2-111.
. Original 3.4-1 thru '3. 4-5...... Original i
)
2.1-1.
. Original 4-1 thru 4-11.
Original
$U 2.2-1 thru 2.2-5....... Original 4.1-1 thru 4.1-15.
Original 357 i
Amendment 2 C
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Docket No. 50-312
~ { F E F F ECTIV E PAGES April 15, 1963 Amendaient No. 2 Page or Fig. No.
Issue Page or Fig.'No.
Issue a
Fig. 4.1-1.
. Original Fig. 5.7-1.
. Amendment 2 Fig. 4.1-2.
. Amendment 2
- 5. 8-1.
. Original 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-ii.
. 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.
. Origin'al 6.1-8.
. Original Fig. 4.2-1.
. Amendment 2 6.1-9 thru 6.1-10.
. Amendment 2 Fig. 4.2-2 thru 4.2-8.
. Original 6.1-11.
..\\mendment 1 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.
. Original 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 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 u
j
}5.1-2.
e ~
. Original 7-i.
. Amendment 2 rc )
v 5.1-3.
. Amendment 1 7-11.
. Amendment 1 5.1-4 thru 5.1-9.
. Original 7-iii.
. 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 17P-30........... Amendment 1 7.2-1 thru 7.2-5.
. Original 5.'l-)l............ Original 7.2-6.
. Amendment 2 5.1-32 thru 5.1-33.
. Amendment 1 7.2-7 thru 7.2-8.
. Amendment 1 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 5.2-1 thru 5.2-5.
. Original 7.3-1 thru 7.3-3.
. Original 5.3-1.
. Original 7.3-4 thru 7.3-7.
. Amendment 2
- 5. 4-1.
. Amendment 1 Fig. 7.3-1.
. Amendment 2 5.4-2 thru 5.4-5.
. Original Fig. 7.3-2.
.' Original 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.
. Origin 6'i 5.6-3 thru 5.6-6.
.Amensmcat 2 8-1 thru 8-11
. Amendmen't 2 Fig. 5.6-1.
.\\.
... Amendment 2 8.1-1.
. Original
- 5. 7-1.
. Amendment 2 8.2-l'thru 8.2-18 ?.. '.. Amendment 2 1.7-2.
. Original Fig. 8.2-1 thru 8.2-3
. Amendment 2 O.. i
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Amendment 2
r-Docket No. 50-312
'^N E
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LIST OF April 15' 1968
[xV E F F ECTIV E PAGES Amendment No. 2 Issue Page or Fig. No.
Issue Page or Fig. No.
r 8.3-1 thru 8.3-3...... Amendment 2 12-1.
. Amendment 1 8.3-4.
........ Amendment 1 12.1-1.
. Original Original' 12.2-1 thru 12.2-2.
. Amendment 2 8.4-1.
. Original Fig. 12.2-1.
. Amendment 2 9-1.
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.
. Amendment 2
. Original 9.1-r thru 9.1-2.
. Amendment 2 12.4-1.
9.1-3.
. Original 12.5-1.
. Original 9.1-4 thru 9.1-7.
. Amendment 2 12.6-1.
..... Original
. Original 12.7-1.
. Original 9.1-8.
. Original
. Amendment 2 13-i..
Fig. 9.1-1.
9.2-1.
. Original 13.1-1 thru 13.1-2.
. Original Amendment 2 9.2-2 thru 9.2-7.
. Amendment 2 13.1-3.
. original
... Original 13.2-1.
9.2-8....
.... Amendment 2 13.3-1.
. original 9.2-9....
Fig. 9.2-1.
.. Original 14-1.
........... Original 9.3-1 thru 9.3-6.
. Amendment 2 14-11 thru 14-iii.
Amendment 1 Fig. 9.3-1 thru 9.3-3.
. Amendment 2 14-iv thru 14-viii
. Original 9.4-1 thru 9.4-4.
. Original 14.1-1 thru 14.1-7
. Original s
((j,-j Fig. 9.4-1.
. Amendment l-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. 9.5-1 thru 9.5-2.
. Amendment 2 14.1-17 thru 14.2-19.
. Original
. Amendment 2 14.1-20.
Amendment 2 9.6-1.
9.6-2 thru 9.6-7.
. Original Fig. 14.1-1 thru 14.1-21.
. Original
. Original.14.2-1.
. Original l, Fig. 9.6-1.
. Original 14.2-2.
Amendment 2 9.7-1 thru 9.7-2.
. Amendment 2 14.2-3.
Amendment 1
'"=
ig. 9.7-1.
. Original 14.2-4.
. Original 10-1.
10.1-1..........
. Original 14.2-5 thru 14.2-6.
Amendment 1 14.2-7 thru 14.2-9.
. Original
- 10. 2-1........... Amendment 2 Amendment 2 10.2-2.
. Original 14.2-10.
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 1 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
. Amendment 2 14.2-33.
Amendment 2 11.1-5......
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.
..Aaqndment 1 14.2-38.
Amendment 2 11.2-2 thru 1122-5...... Original 14.2-39.
. original p
11.2-6'........... Amendment 1 Fig. 14.2-1 thru 14.2-18.
. Original h
11.2-7 thru 11.2-11.
. Original Fig. 14.2-19 thru 14.2-20. Amendment 2 11.3-1..
. Original F ig. 14.2-21 thru 14.2-28.
. Original (g
359 E
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,m Docket No. 50-312
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LIST OF April 15, 1968
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E F F ECTIV E PAGES Amendment No. 2 Issue Page or Fig. No Issue Page or F,ig. No.
Fig. 14.2-29.
. Amendment 2 Preliminary Projections to Fig. 14.2-30 thru 14.2-31.
. Original 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. 14.2-35.
. 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 thru 14.3-8.
. Amendment 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.,,.. 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, 1 thru 14.
Original Fig. 14.3-4 thru 14.3-5.
Amendment 1 Addendum to Seismic Hazard Fig. 14.3-6.
. Original at the Clay Site-l sheet Original Amendment 1 Seismic Hazard at the Fig. 14.3-7.
14.4-1 thru 14.4-2.
. Original Sierran Sites Area 15-1 thru 15-5.
...... Original 1 thru 10.
. Amendment t
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1 2E Soil and Foundations Y];p x
Appendix 1 Table of
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Contents.
Amendment 2 Investigation Report lA Answers to Questions.
Amendment 1 2E-1 thru 2E-11, LA-1 thru 1A-14.
Amendment 1 Fig. C-119-E thru Ori inal b
Fig. lA.2-1........ Amendment 1 C122-E.
Amendment 1 Report of Laboratory LA-l7 thru 1A-23.
Amendment 2 Testing-1 thru 9, 3 Tables, Fig. 1 thru 2 and 9, curves 1B,qpality Assurance Amendment 1 i thru 7.
Original
"'ppgrations.
1B-13*...........
. Original 2F Meteorological Station IB-2 thru 1B-4.
. Amendment i 1B-5.
Amendment 2 2G Storage Reservoir Criteria Fig. IB-1.
. Amendment l 2CA-1.
. Amendment 1 1C Rancho Seco Project
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Engineering Staff.
Amendment 2 2H Answers to Questions IC-1 thru IC-4.
. Amendment 2 Fig. LC-1.
Amendment 2 Letter pg. I and 2.
. Amendment 2 Fig. 2H.2-1 thru 2H.2-2. Amendment 2 Appendix 2 Table of Amendment 2 Appendix 3 Table of Contents.
. Amendment 1 Contents.
2A Final Report i thru vi.
. Original 3A Answers to Questions I thru 63.
. Amendment i Fig. 3A.2-1 thru,.3A.2-3.. Amendment 1 Supplement ys, 1 thru 18.
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Amendment 2 Fig. 3A.4-1.
. Amendment 1 3A-15 thru 3A-23.
. Amendment 2
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2B Southeast Area Plan
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(Bound)-14 pages.
. Original Fig. 3A.14-1.
. Amendment 2 f.
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Amendment 2 iP
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LIST OF A ril 15, 1968 P
E F F ECTIV E PAGES Amendment No. 2 L
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 5H-1 thru 5H-5.
Original 4A-1 thru 4A-6 SI Containment structure Fig. 4A.1-1 thru 4A.1-15. Amendment 1 Instrumentation Appeadix A 51-1.
