ML16194A165
| ML16194A165 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/12/2016 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16193A354 | List: |
| References | |
| Download: ML16194A165 (437) | |
Text
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0503W-1
*TWO: DIGIT"~tABLE OF: CON.TENTS
- Section
- 1. 0
1.1 INTRODUCTION
AND GENERl);L DESCRIP'.J;iON OF PLAI$1;: INTRODUCTION 1.2 GENERAL PLANT DESCRIPTION 1.3 COMPARISONS 1.4 IDENTIFICATION OF.AGENTS AND CONTRACTORS
*' £ ,.
- 1. 5 REQUIREMENTS FOR FURlllER TECHNICA~L*INFORMAT~p:til:
- 1. 6 MATERIAL INCORPORATEP.:BY REFERENCE*
1.7 DRAWINGS 1.8 NRC REGULATORY GUIDES 1.9 OTHER CONCERNS AND'CptlMITMENTS l.9A TMI RELATED REQUIREMENTS 2.0 SITE CHAltACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.2 NEARBY INDUSTRIAL. TRANSPQRTATJON,.***Mfil*MILITARY FACILITIES.* 2.3 METEOROLOGY 2.4 HYDROLOGY 2.4A EROSION ESTIMATES
- 2. 5 GEOLOGY, SEISMOLOGY AND GEOTECHNICAL' . ENGINEERIUG 2.SA BORING LOGS & DATA SUMMARIES 2.5B FLORIDA EARTHQUAKE Oif.:-OCTOBER 27. 19!:f
- 3. 0 DESIGN CRITERIA-STRUCTURES, COf1PONENJ.'.S. EQJJf:P,~T AND-SYSTEMS*
3.1 CONFORMANCE WITH NRC GENERA~ DESIGN CRITERIA
- 3. 2 CLASSIFICATION OF STRU.£'.f'p~rns' SYSTEM§ ¥iD COMPONENTS .
0151F/l I
0503W-2
' .: Two DIOO!Yf:f'.#BtE OF CON:TENTS (Cont'd)
Section 3.3 3.4 WIND AND TORNADO LOADINGS WATER LEVEL (FLOOD) DESIGN 3.5 MISS ILE PROTECTION " ~. :::~- .-._":-,' a;,; 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE RUPTURE OF PIPING
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3.6A HIGH ENERGY PIPE RUPTURE ANALYSIS - INSIDE CONTAINMENT 3.6B HIGH ENERGY PIPE RUPTURE ANALYSIS - OUTSIDE CONTAINMENT 3.6C PIPE WHIP RESTRAINTS AND BREAK LOCATIONS
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3.6D STRUCTURAL DETAILS OF PIPE WHIP RESTRAINTS 3.6E MAIN STEAM AND FEEDWATER ANALYSIS
- 3. 61i.QG~i. . C~ .bMQDEWR ;bERGY{I!iEPlNGTFiULURE ANALYSIS
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3.7 SEISMIC DESIGN 3.8 DEs4Gi-S :ot;, .CATEGORY I ~.~TRUCTURES 3.SA EVALUATION OF CONCRETE MASONRY WALLS 3.9 MECHANICAL SYSTEMS AND COMPONENT$
- 3. 9A OPERABILITY CONSIDERATtiJNS; FOR*"BE-!-SMIC CATEGORY I ACTIVE PUMPS AND VALVES 3.9B CONCRETE EXPANSION ANCHOR DESIGN 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND
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- t.I.~ :*. 1.oAt.G.l r:.L ~£ili:E--TiRil~.1:-.i!'oR."/:SErst1tc, QuA1LrFICATION oF sEisMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT,'AND THEIR SUPPORTS 3.11 ENVIRONMENTAL QUA:t:.'i:FICATION 4.0 REACTOR 4 .1 SUMMAR¥- DES<m~i'i'ION *; /*. .J,-
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* :0151F/2 II Amendment No 6 (4/91)
0503W-3
- Section 4.2 Title FUEL SYSTEM DESIGN ',~
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4.3 NUCLEAR DESIGN 4.4 THERMAL AND HYDRAULIC DESIGN 4.5 REACTOR MATERIALS'..\:'.
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4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS
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5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS l'!'\ _t. . 5.1
SUMMARY
DESCRIPTION
~ 1\3~( -~l_~ -!-*~~l.V.~ *r~_;'*._._....iJJY ::r_~*;... (:.j * .f 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY (RCPB)
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5.2A OVERPRESSURE PROTECTION FOR ST- LUCIE UNIT 2 - PRESSURIZED WATER REACTOR JJ:'-._ .:' i,;,)-:_,,~:! ..... ,,.,- __M WJ:.:J. :~;t *:'- .. :: 5.2B ANALYSIS OF ST. LUCIE' UNI!J[:J. l;{AT:tJRAL::*:tU!§'.JU~TIQN COOLDOWN _: WITHOUT UPPER HEAD VOIDING AND ST. ~UCIE UNIT 2 CONDENSATE STORAGE TANK REQUIREMENTS ~ >~:;.c: .. -' '.**F*::T_} *: .t: 5 .2C ST. LUCIE UNIT 1 NATURAL QIRCULATION_CpOL'@.liN* 5.3 REACTOR VESSEL . l. . :. :* 5.4 COMPONENT AND *'SUBSYSTEM*DESIGN*_:2 ,~S\{-;_,. , -. * "'ff' lf. *:
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6.0
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6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.2 CONTAINMENT SYSTEM_S Ji: :;: -- ~: ~-*i,/* A..~_}~}T:. . t ;::!, -~~.:.:.. ~Jl?..~** ..... 6.2A EBASCO MODIFICATIONS TO THE --CONT~f!'-L.:'!':J1~~ _::~6- COMPUTER CODE 6.2B WATEMPT-A COMPUTER CODE-* TO QALG_Ul::&~T~:- 1'.J;~J~-~~~p BUILDLNG, ANNULUS TRANS IENX: .:.::.. . ~ ,._,,, . __ ~IJ :_* _,.._,-_:f'- .-_ -~I~V';_.~~~-;;,_::i-' 6.3 EMERGENCY CORE COOLING_-. SYST~ .;___.;;;,\" " ' :- .,.- *,i}1:1 6.4 HABITABILITY SYSTEMS 6.5 FISSION PRODUCT REMOVAL AND :C0Nn~¥ s~s~~§r-;_;::* . I.
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6.6 INSERVICE INSPECTION OF QUALITY GROUP B AND C COMPONENTS 7.0 INSTRUMENTATION AND CONTROLS
7.1 INTRODUCTION
0151F/3 III *Amendment No_6 .. ' .. (4/91)
0503W-4 1
-' TWO DIGIT."'TABLE OF3CONTENTS (Cont'd)
Section 7.lA 7.2 RPS MATRIX POWER REACTOR PROTECTIVE SYSTEM SUPPL~~ISOLATION QUALIFICATION 7.3 ENGINEERED SAFETY FEATURESVS~S~EMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION
- 7. SA SAFETY ASSESSMENT 'SYSTEM .. iii
. :n;;rr.;:;Y7._6 .:~~~l~. '~LL OTHER SYSTEMSX REQUIREID }'.QR SAFETY
- 7. 7 CONTROL SYSTEMS NOT REQUIRED": FOR SAFETY
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8.1 INTRODUCTION
- -~
8.2 OFFSITE POWER SYSTEM rl f"
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8.3 ONSITE PQWER_SYSTEM 9.0 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONING. HEATING. COOLING AND VENTILATION SYSTEM 9.5 OTHER AUXILIARY STEAMS 9.SA FIRE PROTECTION PROGRAM REPORT ., 10.0 STEAM AND PQWER CQNVERSION SYSTEM 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM
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- r, 10.4 OTHER FEATURES OF THE STEAM AND POWER CONVERSION SYSTEM 10.4.9A AUXILIARY FEEDWATER SYSTEM REQUIREMENTS EVALUATION 10.4.9B AUXILIARY FEEDWATER SYSTEM RELIABILITY ANALYSIS 0151F/4 IV Amendment No 6 (4/91)
I
- 0503W-5 TWO DIGI'I,'iTA:'BLE OF CONTENTS (Conlt 'd)
- Section 11.0 Title RADIOACTIVE WASTE'* MANAGEMENT .;._
*** ~-.~~*~*-- *-:~.. f 11.1 SOURCE TERMS ...
I -_ 7 .,_ 11.lA DERIVATION OF RESIDENCE TIMES ___':. .1**: 11.2 LIQUID WASTE SYSTEMJ~-- 11.3 GASEOUS WASTE SYSTEM c~: ' . 11.4 SOLID WASTE MANAGEMENT SYSTEM _ ::
.1 11.5 PROCESS AND EFFLUENT g&ADIQLOGICAL* MONITQRING AND SAMPLING SYSTEMS 12.0 RADIATION PROTECTION\_:_::~*** .. -;:r~:r,i::_ __
12.1 ENSURING THAT OCCUPATIONAL RADI'ATIQH.EXPOSURES ARE AS LOW bS REASONABLY ACHIEVABLE (ALARA) - -* .. ,. 12.2 12.3 12.3A RADIATION SOURCES RADIATION PROTECTION DESIGN FEATURES TM! SHIELDING STUDY
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i), 12.4 DOSE ASSESSMENT 12.5 HEALTH PHYSICS PROGRAM 13.0 CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE
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13.2 TRAINING PROGRAM 13.3 EMERGENCY PLAN 13.4 REVIEW AND AUDIT
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13.5 PLANT PROCEDURES 13.6 PLANT RECORDS
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14.0 INITIAL TESTS AND OPERATIONS r: !.J..J. f1[l .. f .! ;r ~~ ** *~: : 15.0 ACCIDENT ANALYSIS - ORGANIZATION AND METHODOLOGY 15.1 INCREASED HEAT REMOVAL BY THE SECONDARY SYSTEM
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0151F/5 v Amendment No 6 (4/91)
0503W-6 TWO DIGIT TABLE OF CONTENTS (Cont'd) Sect~9n 15~~4 15.3 DECREASED HEAT REMOVAL BY THE SECONDARY SYSTEM DECREASE IN REACTOR COOLANT FLOW RATE 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.5 INCREASE IN REACTOR COOLANT SYSTEM INVENTORY 1~~6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.7 RADIOACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 15.8 PRIMARY SYSTEM PRESSURE DEVIATION 15.9 ANTICIPATED TRANSIENTS WITHOUT SCRAM (A!WS)
+s.10 STATION BLACKOUT ANALYSIS 1s..1+ CESEC CYCLE 1 16.0 TECHNICAL 1.SPECIFICATIONS 17.0 17.1 17.2 QUALITY ASSURANCE QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION QUALITY ASSURANCE DURING THE OPERATING PHASE 0151F/6 VI Amendment No 6 (4/91)
812-FSAR SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Amendment 210A. l-1 6 210A.2-1 6 210A.3-1 6 210A.4-1 9 210A.5-1 6 210A.6""."1 6 210A. 7-1 6 210A.8-1 6 210A.9-1 6 210A.10-1 6 210A.11-1 6 210A.12-1 6 210A.13-1 6 210A.14-i 6 210A.15-1 6 210A.16-1 6 210A.17-1 6 210A.18-1 6 201A.18-2 6 210A~ 18-3 6 210A.18-4 6 210A.18A-1 9 210A.18A-2 9 210A.18A-3 9 210A.18A-4 9 210A.18A-S 9 210A.19-1 6 210A. 20-1 6 210A. 21-1 6 210A. 22-1 6 210A.23-1 6 210A.24-1 6 210A.25-1 6 210A.26-1 6 210A.28-1 6 210A.29-1 6 210A.30-1 11 210A. 31.1-1 6 210A. 31. 2-1 6 210A. 31. 3-1 6 210A.32-1 6 210A.33-1 6 210A.34-1 6 210A.35-1 6 210A.35-2 10 210A.35-3 6 F210A.35-1 6 RNQ-2 Amendment No. 12, (8/82)
SL2-FSAR
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NRC QUESTIONS Amendment F210A.35-2 6 F210A.35-3 6 F210A.35-4 6 F210A.35-5 6 210A.36-l 6 210A.37-l 10 210A.38-l 6 210A.39-l 6 210A.40-l 6 210A.40.l-l 7 . - .. 210A.40.l-2 7 210A.41-l 6 210A.41.1-l 10 210A.41.l-2 6 210A.42-l 6 F210A.42-l 6 . *:,:.- F210A.42-2 6 F210A.42-3 6 F210A.42-4 6
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F210A,,42-5 6 .- F210A.42-6 6 F210A.42-7 6 F210A.42":"8 6 *" F210A.42-9 6 ... 210A.42.l..;.l 6 . - 210A.43-l 6 210A.44-l 6 210A.44-2 6 210A.45-l 6 210A.46-1 6 210A.47-1 6
- 210A.48-1 6 210A.49-l 6 210A.50-l 6 210A,.50-2 6 210A.50-3 6 210A.S0-4 6 210A.50-5 6..
210A.50-6 6 210A.50-7 6.. _ 210A.50-8 6 F210A.50-l 6 F210A.50-2 6 F210A,,50-3 6 .. F210A.50-4 (Sheet l of 4) 6 F210A.50-4 (Sheet 2 of 4) 6 F210A.50-4 (Sheet 3 of 4) 6 .. RNQ-3 Amendment No o 10, (6/82)
SL2-FSAR LIST OF EFFEr,TIVE PAGE8 (Cont'd) F210A.50-4 (Sheet 4 of 4) 210A.51-l NRC OUESTIONS Amendment 6 6 210.A.52-l 6 210A. 53-1 6 210A. 54-1 10 210A.55-1 6 210A.56-1 6 210A.56-2 12 210A.56-3 6 210A. 56-4 6 210A.56-5 6 F210A.56-1 6 F210A.56-2 6 210A.57-1 13 2 lOA. 58-1 13 210A. 58-2 13 210A.58-3 13 F210A.58-1 13 210A.59-l 13 F210A.59-1 13 F210A.59-2 13 F210A. 59-3 13 F210A. 59-4 13 210A.60-l 13 210A.61-1 13 F210A.61-1 13 F210A.61-2 11 F210A. 61-3 13 F210A.61-4 13 F210A.61-5 13 F210A.61-6 13 F210A.61-7 13 F210A.61-8 13 F210A.61-9 13 F210A.61-10 13 F210A.61-11 13 F210A.61-12 13 F210A.61-13 13 F210A.61-14 13 F210A.61-15 13 F210A.61-16 13 F210A.61-17 13 F210A.61-18 13 F210A.61-19 13 210.A.62-l 13 RNQ-4 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Page Amendment 210A.62-2 13 210A.62-3 13 210A.62-4 13 210A.62-5 13 F210A.62-1 13 210A.63-1 13 210A.64-1 13 210A.65-1 13 210A.65..-2 13 220.1-1 5 220.2-1 5 220.3-1 .5 220.3-2 5 220.4-1 5 220.5-1 5 220.6-1 10 220.7-1 5 220.8-1 5 220 .9-1 5 220.10-1 5 220 .10-2 12 220.10-3 12 220 .10-4 5 220 .10-5 .12 220.10-6 5 F220.10-1 12 F220.10-2 12 220.11-1 5 220.12-1 5 220.13-1 5 220.14-1 5 220.15-1 5 220 .16-1 5 220.16-2 5 220 .16-3 5 F220.16-1 5 F220.16-2 5 F220.16-3 5 F220 .17-1 5 220.18-1 5 220.19-1 5 220.20-1 5 220.21-1 5 220.22-1 6 220.22-2 6 220.23-1 5 220.24-1 5 RNQ-5 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NPC QUESTIONS Page Amendment 220.25-1 5 220.26-1 6 220.26-2 6 220.27-1 5 220.28-1 5 220.28-2 5 220. 29-1 12 220.29-la 12 220. 29-2 5 220.29-3 5 220. 29-4 5 220.29-5 5 220.29-6 12 220. 29-7 5 220.29-8 5 220.29-9 5 220.29-10 5 220.29-11 .5 220.30-1 12 220.31-1 12 220.32-1 5 220. 33-1 11 F220.33-1 11 F220.33-2 11 220. 34-1 12 220.35-1 5 220.35-2 10 220.35-3 10 220.35-4 10 220. 35-5 10 220.35-6 10 220.35-7 10 220.35-8 10 220.36-1 5 220.37-1 5 220.37-2 13 220.37-3 13 220.37A-l 6 220 .38-1 6 240 .1-1 4 240.2-1 12 240. 3-1 7 240.4-1 4 240.4-2 11 240 .5-1 4 240.6-1 4 240 .6-2 4 240. 7-1 11 F240.7-l 4 RNQ-6 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Page Amendment 241.1-1 6 241.1-2 6 241.1-3 6 241. 2-1 6 F241. 2-1 6 241.3-1 6 241.4-1 6 241.5-1 6 241.5-2 6 241.5-3 6 241.5-4 6 241.5-5 6 241.5-6 6 241.5-7 6 F241.5-1 6 F241.5-2 6 241.6-1 6 241.6-2 6 241. 7-1 6 241.8-1 6
- 241.8-2 241.8-3 241.9-1 241.10-1 251. 2-1 251.3-1 6
6 6 6 6 6 251.4-1 12 251.4-2 12 - 251.4-3 6 251.4-4 6 251.4-5 6 251.4-6 6 251.5-1 6 251.6-1 6 251. 7-1 6 251. 7-2 6 251.8-1 6 251.8-2 6 251.8-3 6 251.8~ 6 251.9-1 6 251.10-1 6 251.10-2 6 251.10-3 6 251.10-4 6 252.1-1 6 252.1-2 6 252.1-3 6 RNQ-7 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Page Amendment 260.1-1 6 260.2-1 6 260.3-1 6 260.4-1 6 260.5-1 7 260 .5-2 7 260.5-3 7 260.5-4 7 260.5-5 12 260. 5-6 12 260.5-7 7 260.5-8 12 260.5-9 7 260.5-10 7 271.1-1 9 271. 2-1 9 271. 3-1 9 2 71.4-1 9 271.5-1 9 271.6-1 9 280. 1-1 q 280 .1-2 9 280. 2-1 9 280.2-1 9 280.3-1 12 280.4-1 9 280.5-1 9 280.6-1 11 280.7-1 12 280 .8-1 9 280.9-1 9 280.10-1 9 280.11-1 9 280.12-1 9 280.13-1 9 280.14-1 12 280.15-1 12 280 .16-1 12 280.17-1 12 280 .18-1 9 280.19-1 9 280.20-1 1.1 280. 21-1 9 280.22-1 11 280.23-1 9 280.24-1 9 280.25-1 9 280.26-1 9 RNQ-8 Amendment No. 13, (2/83)
SL2-FSAR
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}. RC QPESTIONS Amendment 280.27-1 9 280.27-2 9 280. 28-1 9 280. 29-1 9 280. 29-2 q 280.30-1 9 280.30-2 q 280.30-3 9 280.30-4 q F280.30-l 9 F280.30-2 9 F280.30-3 9 F280.30-4 9 F280.30-5 9 281.1-1 7 281.1-2 7 281.1-3 3 281.2-1 4 281. 2-2 11 281. 2-3 6 281. 3-1 11 281.3-2 3 281.4-1 3 281.5-1 3 281.6-1 3 281. 7-1 3 281.8-1 7 281. lA-l 7 281. 2A-l 7 *282.1-1 6 282.1-2 7 F282.l-l 7 282.2-1 6 282.3-1 6 282 .4-1 6 311.1-1 6 311.2-1 12 320.1-1 6 320.1-2 6 320.1-3 6 320.1-4 6 320.1-5 6 320.1-6 6 320.1-7 6 320.1-8 6 320. 1-9 6 320.1-10 6 320.1-11 6 RNQ-9 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) Page 320.1-12 320.1-13 J\TRC OUFSTIONS Amendment 6 6 320.1-14 6 320.1-15 6 320.1-16 6 320.1-17 6 320.1-18 6 320.1-19 6 320.1-20 6 320.1-21 6 320.1-22 6 320.2-1 6 320.2-2 6 320.3-1 6 320.4-1 6 320.5-1 6 320.6-1 6 320 .6-2 . 6 320.7-1 6 320.7-2 6 320.7-3 6 320.7-4 6 320.7-5 6 320.7-6 6 320.8-1 6 320.8-2 6 320.8-3 6 320.8-4 6 320.8-5 6 320.9-1 6 320.9-2 6 320.10-1 6 320.11-1 6 320.12-1 6 320.13-1 6 320 .14-1 6 320.15-1 6 410.01-1 4 410. 02-1 4 410.03-1 4 410.04-1 4 410 .05-1 4 410.06-1 4 410.07-1 11 410. 07-2 4 410.08-1 4 410.09-1 4 RNQ-10 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Page Amendment 410 .10-1 13 410.10-2 13 410.10-3 13 410.10-4 13 410.10-5 13 410.10-6 13 F410.10-1 11 410.11-1 /J. 410.11-2 4 410.11-3 11 410 .11-4 'l+ 410.12-1 /J. 410.13-1 4 410 .14-1 11 410.15-1 4 410.16-1 4 410.17-1 11 410.17-2 4 410.18-1 4 410 .19-1 4 410.19-2 11. 410 .19-3 4 410.20-1 11 410.21-1 4 410. 21-2 4 410.22-1 4 410.23-1 4 410.24-1 4 410.25-1 6 410.25-2 4 410.26-1 4 410.27-1 6 410.27-2 11 410.28-1 4
*410.29-1 11 410.30-l 4 410.31-1 4 410.32-1 6 410.32-2 6 410.32-3 6 410.33-1 11 410.33-2 11 410.33-3 11 410.34-1 6 410.35-1 6 410 .36-1 6 410.37-1 6 410 .38-1 11 RNQ-11 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) Page 410.39-1 410.40-1 NRC QUESTIQ}TS Amendment
. 12 Cl 410 .41-1 9 410.42-1 9 410.43-1 9 410.44-1 9 410.4 5-1 9 410 .46-1 9 420.1-1 5 420.1-2 7 420.1-3 7 420.1-4 7 420.1-5 11 420.1-6 11 420. l-6a 11 420.1-7 5 420.1-8 11 420.1-9 11 420.1-9a 11 420.1-10 5 420.1-11 5
420.1-12 9 420.1-13 9 420.1-14 5 420.1-15 5 420.1-16 5 420.1-17 5 420.1-18 5 420.1-19 12 420.1-20 5 420. 2-1 5 420.2-2 5 420.3-1 9 420.3-2 9 . 420.3-3 . 11 420.3-4 9 420~3-5 9 420.3-6 7 420.4-1 7 420.4-2 12 420.4-3 7 420.4-4 7 420.4-5 7 420.4-6 9 420.4-7 9 420.4-8 7 420.4-9 9 420.4-10 11 RNQ-12 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Page Amendment 420.4-lOa 11 420 .4-11 I 2. 420.4-12 7 420.4-13 12 420.4-14 9 420.4-15 9 420.4-16 9 420.4-17 11 420.5-1 9 420.5-2 9 420.5-3 9 420.5-4 9 420.5-5 9 420.5-6 12 420.s..:.1 9 420.5-7a 9 420.6-1 5 420.6-2 5 420.6-3 5 420.7-1 12
- 420. 7-2 420.7-3 420.7-4 420.7-5 420.7-6 420.7-7 12 11 11 11 11 5
420.7-8 12 420.7-9 12 420.7-10 9 420.7A-i 9 420.7A-l 11 420.7A-2 11 420.8-1 5 420.9-1 12 420.10-1 11 420.10-2 11 420.10-3 11 420.11-1 5 420.12-1 5 420.13-1 5 420.13-2 5 420.13-3 5 420.14-1 12 420.14-2 12 420.14-3 7 420 .14-4 11 420.14-5 7 420 .14-6 7 RNQ-:12a Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NlW OUESTIONS Page Amendment 420.15-1 7 420.16-1 12 420.17-1 5 420.18-1 11 F420.18-l 5 420 .19-1 5 420.19-2 5 420.20-1 5 420.21-1 5 420.21-2 5 420. 22-1 5 420.23-1 5 420. 24-1 5 420.25-1 5 420.26-1 5 420.27-1 5 - 420.28-1 5 420. 29-1 5 420.30-1 5 420.31-1 5 420.32-1 12 420.33-1 5 420.34-1 11 420.35-1 5-420.36-1 5
.420.36-2 12 420.36-3 7 420.36-4 6 420.36-5 11 420.37-1 5 420. 38-1 5 420.38-2 5 420.39-1 5 420.39-2 11 420.39-3 5 420.40-1 5 420.40-2 5 420.40-3 11 420.41-1 12 420.41-2 5 420.41-3 5 420.41-4 5 420.41-5 5 420.41-6 11 420.41-7 11 420.41-8 420.41-9 420.41-10 12 11 11 RNQ..:12b Amendment No. 13, (2/83)
SL2-FSAR l,IST OF EFFECTIVE PAGES (Cont'd) NRC OUESTIQ1'TS Page Amendment 420.41-11 11 420.41-12 11 420.41-13 11 420.41-14 11 420.41-15 11 420 .41-16 . 5 420.41-17 11 420.41-17a 11 420.41-18 7 420.41-19 5 420.41-20 5 420.41-21 11 420.41-21a 11 420.41-22 5 420.41-23 5 420.41-24 5
.420.41-25 11 420.41-26 .