Original The Properties and Micro-5J Answers to Questions structure of Spray-Quenched 5J-1 thru SJ-2.
. Amendment 1 Thick-Section Steels Fig. 5J2-1.
. Amendment 1 15 pages..
. Amendment 2 Original Appendix 6 Table of 4A-1.
. Amendment 1 Contents
. Amendment 1 Appendix B, B & W Data 6A Answers-to Questions (2 pgs.)........Ame'ndment 1 6A-1.
. Amendment 1 6A-2.
. Amendment 2 4A-13 thru 4A-18.
. Amendment 2 6A-3 thru 6A-5..
. Amendment 2 Appendix 5 Table of Contents
. Amendment 1 Fig. 6A.8-1.
. Amendment 2
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. Amendment 2
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5A Structural Design Bases 5A-L thru 5A-5.
. Amendment 2 Fig. 6A.16-l thru Fig. 5A-6 thru 5A-7.
. Amendment 1 6A.16-2.
. Amendment 2 Fig. SA-1 thru SA-2.
. Amendment 1 Appendix 7 Table of SB Justification of Contents
. Amendment i Structural Proof Test -
7A Answers to Questions 7A-1.
. Amendment 1 Pressures
,,.3B-1 thru 5B-3....... Original 7A-2 thru 7A-9.
. Amendment 2 Appendix 8 Table of i
.-5pSpecificationforSplicing Reinforcing Bar Using the Contents
. Amendment 2 Coldwell Process 8A Answers to Questions SC-1 thru SC-3....
. Amendment 2 SD Turbine Generator Missiles Appendix 9 Table of 1 thru 10.
Original Contents.
. Amendment 2 4 sheets of Parts Drawings 9A Answe,rs to Questions' _
SE Justification for Load 9A-1.
. '.. Amendment 2
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9A-2.
. Amendment 1 Factors SE-1 thru SE-2.
. Amendment 2 5F Justification for Yield Appendix 10 Contains nothing Reduction Factors.
Appendix il Table of 5F-1 thru 5F-2...... Original Contents
. Amendment 1 i
SG Description of the Finite 11A Answers to Questions Element Techn\\que Used in#'
. Amendment 1 Containment St'ructural Fig. 11A.1-l.
. Amendment I g
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Analysis Fig. 11A.1-2.
. Amendment 2
,_s 5G-1 thru 5G-2 Original 11A-3 thru 11A-6.
. Amendment 1
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LIST OF April 15' 1968 (s
E F F ECTIV E PAGES Amendment No. 2
- 1 Page or Fig. No.
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Issue l
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 1 15A-1 thru 15A-2.
. Amendment 2 12A-5 thru 12A-10.
. Amendment 2 Fig. 12A.5-1.
. Amendment 2 12A-ll.
. Amendment 2 i
Fig. L2A.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 Questions 13A-1 thru 13A-3.
. Amendment 2 Appendix 14 Table of Contents.
. Amendment 1 14A Answers to Questions 14A-1.
. Amendment 1 14A-2.
. Amendment 2 14A-3 thru 14A-6.
. Amendment 1 kg,/
. Amendment 2 (v
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
. Amendment 1 F)$. 14A.11-1 thru 14A.11-2.
. Amendment 1 14A-21 thru 14A-22
. Amendment 1 14A-23 thru 14A-29.
.Amandment 2 Fig. 14A.18-1.
. Amendment 2
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. Amendment 2 Fig. 14A.19-1.
. Amendment 2
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. 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.
. Amendment 2 Fig. 14A.26-1.
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. Amendment 2 14A-37 thru 14A-41.
. Amendment 2
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TABLE OF CONTENTS s
VOLUME I l.
INTRODUCTION AND
SUMMARY
Section Page
1.1 INTRODUCTION
1.1-1 1.2 DESIGN HIGHLIGHTS 1.2-1 1.2.1 GITE 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 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 1 - HYDRAULIC AND THERMAL DESIGN 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 COOLANT SYSTEM - CODE REQUIREMENTS 1.3-8 (f-l.3.6 ITEM 6 - PRINCIPAL DESIGN PARAMETERS OF THE REACTOR (Li 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 1.3-9 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 1.4.6 CRITERION 6 - REACTOR CORE DESIGN (CATEGORY A) 1.4-5 1.4.7 CRITERION 7 - SUPPRESSION OF POWER OSCILLATIONS (CATEGORY B) 1.4-6 1.4.8 CRITERION 8 - OVERALL POWER COEFFICIENT (CATEGORY B) 1.4-7 1.4.9 CRITERION 9 - REACTOR COOLANT PRESSURE B0UNDARY (CATEGORY A) 1.4-7 1.4.10 CRITERION 10 - CONTAINMENT (CATEGORY A) 1.4-8 1.4.11 CRITERION 11 - CONTROL ROOM (CATEGORY B) 1.4-8 p
t hh 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 SYSTDIS (CATEGORY B) 1.4-12 1.4.15 CRITERION 15 - ENGINEERED SAFETY FEATURES PROTECTION SYSTEMS (CATEGORY B) 1.4-12 1.4.16 CRITERION 16 - MONITORING REACTOR COOLANT PRESSURE BOUNDARY (CATEGORY B) 1.4-12 1.4.17 CRITERION 17 - MONITORING RADI0 ACTIVITY RELEASES (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 SYSTEMS (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 - DIERGENCY POWER FOR PROTECTION SYSTEMS (CATEGORY B) 1.4-17
)
1.4.25 CRITERION 25 - DEMONSTRATION OF FUNCTIONAL OPERABILITY OF PROTECTION SYSTEMS (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 1.4.31 CRITERION 31 - REACTIVITY CONTROL SYSTEMS MALFUNCTION (CATEGORY B) 1.4-20 1.4.32 CRITERION 32 - MAXD1UM REACTIVITY WORTH OF CONTROL RODS (CATEGORY A) 1,4-20 1.4.33 CRITERION 33 - REACTOR COOLANT PRESSURE BOUNDARY CAPABILITY (CATEGORY A) 1.4-21 1.4.34 CRITERION 34 - REACTOR COOLANT PRESSURE BOUNDARY RAPID PROPAGATION FAILURE PREVENTION (CATEGORY A) 1.4-21 1.4.35 CRITERION 35 - REACTOR COOIANT PRESSURE BOUNDARY BRITTLE FRACTURE PREVENTION (CATEGORY A) 1.4-22 1.4.36 CRITERION 36 - REACTOR COOLANT PRESSURE BOUNDARY SURVEILLANCE (CATEGORY A) 1.4-22 nr3 UUL
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ii Amendment 3 i
Section Page 1.4.37 CRITERION 37 - ENGINEERED SAFETY FEATURES BASIS FOR DESIGN (CATEGORY A) 1.4-22 1.4.38 CRITERION 38 - RELIABILITY AND TESIABILITY OF ENGINEERED SAFETY FEATURES (CATEGORY A) 1.4-23 1.4.39 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 1.4.42-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 1.4.44 CRITERION 44 -~ EMERGENCY CORE COOLING SYSTEMS CAPABILITY (CATEGORY ~A) 1.4-26 1.4.45 CRITERION 45 - INSPECTION OF EMERGENCY CORE COOLING SYSTEMSL(CATEGORY A) 1.4-27 1.4.46' CRITERION 46 - TESTING OF EMERGENCY CORE COOLING
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SYSTEMS COMPONENTS (CATEGORY A) 1.4-27 1.4.47 CRITERION 47 - TESTING OF EMERGENCY CORE COOLING SYSTEMS (CATEGORY A) 1.4-28 1.4.48 CRITERION 48 - TESTING OF OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS (CATEGORY A) 1.4-28 (d
CRITERION 49 - CONTAINMENT DESIGN BASIS (CATEGORY A) 1.4-28
'1.4.49 1.4.50 CRITERION 50 - NDT REQUIREMENT FOR CONTAINMENT MATERIAL (CATEGORY A) 1.4-29 1.4.51 CRITERION 51 - REACTOR COO 1 ANT PRESSURE BOUNDARY OUTSIDE 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 ISOLATION 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 1.4.56 CRITERION 56 - PROVISIONS FOR TESTING OF PENETRATIONS (CATEGORY A) 1.4-32 1.4.57 CRITERION _57 - PROVISIONS FOR TESTING OF ISOLATIONS VALVES (CATEGORY A) 1.4-32 1.4.58 CRITERION 58 - INSPECTION OF CONTAINMENT PRESSURE-REDUCING SYSTEMS (CATEGORY A) 1.4-33 1.4.59 CRITERION'59 - TESTING OF CONTAINMENT PRESSURE-REDUCING
. SYSTEMS COMPONENTS (CATEGORY A) 1.4-33 1.4.50 CRITERION 60 - TESTING OF CONTAINMENT SPRAY SYSTDIS (CATEGORY A) 1.4-34 1.4.61 CRITERION 61 - TESTING OF OPERATIONAL SEQUENCE OF CONTAINMENT PRESSURE-REDUCING SYSTDiS (CATEGORY A)- -
1.4-34
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Section Page 1.4.62 CRITERION 62 - INSPECTION OF AIR CLEANUP SYSTEMS (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 1.4.66 CRITERION 66 - PREVENTION OF FUEL STORAGE CRITICALITY
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(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 69 - PROTECTION AGAINSI RADI0 ACTIVITY PILEASE FROM SPENT FUEL AND WASTE STORAGE (CATEGORY B) 1.4-37 1.4.70 CRITERION 70 - C0hTROL OF RELEASES OF RADI0 ACTIVITY TO THE ENVIRONMENT (CATEGORY B) 1.4-38 1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS 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 R0D CLAD FAILURE 1.5-2 1.5.4 HIGH BURNUP FUEL TESTS 1.5-3 1.5.5 INTERNALS VENT VALVES 1.5-3 1.5.6 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 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 14' A U L -+
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Section Page
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2.3 METEOROLOGY 2.3-1 2.
3.1 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 TOP 0 GRAPHY 2.4-1 2.4.3 TERMINAL DISPOSAL OF STORM RUN0FF 2.4-1 2.4.4 HISTORICAL FLOODING 2.4-1
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2.4.5 PREDICTION OF LAND URBANIZATION 2.4-1 2.4.6 GROUNDWATER 2.4-3 2.5 GE0 LOGY 2.5-1 2.6 SE!SMOLOGY 2.6-1 2.7 S0ILS 2.7-1 2.8 SITE ENVIRONMENTAL RADIOACTIVITY PROGRAM 2.8-1
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2.8.1 GENERAL 2.8-1 s-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 REACTOR DESIGN 3.2-1 3.2.1 GENERAL
SUMMARY
3.2-1 3.2.2 NUCLEAR DESIGN AND EVALUATION 3.2-2
.i 3.2.3 THERMAL AND HYDRAULIC DESIGN AND EVALUATION.
3.2-29 l
3.2.4 MECHANICAL DESIGN LAYOUT 3.2-70 3.3 TESTS AND INSPECTIONS 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 ASSEMBLY, CONTROL R0D ASSEMBLY, AND CONTROL ROD DRIVE MECHANICAL TESTS AND INSPECTION 3.3-5
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3.3.4 INTERNALS TESTS AND INSPECTIONS 3.3-10
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3.4 REFERENCES
3.4-1 005 Amendm'ent-3 v
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VOLUME II Section 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 4.1.3 EXPECTED OPERATING CONDITIONS 4.1-7 4.1-8 4.1.4 SERVICE LIFE 4.1.5 CODES AND CLASSIFICATIONS
- 4. 1-15 4.2-1 4.2 SYSTEM DESCRIPTION AND OPERATION 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 4.3.3 SYSTEM INTEGRITY 4.3-9
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4.3.4 PRESSURE RELIEF 4.3-9
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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 COOLANT SYSTEM TESTS AND INSPECTIONS 4.4-3 4.4.3 MATERIAL IRRADIATION SURVEILLANCE 4.4-4 4.5-1
4.5 REFERENCES
5.
CONTAINMENT SYSTEM 5.1 STRUCTURAL DESIGN 5.1-1 5.1.1 CENERAL 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 5.2.1 TYPES OF PENETRATIONS 5.2.2 DESIGN OF PENETRATIONS 5.2-3 3
5.2.3 INSTALLATION OF PENETRATIONS 5.2-5
',j 5.2.4 TESTABILITY OF PENETRATIONS AND WEID SEAMS 5.2-5
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Amendment 4
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Section Page 5.3 CONTAINMENT ACCESSIBILITY CRITERIA 5.3-1 5.4 CONSTRUCTION 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 5.4.3 CONSTRUCTION MATERIALS INSPECTION AND INSTALIATION 5.4-2 5.4.4 SPECIFIC CONSTRUCTION TOPICS 5.4-7 5.5 CONTAINMENT SYSTEM INSPECTION, TESTING, AND SURVEILIANCE 5.5-1 5.5.1 TESTS TO ENSURE LINER INTEGRITY 5.5-1 5.5.2 STRENGTH TEST 5.5-3 5.6 ISOIATION 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 5.9 SYSTEM DESIGN EVALUATION 5.9-1
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ENGINEERED SAFEGUARDS 6.1 EMERGENCY INJECTION 6.1-1 6.1.1 DESIGN BASES 6.1-1 6.
1.2 DESCRIPTION
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 6.
3.1 INTRODUCTION
6.3-1 6.3.2
SUMMARY
OF POSTACCIDENT RECIRCUIATION 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
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Section Page 7.
INSTRUMENTATION AND CONTROL 7.1 PROTECTION SYSTEMS 7.1-l 7.1.1 DESIGN BASES 7.1-1 7.1.2 SYSTDI DESIGN 7.1-6 7.1.3 SYSTDIS EVALUATION 7.1-17 6
7.2 REGULATING SYSTEMS 7.2-1 7.2.1 DESIGN BASES 7.2-1 7.2.2 SYSTDI 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
SUMMARY
OF ALAPJiS 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
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7.4.8 SYSTDI EVALUATION 7.4-4 8.
ELECTRICAL SYSTEMS 8.1 DESIGN BASIS 8.1-1 8.2 ELECTRICAL SYSTEM DESIGN 8.2-1 8.2.1 ELECTRICAL SYSTEM DESIGN NEIWORK INTERCONNECTIONS 8.2-1 8.2.2 STATION DISTRIBUTION SYSTai 8.2-2 8.2.3 DIERGENCY POWER SYSTDI 8.2-9 8.3 DESIGN EVALUATION 8.3-1 8.3.1 EVALUATION OF THE PHYSICAL 1AYOUT 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 9.1 MAKEUP AND PURIFICATION SYSTDi 9.1-1 9.1.1 DESIGN BASES 9.1-1 9.1.2 SYSTEM DESCRIPTION AND EVALUATION 9.1-2
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9.2 CHEMICAL ADDITION AND SAMPLING SYSTai 9.2-1 9.2.1 DESIGN BASES 9.2-1 9.2.2 SYSTDi DESCRIPTION AND EVALUATION 9.2-1 viii Amendment 3 008
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1 9.3 COOLING WATER SYSTEMS 9.3-1 9.3.1 DESIGN BASES 9.3-1 9.3.2 SYSTEM DESCRIPTION AND EVALUATION 9.3-1 9.4 SPENT FUEL COOLING SYSTEH 9.4-1 9.4.1 DESIGN BASES 9.4-1 9.4.2 SYSTEM DESCRIPIION AND EVALUATION 9.4-1 9.5 DECAY HEAT REMOVAL SYSTEM 9.5-1 9.5.1 DESIGN BASES 9.5-1 9.5.2 SYSTEM DESCRIPTION AND EVALUATION 9.5-1 9.6 FUEL HANDLING SYSTEM 9.6-1 9.6.1 DESIGN BASEE
~9.6-1 9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9.6-2 9.7 STATION VENTILATION SYSTEMS 9.7-1 9.7.1 DESIGN BASES 9.7-1 9.7.2 SYSTEM DESCRIPTION AND EVALUATION 9.7-1 VOLUME III g
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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 SYSTEM CHARACTERISTICS 10.1-1 10.1.3 FUNCTIONAL LIMITATIONS 10.1-1 10.1.4 SECONDARY FUNCTIONS 10.1-1 10.2 SYSTEM DESIGN AND OPERATION 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 RADI0 ACTIVITY 10.2-2 10.3 SYSTEM 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 10.3.4 OVERPRESSURE PROTECTION 10.3-2 10.3.5 INTERACTIONS 10.3-2 10.3.6 OPERATIONAL LIMITS 10.3-2
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r-Section Page 11.