420.41-27 " 5 420.41-28 5 F420.41-l 7 420.42-1 5 420.43-1 5 420.43-2 5 420.43-3 11 420.44-1 12
. 420.44-2. 12 420.44-3 5 F420.44-l 5 F420.44-2 5 420.45-1 12 420.45-2 9 420.46-1 11 420.47-1 5 420.48-1 12 420.49-1 5 420.50-1 5 420.51-1 5 420.52-1 5 420.53-1 5 420.54-1 5 420.54-2 5 420.54-3 6 420.54-4 5 420. 54-5 5 420.54-6 5 420.55-1 6 F420.55-1 5 RNQ-13 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PftGES (Cont'd) Page 420.56-1 420.56-2 tlRC QUESTION8 Amendment 12 6 420A.1-1 13 F420A. l-1 13 420A.2-l 13 F420A.2-l 13 F420A.2-2 13 F420A.2-3 13 420A.3-1 13 420A.4-1 13 420A.5-1 13 420A.6-1 13 420A. 7-1 13 420A.8-l 13 420A.9-1 13 420A.9-2 13 420A. 9-3 13 420A. 9-4 13 420A.9-5 13 420A.9-6 13 420A. 9-7
]3 420A.10-1 13 420A .10-2 13 420A.10-3 13 420A.ll-1 13 F420A.11-1 13 F420A.ll-2 13 420A.12-1 13 420A.13-1 13 420A.13-2 13 420A .13-3 13 420A.14-1 13 420A.15-1 13 420A.16-1 13 420A.17-1 13 420A.18-1 13 430.1-1 4 430.1-2 4 430.2-1 12 430.2-2 4 43Q.3-1 4 430.3-2 12 430. 4-1 7 430.4-2 11 430.4-3 12 430.5-1 4 430.6-1 4 430.7-1 12 RNQ-14 Amendment No. 13' (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Amendment 430.8-1 4 430. 9-1 4 430.10-1 4 430 .11-1 12 430.12-1 4 430.13-1 12 430.14-1 4 430.15-1 4 430.16-1 12 430 .16-2 4 430.17-1 4 430 .18-1 4 430.19-1 4 430.20-1 4 430.* 21-1 4 430.22-1 4 430. 23-1. 4 430.24-1 4 430.25-1 4 430.26-1 12
- 430.27-1 430.28-1 430-28-2 430.28-3 430.28-4 430.29-1 4
12 6 6 6 12
. 430.30-1 4 430.31-1 4 430.32-1 4 430.33-1 4 430.33-2 12 430.33-3 4 430.34-1 4 430.35-1 4 430.36-1 4 430.37-1 4 430. 38-1 4 430.39-1 11 430.40-1 4 430.41-1 4 430.42-1 12 430.43-1 11 430.44-1 l?
430.45~1 4 430.46-1 4 430.47-1 4 430.48-1 4 430.49-1 11 RNQ-15 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) Page 430.50-1 430.51-1
- NRC QUESTIONS Amendment 12 4
430.52-1 4 430.52-2 4 430.53-1 4 430.53-2 4 430.54-1 4 430.54-2 4 430.54-3 4 430.54-4 4 430.55-1 4 430.56-1 12 430.56-2 4 430.57-1 4 430.58-1 4 430.59-1 4 430.60-1 4 430.61-1 4 430. 62-1 4 430.63-1 4 430.64-1 11 430.65-1 4 430.66-1 7 430.66-2 7 430.67-1 6 430.67-2 6 430.68-1 7 430.69-1 7 430.70-1 6 430.70-2 6 430.71-1 7 430.71-2 7 430.72-1 12 430.73-1 11 430.73-2 11 430.73-3 12 430.74-1 7 430.74-2 6 430.75-1 12 430.75-2 7 430.76-1 9 430A. l-l 12 430A.2-l 11 430A.2-2 11 430A.2-3 11 430A.2-4 11 430A.3-l 7 RNQ-16 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd)
. NPC QUESTIONS Amendment 430A.4-1 11 430A.4-2 11 430A.4-3 11 430A.4-4 11 430A.4-5 11 430A.4-6 . 11 430A.4-7 11 430A.4-8 11 F430A.4-1 11 F430A.4-2 11 F430A.4-3 11 F430A.4-4 11 F430A.4-5 11 430A.5-1 7 430A.6-1 6 430A.6-2 6 F430A.6-1 11 430A. 7-1 11 430A. 7-2 11 430A.7-3 11-430A. 7-4 11 430A. 7-5 11 430A.7-6 11 430A. 7-7 11 F430A.7-1 11 430A.8-1 . 13 430A.9-l 13 430A.10-1 13 F430A.10-1 13 430A.11-1 13 F430A.11-1 13 430A.12-1 13 430A.13-1 13 440.1-1 6 440.1-2 6 440.1-3' 6 F440.l-1 6 F440.1-2 6 440. 2-1. 11 440.3-1 6 440.3-2 6 440.3(2)-1 6 440.4-1 6 440 .5-1 11 440.5-2 6 440.6-1 11 440.7-1 7 440.7-2 7 RNQ-16a Amendme.nt No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) 440.8-1 MRC OUESTIONS Amendment 6 440.8-2 11 440.8-3 6 440.9-1 6 440.9-2 6 440.9-3 6 440.9-4 6 440. 9-5 6 440.9-6 6 440.9-7 6 440.9-8 6 RNQ-16b Amendment No. 13, (2/83)
SL2-FSAR
- Page LIST OF EFFECTIVE PAGES (Cont'd)
NRC QUESTIONS Amendment 440.9-9 6
. 440.10-1 6 440.11-1 6 440 .12-1 6 440.12:-2 6 440.13-1 6.
440 .13-2 6 440.13-3 6 440.14-1 12 440.14-2 6 440.15-1 6 440 .16-1 6 440.17-1 6 440 .18-1 7 440.19-1 6 440.19-2 . 6 440.19-3 6 440.19-4 6 440.20-1 6 440.21-1 6 440.22-1 6 440.23-1 11 440.24-1 6 440.25-1 7 F440.25-1 6 F440.25-2 7 440.26-1 6 440.27-1 6 440.27-2 6 440.28-1 6 440.28-2 6 440.29-1 6 440.30-1 6 440.31-1 6 440.32-1 6 440.33-1 6 440.33-2 6 440.34-1 6 440.35-1 6 440 .36-1 6 440.37-1 6 440.38-1 6 440.39-1
- 11 440.40-1 6 440.40-2 6
- 440.41-1 440.41-2 440.42-1 RNQ-17 6
11 6 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) Page 440.42-2 440 .. 43-1 NRC QUESTIONS Amendment 11 6 440.44-1 6 440 .. 44-2 6 440.45-1 6 440 .. 46-1 6 440.47-1 6 440.48-1 6 440.48-2 11 440.48-3 6 F440.48-l 6' 440.49-1 6 440 .. 50-1 11 440.51-1 11 440.51-2 6 440.52-1 6 440.53-1 9 440.53-2 9 440.54-1 11 440.54-2 11 440. 55-l 6 440.56-1 6 440.57-1 6 440.58-1 6 440.59-1 6 440.60-1 6 440.,60-2 6 440.61-1 6 440.61-2 6 440.61-3 6 440 .. 61-4 . 11 440. 61-5 6 440.61-6 6 440.61-7 6 440.61-8 6 440.61-9 6 440.61-10 6 440.62-1 6 440 .. 63-1 6 440.64-1 6 440 .. 65-1 6 440 .. 66-1 6 440. 6 7-1 7 440.67-2 7 440.67-3 7 440.67-4 7 440.67-5 7 RNQ-18 Amendment No. 12, (8/82)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTION$ Page Amendment F440.67-l 7 F440,.67-2 7 440.68-1 6 440.69-1 7 440. 70-1 6 440.70-2 6 440. 71-1 12 440. 72-1 6 440.73-1 6 440 .. 74-1 6 440.74-2 6 440. 75-1 7 440.76-1 6 440 .. 77-1 6 440.78-1 6 440.79-1 6 440.80(a)-1 6 440.80(a)-2 6 440,.80(b)-1 6 440.80(c)-1 6 440.80(d)-1 6 440 .. 80(e)-1 6 440 .. 80(£)-1 6 440 .. 80(g)-1 6 440,.80(h)-1 6 440.80(i)-l 6 440,,80(j)-1 6 F440 .. 80(j)-1 6 440.80(k)-1 6 440.80(k)-2 6 440 .. 80(k)-.3 6 F440 .. 80(k)-l 6 440 .. 80(1)-1 11 440.80(1)-2 6 440.80(1)-3 6 440.80(m)-l 6 440.80(m)-2 6 440.81-1 6 440.81-2 6 440. 81-3 7 440 .. 81-4 7 440 .. 81-5 7 440.81-6 7 440.81-7 7 440.81-8 11 440 .. 81-9 11 F440 .. 81(f)-l 7 F440. 81( £) -'2 7 RNQ-19 Amendment No. 12, (8/82)
SL2,.,FSAR.. .. LIST OF EFFECTIVE
.*. . . PAGER. .'
(Cont'd) Page 440.82-1 440.82-2 NFC*OUF.STIONS Amendment 6 6 440.83-1 13 440.84-1 13 440.85-1 13 440.a5:...2 13 F440.85-1 13 440 .86-1 13 440.87-1 13 440.88-1 13 440.89-1 13 440.90-1 13 440.91-1 13 440.92-2 13 F440.92-1 13 440.93-1 13 440.94-1 13 440.95-1 13 F440.95-1 13 440.96-1 13 440.97-1 13 440.98-1 440.99-1 440.100-1 440.101-1 440.101-2 440.102-1 440.102-2 13 13 13 13 13 13 13 440.102-3 13 440.103-1 13 440 .104-1 13 440.105-1 13 440.106-1 13 440.107-1 13 440.108-1 13 440.108-2 13 440 .108-3 13 440.108-4 13 440 .108-5 13 440.108-6 13 440.108-7 13 440.109-1 13 440 .109-2 13 F440.109-1 13 440 .110-1 13 440.111-1 13 440 .112-1 13 440.112-2 13 440.113-1 13 RNQ-20 Amendment No. 13, (2/83)
SL2-FSAR
- Page LIST OF EFFECTIVE PAGES (Cont'd)
NRC OUESTIONS Amendment 440.114-1 13 440.115-1 13 440.116-1 13
*440 .117-1 13 440.118-1 13 440.119-1 13 440.120-1 13 440.121-1 13 440.122-1 1.3 440 .123-1 11 440.124-1 1.3 440.125-1 1.3 440.126-1 13 440 .127-1. 13 440.128-1 13 440.128-2 13 440.128-3 13 440.129-1 13 440.130-1 13 440 .131-1 13 440.132-1 13 440.133-1 13 440.134-1 13 440 .135-1 13 440.136-1 13 440.137-1 13 440.138-1 13 440.139-1 13 440.140-1 13 440.141-1 13 440.142-1 13 440.143-1 13 440.143-2 13 440.143-3 13 440.143-4 13 440.144-1 13 440.145-1 13 440 .146-1 13 440.147-1 13 440.148-1
- 13 440.149-1 13 440.149-2 13 440.149-3 13 440.149-4 13 440.149-5 13 440.149-6 13 440.149-7 13 440.149-8 13 RNQ-21 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC OUFSTIONS Page Amendment 440.149-9 13 440 .149-10 13 li.40.149-11 13 440.149-12 13 440.149-13 13 440 .149-14 13 440.149-15 13 . F440 .149A-l 13 F440.149A-2 13 F440 .149A-3 13 F440.149A-4 13 F440 .149A-5 13 F440.149A-6 13 F440 .149A-7 13 F440.149A-8 13 F440 .149A-9 13 F440.149A-10 13 F440 .149A-ll 13 F440.149A-12 13 F440.149B-l 13 F440.149B-2 13 F440.149B-3 13 F440.149B-4 13 F440 .149B-5 13 F440.149B-6 13 F440.149B-7 13 F440. 149B-8
- 13 F440.149D-l 13 F440.149D-2 13 F440 .149D-3 13 F440.149D-4 13 440.150-1 13 440.150-2 13 440 .150-3 13 440.150-4 13 440 .150-5 13 440.150-6 13 440.150-7 13 440.150-8 13 F440.150-l 13 F440.150-2 13 F440.150-3 13 F440.150-4 13 451.1-1 6 451.1-2 6 451.1-3 6 451.1-4 6 451.1-5 6 RNQ-22 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUFSTIONS Page Amendment 451.1-6 6 451.1-7 6 451.1-8 6 451. 2-1 6 451. 2-2 6 451. 2-3 6 451. 2-4 6 451. 2-5 6 . 451.2-6 6 451.2-7 6 451.3-1 6 451.3-2 6 451.4-1 6 451.5-1 6 451.6-1 6 451. 7-1 6 451. 7-2 6 451. 7-3 6 451.8-1 6 451.8-2 6 460.1-1 6 460.2-1 12 460.2-2 8 460.2-3 6 F460.2-l (\ 460.3-1 6 460.4-1 12 460.5-1 6 460.6-1 11 460. 7-1 6 460.8-1 6 460.9-1 12 460.10-1 6 460.11-1 6 460.12-1
- 6 460.13-l 6 460.14-1 6 460.15-1 6 460.15-2 6 460.16-1 6
.460.17-1 6 460.18-1 6 460.19-1 6 460.20-1 8 460.21-1 6 460.22-1 6 460.23-1 6 4 71.1-1 4 RNQ-23 Amendment No! 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd) Page 471.2-1 471. 3-1 MRC QUESTIONS Amendment 4 4 471.4-1 6 4 71.5-1 6 471. 6-1 6 4 71. 7-1 4 471.8-1 6 471.9-1 4 471.10-1 4 4 71.11-1 6 471.12-1 7 4 71.13-1 6 471.14-1 6 4 71.15-1 12 471.16-1 12 4 71.17-1 7 471.18-1 7 4 71.19-1 6 [~71.20-1 6 471A.l-1 6 471A.2-1 6 480.i-1 6 480.2-1 6 480.3-1 6 480.3-2 6 480.3-3 6 480.4-1 6 480.5-1 6 480.6-1 6 480. 7-1 6 480.8-1 6 480.9-1 6 490.1-1 3 490.1-2 7 490.lA-1 7 490. lA-2 7 lf90.1A-3 7 490. lA-4 7 490.lA-5 7 490.lA-6 11 490.lA-7 7 490. lA-8 7 490. lA-9 7 490.lA-10 10 490. lA-11 11 490. lA-12 7 490. lA-13 7 490. lA-14 11 RNQ-24 Amendment No. 13, (2/83)
SL2~FSAR LIST OF EFFECTIVE PAGES (Cont'd) N~C QUESTIONS Page Amendment 490. lA-15 11 490. lA-16 7 490. lA-17 7 490. lA-18 7 490. lA-19 11 490.1A-19a 11 490. lA-20 7 490.lA-21 .11 . 490. lA-22 11 490. lA-23 11 490.lA-24 7 490.lA-25 7 490. lA-26 7 490.lA-27 12 490.lA-28 7 490. lA-29 7 490. lA-30 7 490.lA-31 lf'I 490. lA-32 7 490.lA-33 9
- 490. lA-34 490. lA-35 490. lA-36 490. lA-37 490. lA-38 490. lA-39 11 11 7
7 7 7 490. lA-40 11 490.lA-41 7* 490. lA-42 7 490. lA-43 11 490. lA-44 7 490. lA-45 7 490.lA-46 7 490. lA-47 7 490. lA-48 7 490. lA-49 7 490.lA-50 7 490.lA-51 7 490.lA-52 7 490. lA-53 13 490. lA-54 10 490. lA-55 7 490.lA-56 13
- 490 .1A-56a 13 490. lA-57 13 490.lA-58 13 490. lA-*59 13 490.l.A-60 11 RNQ-25 Amendment No. 13, (2/83)
SL2.,.FSAR. LIST OF EFFECTIVE PAGES .(Cont'~) Page 490. lA-61 490. l.A-62 NRC QUESTIONS Amendment 7 7 490.lA-63 7 490. lA-64 7 490. lA-65 7 490. lA-66 7 490. lA-67 7 490. lA-68 10 490. lA-69 12 490.lA-70 10 490.lA-71 7 490. lA-72 12 490. lA-73 7 490. lA-74 7 490.lA-75 7 490. lA-76 7 490.lA-77 7 490.lA-78 7 490. lA-79 7 490. lA-80 10 490. lA-81 7 490. lA-82 7 490.lA-83 12 490. lA-84 7 490. lA-85 7 490.lA-86 12 490. lA-87 11 490.lA-88 7 490.lA-89 7 490. lA-90 7 490. lA-91 7 490.lA-92 12 l~90.1A-93 10 490. lA-94 12 490.lA-95 12 490. lA-96 10 490.lA-97 13 490.lA-98 10 490. lA-99 10 490.* lA-100 10 490. lA-101 7 490. lA-102 7 490.lA-103 11 490.lA-104 7 490. lA-105 7 490.lA-106 7 490. lA-107 7 490. lA-108 7 RNQ-26 Amendment No. 13, (2/83)
SL2*-FSAR LIST OF EFFECTIVE PAGES (Cont'd) NRC QUESTIONS Page Amendment 490. lA-109 12 490. lA-110 12
- 490. lA-lll 10 490. lA-112 7 490. lA-113 7 490. lA-114 7 490. lA-115 7 490. lA-116. 10 490. lA-117 7 490.lA-118 7 490.lB-l 7 490.lB-2 7 490.lB-3 7 490.lB-4 7 490.lB-~ 7 F490.1B-1 7 F490.1B-2 7 491.1-1 3 491". ?.-1 11 491.3-1 11
- 491.4-1 492.1-1 492.1-2 492.2-1 492.3-1 492.4-1 3
7 7 7 7 10 492.4-2 7 492.5-1 7 492.6-1 7 492.7-1 11 492.7-2 7 492.8""".l 7 492.9-1 7 492.9-2 10 492.10-1 7 492.11-1 7 492.11-2 7 492.11-3 7 492.11-4 7 492.11-5 . 7 F492. ll-l 7 F492 .11-2 7 492.12-1 7 492.13-1 7 492.14-1 7 492.14-2 7 492.14-3 7 492.15-1 11 RNQ-27 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cqnt'd) Page 492.15-2 L~92 .15-3 NRC OUESTIONS
- Amendment 11 12 492.16-1 7 492 .17-1 10 492.18-1 10 F492.18-1 10 F492.18-2 10 492.19-1 10 492.20-1 10 640.1-1 6 640.2-1 6 640.3-1 6 640.4-1 6 640.4-2 6 640.4-3 6 640.4-4 6 640.4-5 6 640 .5-1 6 640.6-1 6 640. 7-1 6 640.8-1 6
640.9-1 6 640.10-1 6 640.10-2 6 640.10..:.3 6 640.10-4 6 640.10-5 6 640.11-1 6 640.11-2 6 640.11-3 6 640.11-4 11 640.11-5 6 640.11-6 6 640.12-1 6 640.12-2 6 640.12-3 6 640.12-4 6 640.12-5 6 640.12-6 6 640.12-7 6 640.12-8 6 640 .12-9 6 640.12-10 6 640.12-11 6 640.13-1 6 640 .14-1 6 640.14-2 RNQ-28 6 Amendment No. 13, (2/83)
SL2-FSAR LIST OF EFFECTIVE PAGES (Cont'd)
- 640.15-1 640.16-1 NRC QUESTIONS Amendment 6
6 640.17-1 6 640.18-1 6 640.18-2 6
- RNQ-29 Amendment No. 13, (2/83)
... Question No. l SL2-FSAR Table 1.8-1 should indicate the extent to which the applicant (1.8) intends to comply with all applicable NRC regulatory guides and should indicate any proposed exceptions to the regulatory posi-tion.
Response
As indicated in FSAR Section 1.8, Table 1.8-1 provides a listing of all applicable regulatory guides with cross-references for those regulatory guide subjects discussed in particular subsections. Subject to the implementation dates therein, regulatory guides issued on or before May 2, 1977 (Construction Permit date for St Lucie Unit 2) are considered to contain the recommendations that are applicable to the design of this plant. No FSAR revisions are required. Ql-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
2 (1.9.l) Section 1.9.l and the remainder of the FSAR should address the requirements given in NUREG-0737, TMI-Related Requirements for New Operating Licenses.