RADIOACTIVE WASTES AND RADIATION PROTECTION 11.1 RADI0 ACTIVE WASTE RANDLING 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
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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 REFE RENCES 11.3-1 12.
CONDUCT OF OPERATIONS
12.1 INTRODUCTION
12.1-1 12.2 ORGANIZATION AND RESPONSIBILITY 12.2-1 12.3 PERSONNEL TRAINTNG 12.3-1 12.3.1 TRAINING INITIAL PLANT STAFF 12.3-1
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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 r -
Amendment 4 x
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Section 14.
SAFETY ANALYSIS 14.1 CORE AND C00IANT BOUNDARY PROTECTION ANALYSIS 14.1-1
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14.1.1 ABNORMALITIES 14.1-1 14.1.2 ANALYSIS OF EFFECTS AND CONSEQUENCES 14.1-2 14.2 STANDBY SAFEGUARDS ANALYSIS 14.2-1 14.2.1 SITUATIONS ANALYZED AND CAUSES 14.2-1 14.2.2 ACCIDENT ANALYSES 14.2-1 14.3 ENVIRONMENTAL CONSEQUENCES OF HYPOTHETICAL ACCIDENTS 14.3-1
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14.3.1 GENERAL APPROACH 14,3-1 14.3.2 STEAM GENER\\ TOR 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 R0D EJECTION ACCIDENT 14.3-4 14.3.7 WASTE GAS TANK RUPl'URE 14.3-4 14.3.8 LOSS-OF-COOLANT ACC DENT 14.3-5 14.3.9 MAXIMUM HYPOTHETICAL ACCIDENT 14.3-6 14.3.10 IODINE REMOVAL SENSITIVITY ANALYSIS 14.3-10 O
14.3.11 POPULATION DENSITY CONSIDERATIONS 14.3-12
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14.4 REFERENCES
14.4-1 15.
TECHNICAL SPECIFICATIONS C
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VOLUME IV APPENDIX 1 1A ANSWERS TO QUESTIONS IB QUALITY ASSURANCE OPERATIONS
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APPENDIX 5 5A STRUCTURAL DESIGN BASES 5B JUSTIFICATION OF STRUCTURAL PROOF TEST - PRESSURES SC SPECIFICATION FOR SPLICING REINFORCING BAR USING THE CADWELD PROCESS SD TURBINE GENERATOR MISSILES SE JUSTIFICATION FOR LOAD FACTORS
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5F JUSTIFICATION FOR YIELD REDUCTION FACTORS SG 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 (A
APPENDIX 6 6A ANSWERS TO QUESTIONS APPENDIX 7 7A ANSWERS TO QUESTIONS APPENDIX 8 8A ANSWERS TO QUESTIONS
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13A ANSWERS TO QUESTIONS APPENDIX 14 14A AN3WERS TO QUESTIONS APPENDIX 15 15A
' ANSWERS TO QUESTIONS 014
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1A ANSWERS TO QUESTIONS i
1B QUALITY ASSURMCE OPERATIONS
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QUESTION Discuss in detail the scope of the following research, develop-1A.2 ment, or test programs including projected completion dates for various phases of the programs and test equipment descriptions.
To the extent possible, results of the programs to date should be stated.
1A.2-1 Thermal design, including DNB and flow distribution.
(Will loss of a core barrel check valve be simulated in the flow tests?)
1A.2-2 Control rod drives.
1A.2-3 Steam generator including blowdown tests.
(Discuss the desirability of insulating or otherwise maintaining the shell at a high temperature to simulate the thermal transient that might be experienced in the actual gener-ator during secondary system blowdown.)
1A.2-4 Core barrel check valves.
(Discuss the program for test-ing the valves or a scaled prototype under operational and accident flow and temperature conditions including vibrational effects during operation and mechanical forces during blowdown.)
1A.2-5 Material tests at high burnup.
(Discuss which material properties are critical, the results expected and the s
manner in which the results will be used. Could signifi-
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cant data of a confirmatory nature be obtained by remov-ing and testing fuel from the reactor environment at intervals in the future.
If other test programs, cur-rently in progress, are relied on for fuel rod failure mechanisms, describe the scope and schedule of these tests and compare your requirements in detail.)
ANSWER 1A.2-1 Thermal Design Refer to 3.3.2 a.
Departure from Nucleate Boiling Heat Transfer Investigation In the late fall of 1961, The Babcock & Wilcox Company began the design and construction of a large heat transfer facility for the purpose of doing DNB testing at power reactor operat-ing conditions.
In this facility, which is located at the B&W Research Center, Alliance, Ohio, testing over a wide range of variables covering practically all of the situations one might expect to encounter during normal and expected transient operation of water cooled reactors is possible.
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Amendment 1 lA-2 017 t
Docket 50-312
,s Amendment No. 1 (f}
February 2, 1968 y,,,
QUESTION What is your criterion for a minimum shutdown margin during 1A.1 operational transients?
ANSWER Minimum Shutdown Margin - Operational Transients Refer to 1.4.29 The reactor is designed to meet the criterion that it can be shut down to the hot suberitical condition with a margin of at least 1% Ak/k with one control rod stuck out of the core.
The evaluation of operational transients - such as moderator dilution without rod motion, loss of pumping power, and rod withdrawal - has shown that this margin is not changed by these transients, because the reactor returns to the hot subcritical condition at the end of the transient.
This margin at the hot shutdown condition also provides sufficient shutdown reactivity to keep the reactor subcritical in accident-induced transients which cool the reactor coolant to lower temperatures, such as a steam line failure.
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cold walls, inter-channel mixing, and instabilities.
Data from these assemblies are still being analyzed, f'
and work is progressing on a new DNB correlation.
Results to date indicate that the analytical methods used in the design of the reactor core are conservative and that no critical areas exist.
The Babcock & Wilcox Company intends to maintain an active and aggressive program in the field of DNB heat transfer.
Some of the principal programs which will be conducted in the near future are outlined below along with a tentative schedule:
(1) A 9-rod assembly with the same rod diameter, pitch spacing and spacer grid used in the fuel assembly will be tested. A nonuniform axial power generation profile will be employed over six feet of the bundle length.
The power profile will be representative of that portion of the core experiencing the most severe heat transfer conditions and the most probable location of a DNB.
Testing of this assembly is scheduled for the first two quarters of 1968.
(2) An annular specimen with nonuniform power distribution on the outside tube will be tested for additional verif-ication of the effects of length and nonuniform power 3
generation. Power may be supplied to various portions j
of the specimen so that length effects up to the full 12-foot long test region of the specimen may be examined.
Testing for the annular specimen is expected to begin in the fourth quarter of 1968.
(3) A 9-rod bundle test employing nonuniform radial power distribution will be tested in 1969. A definitive pro-gram and schedule for this series of tests is not formulated.
(4)
Depeh.dingupontheresultsobtainedfromtheprevious tests, additional tests will be devised as part of the continuing basic heat transfer and core optimization program. Tests under consideration are for additional radial and axial power distributions, larger test assem-blies, investigation of different grid designs, and transient simulation, b.
Mixing Studies Related to the studies for DNB are additional programs con-ducted to determine the degree of mixing in the fuel rod channels. Flow tests involving a 4-rod assembly have been conducted to determine mixing effects. Flow tests on a mockup of the outer two rows of fuel rods and the can panels s
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1A-4 Amendment 1 0n7
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( f"T The facility is supplied with 1.8 megawatts of d-c power
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and a fully. automated data acquisition system it can be e
operated within the following limits:
4 Pressure - 100 to 2,700 psia Inlet subcooling - 20 to 250 F 6
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Mass velocity - 0.2 x 10 to 3.5 x 10 lbs/hr-ft in a 9-rod assembly.
Present specimen size rod assembly with a heated length of six feet.
DNB detection with thermocouples (resistance measurement back-up)
Flow - 150 gpm at 295 ft head Since the loop has been completed, a variety of experiments have been performed to gain better understanding of the DNB phenomema and to develop empirical relationships necessary for the design of water reactors. Among the experiments i
completed to date have been tests on:
(1) Single tubular specimens with both uniform and nonuni-
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form power distributions. Nonuniform axial peak-to-U average powers as high as 1.9, simulating inlet and out-let peak locations, have been included in the tests.
These tests were conducted as a function of shape,
. length, and system parameters. On the basis of these tests, power shape factors for application of the test results to reactor design have been determined, and it was concluded that simulation of reactor axial power shapes.could be achieved with confidence in test bundles of shorter length than actual reactor fuel assemblies.
(2) Annular specimens with various combinations of inner and outer wall heat generation and nonuniform axial power
. distributions were also tested.
It was determined that the results obtained with the annular data correlated very well with data taken on bundles. Analytical work done on the tubular and annular specimens has formed the bnsis for the bundle size and power generation shape to be used in a future test bundle described below.
(3) A 9-rod test assembly with a uniformly heated length of six feet, simulating the reactor fuel rod diameter, pitch spacing, and spacer grid details, has been tested.
l This was the first test approaching actual geometrical conditions as well as operating conditions for the core.
A Of principal interest were the effects of spacer grids, j
Amendment 1 1A_3
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top and bottom of each assembly. P essure sensors and 4
thermocouples are provided in all fuel assemblies to deter-mine the flow distribution at the core inlet. Additional pressure sensors and thermocouples are provided in other portions of the vessel and core so that overall mixing and pressure drop determinations may be made.
Preliminary investigations in the model and the analysis described in the appendix to Section 3.2.4 herein (Item 5.1.5) indicate that it will not be necessary to simulate the inter-nal check valve construction, operation, or any malfunctions in the vessel model flow tests.
Testing should be completed in mid-1968.
Refer to 1A.2-2 Control Rod Drives 3.3.3 a.
Component Tests The purpose of this program is to seek out potential material and/or design problems prior to production unit testing. The component test program consists cf:
(1)
Evaluation of various grades of Graphitar-bearing mate-rials in an autoclave at 1,600 psi, 600 F, and water chemistry with 13,000 ppm H B0. The bearing materials 3 3 s
are statically loaded against a 17-4 PH shaft such that
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the developed stress is greater than will be present in the actual control rod drive.
(2) Environmental dynamic gecr and bearing tests under loads equivalent to the control rod drive operating conditions.
In this test, the bevel gears, the pinion, and the bear-ings supporting these gears are being tested in an auto-clave at 2,250 psi, 400 F, and reactor water chemistry.
The purpose of this test is to obtain wear characteristics of the gear material combinations and projected life of the bearings.
(3) Simulated drive test. A complete mechanism, which simu-lates the drive with the exception of its overall length, is being tested under no-flow reactor operating conditions of temperature and pressure in an autoclave. An accel-erated wear and life test through a short stroke will be completed in conjunction with the life-testing of the prototype mechanism.
(4) Autoclave testing at reactor operating temperature and pressure of buffer seal, splines, and bearings.
In this test, the spline joints of the drive rod assembly are being tested under static, no-load conditions for corrosion, i
1A-6 Amendment 1
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of two adjacent fuel assemblies have been conducted to deter-4
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mine the friction effects at the perforated wall boundary.
These tests have confirmed that the larger flow cells at the periphery of the bundle compensate for 1.igher equivalent friction adequately. These effects are shown numerically in the appendix to Section 3.2.3.2.4j herein. Additional tests to extend the investigation to larger sizes, and more elabo-j rate geometry for the purpose of confirming the analytical model and value for mixing coefficients, are described below.
(1) A 9-rod mixing test assembly, of the same bundle geometry as the DNB bundle described previously, has been con-structed to determine the degree of mixing present during the DNB investigations. Testing with this assembly is currently in progress and is expected to be completed in the first two quarters of 1968.
(2) A 16-rod assembly with the simulated juncture of four perforated, fuel assembly cans meeti,ng at the corner is under construction. Testing with this assembly will enable one to determine the degree of mixing which occurs between fuel assemblies, and will give more detailed information on velocity distributions and mix-ing in the peripheral cells of the fuel assembly than did the 4-rod tests. The current core analysis considers only mixing within a fuel assembly and does not take h-credit for mixing external to the assembly.
It is V
expected that testing with this assembly will begin in January 1968.
(3) A facility large enough to accept a 64-rod assembly is currently under construction. Tests for this facility are not yet firm, but it is expected that some of the preliminary work for calibration of in-core thermocouples and pressure differential instrumentation will be done in this facility.
Initial plans were to construct a low pressure facility large enough to accept a full size, cross section fuel assembly. This has currently been replaced with the 64-rod assembly, and its need will be re-evaluated. Testing in this facility is scheduled for the second quarter of 1968.
j c.
Vessel Model Flow Tests A 1/6 scale model of the reactor vessel, the internals, and I
the reactor coolant piping from the pumps to the reactor vessel is currently being tested at the Research Center.
Portions of the reactor vessel and internals are constructed of transparent plastic to facilitate visual observation of flow patterns within the vessel. The reactor core is simu-lated in the model with individual fuel assemblies constructed
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of perforated sheet material and calibrated orifices at the
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In addition to the testing described in the above refer-
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ence, secondary system blowdown tests have been carried out, and a reactor coolant (primary) side blowdown is planned when schedule commitments permit. The hot water facility of the Research Center is shared by the Control Rod Drive Tests, the Steam Generator Tests, and other experiments and tests.
Three secondary system blowdown tests have been completed.
The results of these tests have demonstrated the integrity of the steam generator under conditions of rapid depres-surization and large (greater than 200 F), tube-to-shell temperature differentials.
In addition, the results of these tests are used in the development and verification of analytical models for steam system blowdown analyses.
The construction of the test steam generator (including insulation) is such that the thermal time constant of the shell is lower than that of a full-scale unit. This lower time constant results in more rapid cooling of the shell during steam system blowdown than would occur in a full-size unit.
The primary side blowdown test will provide temperature
'g conditions which simulate a thermal transient greater
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than that for the full-scale unit secondary blowdown as well as simulation of the thermal transient for primary blowd own.
Refer to 1A.2-4 Core Barrel Vent Valves 3.3.4 The core barrel vent valves will be designed to relieve the pressure generated by steaming in the core following the LOCA so that the core will remain suf ficiently covered. The valves will also be designed to withstand the forces resulting from rupture of either a reactor coolant inlet or outlet pipe. Testing of the valves will consist of the follcwing:
a.
A full-size valve assembly (seat, locking mechanism, and socket) will be tested at steady-state conditions at the maximum pressure expected to result during the blowdown.
b.
S uf ficient tests will be conducted at zero pressure 3
to determine the frictional loads and clearances in the hinge assembly, the inertia of the valve cover, and the deflections resulting from impact of the cover so that the valve response to cyclic blowdown forces c.sy be determined analytically.
3 OI23 1A-8 Amendment 1
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(5)
Autoclave testing of shortened drive rod assembly under
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static load conditions. This test is similar to the spline testing with the addition of the bevel gear set at the lower end of the drive rod.
(6) Autoclave testing at reactor temperature and pressure of the bevel gears, bearings, shortened rack, and pinion gear under vibratory loading of the rack to determine the fretting characteristics of the gear train. This test is a static load test, b.
Full Scale Prototype Testing Under No Flow Conditions This test will be performed in an autoclave permitting full stroking in room temperature water and ac reactor operating conditions of temperature, pressure, and water chemistry.
The cold tests will be utilized only as an initial checkout of the drive prior to temperature and pressure testing. The centrol rod will be simulated with a dummy weight. Misalign-ment will be introduced to note its effect on wear and overall performance of the drive mechanism, c.
Full Scale Prototype Testing at Reactor Operating Conditions of Temperature, Pressure, and Flow The full life test program as defined in the PSAR will be
('7-g) conducted under this test. A prototype control rod and fuel g
assembly will be used in order to establish the complete drive train assembly.