Response
See revised FSAR Subsection 1.9.l . Q2-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
3 (2.1.2.1) Section 2.1.2.l states "FP&L controls the use of all land and water inside the site boundary (property) lines." Part of the exclusion area appears to be outside the property lines and extends into a body of water (Figure 2,1-2). Address the infor-mation requested in Section 2.1.2.l of Regulatory Guide 1.70 Revision 3.
Response
Arrangements have been made by FP&L to control traffic on the highway and waterway, in case of emergency, to protect the public health and safety. FSAR Subsection 2.1.2.l provides a reference to applicable sections of the St Lucie Plant Emergency Plan where emergency support agreements with various agencies are provided. No FSAR revisions are required
- Q3-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
4 Section 2.4.6 of the FSAR states "The areas of the U.S. that are most susceptible to tsunamis are bordered by the Pacific Ocean or Gulf of Mexico. The site is on the Atlantic Coast and there-fore tsunamis are not a phenomenon that ~ould affect the St Lucie Site". Although the Atlantic Coast may not be as suscep-tible to tsunamis flooding as other locations, this is not adequate justification for not evaluating maximum tsunami flood-ing at St Lucie. Provide the information requested in Regula-tory Guide 1.70 *Revision 3, Sections 2.4.6 and 2.4,6,l through 2.4.6.7.
Response
See revised Subsection 2,4,6 of the FSAR for the justification of not needing to evaluate potential tsunami flooding . Q4-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
5 (2.5.4.5) Discuss measures to monitor foundation rebound and heave as specified in Regulatory Guide 1,70 Revision 3.
Response
See revised Subsection 2,5,4.5 of the FSAR
- QS-1 Amendment No, 1, (4/81)
SL2-ER-OL
- Question No.
- 6. Explain your assumption that a delay in St Lucie 2 will preci-pitate a delay in the Martin Co,al Unit 3.
Response
The assumption was: if St Lucie Unit 2 was delayed for reasons of financial, environmental, regulatory, or political issues, then it would be reasonable to assume that a similar delay could occur for Martin Coal Unit 3. No ER-OL revisions are required .
- Q6-1 Amendment No. 1, (4/81)
SL2-ER-OL
- Question No.
- 7. What percentage of St Lucie Unit 2 is currently completed (specifically, what portion of the $925 million estimated capital cost has been spent)?
Response
The January 1981 estimated capital cost for St Lucie Unit 2 is
$1.1 billion. Of this amount, $661,921,000 or 60.2 percent of the estimated cost has been spent to date.
No ER-OL revisions are required . Q7-1 Amendment No. 1, (4/81)
SL2-ER-OL
- Question No.
- 8. Provide assumptions and trace through the calculations per-;- formed in your conclusions in Subsection 8.1.2 that*, "The operation of St Lucie Unit 2 will result in an annual savings of an estimated 8.5 million barrels of crude oil per year. This annual saving translates into a dollar saving of $137 million per year (1978 delivered price)".
Response
The methodology used* to calculate the fuel oil required to replace other types of generation is as follows: bbls/yr= (Unit MWH rating/hr)x(24hrs/day)x(days/yr)x(capacity factor) Fossil eq. MWH/bbl Assuming: St Lucie Unit 2 rating: 850 MW gross Capacity factor: .72 Fossil eq. MWH/bbl: .63 bbls/yr= (850 MWH/hr)x(24 hrs/day)x(365 days/yr)x(.72)
.63 bbls/yr= 8.5 million This annual saving today translates into a dollar saving of $306 million per year (January 1981 fuel price 0£$36.00/bbl).
See amended ER-OL Subsection 8.1.2 .
- QB-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
9 (3.4.1.1) Per Regulatory Guide 1.70 Revision 3, describe the procedures required to bring the reactor to a cold shutdown for the flood conditions identified in Section 2.4.14.
Response
As indicated in Subsection 2,4.14, flood protection requirements are part of the plant Technical Specifications. The flood protection technical specification will include reference plant
- procedures utilized during a hurricane watch or a hurricane warning.
No FSAR revisions are required . Q9-l Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
10 (3.5.1.1) Per Regulatory Guide 1.70, identify in FSAR Figure 3.5-3 the missiles to be protected against for all equipment.
Response
FSAR Figure 3.5-3 has only been provided in order to show, in an elevation view, the turbine low trajectory missiles. Section 3.5 identifies those missiles to be protected against. No FSAR reviiion~ are required
- Ql0-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
11 (3.5.3) Per Regulatory Guide 1.70, discuss the potential for generating secondary missiles by spalling and scabbing of concrete barriers.
Response
See revised Subsection 3.5.3,1,1 .
- Qll-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
12 (3.8.5.l) Per Regulatory Guide 1.70, discuss the effect of waterproofing membranes on the capability of the foundation to transfer shears.
Response
See revised FSAR Subsection 3.8.5.5 *
- Q12-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
13 (3.9.4.2) Per Regulatory Guide 1.70, provide a discussion of NRC general design criteria, regulatory guides, and positions that are applied in the design, fabrication, construction, and operation of the CEDM.
Response
The following NRC design criteria, regulatory guides and posi-tions were applied to the design, fabrication, construction and operation of the CEDM.
- 1) General Criteria Per 10CFR50 Appendix A; GDC-1, 2, 4, 14, 15, 23, 26, 30 and 31.
- 2) Regulatory Guides 1.29, 1.31, 1.37, 1.44, 1.48, 1.71, and 1.85.
3.) NRC Position: MTEB 5-1. No FSAR revisions are required *
- Q13-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
14 (3.11.l) Per Regulatory Guide 1.70, provide in FSAR Table 3.11-1 chemical and vibration (non-seismic) definitions. Res pones FSAR Subsection 3.11,S,l provides a description of the chemical environment to which the components of the engineered safety features system and associated safety related components inside the containment are exposed to. As indicated in Subsections 3.9.3.2.la and 3.9.3.2~2, active pumps and valves are designed to take into consideration vibration resulting during normal operation and faulted conditions. Active valves and pumps are subjected to factory, include vibration tests. After these pumps and valves are installed in the plant, they undergo the cold hydro tests and hot functional tests. Also, Subsection 3,9,2,l describes the preoperational test program where major equipment are observed for excessive vibration. No FSAR revisions are required *
- Ql4-l Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
15 (4.3.2.4) Per Regulatory Guide 1.70, provide in your discussion of control requirements the effects of pH, permitted rod insertions at power and error allowances, and the required and expected shut-down margin as a function of time in cycle (including uncertain-ties in shutdown margins and experimental confirmation from operating reactors).
Response
Normal changes of pH and coolant chemistry which would be expec-ted to occur during normal operation will have no measurable effect on core reactivity. In the unlikely event that gross changes occur in core chemistry due to an abnormal operating mode, it is possible for material from other parts of the pri-mary system to be deposited on the fuel. The small reactivity changes associated with an event of this type can easily b~ accommodated either through changes in the dissolved boron concentration or through changes in CEA position. Figure 4.3-40 shows a typical power dependent CEA insertion limit curve (PDIL) which specifies the allowable CEA insertion for any given power level. Figure 4.3-40 was assumed in deter-mining initial conditions for the transient analyses presented in Chapter 15. A final PDIL, which has been verified to be consistent with all aspects of the safety analysis, will be incorporated in the Technical Specifications. The uncertainty allowances associated with each component of the shutdown margin calculation are discussed in Subsection 4.3.3 which ai'so includes experimental confirmation of these results from operating reactors. Table 4.J-6 shows the calculational uncertainty involved in the calculation of the reactivity defect between hot full power and hot zero power at EOC4. The uncer-tainty allowance on moderator defect is +10 percent; the allow-ance on fuel temperature defect is +15 percent; the allowance on rod worth is 2:_10 percent.
- Shutdown margin requirements are largest at end-of-cycle because the MTC is the most negative at that time. Furthermore, avail-able rod worth is less during an equilibrium cycle than for any other cycle. Consequently the data which is presented for end-of-cycle four (cycle four is the first equilibrium cycle) is the most limiting possible for a nominal 3-batch annual refuel-ing fuel management scheme. Table 4.3-7 illustrates that even for the EOC4 case the available rod worth (after assuming that the most reactive CEA is stuck out of the core) exceeds the required reactivity worth by 1.21 percent ~P*
- No FSAR revisions are required.
QlS-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
16 (4.4.4.2)
~er regulatory guide 1.70, discuss the effect of partial or total isolation of a loop on the core hydraulics evaluation.
Response
Partial isolation can be experienced when electrical power to a main reactor coolant pump is lost (total isolation of a cooling loop is not possible). The Technical Specifications will indicate that reactor core power generation will only be permitted with four pump opera-tion. The "low flow trip" will shutdown the reactor if flow to the vessel is reduced due to a loss of one of the coolant pumps during power operation (see Section 7.2). However, part-loop operation is permitted during plant start-up and shut-down. Hydraulic loads on various reactor internal components were computed for all part-loop configurations to assure the integri-ty of the components. These part-loop loads are included in the analysis of Subsection 4.4.2.6.3. Subsection 4.4.4.2.5 is added to the FSAR to include the above response. I Ql6-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
17 (4.5.2.3) Per Regulatory Guide 1.70 Revision 3, discuss conformance to the requirements of the ASME B&PV code.
Response
The nondestructive examination procedures used for the exa~ina tion of tubular products and fittings used as elements of the reactor internals are in compliance with the Class 1 require-ments of Section I I I of the ASME Code, 1971 issue, including 1972 Summer Addenda. * ' FSAR Subsection 4,5,:2.,3 is amended to reflect the above response .
- Ql7-l Amendment No, 1, (4/81)
SL2-FSAR
- Question No.
18 (3.5.1.2) Per Regulatory Guide 1.70 Revision 3, discuss missiles due to gravitational effects. Include a list of all such missiles.
Response
In.accordance with the requirements of SRP Section 3.5 . .1.2, "Internally Generated Missiles (Inside Containment)," credible primary missiles are identified in Table 3.5-4. This table also identifies the structure, or the shield wall, which.contains the
.' ... '~ * ' pot~ntial missiles within the c:onfined area to prevent damage to the safety related equipment.
The following design critetia, procedures and controls* have been implemented to avoid damage to safety related equipment from potential gravity missiles inside the contairiment.
- 1) Structural steel inside the containment is designed for. the SSE,.
- 2) All Class lE electrical equipment and associated raceways (cable trays, conduits and boxes) located inside the con-tainment is seismically supported. Maximum use of existing se~smic Category I steel is utilized to support nonsafety raceway systems*. Non-Class lE electrical equipment and raceways are not seismically supported when an analysis demonstrates that the falling of this equipment will not endanger any Class lE equipment.
- 3) All H&V ducts inside containment are seismically supported to prevent gravity missiles. ,,
- 4) Non-seismically supported piping has been routed away from safety related equipment.
No FSAR revisions are required *
- Q18-l Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
19 Per Regulatory Guide 1.70 provide the following for ASME Code Class 1 components supports:
- a. A summary description of mathematical or test models used,
- b. Methods of calculation or test, including simplifying*as-sumptions, identification of method of system and component analysis used, and demonstration of their. compatibility in the case of components and supports designed to faulted limits.
Response
The analysis of CEDMs is summarized in Subsection 3.9.4 and models used are summarized in Subsection 3.9.1. Information on reactor internals is given in Subsection 3.9.5 of the FSAR. The models used in the analysis of the reactor internals are de-scribed in Subsections 3.7.3.14 and 3.9.2 of the FSAR. I5 The analysis of major reactor coolant system components and supports is summarized in Subsections 3.7.3.1.2, 3.9.3.1.2, .and 3.9.1. FSAR Subsection 3.9.3.1.2 is submitted with this amendment *
- 19-1 Amendment No. 5, (8/81)
SL2-FSAR
- Question No.
20 (3.6) FSAR Section 3.6 cannot be reviewed until the appendices "To be supplied in a later amendment" are supplied. Provide the ap-pendices or supply the approximate date of the amendment to be provided.
Response
Appendices in Section 3.6 of the FSAR will be available as follows: Appendix Amendment to be provided by 3.6A May 29, 1981 3.6B May 29, 1981 3.6C May 29, 1981 3.6D May 29, 1981 3.6E May 29, 1981 3.6F May 29, 1981 Q20-l Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
21 Per the requirements of Regulatory Guide 1.70 address or (7.1.2) reference compliance with Regulatory Guides 1.47, 1.53, 1.62, 1.63, 1.68, l.?3, 1.75, 1.80, 1.89, 1.97, 1.100, 1.118 in Subsection 7.1.2. *
Response
FSAR Subsection 7.1.2~2 provides a discussion for Regulatory Guides 1.47, 1.53, 1.62, 1.63, 1.68, 1.73, 1.75, 1.80, 1.89, l.97, 1.100, _and 1.11~ (refer to pages 7.1-7 through 7.1-11). No FSAR revisions are required *
- Q21-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
22 (7.2.2) Per the requirements of Regulatory Guide 1.70, provide a discussion of a spurious control rod withdrawal transient in Subsection 7.2.2.
Response
.The effect of the Reactor Protective System on plant transients is addressed in the following Subsections:
7.2.2.1.1 Moderate Frequency Events and Infrequent Events 7.2.2.1.2 Limiting Faults 1.2.2.5 Effects of Other Associated Functions Subsections 7.2.2.1.1 .and 7.2.2.1.2 refer to Chapter 15 (FSAR Subsections 15.0.l and 15.4.2) for analysis of transients, one j5 of which is the Sequential CEA Withdrawal event. No FSAR revisions are required *.
- 22-1 Amendment No. 5, (8/81)
SL2-FSAR
- Question No.
23 The information listed as 11 Later 11 for Table 7.* 2-5 and Figure (7 .2) 7.2-7 will be required to commence the review of the applicable sections.
Response
The missing information in Table 7.2-5 and Figure 7.2-7 is being supplied with this amendment
- Q23-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
24 Per thP. requirements of Regulatory Guide 1.70, provide the (7 .5) accuracy of all the instruments listed in Table *7.5-1 Respons~
*rable 7 .5-1 will be amen.ded to include the instrument loop accu-racies by June 30, 1981 *
- Q24-1 Amendment No. 1, (4/81)
SL2-FSAR Question No *
- 25 (8.1)
Response
Per the requirements of Regulatory Guide 1.70, provide or reference compliance with Regulatory Guide 1.131 and IEEE Std. 387 in Section 8.1 Refer to revised FSAR Subsections 3.1.2 and 8.3.1.2.3 for a I 5 discussion on compliance to Regulatory Guide 1.131 and IEEE 387 *
- 25-1 Amendment No. 5, (8/81)
SL2-FSAR .* Question No. 26 Figures 10.1-1, 10.1-2 and 10.1-3 are not legible (the P&I draw-oo .1 > ing print reduction is too small). Provide legible figures.
Response
Figures 10.1-1, 10.1-2 and 10.1-3 have been redrawn and resub-mitted in Amendment 0 (refer to Figures 10.1-la and lb, 10.l-2a and 2b, and 10.1-Ja and 3b). No *FSAR revisions are required *
- Q26-1 Amendment No. 1, (4/81)
- 1. SL2-FSAR
' Question No *
- 27.
(10.2.3)
Response
Provide the material chemical analysis of the turbine disk and rotor.
'!he turbine disk and rotor are heat treated nickel-chromium-m:olybdenum-vanadium alloy steel procured to turbine manufacturer's specifications that define the manufacturing method, heat treating process, and the test and inspection methods. As indicated in Subsection 3.5.1.3.3.2, inspection and tests conducted at the forgi'ng manufacturer's plant include a ladle analysis of each ~eat of steel for chemical composition to determine that the composition is within the limits defined by the specification. These specifications exceed the requirements.
, , of corresponding ASTM specifications.
No FSAR revisions are required
- Q27-l Amendment No. 8, (1/82)
SL2-FSAR Question No *
- 28 (10.2.3)
;Response Identify the specific fracture mechanics analytical metho.ds used and attendant key assumptions. i
*)
Details of the fracture mechanics analysis techniques applied to 8 the St Lucie Unit 2 turbine are given in the reference cited below. No FSAR revisions are required
- G.T. Campbell,. "Techniques for Fracture Mechanics Analysis of Nuclear Turbine 8 Discs," Westinghouse Steam Turbine Division, September 1974 *
- Q28-1 Amendment No. 8, (1/82)
SL2-FSAR
- Question No
- 29 Give the high temperature stress-rupture material properties of (10.2.3) the high pressure turbine rotor. Describe the method used to obtain the properties.
Response
Probability and fatigue crack growth rate data and stress rupture data of the high pressure rotor are given in Reference 4 I8 of Section 3.5. No FSAR revisions are required *
- Q29-1 Amendment No. 8, (1/8~)
SL2-FSAR Question No
- 30 Describe or reference the specific criteria used to insure (10.2.3) protection against brittle failure of the low-pressure turbine disks. Include detailed information on ductile-bri~tle trans-ition temperature (NDT or FATT) and minimum operating temper-ature.
Response
As indicated in Subsection 3.S.1.3.3.2, the tensile, Charpy V Notch impact and FATT properties are determined from specimens removed from the discs at specific locations. The test method used for determining these specific mechanical properties are defined by ASTM A-370. No FSAR revisions are required *
- Q30-l Amendment No. 8, (1/82)
SL2-FSAR Question No., 31 Provide the following design information for low-pressure disks (10.,2.3) and high-pressure rotors:
- 1) The tangential stress due to centrifugal loads, interference fit, and thermal gradients at the bore region at normal speed and design overspeed, and 2) The maximum tangential.and radial stresses and their location.
Response
'11te turbine is designed to withstand normal conditions, anticipated transients or accidents resulting in turbine trip.
The turbine disk design is based on the turbine disk design criteria in Standard Review Plan 10.2.3, II.5 (11/75) and our extent o*f compliance is as follows: a) '11te calculated overspeed upon a loss of load is less than 110% of rated speed. Adding 5% to this speed as required in SRP 10.2.3 gives approximately 115% of rated speed. The turbine rotor is designed and tested to 120% of rated speed. b) '11te combined stresses of low pressure disks at design overspeed due to centrifugal.forces, interference fit and thermal gradients do not exceed 0.75 of the minimum 3 specified yield strength of the materials. Since the high
- c) pressure rotor is a solid forging and not a disk design the acceptance criteria in Section II.5 of SRP 10.2.3 does not apply.
'11te turbine shaft bearings are designed to withstand normal operating loads anticipated transients or accidents resulting in turbine trip.
d) '11te rotors are designed so that the response levels at the natural critical frequency of the turbine shaft assemblies are controlled between 0 and 20% overspeed, so as to cause no distress to the unit during operation. e) nie bore of the high-pressure rotors can be inspected in service with the rotors removed from the cylinder. The rims of the low-pressure discs can also be inspected. '11te bores and keyways of the low pressure rotors can be inspected by means of Ultrasonic techniques. Reference
-See FSAR Subsection 10.2.3.1 .
- 31-1 Amendment No *. 3, (6/81)
SL2-FSAR
- Question No.
32 Describe the ins~rvice inspection program for main steam stop (10.2.5) and control valves and reheat stop and intercept valves as applicable.
Response
The requirements of the ASME Code Section XI do not pertain to these valves inspection and testing of the valves is performed in accordance with the manufacturer's reconnnendations and FP&L maintenance and inspection philosophy. No FSAR revisions are required *
- Q3_2-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
33 (l0.3.4)
Response
Discuss or reference* provisions made to allow for inservice inspection of main steam lines. Refer to revised FSAR Subsection 10.3.4 *
- Q33-l Amendment No. 1, (4/81)
SL2-FSAR
- Q~estion No.
34 What conductivity limits of the cooling water are permitted and (l0.4.1.3) how long may the condenser operate in a degraded, contaminated
' condition without affecting the condensate/feedwater quality?
Response
C-E's normal specification for cation conductivity within the condensate and feedwater is ,0.5 µmhos/cm. System operation within a range between o.s to 1.5 µmhos/cm is acceptable; how-ever, it is *considered abnormal. During periods of abnormal chemistry, the steam generator blowdown rate must be increased as well as the sampling frequencies along the secondary system. A fault condition exists if abnormal conditions are exceeded and a plant shutdown should be commenced if these conditions persist longer than four hours. Table 10.3-3 and Subsection 10.4.1 are amended to reflect the
*above *
- Q34-1 ' Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
35 (10.4.7) Identify and describe all manual operating transients that could cause the water level in the steam generator to drop below the sparger or uncover the feedwater nozzles (J-tubes).
Response
The following operating transients can cause the steam generator sparger to be uncovered. (1) Reactor/Turbine trip from high power levels. (2) Loss of one main feedwater pump or a condensate pump above 50 percent power. (3) Operator error during manual mode. (4) Equipment malfunction Transients involving loss of heat removal by the secondary system are described in Section 15.2. No FSAR revisions are required *
- Q35-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
36 (10~4.7). Provide main feedwater p1p1ng isometric drawings (from the steam generators to the restraint on the upstream side of the isola-tion valve outside the containment) for both steam generators *. Include pipe sectional lengths between bends and horizontal and vertical runs. Show *the auxiliary feedwater inlet location.
Response
A main feedwater p1p1ng isometric drawing has been added to Section 10.4 as Figure 10.4-13. See_ revised Subsection lQ.4.9.3
*of the FSAR. ., .
- Q36-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
37 .Provide a drawing of the steam generator feedwater spa~ger with* (10.4.7) J tubes. Give pertinent dimensional data and show details from the sparger inlet including penetrations through the steam gen-erator wall.
Response
Figures 5.4-16 and 17 h*ave been added to Section 5.4 (See amended Subsection 5.4.2.1.2)
- Q37-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
38 (10.4.9.2) State the maximum length of time the plant can stand without normal feedwater and the minimum auxiliary feedwater flow rat~ required after this time period.
Response
The maximum time the plant may stand without normal feedwater depends on the plant heat load. Following a trip from a low power history, the boil dry time would be very long. It is unlikely however that the operator would wait in excess of 30 minutes prior to initiating auxiliary feedwater flow. For . transients from higher power levels, see Chapter 15 analyses. No FSAR_revisions are require9 .
- Q38-l Amendment No. 1, (4/81)
- SL2-FSAR
- Question No.