Inasmuch as the rack and pinion drive concept described in Section 3.2.4.3.2 herein is somewhat different from the first rack and pinion drive tested at the Research Center, the test program which has been outlined above provides an extension of previous tests to establish verification of drive perform-ance and adequacy. The previous test program verified the basic material selection, the snubber design, and the buffer seal concept for use with a rack and pinion drive.
The components test program (items a and b) is scheduled for completion by the end of December 1967.
Hot loop testing (item c) at full flow conditions of the prototype drive will begin in October 1967, and continue until the end of December 1967.
Refer to 1A.2-3 Steam Generator 4.4 The basic steam generator test program is discussed in detail in Appendix 4A of the Duke Power Company PSAR (Dockets 50-269, 270, and 287).
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Amendment 1 1A-7
The fuels irradiation program will test fuel specimens at design temperatures and at exposures in excess of those obtained in the fuel rod.
The specimens irradiated to the design burnup are scheduled to be completed in mid-1969, well in advance of reactor opera. ion. The pro-gram will provide information on the swelling rate of U02 as a function of burnup, density, heat rate, and cladding restraint.
Fuel specimens will be operated at heat rates up to 21.5 kw/f t, which is in excess of the peak specific power in the core. The burnup will range up to 75,000 W D/TU. The fuel rods will operate with a cladding surface temperature of 650 F.
A program has been carried out to determine the effects of irradiation on the mechanical properties of Zircaloy-4.1 Tests were conducted to temperatures as high as 775 F.
The summary of results from this program is as follows:
a.
The room temperature tensile and yield strengths of Zircaloy-4 increased with total neutron exposure for irradiation temperatures up to 650 F.
The rate of increase was greater at lower irradiation temperatures.
This increase in strength was accompanied by a decrease in the total and uniform elongations.
b.
The room temperature yield and tensile strengths of
)
the specimens irradiated at 775 F were somewhat lower than those of the specimens before irradiation. These changes in properties, however, were not significantly different from those observed in specimens aged out-of-pile for like periods of time, c.
The room temperature uniform elongation values for both annealed and cold-worked material were approxi-mately 2 percent after neutron irradiation at 130 F 1
to an exposure of 4.5 x 10 9 nyt (E > 1 Mev).
d.
A difference in irradiation behavior was noted for the longitudinal and transverse specimens, narticu-larly after irradiation at 775 F.
At this semperature the tensile strength in the transverse direction con-tinued to increase whereas in the longitudinal speci-mens the strength decreased.
A summary of capsule specimens is given in Table 1A.2-1 and a tentative schedule is presented in Figure 1A.2-1.
The following is a description of the research now in progress at B6W that is related to the current reactor design.
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1A-10 Amendment 1
The valve assembly will be pressurized to determine
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what pressure differential is required to cause the c.
A determination of the valve to begin to open.
pressure differential required to open the valve to its maximum open position will be simulated by mechanical means.
A valve assembly will be installed and removed d.
remotely in a test stand to judge the adequacy of handling equipment.
Since the temperature differential existing across the valve assembly during normal operation in the reactor is only approximately 55 F, and since the same material is used for the valve seat, socket, and cover, there is no need to conduct tests at elevated temperatures.
The valves are located in a region of relatively low velocity and turbulence, and preliminary analysis indi-cates that there is insufficient energy in the coolant to cause vibrational problems. Therefore, no testing to prove the vibrational adequacy of the valve is planned.
Testing should be completed by January 1969.
Refer to 1A.2-5 High Burnup Fuel Tests
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The design of fuel rods for pressure cycles and thermal 3.2.4.2.2 gradients are amendable to analysis, based on out-of-pile In determining the behavior of materials properties.
under the influence of accumulated irradiation the prop-erties of interest are uranium dioxide growth rates under restraint by tubular cladding, and the ability of the cladding to absorb strain without failure at reactor J
operating conditions.
A detailed report of sources of information for the irradiation of clad and fuel has been presented in the In addition to the PSAR, 3.2.4.2.2 plus references.
PSAR references, irradiation of fuel assemblies or par-tial fuel assemblies with Zircaloy-clad 002 is in prog-These ress in the Saxton and Big Rock Point reactors.
data will demonstrate the behavior of fuel assemblies
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under the combined effects of irradiation, pressure cycles, thermal gradients, reactor coolant environment, J
and fuel-clad restraints.
B&R-is conducting a program to obtain a better understand-ing of fuel growth rates and irradiation effects on clad-i ding, the influence of hydrogen on cladding, and fission gas release at high burnup for the specific design burnup f)
projected for peak power regions in the reactor.
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4 026 1A-9 Amendment 1
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Task III - High Burnup Fuel Irradiation O-)
The primary purpose of the Hi h Burnup Program is to F
determine the swelling rate of UO2 as a function of burnup using fuel rods of the same design as the core.
In addition to determining the swelling rate, the effect of several other variables including the density, heat rate, and cladding restraint will be investigated.
The program consists of capsules some of which will operate at a heat rate of 18 kw/ft and others at a heat rate of 21-1/2 kw/ft. The pellets, other than U-235 con-tent, will conform to the reactor fuel specifications.
The burnup will range from 10,000 to 75,000 FWD /TU with six capsules exceeding 45,000 FMD/TU. The capsules will not operate with an external pressure.
However, two different cladding thicknesses, 0.015 end 0.025 in will be used to vary the restraint offered by the cladding.
The f uel rods will operate with a cladding surface temp-erature of 650 F.
The diametral gaps between the pellets and cladding will vary from 4-5 to 7-8 mils, to give smeared densities of about 92.3 and 90.8 percent, respec-tively. These gaps and smeared densities are consistent with the fuel rod specifications. The insertion date for the first capsule was September 5,1967.
The tests are oriented toward the determination of the behavior of materials in an irradiation environment and
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to determine the optimum geometric and material proper-ties for the specific application. The information is essential for advancement of the art, but is not con-sidered critical in the sense that all of the programs must be completed to insure safe operation.
Removal and testing of fuel taken from operating reactors at intervals during operation is not considered necessary.
The data on hand, plus programs which are currently under way, should satisfactorily provide the information neces-sary for assurance of safe operation within the limits required.
REFERENCE 1 Mechanical Properties of Zircaloy-4 Af ter Irradiation at 130, 650, and 775 F, TP-299, April 1967.
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1A-12 Amendment 1
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Material Irradiation Testing Program
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The purpose of this program is to determine the effects of irradiation on the core components of a central sta-tion power reactor. The program is divided into three tasks:
Task I Low Burnup Fuels Irradiation Task II Zircaloy-4 Irradiation Task III High Burnup Fuels Irradiation Task I - Low Burnup Fuels Irradiation The primary objective of this task is to investigate the dimensional stability of pellet-type fuel rods when irradiated at current and future IVR operating conditions.
The program consists of capsules, some of which are designed to operate at 21-1/2 kw/f t.
The cladding for these capsules will operate at a surface temperature of about 640 F.
All of the capsules will be irradiated, when possible, for one complete cycle of the BAWTR.
Under normal operation, this will amount to about 25 p
EFPD and a burnup of about 3,500 to 4,000 LWD /TU.
The irradiation of capsules, initially operated at a heat rate of 25-25.7 kw/ft, has been completed. Some capsules received as much as 609 power cycles at 22.8 to 24.6 kw/ft. Hot cell examination is underway.
Task II - Zircalov-4 Irradiations The Zircaloy-4 cladding in the core operates with outside and inside surface temperatures as high as 650 and 800 F, respectively. A program was therefore designed to deter-mine how the mechanical properties of Zircaloy-4 are affected by irradiation at these temperatures.
C Longitudinal specimens cut from 0.425-in diameter Zircaloy-4 tubing are used to determine the properties in the longitudinal direction. Ring specimens and flat-tened rings conforming to dimensions of the longitudinal specimens are used to determine the properties in the transverse direction. Some of the tensile specimens were charged with 250 to 400 ppm hydrogen prior to 2
irradiation.
Irradiation of the two 300-day capsules is continuing without any operational difficulties. As of June 30, 1967, these capsules had achieved an exposure of 306 EFPD.
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, V 028 Amendment 1 1/-11
Docket 50-312
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Amendment No. 1 February 2, 1968 QUESTION We believe that research and development above that which you 1A.3 have indicated will be required to justify the use of core barrel check valves as a solution to the steam bubble problem.