39 In Table 11.2-8 indicate the basis for RCS fraction for sample (11.2.2) and laboratory drains.
Response
The basis for the sample and laboratory sink drains is ANSI-Nl99, November 1975. Table r1.2-8 has been revised to ieflect the above
- Q39-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
40 Provide P&I drawings that show system interconnections (11.4.2) and seismic and quality group interface.
Response
As indicated in FSAR Subsection 11.4.2, Figure 11.4-1 is a pro-cess flow diagram for a portable Solid Waste Management System that would be utilized by an outside contractor selected to per-form this work~ As shown on Figures 11.2-7* and 11.2-6, the per-manent plant piping from the was*te concentrator and spent resin tank, respectively, are .classified as nonsafety and nonseismic Category I lines. No FSAR revisions are required . Q40-l Amendment No. 1, (4/81)
SL2-FSAR
- Question No,.
41 Reference or describe the proposed means of assigning shift (13.1.2.3) responsibilities for implementing the radiation protection program on a round-the-clo~k basis.
Response
St Lucie Uni.t 1 presently schedules qualified radiation protection men (RPM) on shift 24 hours per day, seven days per weeke This schedule :will be expanded to include St *. Lucie Unit 2 when necessary.
. -- ' ~
See revised Subsection 13.1.2.3 of Amendment 1
- Q41-l Amendment No. 1, (4/8)
SL2-FSAR
- Question No.
42 The FSAR states "The startup testing program is developed (14 ~ 2 .1) using the recommendations of Regulatory 1.68. *Preopera-tional and Initial Startup Test Program for Water Cooled Power Reactors, 11/73 (RO)." Your test program ~hould meet the re-quirements of Regulatory Guide 1.68, Revision 2.
Response
FP&L intends to comply with Regulatory Guide 1.68, Revision 2 as addressed in Chapter 14 ~f the FSAR. See revised Chapter 14 of Amendment 1 *
- Q42-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
43 (15) Chapter 15 does not contain or reference either sufficient backup information and data or justification of the accident analysis methodology for the staff to proceed with our review. This should be provided with our review. This should be provided for the following areas:
- 1. The five frequency groups and the acceptance guidelines assigned to these groups,
- 2. Event identification and combinations in conjunction with the
*assigned frequencies and associated conservatism,
- 3. The selection of the limiting event or event combinations for analysis within a group, and
- 4. The evaluation of parameters that may affect 'the performance of barriers (i.e., containment, filters, etc.) that restrict or limit the transport of radioactive material to the public, and information to fully substantiate the dose analysis and conservatism so as to allow an independent analysis t.o be performed by the NRC staff as specified in Section 15 of Regulatory Guide 1.70 Revision 3.
R~spo~se
- The overall* introduction to Chapter 15 and the introductions to each major analysis section already address this question.
Therefore, Combustion Engineering proposes that a meeting be held between the NRC staff and our engineering personnel to present and explain our safety analysis methodology in more detail. After such a meeting, the NRC staff could request formal resolution of specific areas of concern. No FSAR revisions are required .
- Q43-1 Amendment No. 1, (4/81)
SL2-FSAR
- Question No.
44 (15) Provide the design basis LOCA analysis as specified in Regulatory Guide 1.70, Revision 3 and Standard Review Plan 15.6.5 and associated appendices.
Response
The design basis LOCA analysis which demonstrates compliance with NRG criteria 10CFR50.46 is contained in Chapter 6 of the FSAR. The minimum containment pressure analysis is found in Subsection 6.2.1.5. Emergency Core Cooling System performance is presented in Sectio~ 6.3, which contains the analysis results for the large break LOCA (Subsection 6.3.3.2). The analysis results of the small break LOCA (Subsection 6.3.3.3) are being submitted to the NRG along with this amendment. The long term cooling plan (Subsection 6.3.3.4) will be submitted in June 1981. No FSAR revisions are required .
- Q44-1 Amendment No. 1, (4/81)
SL2-FSAR
- Qiestion No.
210A.l Details of the actual limited displacement break flow area and the actual break separation ~ime at any circumferential break location are needed for this specific plant.
Response
Details of the actual limit~d displacement break flow area and the actual break separation time at any circumferential*break location are provided in revised FSAR Subsectlon 3.6.2. See revised FSAR Subsection 3.6.2. 210A.1-1 Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
210A.2
Response
.rt must be demonstrated that St. Lucie plant analysis system parameters fall within the design envelope of CENPD-168, Revision 1.
The system parameters of the St. Lucie 2 plant fall within the design envelope of CENPD-168, Revision 1. FSAR Subsection 3.6.2. has been revised. See revised FSAR Subsection 3.6.2 *
- *.210A.2-1 Amendment No. 6, (9/81)
SL2-FSAR
- Qiest ton No.
210A. 3* Assurance should be provided that the criteria used to predict break'location, as referenced to CENPD-168A, Revision 1 is used for reactor coolant system piplng only. If this criteria is used for piping other than the RCS, additional justification must be provided.
Response
CENPD-168A, Revision l ls used to predict break locations for Reactor Coolant loop piping only. For all other high.energy piping, as indicated in Subsection 3.6.2.1.l(e), pipe whip analysis is performed based on "break anywhere" criteria, and jet impingement analysis is performed for breaks postulated in accordance with the criteria given in Subsections 3.6.2.2.l(b), (c) an~ 3.6.2.2.2. No FSAR change is required *
- 210A. 3-1 .Amendment No. 6, (9/81)
.SL2-FSAR Question No *
- 210A.4
Response
Additional items not covered by CENPD-168, and which should be provided for the reactor coolant system of the plant are the pipe whip restraint parameters such as stiffness values and gap sizes. The pipe whip restraint parameters (i.e., stiffness values) are provided in revised FSAR Subsection 3.6.2.1. Gap sizes for reactor vessel inlet break are also provided. I9 See revised FSAR Subsection 3.6.2.1 *
- 210A.4-1 Amendment No. 9, (3/82)
SL2-FSAR
- Q.lest ton No.
210A. 5 The applicant has made a commitment to provide the following item in a future amendment to the FSAR. High Energy pipe rupture analysis inside containment (Appendix 3.6A). Response. All information requested above has been supplied in Amendment 3 of the St. Lucie 2 FSAR (issued June 81). No FSAR change is required *
- 210A.5-l Amendment No. 6, (9/81)
SL2-FSAR (pestion No *
- 210A.6 The applicant has made a commitment to provide the following item in a future amendment.to the FSAR.
High Energy pipe rupture analysis outside containment (Appendix 3.6B).
Response
All information requested above has been supplied in Amendnient 1 of the St. Lucie 2 FSAR (issued -June 81). No FSAR change is required *
- 210A.6-1 Amendment No. 6, (9/81)
c,
. SL2-FSAR
- Qiestion No.
- 210A. 7 The applicant has made a commitment to provide the following item tn a future amendment to the FSAR..
Pipe whip restraints and break location (Appendix 3.6C).
Response
~11 information requested above has been supplied in Amendment 3 of the St. Lucie 2 FSAR (issued June 81).
Mo FSAR change ls required *
- 21()A.7-1 Amendment No *. 6, (9/81)
SL2-FSAR
- ~estion 210A.g No.
The applicant has made a commitment to provide the following item in a future amendment to the* FSAR. Structural details of the pipe whip restraints (Appendix 3.6D).
Response
All information requested above has been supplied in Amendment 3 of the St. Lucie 2 FSAR (issued June 81)* No FSAR change is required *
- 210A.8-1 Amendment No. 6, (9/81)
SL2-FSAR
- Q.iestlon No.
210A. 9 The appl leant has made a commitment to provide the follo111ing item ln a future amendment to the FSAR. Main Steam and feedwater dynamic analyses (Appendix 3~6E).
Response
All information requested above has been supplied ln Amendment 3 of the St. Lucie 2 FSAR (issued June 81). No FSAR change is requlred *
- 210A.9-l Amendment No. 6, (9/81)
SL2-FSAR Q.testion No *
- 210A.10 The applicant has ma1e a commitment to provide the following item in a future amendment to the FSAR.
Moderate Energy analysis (Appendix 3.6F).
Response
All information requested above has been supplied in Amendment 3 of the St. Lucie 2 FSAR (issued June 81). No FSAR change is required *
- 210A.10-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A. l.l, The FSAR Subsection 3.6.2.2 should be clarified to show that the requirement of 0.8 (Sh + SA) is- based on the sum of Equations (9) and (10) of Paragraph NC-3652 of the ASME B&PV Code, Section III and not Equation (9) and (10) individually.
Response
The stress criteria of 0.8 (Sh + SA) is compared against the sum of Equations (9) and (10) of Paragraph NC-3652 and ND-3652 of the ASt'1E Code Section III to determine break locations. Refer to revised FSAR Subsection 3.6.2.2.l(c)(2) *
- 210A.11-1. Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 2 lOA.12
Response
Provide criteria for postulated pipe breaks in both high and moderate energy piping systems in the containment penetration area. High energy pipe breaks and moderate energy leakage cracks are postulated based on the criteria of Subsections 3.6.2.2.1(6), 3.b.2.2.2 and 3.6.2.4 for piping located in the penetration area. The pipe rupture analysis does not take credit for any break exclusion region. Refer to revised FSAR Subsection 3.6.2.S *
- 210A.12- l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.13
Response
Provide the basis for the 0.8 SA criteria for expansion stresses which is stated in Subsection 3.6.2.2.2(2) of the FSAR. The basis for 0.8 SA criteria for expansion stresses stated in Subsection 3.6.2.2.2(2) of the FSAR is as per Appendix B (Giambusso criteria) to BTP APCSB 3-1 item 2.(b).(2). Refer to revised FSAR Subsection 3.6.2.2.2 *
- 210A.13-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.14 Provide a listing of the high energy systems that are considered for pipe rupture analysis. In addition provide a summary of the results of the analyses of these systems to demonstrate that essential systems, components, and supports will not be impaired as a result of high energy pipe breaks.
Response
High energy syst~m considered for pipe rupture analysis are: Inside Containment
- 1. Main Steam and Feedwater
- 2. Reactor Coolant (includes pressurizer surge, spray and relief)
- 3. Safety injection (all lines pressurized by the safety injection tanks)
- 4. Shutdown Cooling (high energy portion only)
- 5. Chemical and volume control (letdown and charging)
- 6. Steam generator blowdown Outside Containment
- 1. Main Steam and Feedwater
- 2. Chemical and Volume Control (letdown and charging)
- 3.
4.. 5. 6. Steam generator blowdown Auxiliary Steam System Auxiliary feedwater system Steam supply to Auxiliary Feedwater pump Tne results of the analyses of tnese systems are presented in Appendices 3.6A and 3.6B. No FSAR change is required *
- 210A.14-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.15
Response
When longitudinal breaks are postulated, assurance must be
.provided that they are chosen in the location that is likely to cause the maximum damage.
Longitudinal breaks are postulated to occur at any location about the circumference of the pipe. This method identifies the' location that causes the maximum damage. Refer to Subsection 3.6.2.3(b)(2) of the FSAR. No FSA~ change is required *
- 210A.15-l Amendment No *. 6, (9/81)
SL2-FSAR
- Question No
- 210A.16 The following information is required as it pertains to the subsystem analysis before our review can be completed.
(1) The method for determining that an adequate number of degrees of freedom were used in the dynamic modeling to determine the response of all Category I and applicable Non-Category I structures and plant equipment.
Response
For piping systems, adequate mass points and corresponding dynamic degrees of freedom are selected and distributed to provide for appropriate representati'on of the dynamic characteristics of the subsystem. As indicated in Subsection 3.7.3.1.1.2, the maximum spacing of the mass points does not exceed one half the distance for which the frequency of a simply supported beam would be 20 Hz. Each mass points, except for points indicated as restrained in a given direction, have 3 linear degrees of freedom. Therefore, the degrees of freedom exceed twice the number of mode with frequencies less than 33 Hz. The dynamic models of the cable tray and HVAC duct with their respective support structures were constructed with an adequate number of mass points in order to simulate the dynamic behavior of the subsystem. The number of mass points in the dynamic model is adequate because the number of degrees of freedom exceeds twice the number of modes with frequencies less than 33 Hz. NSSS vendor supplied subsystems are modeled with sufficient masses (dynamic degrees of freedom) such that inclusion of additional degrees of freedom results in less than a 10 percent increase in responses (Analysis of the pressurizer, surge line, and spray line also meets the alternate criteria that the number of degrees of freedom is greater than twice the number of modes with frequencies less than 33 Hz). No FSAR change is required *
- 210A.16-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.17
Response
P.rovide justification that a sufficient number of modes were considered to assure* participation of all significant modes is required. The criterion for sufficiency in number of dynamic modes is that the inclusion of additional modes does not result in more than 10 percent increase in response. In general, this can be satisfied by including all the dynamic modes below 33 Hz, if the highest mode calculated below 33 Hz has already fallen into the flat rigid response region of the corresponding response spectra, the effect of the remaining high modes are taken care of by adding the dynamic analysis result with an equivalent static solution in *sRSS summation. Dynamic analysis of cable tray/HVAC duct-support systems has combined all modes in the flexible region together with residual terms accounting for higher modes in the rigid region. Criterion used to assure that sufficient modes are.included in the analysis of NSSS vendor supplied subsystems is that the inclusion of additional modes results in less than a 10 percent increase in response. Analysis of the coupled components of the RCS included all modes less than 50 Hz (Subsection 3.7.3.l.2.3(b)) *
- No FSAR change is required *
- 21 OA.17-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.18
Response
Provide the methods used to handle the relative displacements of Category I supports. (1) The following is a summary of the method used to handle the relative seismic displacement of support in piping systems.
- a. The relative seismic displacements between supports/restraints installed on the same building structure are normally* negligible in *the stress analysis.
- b. The relative seismic displacements between supports/restraints located in two buildings on separate
*mats are to be derived from the combination of co-directional maximum absolute seismic displacement of the two buildings at the supporting elevation by SRSS (square root of the sum of the square*s) method.*
Since the adjacent buildings on separate mats are dynamically analyzed as decoupled systems, the real relative seismic displacement between piping restraints on separate buildings are not readily available. As the soil interactions are incorporated in the dynamic analysis of the buildings, the maximum absolute seismic displacements of each building available for piping stess analysis usage*normally contain.considerable soil movement which is expected to be closely in-phase for locations between the adjacent buildings. This is evidently demonstrated in the Maximum Absolute Vertical SSE Seismic Displacement Tables for Reactor Containment ~uilding (RCB) and Reactor Auxiliary Building (RAB)(Table 210A.18-l). The values for RCB mat and the top of the steel containment are 0.0304 feet and 0.03067
.feet respectively while the value of RHB mat and top mass of RAB are 0.02108 feet and 0.02145 respectively. The displacements for masses in between are approximately in the same magnitude.
- A relative seismic displacement between RCB and RAB mats calculated by Absolute Sum Combination of the two maximum absolute displacements (ABS) will result in a value of
'0.05148 feet. This essentially represents an assumption of complete out-phase of the adjacent mats each at its peak movement at the same instant. In consideration of similarity in surrounding soil properties, closeness of spacing and the seismic wave motion in ground, this is very unlikely to happen. Furthermore, the fundamental frequencies of the adjacent building which normally
- dominate the displacement response are also not the same.
210A.18-1 Amendment No. 6, (9/81)
SL2-FSAR The Application of Absolute Sum Combination of the_ two maximum absolute building seismic displacement onto the interface piping will result in a piping system design overly emphasized in relative displacement condition. This, by the nature of flexibility requirement, tends to undercut the conservativeness usually existed in normal seismic protection design on the basis of seismic response of inertia effect which, being accounted for the primary pipe stresses, is more probable to cause structural failure. The SRS delineated in our Response lb, to Question 210A.18 offer a better balanced seismic protection design for a combined effect of seismic inertial and displacement on interface piping. It is recognized that the pattern of the relative horizontal movement, between building at high elevation is usually hard to predict. A less conservative approach is to consider the complete out-phase of the two building structural response along with the close in-phase of the two adjacent mats. Therefore, the relative seismic displacements between support/restraints located on two buildings on separate mats can be derived by taking the absolute sum of each relative building seismic displacement towards each mat and the difference of the mats displacements. This is assigned as .:\ABSR, in the attached table for comparison of combination methods for evaluating relative seismic displacements. As all the piping penetrations are located at the lower part of the building,. the strong contribution from the soil movement to the relative displacements between the pipe restraints is anticipated. The Relative Seismic Displacement derived by SRSS method ( .:\SRSS)m as shown in the comparison table, is considerbly higher than 6ABSR.
- c. The relative seismic displacements between supports/restraints located in two buildings on a common mat or attached to two structures within the same building are to be derived by taking the square root of the sum of the squares of each relative seismic displacement towards a common reference.
- d. For piping connecting to equipments or primary piping systems of which the available maximum seismic displacements are relative to the base support of the major equipments, the base supports are to be selected as the common reference. The maximum seismic displacement of the subject piping restraint system are to be converted into relative displacements towards the base support of connecting nozzle and the piping restraint system are then to be derived by taking absolute addition of the two relative seismic displacements which in turn are relative to the base support of the equipment.
210A.18-2 Amendment No. 6, (9/81)
SL2-FSAR
- e. The piping sys~em will be analyzed separately with
- relative seismic displacement input in each of the three orthoganal coordinate directions. The resultant response (such as pipe stress, moment, force, etc) are obtained by taking SRSS of the response corresponding to each coordinate direction.
(2) Relative displacement among supports located at different floor elevations are not considered in cable tray and HVAC duct seismic analysis. Ducts are provided with flexible jo~nts to accommodate relative displacement.of the supports. (3) For the coupled components of the RCS the relative support displacements are applied directly in the time history analysis methods described in Subsection 3.7.3.1.2.3. For NSSS vendor supplied multiply supported subsystems analyzed by response spectrum methods, relative support displacements are applied statically in the most unfavorable manner. No FSAR change is required *
- 210A.18-3 Amendment No. 6~ (9/81)
SL2-FSAR Question No
- 210A.18A Please prepare a comparison of maximum differential seismic displ'acements of structures met on a common mat, (DBE time history vs design values) for St Lucie Unit No 2 o
Response
On St Lucie Unit 2, the design values for displa.cement of structures not on a common mat were computed by a SRSS summation. of maximum response spectra displacements. The maximum response spectra values a re ba~ed on enveloping maximum response spectra displacements of each building considering a range of various soil properties. In response to the NRC's request for justification of these design values, various scoping studies have been performed to determine appropriate values for maximum expected displacements at St Lucie Unit 2. These values are compared to those actually used in the pipe stress analysis calculations. Scoping Studies Summary
- 1) The most realistic indication of the true differential seismic displacements can be seen by comparing the time history values between two points on different structures and calculating the maximum difference. Inherent in this comparison are two assumptions for adjacent structures:
- a.
b. Motion of the two structures begin in-phase. Differential ground motion between the two structures are negligible. Table 210A.18A-l indicates these maximum difference values for two cases: for piping system going from elevation 28'-0" in the reactor building (RB) to elevation 18'-6" in the reactor aux building (RAB) and for a system from elevation 48'-0" in the RB to 42'-6" in the RAB~ These two cases are typical for piping between these two buildings. Note that the time history values are in all cases substantially less than the St Lucie Unit 2 design values.
- 2) In order to account for any additional displacements resulting from inaccuracy due to the two assumptions stated above, the time history values have also been combined by maximum summations of the values as a function of time. As would be expected, these absolute sum displacements are greater than those calculated previously. However, as shown in Table 210A.18A-l, these values .are in all cases less than the St Lucie Unit 2 design values.
- 3) An additional parameter that could be considered is variation in time interval. The time history earthquakes were generated using a time interval of .005 sec. This could be "spread" by 20 percent analagous to the peak spreading employed for response spectra. Thus time histories with intervals from
.004 to .006 sec would be considered. This spreading would account for uncertainties in design assumptions.
210A.18A-1 Amendment No. 9, (3/82)
SL2-FSAR Table 210A.18-2 indicates that even assuming 20 percent time interval spread, the maximum differential displacements are in all cases less than the St. Lucie Unit 2 design values.
- 4) Finally, the displacement values from the "spread" time history have been combined by the maximum sum method. As shown in Table 210A.18A-2 the design values are in all cases within 15 percent of the values determined by this technique.
The maximum difference is less than l/'C", Discussion Seismic displacement is a secondary stress when applied to pi.ping systems and is combined with other stresses to determine the total effect. For supports and penetrations, seismic displacement is a primary load. However, this is combined with several other primary loads. Variations in the magnitude discussed above would have negligible impact on the total system and would therefore not require any further evaluation. . It should also be noted that the deviation in displacement values exists only for the .006 sec time interval. The .006 sec time interval reflects a 20 percent increase above the base case of
.005 sec to correspond to the spectra broadening that was used on St Lucie Unit 2. However, this interval could be reduced to 15 percent and be in compliance with the Standard Review Plan 3. 7. 2 Conclusion (11/24/75). The effect would be to reduce or eliminate the deviation between the time history value and the design value.