Further consideration should be given to testing (1) vibration effects on the valves (caused by core barrel vibrations) and (2) flow charac-teristics in the reactor after loss of a valve. We believe that if the loss of a valve is not detectable, the DNB ratio at the overpower condition af ter loss of a valve must be not less than 1.3 (based on the W-3 correlation).
ANSWER In order to investigate vibration of the vent valves caused by Refer to core barrel vibrations, it was assumed that the core support 3.3.4 shield would excite the disc at a frequency where the shield mode shape corresponded to an 8-valve configuration. This frequency is 125 Hz and is substantially below the lowest resonant fre-quency of the disc, i.e. 1500 Hz.
This large difference in fre-quency indicates that vibratory motions transmitted from the core support snield to the disc will not be amplified by the disc and will not exceed transmitted motions from the shield, which our preliminary analysis indicates will be less than 0.005 inch.
Other more rigorous, but more time consuming, analytic methods are being pursued in order to confirm the vibratory motion of the
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shield. Assuming the worst case of the disc being force-excited at 125 Hz, the amplitude of the disc would have to exceed 0.025 inch in order to develop an inertial force which would exceed the pressure load of 2-1/2 tons (based on 31.5 psi) which acts to keep the valve shut, at full flow. Therefore, it is not possible to transmit sufficient high frequency vibratory power from the coolant stream to cause the shield to vibrate at an amplitude of 0.025 inch.
It is concluded that even under the most pessimistic assumptions, excitations from the core support shield cannot cause the valve to open or vibrate. Therefore, it is not neces-sary to perform a vibration test which would attempt to vibrate the valve by simulating the postulated excitation.
The DNB ratio in the hot channel at the maximum overpower with a vent valve disc off will be high enough to insure that there is a 99 percent confidence that at least 94.5 percent of the popula-tion of all such channels are in no jeopardy of experiencing a DNB. This degree of protection is consistent with Paragraphs 3.1.2.3 and 3.2.3.1.1 of the PSAR.
It will be demonstrated in the final design that the DNB ratio in the hot channel with the flow resulting from the loss of one vent valve disc will not be less thar 1.3 using the W-3 correlation.
A preliminary sensitivity analysis using postulated worst case parameters has been made for the reduced flow. The results of this analysis are described in the appendix to Section 3.2.4 j
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1970 031 FIGUURE 1A. 2-1 IIIGli BURNUP IIRRADIATIO:' PROGREi SCllEDULE (BAS. SED ON B/4*-DI-192),
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herein where the DNB ratios for full and reduced flow are
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as follows for various reactor powers:
Percent Rated Power DNBR (Full Flow)
DNBR (Reduced Flow) 100 1.76 1.68 107.5 1.53 1.44 112 1.40 1.30 114 1.34 1.24 The minimum DNB ratio of 1.24 resulting from the analysis at 114 percent power for the postulated worst case is large enought to ensure a DNB ratio of not less than 1.30 for final design con-ditions. The postulated worst case, used for sensitivity analy-sis, is not the design condition but a case with heat transfer and mechanical conditions much more severe than expected in the final design. This is demonstrated by a comparison of the nom-inal and postulated worst case as shown in the appendix to Section 3.2.3.2.4 hetein where the W-3 DNB ratios are as follows at rated flow conditions and 114 percent power:
Cell Type Nominal DNB Postulated Worst Case DNB y
(m,)
Corner 1.85 (1.71) 1.34 (1.24)
Wall 1.89 1.38 Unit 1.89 1.46 The minimum DNB ratios occurring in the corner cell for the two conditions at reduced flow due to loss of a vent valve disc are shown in parentheses above. The final design DNB will be within the limits of 1.71 to 1.24 shown.
It is expected that a value greater than 1.30 will result from final evaluation of the com-bination of the following significant factor s:
(1) Mixing coefficient of 0.03 to 0.07 at design conditions compared to 0.01 used in the preliminary analysis.
(2)
Statistical determination of nechanical tolerances in lieu of minimum conceivable dimensions.
(3) A more accurate determination of the hot channel local peak-ing factor of 1.095 shown in Figure 3.2-55 of the PSAR con-sidering:
(a) the statistically determined water gap, and (b) the excess metal in the solid can section surrounding the corner pin. The final value is expected to be about 1.06.
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1A-15 Amendment 1
(4)
Application of final vessel and core flow distribution tests results instead of the hot to average fuel assembly flow ratio of 85 percent assumed for the worst postulated Case.
(5) The statistical comparison of the multiple rod fuel assembly heat transfer test data with the singic channel data that currently forms the basis for the W-3 correlation.
A consideration of the final thermal-hydraulic design compared with the preliminary postulated worst case and the mechanical integrity of the vent valve indicates that it is very unlikely that the core will be subject to an unsatisfactory heat transfer condition.
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1A-16 Amendment 1
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\\[~~)N QUESTION Update the discussion of your proposed design with respect to
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1A.4 its conformance to the Commission's Proposed General Design (DRL 1.1) Criteria.
Include in this discussion the impact of the several design changes made in your facility.
ANSWER Response to the Commission's Proposed General Design Criteria including discussion on the impact of the several design changes made are presented in Section 1.4 of the PSAR. Those criteria which reflect changes are 7, 10, 11, 22, 38, 44, 46, 52, 59, 61 and 62.
QUESTION Describe cach of your research and development programs with a LA.5 proposed schedule for obtaining the desired information.
(DRL 1.2)
Include, as appropriate, when the design of the associated feature must be frozen in order to meet the schedule for con-struction of the Rancho Seco Plant.
ANSWER Research and development programs that will provide information to complete the final detail design of some of the components or to demonstrate the capability of the design for future (d'g 3 operation at a higher power level are summarized in Section
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1.5 of the PSAR. Further discussion of research, development
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or test program is provided in the answer to Question 1A.2 in Appendix LA of Amendment 1.
Additional information and dis-cussion is provided below:
a.
Once-Through Steam Generator Testing necessary to prove the adequacy of the once-through steam generator design for service at the initial power level and to confirm the size and configuration of the units has been completed. These programs were described in Appendix 4A of the Oconee PSAR (Docket Nos. 50-269, 270 and 287) and in the Rancho Seco PSAR, Appendix 1A, Question LA.2.
Steady state and load changing operations using once-through steam generator models have demonstrated the ability of the unit to follow transients and the interac-tion of the control system with the water level, steam pressure and flows. Primary and secondary blowdown tests on the models have demonstrated the integrity of the units under conditions of rapid depressurization and large tube-to-shell temperature differentials. The results of the blowdown tests are being used in the development and verification of analytical models for steam system blow-down analyses.
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Amendment 2 1A-17
b.
_ Control Rod Drive Unit These programs have been described in Section 3.3.3.4 and Appendix 1A, Question lA.2 of the PSAR. Some of the results of those programs will be discussed in this reply.
The development and testing of the rack and pinion drive is being conducted under three separate programs:
1.
Full-scale prototype testing under no-flow conditions.
2.
Full-scale prototype testing at reactor operating conditions of tempera ture, pressure, and flow.
3.
Components testing.
The no-flow prototype testing is performed in an autoclave in which the reactor conditions of control rod stroke, temperature, pressure, and water chemistry are duplicated.
The tests are performed with a dummy weight equivalent to the weight of the control rod assembly attached to the rack.
The objectives of this testing were to verify the design concept and to obtain a preliminary verification of the trip insertion time.
The mechanism was subjected to approximately 100 full-stroke cycles and 100 trip cycles simulating both hot and cold reactor conditions.
This testing confirmed that the design and the mechanical arrangement met the objectives. The time for 2/3 insertion was less than 1.2 seconds; the snubber design worked properly, and the buffer seal did not impair trip capability.
Further testing was conducted which included a complete life test of full-stroke cycles and trip cycles simulating reactor operating conditions with maximum tolerance mis-alignment. Examination of componencs after the test indicated that the wear observed was acceptable on all components except the miter gear which although badly worn continued to operate satisfactorily.
The control rod drive life testing program will be con-tinued after the mechanism has been refurbished and modi-fled to incorporate a new miter gear utilizing 17-4 PH nitrided or Haynes 25 metal.