Based on the comparisons presented in Tables 210A.18A-l and 210A.18A-2 and the discussions above, the specific seismic displacement design values developed for St Lucie Unit 2 are acceptable. This conclusion would apply to all the piping systems which penetrate the containment and Shield Building listed in FSAR Table 6.2-52, except as indicated in Table 210A.18-3. 210A.18A-2 Amendment No. 9, (3/82)
SL2-FSAR TABLE 210A.18A-l
- COMPARISON OF TIME HISTORY DISPLACEMENTS TO DESIGN VALUES Displacement between RB El 28'-0" & RAB El 18'-6" Direction Max ABS Sum Maximum Difference Design Value N-S Q 00484 I 0.0462' 0.0515' E-W 0.0462' 0.0426" 0.0522' Vert 0.0371' 0.0286' 0.0372' Displacement between RB El 48'-0" & RAB El 42'-6" N-S 0.0615' 0.0593' . 0 .0649' E-W 0.0591' 0.0531' 0.0625' Note:
Maximum absolute summation is the maximum value found by adding the absolute .....,., value of the displacement of one building to the absolute value of the '* displacement of the other building both taken at the same time over the duration of the seismic event. *,'
- Maximum difference is the maximum value found by algebraic subtraction of the displacements of one building to the displacement of the other building both taken at the same time over the duration of the seismic event *
- 210A.18A-3 Amendment No. 9, (3/82)
SL2-FSAR TABLE 210A.18A-2 EFFECT OF T VARIATION ON COMPARISON OF TIME HISTORY DISPLACEMENTS TO DESIGN VALUES
~T ABS Sum (ft) Max. Diff. (ft) Design Values N-S .004 .0457 .0391 .0515 I Displacement betw RB El 28'-0" 0005 .0484 .0462 .o 515 I & RAB El 18'-0" .006 .0567 .0414 ,0515 I Vert .004 .0357 .0309 .0372' Disp betw RB El 28'-0" & .005 .0371 .0286 .o 3 72 I RAB El 18'-6" .006 .0431 .0364 .0372' E-W .004 .0410 .0392 ,0522 I Disp betw RB El 28'-0" & .005 .0462 .0426 ,0522 I
- RAB El 18'-6" N-S Di sp betw
*RB El
.006
.004
.0590
.0551
.0394
.0496
,0522 I
.064 9 I 48'-0" & .005 .0615 .0593 .0649' RAB El 42'-6" .006 .0713 .0528 .0649' E-W .004 .04806 .04806 .0625 I Di sp betw RB El 48'-0" & .005 .0591 .0531 .0625 I RAB El 42'-6" .006 .0728 .0502 .o 625'
- 210A.18A-4 Amendment No. 9, (3/82)
*
- SL2-FSAR TABLE 210A.18A-3 SEISM1C DISPLACEMENT BETWEEN BUILDINGS
*Penetration fl. Elevation Pipe Size System Reason For Not Analyzing Penetration 10 54'-0 48" Main Containment Purge Flexible Connection in ductwork in RAB immediately downstream containment isolation valve 11 33'-0 48" Main Containment Purge HVAC duct ends approx 5 ft. into RAB HVAC duct ends approx 2 ft. into RB 12B' 34'-0 Spare Penetrates concrete shield wall only 25 27'-6 36" Fuel Transfer Tube No piping connections on either FHB side or containment side 45, 53 25'-0/28'-0 3/8" Instrument tubing Is analyzed for ABS Sum displacement 55 25'-0 Spare No piping 56 26'-0 8" HVAC Mini Purge Inlet HVAC duct ends approx. 5 ft. into RAB HVAC duct ends approx 2 ft. into RB NI 57 51'-0 8"
~ HVAC Mini Purge Outlet Flexible connection on ductwork in RAB immediately .... downstream containment isolation valve ~I V1 58 51'-0 Spare Blind Flanged 59 56'-0 31" Shield Bldg Vent System Flexible connection on duct work in RAB immediately downstream containment isolation valve 60 56'-0 31" Shield Bldg Vent System Flexible connection on ductwork in RAB immediately downstream containment isolation valve 61 51'-0 Spare Blind flanged 62 51'-0 3/4" Instrument Air to Is analyzed for ABS Sum displacement Construction Hatch 65 58'-0 Spare Blind flanged 66 58'-0 Spare Blind flanged 67 24" Containment Vacuum Relief Penetrates Steel Containment only z 0 68 24" *Containment Vacuum Relief Penetrates Steel Containment only
*Shown on drawing 2998-G-213 Sheets 3 and 4, Revisions 4 and 5 respectively
SLZ-FSAR TABLE 210A.18-l
- - a>MPARISON OF COMBINATION METHODS FOR RELATIVE SEISMIC DISPLACEMENTS VERTICAL SSE MAXIMUM ABSOLUTE SEISMIC RELATIVE SEISMIC DISPLACEMENT (ft.) DISPLACEMENT (ft.)
Reactor Containment Reactor Auxiliary Building MASS II MASS II AXA AXB AABS ASRSS AAB SR 6 (top of PCV) 0.03067 1 (top) o. 02145 o. 05212 0.03743 0.00996 16 (Mat) 0.03040 23 (Mat) 0.02108 0.05148 0.03699 0.00932 N-S MAXIMUM ABSOLUTE SEISMIC RELATIVE.SEISMIC SSE DISPLACEMENT (ft.) DISPLACEMENT (ft.) Reactor Containment Building MASS II MASS II AXA 19(EL. 44'-0") 0.04066 3 (EL.42'-5") 0.0402 0.0808 0.05717 0.03646 25 (Mat) 0.02783 5 (Mat) 0.0222 0.05003 0.03560 0.00563 (1) SRSS = (AXA 2 +AXB 2)1/2 (2) ABSR = l<AXA) - (AAmat>I + l<AXB-AB mat)j
+ IC AA mat -AB mat)j (3)
Where AXA Maximum Absolute Seismic displacement of building A at restraint location. AXB Maximum .Absolute Seismic displacement of building B at restraint location *
- Maximum Absolute Seismic displatement of the mat.
210A.18-4 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.19
Response
Provide information on how significant effects such as piping interactions, externally applied structural restraints, hydrodynamic loads and non-linear responses are accounted for. The effects of piping i~teractions and hydrodynamic loads are considered in the analysis of Safety Class 1, 2 and 3 components. The safety-related equipment is designed to withstand the piping interaction loads that could be imposed on these components. These loads are combined with the other plant loading in accordance with FSAR Table 3.9-6. The equipment is analyzed to assure that under these loadings the operability will be assured. The summary of results for these active components is provided in FSAR Appendix 3.9A. Interaction of the RCS main loop piping and the major components is accounted for directly in the time history analysis of the composite coupled model. Treatment of hydrodynamic effects and non-linear response of the reactor internals and fuel is discussed in Subsection 3.7.3.14. See revised FSAR Subsection 3.7.3.14 *
- 210A.19-l Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A.20 If the equivalent static load method was used, justification must be provided that the system can be represented by a simple model and that ~he relative motion between support points is accounted for.
Response
The piping system stress analysis does not utilize the equivalent static load method. The piping analysis utilizes the modified equivalent static. load method which is described in the response to Question 210A.23 *
. Cable tray and HVAC duct seismic supports were designed in the following manner:
A seismic response analysis was performed on a 3-D model, which represented a typical multiple span of cable tray (HVAC duct) and its supports. Each tray (duct) support in the model was assigned
.a fundamental frequency of 16 to 18 Hz in three directions. The results of this analysis was an amplification factor which was then used to determine the static equivalent "g" values :for design of individual cable tray (HVAC duct) supports.
- Each support is designed to have a fundamental frequency of 16 to 18 Hz *
- Regarding considerations of relative motion between supports see discussion for 210A.18 above.
For* NSSS vendor supplied subsystems the equivalent static load method is limited to analysis of components which can be realistically represented as single-degree-of-freedom systems or by simple-beam or frame type models. For multiply supported components the relative motion between supports are applied irt the most unfavorable manner using static analysis procedures and responses are added to those due to inertial effects by the absolute sum method. No FSAR change is required *
- 210A.20-l Amendment No. 6. (9/81)
SL2-FS.AR Question No
- 210A.21 The criteria and procedure given for the modeling of the seismic systems and the criteria for determining whether a component is analyzed as part of a system or independently requires amplification and inclusion of all information required by the SRP. Before our review can be completed on this section, the criteria and procedures actually used must be described.
This should include the modeling procedures used and the criteria for decoupling as outlined in SRP Subsection 3.7.2, paragraph II r. 3.
Response
The criteria and procedure for the modeling of the .seismic system are stated in FSAR Subsection 3.7.2.3. For the reactor building in particular, studies uSing seismic models with and without subsystem are made to ensure the coupling' effect is minimal. Models with major equipment (such as steam generators and reactor vessels) and the supporting structure (i.e. the internal structure) modeled separately and modeled together are constructed and the Computer Code STARDYNE is employed. Dynamic responses such as frequencies, accelerations, and response spectra are c'ompared. The differences are found negligible *
- The reactor internal structure response spectra as shown on Figure 3~7-15 illustrates that the peak acceleration occurs *approximately at 3 Hz. The RCS loop has a fundamental of 10 Hz. Thereby the coupling effect between the reactor building and the RCS loop is insignificant.
No FSAR change is required *
- 210A.21-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.22 A discussion of the methods actually used in determining the fundamental frequencies is required in this FSAR Section. Also explain how the three ranges of equipment/support behavior (rigid, flexible, resonant) delineated are handled in the analysis. A statement or statements is required as to how these matters are considered in the analysis.
Response
For piping systems which are analyzed by either modal response
*spectra method or modified equivalent static load method, the fundamental frequencies are determined by, the stiffness matrix method of natural mode analysis as described in Subsection 3.7.3.1.1.2.2. For piping systems which are analyzed by simplified seismic analysis method, the exact values of fundamental frequencies are not calculated. ~s described in Subsection 3.7.3.1.1.C, the piping are restrained to have fundamental mode periods less than 70 percent of the first mode period of the supporting structure. This was accomplished by comparing and modifying the restraint spacing in design with that of a simply supported beam.
Where feasible, the piping system are arranged to be in the rigid region (i.e., the fundamental frequencies are more than twice the* dominant frequencies of the support structure. If the fundamental
- frequency of the piping system is less than twice but more than 1.43 of the dominant frequencies of the support structure, the modified Equivalent Static Load Method as delineated in Subsection 3.7.3.1.lb is used. The Modal Response Spectra Method is normally used for piping systems in the flexible or resonant region.
Frequencies for the Reactor Coolant System and reactor internals
*are calculated.in accordance with the procedures described in Subsection 3.7.3.1.2.3 (b) and 3.7.3.14, respectively. The three ranges of equipment/support behavior (rigid, flexible, resonant) are not delineated for NSSS vendor supplied subsystems. (Current Subsection 3.7.3.4 describes procedures for NSSS vendor supplied subsystems).
See revised FSAR Subsection 3.7.3.14. *
- 210A.22-l Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A.23 Justification has been provided for the use of the equivalent static load method for piping systems. Similiar justification is needed for all equipment for which this method was used. Also provide clarification on how the modified equivalent static load method differs from the equivalent static load method.
Response
The modified'equivalent static load method as described in FSAR Subsection 3.7.3.1.1.b, is a frequency based static analysis. It is applicable when the piping system is proved to be in the relatively rigid s~de of the dominant frequency of the supporting structure. At first, the fundamental frequency of the piping system is determined by the same stiffness matrix method of natural mode analysis described in FSAR Subsection 3.7.3.1.1.a*.2, then a static analysis is performed using an acceleration value of 1.5 times the maximum value of the applicable floor response spectrum in the period range equal to*or less than the first mode period of the piping system. The equivalent static load method, as we interpret from SRP 3.7.3 Section II, b does not require demonstration of the fundamental frequency of the piping system, equipment etc., a factor of 1.5 is applied to the peak acceleration of the applicable floor response spectrum to obtain the equivalent static load
- As indicated in Subsection 3.7.3.1.1, the seismic analysis of Non-NSSS piping is done by using one of the three following, methods:
a) Modal Response Spectra Method -- This method is based 9n the classical modal analysis which involves the calculatioJ;l of all the significant natural frequencies and their mode, shape vectors and the response combination of these modes of, vibration. b) Modified Eq~ivalent Static Load Method (Simplified dynamic analysis). This method involves the calculation of th~ first mode period of the piping system to determine the applicable value of accelerations which in turn is used in the ! equivalent static analysis. c) Simplified Static Method (chart method) -- This method involves the development of reference. restraint spacing based on preset value of fundamental piping period to preclude the possible resonance with the support structure. The location of restraint on the piping system is determined by comparing the individual selected restraint spacing with the reference restraint spacing *
- No FSAR change is required.
210A. 23-1 Amendment Nd.* 6, (9/81)
SL2-FSAR Question No
- 210A.24 Discuss the approach for combining the loads corresponding to the three components of earthquake motion when the time history method of analysis is used.
Response
For NSSS vendor supplied subsystems analyzed by_ time history methods, maximum components of reaction at all design points are calculated for each separate direction of seismic excitation. Maximum co-directional responses resulting from each of the three orthogonal directions of ground excitation are then combined by the square root of the sum of the squares (SRSS) method. The resultant six load components are applied simultaneously in computing the stresses for each component or*structure. (See Subsection 3.7.~.l.2.4). No FSAR change is required *
- 210A.24-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.25 The criteria to be used in the analysis of multiple supported equipment and components meet the staff requirements as outlined in NRC Standard Review Plan 3.7.3 Section II-9,with the exception that a commitment be made to combine the support displacements in the most unfavorable combinations.
Response
For combination of support displacements of the piping system, see Response to Question 210A.18. For NSSS vendor supplied multiple supported components analyzed by the :response spectrum method, the .support displacements are imposed on the supported item in the most unfavorable combination using static analysis procedures. The responses due to the inertial effect and relative displacement are combined by the absolute sum method. (See Subsection 3.7.3.l.2.3(d) for surge (and spray) line analysis). The analysis of the multiple supported coupled components of the RCS are analyzed using time history procedures with the relative support displacements applied directly as described in Subsection 3.7.3.1.2.3. Note: FSAR Subsection 3.7.3.9(c) currently provides commitment requested by the draf_t SER *
- No FSAR change is required *
- 210A. 25-1 Amendment No. 6, (9/81)
.SL2-FSAR Question No
- ZlOA.26 This section concerning the interaction of other piping with seismic Category I pipi_ng adequately defines how these piping systems are handled when they are a part of the same system.
However, information is required as to how Non-Category I piping systems are analyzed and/or isolated from Category I piping when the systems are in close proximity so that a failure of the Non-Category I piping would not damage the Category I piping.
Response
As required for safe plant shutdown, Non-Category I piping is seismically supported where it passes over seismic Category I piping, valves and valve operators. See revised Subsection 3.7.3.13 *
- 210A. 26-1 Amendment*No. 6, (9/81)
SL2-FSAR Question No *
- 210A.28 A description of the linear vertical analysis and nonlinear horizontal analysis is provided. Verify whether or not a vertical nonlinear analysis is used in the event that the linear vertical analysis indicates that the response of the core may be sufficiently large to lift off the core plate. In case it is used, provide a description of the analysis.
Response
A linear analysis has been completed. Because of the low level of excitation, the fuel does not lift off the core support plate. Therefore, a nonlinear analysis is not required. No FSAR change is required *
- 210A.28-l Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A.29 Provide a commitment that closely spaced modes are considered as per Regulatory Guide 1.92, in the analysis of the reactor internals and the core.
Response
In the analysis of reactor internals and the core, closely spaced modes are considered in accordance with Regulatory Guide 1.92. See revised FSAR Subsection 3.7.3.14. *
- 210A.29-1 Amendment No. 6, (9/81)
0949W-l SL2-FSAR
- Question No.
210A. 30 Piping vibration, thermal expansion, and dynamic effects testing will be conducted during the St. Lucie plant's preoperational and start-up testing program. The purpose of these tests is to confirm that the piping, components, restraints, and supports have been designed to withstand the dynamic loadings and operational transient conditions that will be encountered during service as required by the ASME Section III Code and to confirm that no unacceptable restraint of normal thermal motion occurs. We have identified the following open issues in our review. The issues are identified by sections of the FSAR. Many of the items required by the Standard Review Plan (SRP) Subsection 3.9.2 are covered only briefly or not at all in this section. The SRP Acceptance Requirements II.la through f and items a through d of the review procedures should be addressed before this FSAR section can be considered acceptable. The staff requires a commitment to test all high energy piping and all seismic Category I moderate energy piping including supports and restraints for thermal expansion, steady state vibration, dynamic and transient loads.
Response
- SRP Subsection 3.9.2 Acceptance Requirements Ila through f and Review Procedure items a through d are addressed in the revised FSAR Subsection 3.9.2.1.
The listing of safety-related snubbers is addressed in the St. Lucie 2 Technical Specification document. A list of deflection points is provided in the St Lucie 2 Preoperational Vibration test 11 procedure and thermal expansion test procedure. See revised FSAR Subsection 3.9.2 . 210A.30-l Amendment No. 11, (7/82)
SL2-FSAR Question No.
- 210A.31. l Justify decoupling the horizontal and vertical components of the responses to blowdown loads.
Response
The axial and lateral internals models were uncoupled to provide more spatial detail to account for important structural characteristics in the separate models. There is a separation in the axial and lateral natural frequencies, so that the response characteristics in the separate directions are not coupled. Typical lateral natural frequencies for St. Lucie Unit 2 range from 2 to 25 hertz. Typical vertical natural frequencies range from 25 hertz to 200 hertz. The results of the analyses show the lateral displacements to be small and, when combined with the maximum axial loads, the beam column effects are negligible. Typical peak horizontal relative displacements for St. Lucie Unit 2 are .100 inches. Typical peak vertical displacements are approximately 1/10 of horizontal. See revised FSAR Subsection 3.9.2.5.2.3 *
- 21 OA
- 31. 1-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.31.2
. Response Justify the use *of results of linear analysis for the inherent non-linear problem
- The horizontal and vertical models which were used to determine the LOCA structural responses of the reactor internals were non-linear. The CESHOCK (References 1, 2, 3) code was used to calculate the maximum component loads that resulted from the postulated hot and cold leg breaks. These analyses considered non-linearities such as gaps, damping, friction, hysteresis and coefficient of restitution.
References
- 1. "Topical Report on Dynamic Analysis of Reactor Vessel Internals Under Loss-of-Coolant Accident Conditions with Application of Analysis to CE 800 Mwe Class Reactors,"
Combustion Engineering, Inc., Report CENPD-42, August.1972 (Proprietary).
- 2. Gabrielson, V.K., "SHOCK, A Computer Code for Solving Lumped-Mass Dynamic Systems," SCL-DR-65-34, January 1966.
- 3. "Structural Analysis of the 16 x 16 Fuel Assembly for Combined Seismic and Loss-of-Coolant-Accident Loadings,"
CENPD-178P, Combustion Engineering Propriety Report, October 1976. See revised FSAR Subsection 3.9.2.5.2.3 *
- 210A,31.2-1 Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A. 31. 3 Present a discuss.ion outlining the effects of system flow upon mass and flexibility properties.
Response
The effects of system flow on the dy~amic response of reactor vessel internals are secondary. The hydrodynamic effect of these components is dominated by hydrodynamic coupling and hydrodynamic added mass. Both of these effects are considered in the dynamic response analyses of these components. A detached description of CE methodology for hydrodynamic mass is presented in CENPD-178-P Rev. 1, "Structural 'Analysis of Fuel Assemblies for Seismic and Loss-of-Coolant-Accident Loading" released in August 1981. Additional references which describe\this hydrodynamic mass methodology are listed below. References
- 1. Fritz, R.J. "The Effect of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry, Paper No. 71-VIB-100.
- 2. McDonald, D.K. "Seismic Analysis of Vertical Pumps Enclosed in Liquid Filled Containers," ASME Paper No. 75-PVP-56 *
- See Revised FSAR Subsection 3.9.2.5.2.3 *
- 210A.31.3-1 Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A.32 We find this program acceptable provided the app~icant submits a correlation of the St. Lucie Unit 2 observed vibrational characteristics with the results from the prototype reactors.* l;f the comparison of the observed vibrational characteristics of St.
Lucie with those of the prototype plants indicate the need for any corrective action, the staff will review the: applicant's proposed corrective action for St. Lucie Unit 2 and provide its evaluation in a supplement to this SER.
Response
During the preoperational test program, the internals are subjected to the significant flow modes of normal plant operation. Before and after these flow tests, the internals are fully examined to determine any evidence of excessive vibrations. The observed vibrational characteristics of St. Lucie Unit 2 will be compared to those of the prototype plants as described in Subsection 3.9.2.4. A separate report will be sent to the NRC after the final examination that will contain the results of the program as well as identify any needs for corrective action. No FSAR change is required .
- 210A.32-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A. 33 The discussion of plant conditions in Subsection 3.9.3.1 of the FSAR requires clarification. The Loading Combination Method of response combination and allowable limits should be provided for all ASME Class 1, 2 and 3 components and their supports for each design and service condition.
Response
The plant conditions considered for the seismic qualification of the mechanical components are normal, upset, emergency and faulted. The Design Loading Combination for each plant condition is provided on FSAR Table 3. 9-5 while the corresponding allowable stress limits are provided on FSAR Tables 3.9-6 and 3.9-7. When dynamic loadings are present, the methodology of combining responses met the requirement of NUREG 0484, Revision 1, dated May 1980. As illustrated in the NUREG, a summation of the static loads are combined by the absolute sum method with the combined dynamic loads. FSAR Appendix 3.9A , provides the seismic loading criteria and the results of analysis which illustrates that the actual loads encountered are less than the ASME allowables. The loading combinations and design stress limits for ASME code class and NSSS components (except valves) are presented in Tables 3.9-15 and 3.9-16. The loading combinations and stress limits
- for valves, pumps, and all other class 2 and 3 components are presented in Tables 3.9-17 through 3.9-22.
See revised FSAR Tables 3.9-5, Sa, 15, 16, 17, 18, 19, 20, 21 and 22. Note: The use of emergency limits in Table 3.3-1 for other than the ATWS event is under generic discussion with respect to probability of occurrence of specific events. Refer to Item ,210A.55
- 210A.33-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.34 The methods of combining responses to the various loads listed in Subsection 3.9.3.1 of the FSAR are not defined. We will require a description of the methods used for the combinations of responses to all dynamic loads for all NSSS and BOP supplied ASME Class 1, 2 and 3 equipment, components and their supports. Our position on this issue is outlined in NUREG-0484, "Methodology for Combining Dynamic Responses," Revision 1 dated May, 1980.
Response
See revised loading combinations in Question 210A.33. No FSAR change is required *
- 210A.34-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.35
Response
The response of certain reactor coolant system components and their supports to postulated asymmetric LOCA loads needs to be addressed in accordance with NUREG-0609. Table 210A.35-l provides the status of the evaluation of components, structures and attachments to the RCS when subjected to asymmetric loads. Where the evaluation has been completed, the results have been shown acceptable. See revised FSAR Subsections 4.2.3.1.2.3, 3.7.3, 3.9.1.4,- 3.9.2.5 *
- 2fOA.35-1 Amendment No. 6, (9/81)
SL2-FSAR TABLE 210A.35-l ASSESSMENT OF STRUCTURES/ASYMMETRIC LOADS Assessment Evaluation Component/Structure Status Basis Reference Comments ECCS Piping In Progress Plant Specific FSAR 3. 9.1.4.5 Preliminary analyses pre-Analysis dict acceptable results. Final results will be reported to the NRC via 10 separate correspondence July, 1982. ECCS Piping Supports & Restraints In Progress Plant Specific FSAR 3.9.1.4.5 Preliminary analyses pre-Analysis dict acceptable results. Final results will be reported to the NRC via separate correspondence 10 July, 1982. CEDMS Complete Plant Specific FSAR 3.9.1.4.3 Analyses predict Analysis FSA.R 3.7.3.14 acceptable results. FSAR Amendment 9, March 1982. N ..... 10 0 > Reactor Internals Complete Plant Specific FSAR 3.7.3.14 Results Ul Analysis FSA.R 3.9.2.5 are acceptable. V1 I FSAR 3.9.5.4 Amendment 9, March 1982. N Fuel Complete Plant Specific FSAR 3.7.3.14 Analyses show acceptable Analysis FSA.R 4.2.3.1 result. FSAR Amendment 10, June 1982. IIT
!Z 0
1017W-2 SL2-FSAR TABLE 210A. 35-1 (Cont'd) Assessment Evaluation Component/Structure Status Basis Reference Connnents Reactor Pressure Vessel Complete Plant Specific FSAR 3.9.1.4.1 Complete Analysis Steam Generators Complete Plant Specific FSAR 3. 9. 1. 4. 1 Analysis Reactor Cpolant Pumps Complete Plant Specific FSAR 3. 9 .1. 4 .1 Analysis Reac.tor Vessel Supports Complete Plant Specific FSAR 3.9.1.4.1 Analysis Steam Generator Supports Complete Plant Specific FSAR 3.9.1.4.1 N Analysis ~ w Reactor Coolant Pump Supports Complete Plant Specific FSAR 3. 9 .1. 4 .1 VI I Analysis w Biological Shield Wall Complete Plant Specific FSAR 6.2.1.2 Analysis Steam Gen. , R C Pump Complete Plant Specific FSAR 6.2.1.2 Analysis Compartment Wall Complete Plant Specific Analysis RCS Main Piping Complete Plant Specific z Analysis 0
- DESCRIBE
- C - COMPLETE PLANT GEOMETRY P - IN PROGRESS c
CALCULATE CALCULATE MASS AND ENERGY
- SUB-COMPARTMENT
-- EVALUATE FOUNDATIONS AND WALLS c RELEASES c PRESSURES c
j I OBTAIN
~ SUPPORT LOADS
-- EVALUATE SUPPORTS c c START _
c DEFINE PIPE BREAKS c DESCRIBE ECCS PIPING GEOMETRY -. , , ,r 1, DEFINE STRUCTURAL - OBTAIN ECCS
~
ECCS
~
PREPARE c FOUNDATION PROPERTIES
~
c ANALYSIS DF~CS c COLD LEG PIPE MOTION p PIPING ANALYSIS PIPING EVALUATION REPORTS
;;o m
(I' "Tl r 0
;;o J~ j ~
.. c OBTAIN VESSEL MOTION p
EVALUATION OFCEDMS j ~ j ~ ' "Tl Gi -a-I () .-I 0> SIMPLIFY C: ;;oO r -c CALCULATE MODEL OF EVALUATION
- c m ;;o c
- 0 BLOWDOWN ~ OF Q :::E INTERVALS m~< INTERNALS LOADS mm c p cm
;;o (/I .,, ;;o c AND FUEL ~m~ '
r S?<> )> err >r z_ s: m DEVELOP STRUCTURAL ~ l>O> STRUCTURAL EVALUATION W >VI VI O-<
-I G) c: :r:
2 c s: MODEL OF INTERNALS
-- ANALYSIS OF
-- ANALYSIS OF FUEL OF FUEL
!...~~ -z -I m INTERNALS m () 2
-I c AND FUEL c
-I
;;o
-I 0
"' 3::
2
!=>
t
-0 z>
() Ol
-=
iO Oil
- GEOMETRY AND MOMENT CAPABILITY SIMILAR TO PALO VERDE
- PIPE BREAK+ SSE HEAD VELOCITIES LOWER THAN THOSE FOR PALO VERDE
- SINCE PALO VERDE HAS BEEN DEMONSTRATED ACCEPTABLE, ST. LUCIE 2 CEDM ARE EXPECTED TO BE DEMONSTRATED TO BE ACCEPTABLE
- ANALYSIS IS EXPECTED TO BE COMPLETED BY NOVEMBER 1, 1981 AMENDMENT NO. 6 (9/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 CEDM FIGURE 210A.35-2
- 600 PLASTIC INSTABILITY 510 KIP-INCHES l -----------------=-=----------
480 ASME LEVEL LOAD LIMIT (70% PLASTIC INSTABILITY) Cl) w 360
- c u
z cl..
;;2 I-zw
- !: 240 0
- 120 0 100 200 300 4000 500 ANGLE () X 1000 AMENDMENT NO. 6 (9/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 CEDM NOZZLE MOMENT CAPABILITY FIGURE 210A.35-3
- PRELIMINARY CALCULATIONS INDICATE THAT LINES 1A AND 1B ARE THE MOST SEVERELY LOADED
- COMPARISON OF INPUT MOTIONS WITH OTHER ECCS LINES PREVIOUSLY ANALYZED INDICATE THAT
- 1) PLASTIC ANALYSIS IS REQUIRED.
- 2) RESULTS ARE ANTICIPATED TO DEMONSTRATE ACCEPTABILITY.
- ANALYSIS IS EXPECTED TO BE COMPLETED BY SEPTEMBER 30, 1981 AMENDMENT NO. 6 (9/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ECCS PIPING FIGURE 210A.35-4
- 2 t-1 SUCTION ELBOW TOTAL CISP.
AT MIDDLE SECTION
- - - - ORIGINAL SHAPE
- - - DEFORMED SHAPE EXAGGERATED BY A FACTOR OF 5 EFFECT OF MOMENT OF 60.2x106 IN-LB NOTE: A CALCULATION OF THE DEFORMATION OF THE ST. LUCIE 2 RC PIPING WHEN SUBJECTED TO THE MAXIMUM MOMENT ALLOWED BY SECTION Ill, NB 3652 WAS PERFORMED. THE ATTACHED MATERIAL SHOWS THE RESULTS OF THAT CALCULATION.
AMENDMENT NO. 6 (9/81) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 RC PIPING DEFORMATION FIGURE 210A.35-5
\
SL2-FSAR
- Question No.
210A.36 Provide the stress limits and criteria to limit deformation and assure functional capability for Class 2 and 3 austenitic pipe bends and elbows.
Response
All ASME III Class 2 and 3 austenitic stainless steel pipe bends and elbows were reviewed and evaluated for functional capability based on General Electric Topical Report NEP0-21985. All systems have been verified to be acceptable except these noted below. The stress limit*of 1.5 Sy was exceeded by a piping elbow in both the Containment Spray System and the Fuel Pool Cooling System. The restraint/support system for these piping sys-tems are in the process of redesign to comply with this stress criteria. Two different elbows .on the Intake Cooling Water (ICW) system (CW-76 and CW-77) have a Do/t ratio exceeding the limit iisted ~n the GE report (120 vs. 100). However, the following information is provided to verify the adequacy of this existing piping design:
- 1) The coefficient of 0.75i used in the original stress calcu-lations is 6.445. The modified coefficient value based on NED0-21985 for Do/t > 50 is 19. 98. The primary stresses under faulted conditions calculated with the modified co-efficient of 19.98 are 21,816 psi and 18,142 psi for the 30" lines (CW-76 and CW-77, respectively). These are well below the allowable stress of 1. 5 Sy (i.e. , 41, 850 psi)
- Based on the low stresses in the elbows, no deformation is expected.
- 2) Each of the ICW lines (CW-76 and CW-77) contains a restric-tion orifice which limits the flow for pump protection.
The restriction orifice inside diameter of 12.25 inches represents a reduction in area of approximately 84 percent which is expected to be far greater than the area reduction induced by the piping stress at any bend or elbow. Thus, the restriction orifice is considered to be the limiting element to flow in the piping system. Therefore, since the calculated stress are well below the allowable stress (i.e. 21,816 vs 41 1 850) and since the system orifice is a more critical component for flow restriction than the potential deformation of the pipe bends or elbows, the functional capability of the system is assured. No FSAR change is required . 210A.36-l Amendment No. 6, (9/81)
SL2-FSAR Question No. 210A.37 Subsection 3.9.3.3 of the FSAR should include a more detailed description of the calculation procedures, which were used in the parametric studies for closed discharge systems.
Response
The closed discharge system of the safety and relief valves from the pressurizer are analyzed by a time-history dynamic analysis. As a more conservative, less complex approach, the closed discharge system of the safety and relief valves on safety-related auxiliary system such as Safety Injection System and Chemical and Volume Control System are analyzed by a static analysis. A transient hydraulic force equal to the freely blowing reaction force acting in bot.h directions with a dynamic load factor of 2 is applied to each long straight leg of the piping system for flashing service. For short intermediate straight leg (L) the unbalanced hydraulic transient fo~ce.will be 10 modified by a factor (1.51/c.t) to account for the valve opening time (t), piping Length (L) and the acoustical velocity of the propagating wave (c). For non-flashing liquid discharge system, the same transient hydraulic forces are applied to the valve outlet and the first elbow and the transient hydraulic force in the downstream of the discharge piping are considered in the analysis for the maximum momentum change of the fluid. No FSAR change is required *
- Amendment No. 10, (6/82) 210A.37-l
SL2-FSAR Question No *
- 210A.38
Response
Information should be provided in Subsection 3.9.3.3 of the FSAR relating to the various design and service loading conditions and combinations thereof, and the corresponding stress criteria used in the design for the mounting of pressure relief valves. The Design Stress limits as delineated in Subsection 3.9.3.1.1 and Tables 3.9-6 and 3.9-7 and Design Loading Table 3.9-5 are applicable to the mounting of pressure relief valves. No FSAR change is required *
- 210A.38-l Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A.39 The method of evaluating the structural response of the piping and support system stiffness in the dynamic analysis of these mountings should be discussed in Subsection 3.9.3.3 of the FSAR.
Response
In the dynamic analysis of "the Safety/Relief valve discharge piping system, the same stiffness matrix method as described in Subsection 3.7.3.1.1.2 is used for the representation of the structural response of the piping. Supports are modeled as a spring element with a finite stiffness. No FSAR change is required *
- 210A.39-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A. 40.1 a) b)
Compare AISC allowables with those of ASME Code Appendix 17. Show them comparable. How is reduced material yield strength at elevated
'temperatures addressed when using the AISC code?
c) ASME requires CMTR's and C of C's. What material documentation does AISC require?
Response
a) The allowable stresses in the AISC Code and ASME Code Appendix 17 have been reviewed and are similar in most respects. The following differences are identified:
- 1) At the contact surface of a weld producing a tension load in the through thickness direction of plates and eiements of rolled shapes, the allowable tension stress is limited to 60 percent of yield by AISC and 30 percent of yield by ASME Appendix 17.
Ultrasonic testing of both shop and field full penetration
,tee welds has been specified' for materials used in A/E supplied NSSS component supports where lamellar tearing may be a problem. In addition> ultrasonic examinations is required to be performed on flange prior to welding the flange to the web on built up girders and columns.
The materials used in piping supports are less than one inch thick .in the majority of cases, therefore lamellar tearing is not considered a problem. (Refer to "Significance and Control of Lamellar Tearing of steel plate in the shipbuilding industry" 1979).
- in addition, lamellar tearing generally initiates during or shortly after the welding process. This has not been identified as a problem with piping supports.
- 2) ASME Code Appendix 17 Load Increase Factor is 1.2 Sy/Ft not to exceed 0.7 Su/Ft. Ebasco has used an increase factor of 1.6 across the* board. This is applied to minimize redesign .and refabrication which might support design *.
- 210A.40.l-1 Amendment No. 6, (9/81)
SL2-FSAR The SL2 design criteria for the NSSS supports in the factored (faulted) condition allows the AISC allowables to be increased by 1.6 but less than .96xF or .90x critical buckling where applicable. ME~ requested that the NSSS supports in the faulted condition be reviewed for compliance to ASME III Sub. NF F-1370(a) wherein the
- allowable stresses in Appendix 17 may be increased by a factor ranging from 1.4 to 2.0 depending upon the material tensile and yield stresses. Critical elements in the NSSS supports have been reviewed and the actual design stresses are below those allowed by ASME Ill Sub. NF.
The only material used for seismic Category I pipeline supports for which S /Su exceeds 0.73 is SA-500 Grade B structural tubing. $upport designs hfhre been reviewed and in no cases are normal AISC allowable of 0.6 Sy exceeded. (Note F~=0.6 Sy). b) No reduction in yield strength is taken for application below 700 F, in .accordance with the AISC Manual of Steel Construction (
Reference:
Effect of Heat on Structural Steel). Austenitic material in seismic Category I component supports is for hypochlorination piping supports for the intake cooling water pumps. , The austenitic piping supports are located in the intake structure.* Austenitic material is also used for welded attachments on austenitic pipe and is designed to the same allowables as the pipe. c) Ebasco practice for seismic Category I requires Certificates of Compliance and CMTR's as apropriate. No FSAR change is* required. 210A.40.1-2 Amendment No. 6, (9/81)
- SL2-FSAR
- Question No
- 2l0A. 40. l _a) b)
Compare AISC allowables ~.Jith those of AsME Code Appendix 17. Show them comparable. How is reduced material yield strength at elevated temperatures addressed when using the AISC code? c) ASME requires CMTR's and C of C's. What material documentation does AISC require?
Response
a) The ~llowable stresses in the AISC Code and ASME Code Appendix J. 7 have been reviewed and are similar in most respects. The following differences are identified:
- 1) At the contact surface of a weld producing a tension load in the through thickness direction of plates and elements of rolled shapes, the allowable tension stress is limited to.60 percent of yield by AISC and 30 percent of yield by ASME Appendix 17.
Ultrasonic testing of both shop and field welds has been specified for materials used in A/E supplied NSSS component supports where lamellar tearing may be a problem *
- The materials used in piping suppor.ts are less than one inch thick in the majority of cases, therefore lamellar tearing is not considered a problem. In addition, lamellar tearing is most likely to occur during or shortly after the welding process. This has*not been identified
.as a problem with piping supports.
Further, the requirement to limit these stress to 30 percent of yield has been deleted from Appendix 17 of the ASME Code in the 1980 Edition.
- 2) ASME Code Appendix 17 Load Increase Factor is 1.2 Sy/Ft not to exceed 0.7 Su/Ft. Ebasco has used an increase factor of 1.6 across the board. This is applied to minimize redesign and refaQrication which might suppor~
design *
- 210A. 40 .1-1 Amendment No. 6, (9/81)
SL2-FSAR The SL2 design criteria for the NSSS supports in the factored (faulted) condition allows the AISC allowables to be increased by 1.6 but less than .96xF or .90x critical buckling where applicable. MEb requested that the NSSS supports in the faulted condition be reviewed for compliance to ASME III Sub. NF F-1370(a) wherein the allowable stresses in Appendix 17 may be increased by a factor ranging from 1.4 to 2.0 depending upon the material tensile and yield stresses. Critical elements in the NSSS supports have been reviewed and the actual design stresses are below those allowed by ASME III Sub. NF. The only material used for seismic Category I pipeline supports for which S /Su exceeds 0.73 is SA-500 Grade B structural tubing. Support designs are currently being reviewed to identify and justify cases where load increase factors exceed 1.2 SyfFt* This will be completed by September 11, 1981. b) No reduction in yield strength is taken for application
- _below 700 F, in accordance with the AISC Manual of Steel Construction (
Reference:
Effect of Heat on Structural Steel). Austenitic material in seismic Category I component supports is for hypochlorination piping supports for the intake cooling water pumps. The austenitic piping supports are located in the intake structure. Austenitic material is also used for welded attachments on austenitic pipe and is designed to the same allowables as the.pipe. c) Ebasco practice for seismic Category I requires Certificates of Compliance and CMTR's as apropriate. No FSAR change is required. 210A.40.l-2 Amendment No. 6, (9/81)
- SL2-FSAR
- Question No.
210A.41.l a) b) Justify the use of SRSS for combination of SSEI and SSED in the faulted condition. Provide the faulted allowable stresses for bolts *. c) Compare Table 3.9-5 with the loading tables of Subsection 3.8.3. d) Define the materials for which allowables are given in Tables of Subsection 3.8.3.
Response
a) Where the fundamental frequency of the piping system is beyond the resonant region of the supporting structure the SSE will be combined in the f~llowing manner: SSE =v1ssEr2 + SSED Where the piping fundamental frequency is not beyond the structural resonant region the SSE will be combined in the followin_g manner: *
- SSE = JssErl + lssEDI
- b) Faulted allowables for bolts are established as 1.6 X AISC allowables (normal).
Material. AISC Allowable Tensile Stress Allowable tensile A-325 40 Ksi A-490 54 Ksi stress For Faulted Condition 64 Ksi 86 Ksi
% of Ultimate 53% 58%
(1/2" - l" dia.) 61% (11/8"-11/2" dia.) c) Requirements for component supports are addressed in Subsection 3.8.3. Attached is Table 210A~41.l-l comparing the load combinations in Tables 3.9-5 and 3.8-12. d) Allowable stresses are based on Section 1.5 of AISC which in turn are based on ASTM material values. AISC factors of safety vary from 1.67 to 2.0 on yield strength. The 110 development of factors of safety is documented in the Commentary to the AISC Code
- No FSAR change is required.
210A.41.1-1 Amendment No. 10, (6/82)
SL2-FSAR
- Table 3.9-5
!llormal a) PO+OW TA3LE 210A.41.1-l Corre!:;ponding Table 3.8-12 Load Comb.
- 1) D+L Remarks PO and DW incor-
- 2) Df-L+E porated in L.
Normal b) TO 3) TO incorporated in To* Upset a) Po+DW+OBE* 4) RVO and FVC not ap-b) Po+DW+oBE+ plicable. OBE cor-RVO+FVC+T responds to E. PO, c) TI DW and T incorp. in L. TI incorporated in R and E. Thus, load combinations a,b, and c are enveloped by load combination 4 of Table 3. 8-12. Faulted a) Po+DW+DBE 8) D+L+Ta+R.a+Pa+ RVO and FVC not ap-b) PO+DW+DBE+RVO l.O(Yj+Yr+Ym)+E' plicable. DBE cor-c) Po+DW+nBE+FVC responds to E'. PO d) .Po+DW+FC and DW incorp. in L. e) PO+DW+DBE+FC FC incorp. in Pa, Yr and Yi* Thus, load comoinations a through e are enveloped by load combination 8. *
- 210A.41.1~2 Amendment No. 6, (9/81)
SL2-FSAR
- ~estion 210A.42 No.
Pr~vide the allowable buckling limits for ASME Class 1 linear and
- plate and shell type component supports subjected to faulted condition load. Also provide additional information concerning the design pf support bolts and bolted connections.
Response
All safety-related component supporting structures are designated "seismic Category l*" !Dad combinations and allowable stresses are in accordance w~th Standard Review Plant 3.8.3 and Standard Review Plan 3.8.4. The margin of safety for these structures is inherent in the design equations in the AISC Specifications. For linear and plate and shell type component supports subjected to the accident (faulted) load con~ition, the design stresses are limited to ninety percent of the critical buckling stress as applicable. For design of support bolts and. bolted connections, refer to the above paragraph. CE a) Buckling failure mode of the RCS supports is not credible due to the design characteristics of the supports *
- b)
. c)
The bolts in_ CE scope of supply (Steam Generator Skirt to Sliding Base) are designed to be below 70 percent of ultimate which, for ~he material, is less than 75 percent of yield
- Required Preload of interface Anchor Bolts (S.G. Snubber, Pressurizer Skirt) were specified to Ebasco.
NRC Position: Any support for a Class 1, 2 or 3 component in which the buckling stress)67 percent critical buckling must be justified as to why the margin against buckling failure is sufficient. As a result of discussions of this response during the review meeting, Question 210A.42.l was generated. This question and its response are attached. No FSAR change is required.
- 210A.42-l Amendment No. 6, (9/81)
Cl REACTOR VESSEL SUPPORT
- c o..J VIEWB-B FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 RV SUPPORT FIGURE 210A.42-1
ft INLET NOZZLE i...---------+--- (113-13/32) TO <t OF REACTOR VESSEL-COLD l,T/i/>-:;r-/-r-;.-.,..__j NOTE: THERMAL GROWTH TO HOT CONDITION IS 50°. THE BASE-PLATE MUST HAVE ADEQUATE SURFACE TO ASSURE FULL CONTACT WITH SLIDE, ITEM 4, UNDER COLD OR HOT CONDITIONS. T2 I I
--r;,* __ L .
<t R.V. SUPPORT, SOCKET, SLIDE AND EXPANSION PLATE SECTION C-C SUPPORT ARRANGEMENT AT INLET NOZZLE - 2 REQUIRED AMENDMENT NO. 6 (9/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 RV SUPPORT FIGURE 210A.42.:2
I I
- I I LvJ-N0.28 s
J FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 REACTOR VESSEL ARRANGEMENT FIGURE 210A.42-3
(106-1/4) o.o.
~-----
1, II ' I I 12" SURGE NOZZLE (116) 0.0. FLANGE ELEVATION PRESSURIZER ARRANGEMENT
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 PRESSURIZER & SKIRT FIGURE *21 OA.42-4
- 11+/-1/SSTUD PROTRUSION (158-7/8)
SEE NOTE27 r---.....__ __
© STEAM GENERATOR SLIDING
<> BASE INSTALLATION
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SG SLIDING BASE FIGURE 210A.42-5
COLD GAP TO BE DETERMINED DURING HOT FUNCTIONAL TESTING
- 3/64 = 1/64 HOT GAP@ 100% POWER SEE NOTE 17 SECTION B-B SCALE: 1" = 1'-0" SECTION D*D SCALE: 1" = 1'-0" 1/4 x 45" CHAM -~
TVP
*A*
ITEM 25-9.000 +/-.005 (12*1/2)-ITEM 25 }
..__...,_ _ _ ( 8*1/2)-ITEM 27 WITH RESPECT TO DATUM - - - ELEV.(25.41')
SEE REF. DWG. N0.1 SECTION F-F
- SCALE: 1" = 1'-0" SEE NOTE 6 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SG LOWER KEY FIGURE 210A.42-6
- li,.LUGON ST. GEN.
(10'-7-7/32"1 (7'-0-3/4"1 17*..o.3/4"1 Ii,. LUG ON ST.GEN. SNUBBER SURFACE rt+ 7'..()" COLD I Ii. I ,. I
~)- _y_
I v 749 62'-5"
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SG SNUBBER ARRANGEMENT FIGURE :210A.42-7
(1.8) 4SG LUG HOLE HOT POSITION COLD POSITION "A"+/- 1/8 SEE TABLE 1 TOTAL B~ DETAIL Y SCALE: 1/8
- ITEM NO.
TABLE 1 COLD INSTALLATION DIM "A"+/- 1/24 2 4'-8-1 /32" 3 4' 27/32" 4 4' 25/32" 5 4' 25/32" FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SG SNUBBER FIGURE 210A.42-8
- 1/16+/-1/64 GAP*TYP COLD
.500 TIP EXJ THEOR. REACTOR
*X- VESSEL
. Cl IE-13172-321-001 I COLD CLEARANCE TO NEAREST COLD STRUCTURE-CONCRETE OR METAL - - - - - ' CLEARANCE TO NEAREST 6" HOT OVERLAP STRUCTURE-OF KEV TO EXPAN- CONCRETE OR SION t, ITEM 11 METAL TYP-4 PLACES DETAILZ SCALE: 1" = 1'-0" SECTION A-A SCALE: 1" = 1'-0"
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SG UPPER KEY & LUG FIGURE 210A.42-9
{\
SL2-FSAR
- Question No.