The second life test will be conducted with different stroking specifications than those used on the first life test.
/
03c3 lA-18 Amendment 2
/~S Other prototype testing was conducted in another autoclave \\('"') in which all reactor operating conditions except radiation are duplicated. The complete driveline is established with prototype components, i.e., the fuel assembly, control rod, upper guide tube of the reactor internals, and the drive mechanism. This testing concentrated mainly on the performance characteristics under coolant flows ranging from zero to full flow at reactor conditions of temperature, pressure, and water chemistry. The objective of these tests was to determine the compatibility of the mechanism trip tLme with the specification requirements of 1.4 seconds for 2/3 insertion. After some modification of the pattern of flow holes in the shroud of the upper guide tube, the trip time ranged from 1.31 to 1.4 seconds. Selected componunts testing was performed prior to and in addition to the life testing programs in order to resolve potential material or design problems. These component test programs produced the following results: 1. Provided the basis for the selection of Graphitar ~ bearing material. fs 2. Ascertained the buffer seal injection flow rate. ! ( ) 3. Assured acceptable wear from the revised miter gear combination. 4. The corrosion product buildup in the static test of the splines and bearings has not noticeably affected the resistance to rotation of the system. The program will be completed by August, 1968. By that date the prototype mechanism will have completed the life test program as out' lined in the PSAR and all material ~ problems for the production type mechanisms will have been resolved. c. In-Core Neutron Detectors This program consists of basic physics parametric studies of the detector and mechanical insertion - withdrawal tests of the assembly. The development program has been outlined in Section 7.3.3.3 of the PSAR. Mechanical testing of the assembly has been completed. All parametric studies have been completed except the long term radiation effects and the depletion effects. Results to date have been satis-factory and the performance of the detectors has been demonstrated. As of April 1968 detectors have been irradiated (} in the Big Rock Point Reactor for about 34 months and in 'y[j_) The Babcock and Wilcox Test Reactor for about 23 months. These lifetine tests are continuing, s Amendment'3 1A.19 n 7 f, UJu
d. Core Thermal and Hydraulic Design j These programs have been discussed in Section 3.3.2 and Appendix 1A, Question LA.2 of the PSAR. The initial core power level has been justified on the basis of the W-3 cor-relation which has been approved for the design of several similar pressurized water reactors. With the use of that correlation, only the reactor vessel flow model test data is necessary to further substantiate the core thermal and ~ hydraulic design. Test runs already completed without check valves in the internals have demonstrated the ability to provide adequate flow distribution. Tests including check valves will be completed in 1968. The Departure from Nucleate Boiling (DNB) and mixing studies described in the PSAR are being conducted to support the final thermal design margin on the basis of the B&W cor-relation and to provide for an increase in its rated power output when that increase is requested. Due to the fact that the information produced by these programs has been, is being, or will be used to finalize the detail design of components for Oconee and Three Mile Island units which are scheduled to precede the Rancho Seco unit into com-mercial operation by about two years, the information needed from these programs will be available long before it is needed f} to freeze design details for the Rancho Seco unit. QUESTION If not specifically included in 1.2, describe your program 1A.6 including schedule and acceptability criteria for vibration (DRL 1. 3) testing of the core barrel check valves. ANSWER The testing program for the core barrel check valves (internals vent valves) was discussed in Appendix 1A, Question lA.2. In addition to the testing discussed there B&W is presently work-ing with the valve designer-manufacturer and a vibration testing laboratory on the details of the vibration test of a full scale prototype vent valve. The prototype valve will be mounted in a test fixture which duplicates the method of valve mounting in the core suppccc shield and simulates this local area of the internals. The test fixture with valve installed will be attached to a vibration test nachine and excited sinusoidally through a range of frequencies representative of those antici-pated for the core support shield during reactor operation. The relative motion between the valve disc and seat will be monitored and recorded during test. The test results will be evaluated and, if required, the valve design will be modified prior to valve production to eliminate any adverse disc vibra-tion problems. All testing will be completed by January 1, .)' 1969. UJ/ 1A-20 Amendment 3
QUESTION If not specifically inaluded in 1.2, discuss the programs 1A.7 currently in progresa hat will assure fuel element capability (DRL 1.4) for 55,000 MWD /MTU 1 .wp at the design power densities. ANSWER A high burnup fuel irradiation test program is in progress at B&W, and is dr 'ribed in Question LA.2-5, Appendix LA of the PSAR. A
- e. ale for the program is shown in Figure tA.2-1.
This program includes fuel specimens with repre-sentative cladding thickness, fuel-clad gaps and UO2 densi-ties. Heating rates are representative of maximum heating rates and temperatures in the core. Post-irradiation examination will include profilometer scans to determine permanent clad strain, fission gas release, metallographic examination of fuel and cladding, and confirmation of burnups estimated from flux monitors dsring the test. Maxi-mum target burnup is 75,000 MUW/MTU. Examination will be made at several stages of burnup between 10,000 and 75,000 MRD/MTU to determine the behavior of the fuel and cladding as a function of burnup. The damage criteria for the high burnup test program are that the cladding will not allow fission product release, or the entrance of coolant into the fuel rod which could 3 lead to further damage. Other experiments in the industry ("y have shown that the Ibnit of permanent strain in the ((_,) cladding is approximately 1.5%.1 This, therefore, repre-sents the current upper limit to avoid damage associated with excessive clad strain. Design limits are set at approxi-mately 1%. (See PSAR 3.1.2.4.2.c). The consequence of burnup on fuel rods is that continued fuel growth and fission product release will eventually lead to clad failure due to progressive clad strain. The point of failure is influenced by irradiation-induced changes in the cladding. The program is designed to better under-stand the limit of burnup and allowable strain which can be achieved without clad failure. The program will also assure fuel element capability for 55,000 MWD /NTU burnup at the design power densities. It will also give a better understanding of the burnup limit, or " margin of safety," for fuel rods of representative design when tested at maximum heating rates. REFERENCE 1 Fracture of Cylindrical Fuel Rod Cladding due to Plastic Instability, WAPD-TM-651, April 1967. 038 s i i c.s' Amendment 3 1A-21
l QUESTION Submit the staffing and training plans for SMUD's Nuclear 1A.8 Project Engineering Staf f. (DRL 1.5) ANSWER Tae staf ting and training plans for SMUD's Nuclear Project Engineering Staff are presented in Appendix 1C. The program presented will provide the District with a technically quali-fled engineering staff both during construction and after the plant is operational. QUESTION Discuss the principal design decisions yet to be made that lA.9 require nuclear and steam plant knowledge and which affect (DRL 1.6) nuclear power plant safety. Indicate the approximate dates by which these decisions must be made and to what extent reli-ance will be placed upon contractors for making decisions. Indicate how the training plans for SMUD personnel are orien-tated toward these requirements. ANSWER The principal design decisions which affect nuclear power plant safety have been made and are presented in the PSAR. However, studies are in progress as defined in Section 1.5 )' of the PSAR which may affect the final design. SMUD will review the results of these studies, with the aid of consult-ants if necessary, and initiate any action required to ensure a safe plant. As a matter of policy, SMUD does not rely on contractors to make decisions on matters of safety. Additionally, SMUD does not rely on contractors to make decisions concerning plant reliability, maintainability or operability. SMUD, working with its consultants and contractors, identifies problem areas and calls for proposed solutions. The proposals are then eval-uated by SML'D with expert support as necessary from its con-sultants. Training of SMUD personnel toward these requirements is set-forth in Appendix 1C of the PSAR. 039 lA-22 Amendment 3
QUESTION Your Amendment No.1 provided the SMUD response to applicable 1A.10 questions raised during the review of a sbmilar plant (Metro- ~ (DRL 1. 7) politan Edison). This response used information that was available through November, 1967. Please update your response to these questions by considering applicable information that became available in January, 1968. ANSWER The PSAR has been updated to be responsive through Metropolitan Edison Company's PSAR Amendment No. 6 (Docket No. 50-289). QUESTION Discuss and evaluate your program to experimentally study 1A.11 vibrations in the check valves. (DRL 16.6) ANSWER (See the response to Ques tion 1.3) i i l i l 04'd Amendment 3 1A-23 ..}}