210A.42.l a) What are the design values for buckling stresses? b) Commit to using 2/3 of critical buckling stress as a design limit and justify those cases where it is needed.
Response
a) The design values for buckling stresses are specified in
. Section 1. 5 .1. 3 of the AISC Code. This sec*tion identifies a minimum factor of safety of 1.67 which is in agreement with Appendix XVII of the ASME Code, Article XVII-2110 b).
b) Supports of ASME III Class 1, 2 and 3 components have been reviewed and it has been found that buckling stresses do not exceed 67 percent of critical buckling. No FSAR change is required *
- 210A.42.1-l Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
210A. 43 I~ addition, assurances must be provid~1d that stresses due to thermal expansion, thermal gradient and11 differential support movements have been included.
Response
The design and service conditions for supports and restraints included thermal effects and differential support movements as primary loads. CE - The RCS Analysis includes the effects of .thermal gradient, thermal expansion, and differential movement of supports. For Class 1, 2 and 3 vessels and pumps, nozzle loads include piping thermal expansion loadings. Vessels which are supported at both ends are provided with one fixed support and one slotted support to accommodate the axial thermal growth of the shell. No FSAR change is required *
- 210A.43-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A. 44
Response
We will also require an acceptable response to our request for pre-service inspection and testing information on snubbers. Inspection and testing of snubbers on safety-related,piping or components shall be performed as follows:
- 1) After snubber installation is completed but not more than.six months prior to the start of pre-core hot functional testing, a pre-service visual examinatlon shall be performed to verify that:
a) There are no visible signs of damage or impaired operability as a result of storage, handling, or installation. b) The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifications. c) Snubbers are not seized, frozen, or jammed. d) Adequate swing clearance is provided to allow snubber movement *
- e)
£)
If applicable, *fluid is to the recommended level and is not leaking from the snubber system. Structural connections such as pins, fasteners, and other connecting hardware such as lock nuts, tabs, wire, or cotter pins are installed correctly. If the period between the pre-service ex~mination and the start of hot functional testing exceeds six months due to unexpected situations, re-examination of items (a), (d), and (e) shall be performed. Snubbers which are installed incorrectly or otherwise fail to meet the above requirements shall be repaired or replaced and re-examined in accordance with the above criteria *
- 210A.44-1 Amendment No. 6, (9/81)
SL2-FSAR
- 2) During pre-core hot functional testing, snubber thermal movements for safety-related systems whose operating temperature exceeds 250 F shall be verified as follows:
a) During initial system heatup, snubber expected thermal movement shall be verified for any safety-related system which attains operating temperature. Verification shall be performed at RCS temperature plateaus of approximately 260 F, 360 F, 480 F, and 532 F. Snubber thermal movement shall be observed during cooldown and verified at ambient conditions after cooldown is completed. b) For those safety-related systems which do not attain operating temperature, observation and/or calculation shall be used to verify that snubbers will accommodate the projected thermal movement. c) Snubber swing clearance shall be verified for the temperature plateaus in (~) above
- Any discrepancies or inconsistencies shall be addressed as follows:
a) The snubber in question shall be removed from service or other interim action shall be taken to prevent system damage prior to proceeding to the next temperature plateau. b) The discrepancy or inconsistency shall be evaluated for cause and corrected prior to core load. c) The snubber in question shall be monitored again during heatup for post-core hot functionals to verify that the problem has been resolved. No FSAR change is required. 210A. 44-2 Amendment No. 6, (9/81)
- SL2-FSAR Question No.
210A.45 The thermal deflection problem of dissimilar materials is not covered and there is no information as to the allowable and actual deflections due to the various loading conditions. Design margins for stress, deformation, and fatigue should be presented and should be shown to be equal to or greater than those of other plants of similar design having a period of successful operation.
Response
In response to. the NRC questions pertaining to the design margins for functional reliability regarding the design criteria for non-pressurized components, the components outside the pressure boundaries are the coil stack, the pressure housing shroud, and the cooling shroud. All are designed to be a slip fit over the motor housing and are capable of being removed at temperature. A test was performed to verify this requirement, Dimensions and materials used for the St. Lucie 2 CEDMs are identical to those on operating reactors. All failure modes of non-pressurized active components will not affect the safety function of the CEDM. The coil stack is designed and has been tested to verify its capability .to withstand loss of air coolant flow for up to four hours without loss of function. *
- Parts within the pressure boundary, such as the motor assembly, have been sized for thermal deflections caused by dissimilar material so that clearances ~re available above the maximum design temperature of 650 F.
No FSAR change is required *
- 210A.45-l Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
210A.46 Verify that the control Element Assemblies (CEA's) can be inserted for an inlet break and the reaction can be stopped for an outlet break.
Response
As stated in Subsection 3.9.1.9.3 (Question 210A.35), insert-ability for the design *basis pipe breaks is not required. Pressure boundary integrity will be demonstrated. See NRC Question 210.35 *
- 210A.46-l Amendment No. 6, (9/81)
*sL2-FSAR
- Question No.
210A.47 Respons~ Identify the highest µsage factor and the location where it occurs in the reactor internals. The highest .usage factor for the reac~or internals is found to
- occur.in the core support barrel flange region and is less than
.15.
See revised FSAR Subsection 3.9.5.4.1 *
- 210A.47-1 Amendment No. 6~ (9/81)
SL2-FSAR
- Question No.
210A.48 The information presented in Subsection 3.9.6 of the FSAR does not
~ontain sufficient detail to demonstrate how the applicant intends to implement the in-service testing of pumps and valves requirements of ASME Section XI, "Rule for In-service Inspection of Nuclear Power Plant Components." We will require a description of the applicant's propqsed frogram on this subject. Guidelines on the type of information that we require is contained in Att~clunent 1 of the SER.
Response
The planned inservice inspection and testing programs were developed employing the R.G. l.~6, Revision 1, criteria for
, quality group classification~ and standards (Quality Group A is the same as ASME Class 1, etc.).
Section I. Valve Test Program Outline The valve te~t program shall be conducted in accordance with Subsection !WV of Section XI of the 1980 Edition of the ASME 8oiler and Pressure Vessel Code through Summer 1981 Addenda, except for specific relief reqµ~sted in accordance with 10CFR50.55a(g) (5) (iii) which is identified in Subsection I.G. The period for this valve test program starts April 21, 1983, and ends April 21, 1993. Section II. Pump Test Program Outline The pump test program shall be conducted in accordance with Subsection IWP of Section XI of the 1980 Edition of the ASME Boiler and Pressure Vessel Code through Summer 1981 Addenda, except for specific relief requested in accordance with 10CFR50.55a(g) (5) (iii) wtµch is identified in Subsection II.A. The period for the pump ~est progra~ starts April 21, 1983 and en4s April 21, 1993. No FSAR change is required *
- 210A.48-l Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
210A.49 We will require and acceptable response to our request for additional information on periodic leak testing of pressure isolation valves.
Response
FP&L recognizes that leak tight integrity of primary coolant system pressure isolation valves, whil~ an integral part of IST, will be reviewed in detail separately on an advanced schedule prior to OL. A separate response, as part of !ST, will be transmitted. No FSAR change is required *
- *210A.49-1 Amendment No. 6, (9/81)
SL2-FSAR .Question No. 210A.50 The appltcant should define the stress limits to which the welded attachments in the high and moderate energy piping systems are designed. Also describe the detailed stress analysis or tests which may.have heen performed to demonstrate compliance, including one example with a summary of the results.
Response
Stress analysis of.piping is performed for various loading conditions as presented in Table 3.9-5. This analysis also determines the loads on welder attachments which are used as seismic restraints. Local stresses on piping due to welded attachments are calculated using WRC Bulletin 107. The loc~l stresses are combined with other stresses determined by pipe
- stress a11alysiS in the welded attachment location and ASME code Section III allowable Stress Criteria are satisfied. A sample calculation and the summary of results are attached.
The trunnion analysis utilizes the CYLNOZ computer program to determine the adequacy of the piping system by including the local stresses into the appropriate loading combination. The loading combin'itions and allowable s.tress criteria are provided. The computer pro'lram uses the piping load in global coordinates and
- transforms these loads into the following.components:
(1) (2) (3) Radial/shear loads (P/VL, Ve) Circum./longit. bending (Mc/ML) Torsional moment (Mr) The details of local stress intensity calculation for unit load application for P, VL, Ve, Mc, ML, and Mr are provided. Membrane and bending stresses in hoop direction due to P, Mc and ML are calculated at four locations around the trunnion in outer and inner surface of the pipe. For example, AU and AL represent the stresi;;es at point A upper surface and lower surface respectively. Simtlar calculation for membrane and bending_ stresses in longitudinal direction due to P, Mc, and ML are calculated at the same eight locations of the pipe as described above. Shear stresses due to VL, Ve, and Mr are calculated at these locations. Using the above stresses, stress intensity is calculated at each of these eight locations. The largest stress intensity value is added with pressure stress (if applicable) and normal stress in the pipe, d.ue to corresponding load case and compared with the allowable stress. The results of the analysis conclude that the calculated stress are within the ASME allowable for all lo~ding combinations. See revised Table 3.9-5 *
- 210A.50-1 Amendment No. 6, (9/81)
*
- SL2-FSAR BP TECHNICAL SE:RVICES (BPT)
COPYRIGHT (C) 1979, BPT INTEGRAL WELDED ATTACHMENT ANALYSIS ASME BOILER AND PRESSURE VESSCL CODE SECTION *III CLASS 2 & 3 AND B31.l NUCLEAR ST. LUCIE PLANT UNIT 2 Material Tempera tu re Operating Design SU SM SY SC SH Young's Spec Deg .F Pressure Pressure PSI PSI PSI PSI PSI Modulus PIPE: A-106 GR.B 210. 100. 150. 60000. o. 31810. 15000. 15000. 27.67 TRUNNION: SA-106 GR.B 60000. 20000. 31810. 15000. 15000. 27.67 Run Pipe Trunnion Weld Pad OD Thck. OD Thck. Length Size Type Thck. Alpha 1 Beta 1 Alpha 2 Beta 2 8.625 0.322 3.500 0.216 1.000 o.o F o.o -22.000 o.o 68.000 90.000 TYPE RESTRAINT FRICTION ID = 0 COEFFICIENT OF FRICTION = N/A PT = A FIGR = N PARM OPT = P 0 I N INPUT MNS FX FY FZ MX MY MZ WGT 110. -2. -1301. 13. o. o. o. THERl 656. -587. 100. 6. o. o. o. THER2 421. -5 75. 138. 6. o. o. o. SETL 10. 11. o. o. o. o. o.
§"
(!) DBEl 3836. 4160. 768. 40. o. o. o.
"~
(!) OCCFC 15. 17. o. 1. o. o. o.
"z rt "'CD
~
* *SL2-FSAR NOTES
- 1) Stresses at pipe/attaclm1ent interface evaluated according to Bergen-Patterson report "Trunnion Analysis and Design;" B-P Report No. ER-90377.1-0.
- 2) Stesses in attachment evaluted according to Bersen-Patterson Report "Trunnion Analysis and Design.: B-P Report No. ER-90377.1-0.
- 3) Stress intensity values are calculated in CYLNOZ.
N "'0I z 0
SL2-FSAR SUMM~RY OF LOAD CASES Loa:! Service Code* B-P Case No Level Equipment No Equipment No Equation 1 Design 8 (A-1) SLP + WGT ( SH 2 Nonnal 10 (B-1.1) THERl ( SA 3 Normal 10 (B-1.1) THER2 (SA 4 Nonnal 11 (B-1.2) SLP + WGT + THERl .( SH + SA 5 Normal 11 (B-1.2) SLP + WGT + THER2 ( SH + SA 6 Nonnal lOA (B-1.3) SETL ( 2.4SC 7 Upset 9 (B-2) SLP + WGT ( l.2SH 8 Flnergency 9 (B-3) SLP + WGT ( l.8SH 9 Faulted 9 (B-4) SLP + WGT + OBEl + OCCFC ( 2.4SH 0 *Code refers to ASME Boiler and Pressure Vessel Code, Section III, Subsection NC, 1977. "'0 ..,.I z 0
*
- SL2-FSAR SlN?*fARY OF STRESS CALCULAT LONS A. Run Pipe Load Longitudinal Local Minimum Case P_rc ssure + Stress + Normal Total Al low No. Stress Intensity Stress Stress < Stress Failure 1 893. 1746. 110. 2750. 15 000.
2 o. 312 7. 656. 3783. 22500. 3 o. 3221. 421. 3642. 22500. 4 596. 7 519. 766. 8881. 37500. 5 596. 7313. 531. 8440. 3 7500. 6 o. 51. 10. 61. 36000. 7 596. 1746. 110. 2452. 18000. 8 596. 1746. 110. 2452. 27000. 9 596. 9972. 3961. 14 529. 36000. N 0 0 I
* *SL2-FSAR B. Trunnion Load Combined Axial and Bending Case Initial EQ Shear C. Weld No. FA SFA FBI SFBl FB3 SFB3 Ratio ( Allow Tm ax < FV Stress
< Allow. Failure 1 19086. 584. 19086. 18. 19086. 10. 0.032 1. 00 12. 12724. N/A N/A 2 21193. 45. 57258. 334. 57258. 852. 0.023 1. 00 527. 38172. N/A N/A 3 21193. 62. 57258. 327. 57258. 835. 0.023 1.00 516. 38172. N/A N/A 4 57258. 539. 57258. 316. 57258. 862. 0.030 1.00 529. 38172. N/A N/A 5 57258. 522. 57258. 309. 57358. 845. 0.029 1. 00 518. 38172. N/A N/A 21193. o. 57258. 6. 57258. 16. o.ooo 1.00 IO. 38172. N/A N/A 6
7 19086. 584. 19086. 18. 19086. 10. o.032 1.00 12. 12724. N/A N/A 25448. 584. 25448. 18. 25448. 10. 0.024 1.00 12. 16965. N/A N/A 8 9 38172. 928. 3817 2. 2516. 38172. 6071. 0.249 1.00 3784. 25448. N/A N/A ~ 0 \.n 0 I cr-r;- !t>
- s
~ ct>
- s rt z
0
SL2-FSAR Table computation sheet for local stresss in cylindrical shells with circular attachment.
- 1. Applied Loads**** 3. Geometric Parameters Radial Load, p 1000.000 lb Gamma 12.8929 Ci re. Moment, l11 MC 1000.000 in/lb Beta 0.3688 Long. !foment, M2 l1L 1000.000 in/lb Torsional Moment, MT 1000.000 in/lb Shear Load, vc 1000.000 lb Shear Load, VL 1000.000 lb
- 2. Geometry 4. Stress concentration due to Vessel Thickness T 0.322 in A) Membrane Load KN 1.000 Attachment Radius RO 1.750in B) Bending Load 1.000 Vessel Radius R = RM 4.151 in Interpolated Calculate Absolute Values Stressess (PSI. Around Circumference At Values of Stress and Enter Result AU AL BU BL cu CL DU DL NY/(P/R) = 1.188* KN(NY/(P/R))*P/(R*T) 888.95 -1328.13 -1328.13 -1328.13 -1328.13 -888.95 -888.95 -888.95 -888.95 MY/P 0.032** 1ZB(MY/P)*(6*P/T**2) 1839. 72 -1839. 72 1839. 72 -1839.72 1839. 72 -3090.05 3090.05 -3090.05 3090.05 NY/(l11/A)= 0.575 lZN(NY/(Ml/A))*Ml/(I*R) 280.81. o.o o.o o.o o.o -280.81 -280.81 280.81 280.81 MY/(~l/B)= 0.080 1ZB(MY/(Ml/B))*6*Ml/(I*T) 3023.52 o.o o.o o.o o.o -3023.52 3023.52 3023.52 3023.52
[)" NY/(M2/A)= 1.382 KN(NY/(M2/A))*M2/(I*R) 675.02 -675.02 -675.02 675.02 675.02 o.o o.o o.o o.o "':::~ MY/(M2/B)= 0.024 KB(MY/M2/B))*6*M2/(I*T) 925.30 -925.30 925.30 925.30 -925.30 o.o o.o o.o o.o "'rt Adel algebraically for summation z 0 of Y stresses, Sigma (Y) -4768.17 761. 88 -1567.53 261. 32 -7283.33 4943.82 -674. 67 -541. 61
*
- SL2-FSAR Table computation sheet for local stresss in cylindrical shells with circular attachment.
Interpolated Calculate Ab so lute Values Stressess (PSI. Around Circumference At Values of Stress and Enter Result AU AL BU BL cu CL DU DL NX/(P /R) 1. 775* KN( NX/P /R)) *P / (R *T) 1328.13 -888. 95 -888.95 -888.95 -888.95 -1328.13 -1328.13 -1328.13 -1328.13 MX/P o. 052** KB(MX/P) *( 6*P /T**2) 3090.05 -3090.05 -3090.05 -3090.05 3090.05 -1839.72 1839. 72 -1839. 72 1839. 72 NX/(Ml /A)= 1.236 KN( NX/(Ml /A) )*Ml/ ( l*R) 603.84 o.o o.o o.o o.o -603.84 -603.84 603. 84 603. 84 MX/(Ml/B)= 0.040 KB(MX/(Ml/B))*6*Ml/(I*T) 1509.91 o.o o.o o.o o.o -1509. 91 1509.91 1509. 91 -1509. 91 NX/(M2/A)= o. 563 KN(NX/(M2/A))*M2/(I*R) 275.07 -27 5. 07 -275.07 275.07 275.07 o.o o.o o.o o.o MX/(M2/B)= o. 041 KB(MX/(M2/B))*6*M2/(I*T) 1561. 46 -1561.46 -15*61. 46 1561. 46 -1561. 46 o.o o.o o.o o.o Add algebraically for summation of X stresses, sigma (X) -5815.54 3487.49 -2142.47 914. 72 -5281. 59 1417.67 -1054.10 -394.47 Shear Stress (Load VL) TAUl VL/PIR 564.88 o.o o.o o.o o.o -564.88 -564. 88 564.88 564. 88 N ..... Shear Stress (Load VC) TAU2 VC/PIR 564.88 564.88 564.88 -564.88 -564.88 o.o o.o o.o o.o 0
- " 161.39 161.39 161.39 0 Shear Stress (Torsion MT) TAU3 MT/PIRT 161. 39 161.39 161. 39 161. 39 161.39 161.39 I
00 Add algebraically for summation of shear stress, TAU 726.27 726. 27 -403.48 -403.48 -403.48 -403.48 726.27 726.27 s *** 6187.23 3668.94 2350.41 1107.18 7361.59 4989.39 1615.03 1459. 98 Combined stress intensity, Notes: ~ (1) A = R**2*Beta (2) B = R*Beta (3) I=R*Beta*T (4) Y=PHI (5) .PIR=3.14*RO*T (6) PIRT=6.28*R0**2*T. ~ c.. 8
- NY/(P/R) Yields the membrane PHI-Stresses at locations C and D and the X-stresses at locations A and B.
(1)
- s NX/(P /R) Yields the membrane PHI-Stresses at locations A and B and the X-stresses at locations C and D.
rt
** MY/P Yields the bending PHI-Stresses at locations A and B and the X-stresses at locations C and D.
z MX/P Yields the bending PHI-stresses at locations C and D and the X-stresses at locations A and B. 0 Figures lC-1 and 2C-l are used for MY/P and MX/P at the transverse locations, respectively.
*** S is computed as the largest difference between all pairs of principal stresses, wit.h the normal stress assumed zero.
**** For sign convention, etc., refer to WRC Bulletin 107, March 1979 Revision of August 1965.
FX = + 5580 - 6355 (LB.)
.MX = 0 (FT.-LB.)
FY + 2207
,, FY =
0 (LB.) MY = (FT.-LB.) ( ~MY FZ = 0 (LB.) MZ = 0 (FT.-LB.) r DISPLACEMENTS: FX
\J DX = 0 MX MZ DY = 0 DZ = -0.198 (WORST THER-FZ 0.6 FRICTION FORCE: MAL+ DEAD
'LOAD)
LOADING CONDITION FX (LB.) FY (LB.) FZ (LB.) MX (LB.) MY (LB.) MZ (LB.) WORST THERMAL CASE NO. 1 -783 100 SAFETY RELIEF VALVE THERMAL ACCIDENT CASE NO. 18 -767 138 DEAD LOAD CASE NO. 21 -3 -1301 SETTLEMENT CASE NO. 37 14 0 STEAM HAMMER DBE STATIC "G" LOAD CASE NO. 74 5547 768 DBE ANCHOR RESTRAINT MVT. CASE NO. 72 22 0 AMENDMENT NO. 6 (9/81)
- !'}
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2
SUMMARY
OF LOADS ACTING ON RESTRAINTS - COMPONENT COOLING FIGURE 210A.50-1
JOB# 7446 A. STRESS ISO CC-179-38 REV. ST. LUCIE #2 *
- STRESS CAL.C 2080 REV. 0 STATION FAB # 32 STRESS PT 12 FILE# TA-96 B.STRESS ISO REV.--
SYSTEM COMPONENT COOLING STRESS CALC REV. MARK# CC-2080-12 STRESS PT DIMENSION DATA RUN PIPE: ATTACHMENT: O.D. = 8.625" SIZE= 3.5" OD SCH. 40 WALL THK = .322" LENGTH= 0'-4.15/16" MATERIAL= A-106 GR.B. THK. = .216" MATERIAL= SA-106 GR. B. DESIGN.INFO CODE CS-1 PIPE CATEGORY 3 OPER. TEMP. 210 OF DESIGN TEMP. 250 OF OPER. PRESS 100 PSIG DESIGN PRESS 150 PSIG DESIGN CRITERIA LOADING CONDITION HGR. COMP. STR. STL.
- 1. FAULTED 2 ..
3. RUN PIPE STRESSES PIPE STRESS IN PSI LOADING CONDITIONS STRESS CALC. = 2080 STRESS CALC. = TOTAL STRESS PT. = 12 STRESS PT. = CASE NO. 1 THERMAL EXPANSION 656 CASE NO. 18 THERMAL EXPANSION 421 CASE NO. 21 DEADWEIGHT 110 CASE NO. 37 SETTLEMENT 10 CASE NO. 72 SEISMIC-DBE 15 CASE NO. 74 STATIC SEISMIC DBE INERTIA 3836 AMENDMENT NO. 6 (9/81) FLORIDA POWER & LIGHT COMPANY
- ST. LUCIE PLANT UNIT 2 DATA SHEET INTEGRAL PIPE ATTACHMENT ANALYSIS
.FIGURE 210A.50-2
y DIS.PLACEMENTS:
- Dx = 0.0 Dy Dz
=
=
0.0 0.198" z
~
II
~
_........_____ .. P.P. = 19" ALL RESTRAINTS ARE RIGID RODS - APPLY 75% GIVEN Fx IN EACH Fx RESTRAINT. OUT OF PLANE LOADS DUE TO 'Fx RODS:
- Fz Dz)
= .75 ( P.P.
OUT OF PLANE LOADS DUE TO Fy ROD Fx = .75 0.198
----:rs.- Fx Fx
= 128.
Fz IS ACTUALLY APPLIED 11.25" FROM ct_ OF RUNPIPE BUT APPLIED 7.00" FROM ct_ OF RUNPIPE IN THE MODEL. IN ORDER TO PRODUCE THE CORRECT MOMENTS AT THE OUTSIDE FIBER OF THE RUNPIPE, THE OUT OF PLANE LOADS DUE TO Fy ROD
' ' 1125" - 43125" WILL.BE INCREASED BY THE FACTOR. 7,, ._ . ; ,, . THIS WILL CORRECT THE 4 3 25 MOMENTS BUT MAKE THE SHEAR Vz CONSERVATIVE.
= 11.25" - 4.3125" ( Dz)
Fz 7" - 4.3125" P.P .. Fy
= 2 58 0.198 F
. 50.5 y Fy 98.8 AMENDMENT NO. 6 (9/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 OUT OF PLANE LOADS FIGURE 210A.50-3
Item Quan. Part No. DESCRIPTION Wgt. 1 1 cs B4 S = 4' - 4' (PER 602 STD.I (FIELD CUT TO SUIT) (SA-36) 56 2 *2100 2-5 RIGID TELESCOPING STRUT ASSY. R/3 (P-P = 50Y:zI 3 2003 1-5 END ATTACHMENT R/1 4 2 2600 7-3" </>PIPE CLAMP R/2 5 2 2100 7 RIGID TELESCOPING STRUT ASSY. R/3 (P-P = 19") 6 STR TS 4" x 4" x 5/16 x 4' - 6Y:z LG. (SA-500 GR.Bl 70 7 STR TS 3" x 3" x 3/16 x 6' x 7" LG. (FIELD CUTI (SA-500 GR.Bl 46 8 2 cs 16%" x 3%" x Y." STIFF R. (CLIP CORNERS) (SA-36) 17 MAT'L. TRACEABILITY REQ'D. TYP
\.W 12 (IP) .
~lie'1 I1 Y(+2207)
- ~x
(+5580) (-6355)
~~=-< TYP CLASS I SEISMIC SEISMIC RESTRAINT Stress lso. #CC-179-38/
Stress Cale. #2080/0 Stress *Pt. # 12.. D.W. LOAD= 1301 i--.-1-W8x24 (IP)
--<---W18x50 (IP)
SEE ENLARGED PLAN SHT. 3/4 SECTION A-A
~ NORTH oo REAC- 1800 TOR LOCATION PLAN AMENDMENT NO. 6 (9/81)
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 COMPONENT COOLING SHEET 1 OF 4 FIGURE 210A.50-4
Item lluan. Part No. DESCRIPTION Wgt. 9 2 CST 3" lj>SCH/40 x O' -4 15/16" LG. TRUNNION 6. (SEE DET. SH. 4 OF 4) (SA-106 GR.Bl 10 2 cs 4Y." x 4Y." x Y." R. (SA-36) 6 11 2 cs 3 3/8 x 1 7/8" x 1/4" STIFF R. (CLIP CORNERS) 1 (SEE DET. B-B) (SA-36) 12 1 cs 7" x % x 3" STIFF R. (CLIP CORNERS) (SA-36) 2 APPROX. WGT. 223 W 18 x 50 (IP)
- . ,r
[J==-- . -- ~--- -1 J iiu====----------------------------~' -- I 3/16 ~ =-~ I
~
SECTION B-B FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 COMPONENT COOLING SHEET 2 OF 4 FIGURE '210A.50-4
- "ill( NORTH W 8 x 24 (IP)
W 8 x 24 (IP)
- W 12x31 (IP)
ENLARGED PLAN AMENDMENT NO. 6 (9/S 1) . _FLORIDA POWER & LIGHT COMPANY ST* LUCIE PLANT UNIT 2 COM PON ENT COOL I NG SHEET 3 OF 4 FIGURE 210A.50-4
- } TRUNNION 1.75" I TRUNNION r NOTE 1 c,,
,~--
' I NOTE 2
~~I ______)_*-'-'i 1
--'lo'-_.____.__ _ PROCESS I (\ PIPE-8</>
L__ *__ _jJ 'I O" FOR t ~ %" O" TO 1/8" FOR t> %"
- NOTES:
SECTION A-A
- 1. SPECIFY SHOP OR FIELD WELD ON THE DETAIL DWG.
- 2. USE 22%0 BEVEL FOR TRUNNION SIZES 1" THRU 2%
USE 30° BEVEL FOR TRUNNION SIZES 3" & LARGER. AMENDMENT NO. 6 (9/81) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 COMPONENT COOLING SHEET 4 OF 4 FIGURE 210A.50-4
. SL2-FSAR
- Question No.
210A. 51 Provide the following information for an ASME Code Class 1 piping system identified in FSAR 3.6A. (a) Calculated stress intensity, calculated cumulative usage factor and the calculated primary plus secondary stress range for each point at which these parameters we calculated. (b) Results in tabular form and correlated with node points identified on a sketch of the system.
Response
The calculated stress intensity, accumulated usage factor and primary and secondary stress range is provided in Table 3.6C-l. See new FSAR Tables 3.6C-l and 3.6C-2
- 210A. 51-1 Amendment No. 6, (9/81)
SL2-FSAR Question No
- 210A.52 The postulated guillotine breaks may either be based on complete severence or limtted separation. The applicant should identify which type of break was assumed for all postulated guillotine breaks not included as part of CENPD-168. In case limited separation was assumed, justification should be provided.
Response
For all high energy system failures not inclued in the CENPD-168 report, the pipe rupture and jet analysis conservatively assumed guillotine breaks with complete separation. For limited area breaks see* revised Subsection 3.6.2.1.1. All other guillotine break locations are full area *
- 210A.52-l Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A. 53 The applicant, in describing analytical methods to define forcing functions, should indicate (a) the assumed loading conditions at the time of the postulated pipe break, and (b) the rise time for the initial pulse. If the rise time was greater than one millisecond, justification should be provided.
Response
(a) For high stress points determining the pipe stress analysis considers plant normal and plant upset conditions. The pipe rupture and jet impingement analysis assumes a normal operating condition of 100 percent power prior to the pipe break. (b) The pipe rupture analysis assumed a break opening time equal to or less than one millisecond. No FSAR change is required *
- 210A.53-1 Amendment No. 6, (9/81)
SL2-FSAR
- 210A. 54
Response
What is the status of St. Lucie Unit 2 with respect to the SG feedwater ring event recently experienced at SONGS? Due to design differences between SONGS and St Lucie Unit 2, feedring damage is not expected to occur at St Lucie Unit 2. Test 10 guidelines have been developed to verify that these design differences will preclude feeding damage. No FSAR change is required *
- ---~
210A.54-1 Amendment Noe 10, (6/82)
SL2-FSAR
- Question No.
210A.55 Justify the use of ASME Code Level C service limits for the small feedwater line break transient by showing that the probability of a small feedwater line break is less than or equal to that for an ATWS event (reference: NRC letter James P. Knight to Paul s. Check, dated July 29, 1981).
Response
The response to Question 440.81 (r) shows that for St. Lucie Unit 2 the probability of the small feedwater line break with loss of offsite power, referred to in that r~sponse~ is less than io-7 per plant year (sufficiently low* to justify use of Level C service limits), Note: In the USNRC memo Knight to Check dated July 29, 1981, the review of the event probability* of occurrence is indicated as required to be reviewed by the NRC probabilistic review group (not MEB). The outcome of* tfiat review can affect Question 210A.33 *
- 210A. 55-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 210A.56
Response
Provide information on the use of concrete expansion anchors on St. Lucie Unit 2. The use of concrete expansion anchors.on St. Lucie 2 is limited. Concere anchors are used only on a "last resort" basis. Before using concrete expansion anchors the feasibility of using the following methods are studied:
- 1) embedded plates
- 2) Bridge between embedded plates
- 3) thru bolts If these methods prove unrealistic, concrete expansion joints are used.
When using concrete expansion anchors in a wall, a factor of safety of 15 is used for seismic applications. Prying effects have been studied and are accounted for in the design. Expansion anchors ultimate capacities are established by a field testing program. See Design Criteria (attached), Sample Calculation (attached), and NRC IE Bulletin 79-02 letter Dr. Uhrig to Mr. O'Reilly L-79-180
- dated July 2, 1979 *
- 210A. 56-1 Amendment No. 6, (9/81)
SL2-FSAR
- Question 56 - IE Bulletin 79-02 FLORIDA POWER AND LIGHT COMPANY ST LUCIE UNIT NO. 2 DESIGN CRITERIA
.1 J.:__2 CONCRETE EXPANSION TYPE ANCHORS 12
- 210A.56-2 Amend~ent No. 12, (8/82)
SL2-FSAR
- l. Scope
- This document defines the criteria to be used in the* design of concrete.
expansion-type anchors, meeting the requirements of Ebasco Specification FL0-2998.469, "Drilled-in Expansion Type Anchors in Concrete", for use.in seismic Category I and non-seismic pipe support structures. The loads, load combinations and allowable stresses are defined, and procedures are set forth to account for base plate prying action and shear-tension interaction on the anchors.
- 2. Loads The following loads are considered applicable to the design of concrete expansion anchors:
2.1 D = Dead loads, including the weight of stationary structures, piping and equipment. . 2.2 L Live loads, including any movable equipment loads and loads due to the operation of equipment. 2.3 W Hurricane Load. 2.4 Feqo =Operating Basis Earthquake (OBE), including effects of differential movement of supports. 2.5 Feqs Design Basis Earthquake (DBE), including effects of differential movement of supports. 2.6 Ro = Pipe reaction loads during normal operating or shutdown conditions. 2.7 Ra Pipe reaction loads *under thermal conditions* generated by a
- 2.8 2.9 2.10 T0 Ta postulated pipe break, including R0
- Thermal effects and loads during normal operating or shutdown conditions.
Thermal effects under conditions generated by a postulated pipe break, including T0
- Yr Reactions loads generated by a postulated pipe break,
- including an appropriate factor to account for the dynamic nature of the load.
- 3. Load Combinations The following load combinations shall be used to compute the factored tensions, moments, and shears for use in the design of concrete expansion anchors, the combinations represent those used in the design of concrete structures, with additional load factors added which, in combination with the appropriate capacity reduction factors, result in minimum factors of safety of 4 for static loads, 5 for impactive loads and 15 R2 for vibrating loads.
3.1 l.4D + l.3Ro+ l.3T 0 + 1.71 3.2 l.4D + l.3RQ + l.3T0 + 1.71 + l.7W 3.3 l.2D + 1. 7W 3.4 l.4D + L.3Ro+ 1.3T0 + 1.71 + 3.8Feqo R2 3.5 l.2D + 3.8F
- R2 3.6 1 , OD + 1. OR~q~ L OT 0 + 3 . BF eq s .+ 1 , OL 3.7 l.OD + l.OL + l.O~a + l.OTa +l.25Yr + 3.8Feqs R2 210A. 56-3 Amendment No. 6, (9/81)
SL2-FSAR
- 4. Allowable Stresses The ultimate loads carried by concrete expansion anchors shall be obtained from Specitication FL0-2998-469, Table 4.03, under the columns headed "Static Tensile Ultimate Load" and "Static Shear Ultimate Load," modified by a capacity reduction factor~of 0.25. If the center-to-center anchor spacing is less than 12 anchor' diameters, the ultimate capacity shall be reduced linearly to 50 percent capacity at the minimum center-to-center anchor spacing of 6 anchor diameters.
S. Determination of Prying Forces The presence or absence of prying forces acting on the anchors due to a direct tension force is determined by the following formula: a= 1 0 where T factored direct tensile load per anchor b = distance from the face of the attachment where the load is applied to the centerline of the bolt, minus one-half bolt diameter w width of the base plate contributing to each bolt t = thickness of base plate
- u. = yield stress of base plate l = ratio of the net area to the gross area of the base plate cross section at the bolt line a= ratio of the moment in the base plate at the bolt line to the moment in the base plate at the face of the attachment The presence or absence of prying forces acting on the anchors due to an applied moment is determined by the following formula:
a = 1 0
- --2b d
where M0 = factored applied moment per anchor in tension d = dimension of attachment in the direction of the moment b,w,t,uy,oanda= as defined above 210A. 56-4 Amendment No. 6, (9/81)
SL2-FSAR In the case of combined tension and bending, uy in each, of the above two
- equations shall be reduced so that the same u is computed from each equation, and the sum of the. uy's used is less than or equal to the actual yield stress of the base plate.
If a is found to be zero or less, there is no prying force present.
***If ot is found to* be greater tha~ one, a value of one is to be used.
The bolt force due to a direct tensile force is calculated by the formula: Tb Q = ao = Cl+o:o) a where Q = prying force a = distance from the edge of the base plate to the centerline of the bolt, plus one-half bolt diameter, but not more than l.2Sb a, 0 , T, and b = as defined above If a is zero or less, Q = O. The bolt force due to an applied moment is calculated by the formula: Q = M0 _.,.._a_o_...,._ __,._ b (1 +ao )d+2b a where all terms are as defined above. If a is zero or less, Q = O. Prying forces shall be computed assuming one-way bending in the base plate in each direction. If the base plate is provided with stiffeners between the attachment and the bolt line, the prying force in the direction of the stiffeners may be assumed equal to zero. The total prying force shall be computed as the square root of the sum of the squares of the prying forces in each direction.
- 6. Determination of Anchor Sizes Concrete expansion anchors shall be selected such that:
B v ~ 1 Ba + Va where B = total factored tensile load per anchor, including prying forces, if any Ba= allowab~e anchor tensile capacity, including capacity reduction factor and effects of close spacing V = total factored shear load per anchor Va= allowable anchor shear capacity, including capacity reduction factor and effects of close spacing 210A. 56-5 Amendment No. 6, (9/81)
EXAMPLE OF CONCRETE EXPANSION TYPE ANCHORS DESIGN SUPPORT S7 (CONT.) SEISMIC CLASS I
- FOR ANCHORAGE WITH EXP. BOLTS ly = Iz = 4 x 3.52 = 49 0.01 K-IN X2 0.2K Ix= 4q x 2 = 98 P = 0.2 + (0.13 3 0.04) x3.5 4 49
= o.002K1soLT H = J(o.01 + 0.01 x3.5)2 + (0.07 + 0.01 x3.5)2 4 98 4 98
~
___. lo-
~
CW)
= 0.018K/BOLT 1/2 q_ I I I I ~
CW) LOAD FACTOR ,,;, 3.8 - ~
,___. ~
~
B = 0.062x3.8 = 0.24K/BOLT 1% ,_ '- 3%
- - -- 3% - -- - 1%
v = o.018x3.8 = o.01K1soLT CHECK PRYING ACTION FOR 1/2" q, BOLT a=! (4x0.19x2.0625 _1) = .!(1.254 _ ) _ _ _ _ _ ---::,- _. _ _ _ _ _ 1
-'1' 6 . 5x0.52aYT 6 aYT . \..!..)
a= !(4x0.247x2.0625 _ 2x2.0625 _ )=.1(0.686 _ 2 1 737
)- ______ f2' 6 5x0.52ayM 1x2.375 2.375 6 aYM1 * . ~
a=1(4x0.076x2.0625 _ 2x2.0625_1) = .1(0.211 -2. 737)- _ -. __ ----@ 6 5x0.52ayM 3x2.375 2.375 6 aYM3 FROM
\V f3\
f.I\ & \V 1.254 = 0.211 -1.737- - 0 YT aYM3 . .
- - - - - - -* - - --© AMENDMENT NO. 6 (9/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 INSTRUMENT SUPPORT FIGURE 210A.56-1
- SUPPORT S7 (CONT.)
FROM & aYM1=3.25ayM3--------0 ayT + aYM1 + aYM3 = 36 aVT = 3 6 - 4 . 2 5 a V M 3 - - - - - - - - - - - © SOLVE© & aYMa = 0.119 aYM1 = 0.387 aYT = 35.494
*ex< o NO PRYING ACTION Ba= 0.25x 7.1 = 1.775K/BOLT
- Va = o.25 x 6 = 1.5k/BOLT JL + )!__
Ba Va
= 0.24 + 0.07 1.775 O.K.
1.5
<1 BY INSPECTION, 1/2" fl IS O.K.
USE 4-1/2" </>EXP. BOLT WITH fl 10 x 10 x 1/2 AMENDMENT NO. 6 (9/81) FLORIDA POWER & LIGHT COMPANY ST~ LUCIE PLANT UNIT 2 INSTRUMENT SUPPORT FIGURE 210A.56-2
SL2.;.FSAR
- Question No.
210A. 57 FP&L will confirm that a safety factor of.15 was used across the board on safety-related application piping hangers*. 'This confirmation will be submitted formally to the NRC.
Response
A safety factor of 15 was used across the board (i.e. applied to* dead, therm.al and seismic loads) in t~e* design of the small-bory safety-related pipe supports which require the use of expansion anchors. There are 79 such supports at St. Lucie Unit 2. All the small-bore safety-related pipe supports were design~d by Ebasco Engineering. The large-bore safety-related pipe supports at St;. Lucie Unit 2
*were designed by Bergen-Paterson. Thirty-three of .these supports required the use of concrete expansion anchors. It ha& been determined that Bergen-Paterson did not use a factor of safety of 15 in their design of these supports. In order to verify the design adequacy of expansion anchor applications for large-bore safety-related pipe supports, FP&L proposes to analyze IQ of the 33 Bergen-Paterson designs which represent the worst cases u~ing the ANSYS computer program. The results will be compa~ed to the
- ultimate concrete expansion anchor capacities to determine the actual factors of safety .
210A.57-1 Amendment No. 13, (2/83)
SL2-FSAR Question No *
- 210A.58
Response
FP&L will do a calculation of the anchorage design using a "non-proportional" uy and submit the results to the NRC
- The previously submitted hand calculation sample has been revlsed to reflect the use of a single value Y' as requested by the NRC. See Table 210A.58-l and Figure 210A.58-1 *
- 210A.58-1 Amendment No. 13, (2/83)
SL2-FSAR TABLE 210A. 58-1 INSTRUMENT SUPPORTS Support S7 For Anchorage with Expansion Bolts (See Figure 210A.58-l)' Iy = Iz = 4 x 3.5 2 = 49 Ix = 49 x 2 = 98 0.2 (0.13 + 0.04) x 3.5 k p = + 49 = 0.062 /Bolt 0.01 x 3. 5)2 H = 98 o.018k/Bolt I.Dad Factor = 3.8 B = 0.062 x 3.8 = 0.24k/Bolt v = 0.* 018 x 3.8 = o.01k/Bolt Check prying action for 1/ 2" 0 Bolt
- ~ = io - 2 x 1
a1 = o.8625 10 11 R- = o.86 2 s (4 x 0.19 x 2.065 5*x o.5 2 x 36
- 1\ = -1.12 I <0
~ 4 x 0.247 x 2.0625 - 2 x 2.0625_ -1\ = 3.15 <0 a2 = 0.8625 \s x 0.52 x 36 x 2.375 2.375 J
*~. 4 x 0.076 x ~.0625 - 2 x 2.0625 -1) = 3.17 (0 a3 =~~ x o.s2 x 36 x 2.375 2 .37 5 No Prying Action Ba= 0.25 x 7.1 = 1.775k/Bolt Va = 0.25 x 6 = 1.5k/Bolt B + V = 0.24 + 0.07 ( 1.0 Ba Va T:iT5 T.r
- 210A.58-2 Amendment No. 13, (2/83)
SL2-FSAR TABLE 210A.58-l (Cont'd) Support 87 (Cont'd) By inspection, 1/2" t i s o.k. Use 4 - 1/2" 0 Exp. Bolt with*i.lO~x 1/2 x 0 1 .-10 210A. 58-3 Amendment No. 13, (2/83)
- 0.01K-IN X1 0.13K-IN (A) 1%"
'f
- I I -
3%" j ~
'f
.J '
1/2 R.-> I I 3%"
'l 1%"
1%"-,... - 3%"-
- p
-.3%" -
- - 1%"
(B) AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 INSTRUMENT SUPPORT FIGURE 210A.58-1
SL2-FSAR Question No
- 210A.59 FP&L will determine if it is necessary to develop a sample calculation for base plates with other than four bolts for either single axis or biaxial bending.
Response
Seven cases of support designs utilizing an expansion anchor detail of other than 4 bolts have been identified. Four of these designs, judged to be the most severe, were analyzed using the AN SYS finite element computer program. All but one resulted in a factor of safety of at least 15 for bolt tension and shear. The sole exception was a factor of safety 13.1 for bolt tension. Ultimate capacity values used are based on actual tests performed at site. See Figures 210A.59-l, 2, 3 and 4 *
- 210A.59-1 Amendment No. 13, (2/83)
- REF. AXES z
0 0 0 tx*::
~j R.. %" x 12" x 21"
/x
~~
N~ enI %"</>PHILLIPS
~
0 WEDGE ANCHOR 1, FORCE IN KIP y TS 3x3x~
. 'f MOMENT IN KIP-IN.
0 0 0 :: -~
~'
Y' ,..... 1%"
,~ 0'-9 ..,. 0'-9 1%" Z' APPLIED LOADS:
BASED ON F.S. OF 15 MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ALLOW. BOLT ALLOW. BOLT (TENSION) (KIP) (KIP) (KSI) TENSION SHEAR 0.47 0.22 1.91 0.77 0.99 COMPUTER OUTPUT: . MZKAJD4 dt. 6/29/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (CW-3006-6) FIGURE 210A.59*1
- REF. AXES y
tx*--
~
*1/:- 0 1-0 0 Q
-- R.. 5/8" x 10" - 18" N
I CD 0
-TS3x3x%-
Q. 1 0 0 I
~
Q
%"¢PHILLIPS WEDGE ANCHOR FORCE IN KIP CjJ MOMENT IN KIP-IN.
I Q
-- f Y' ,.. o*-a .. ,.. 0'-6 . ,.. 0'-6 .., .. o*-a .. , Z' APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 8 0.0 -0.616 0.90 2.776 0.0 0.0
- RESULTS:
0 MAX. BOLT LOAD 0.0 MAX. BOLT SHEAR
-0.412 0.90 MAX. R.. STRESS 1.796 0.0 BASED ON F.S. OF 15
~
0.0 ALLOW. BOLT ALLOW. BOLT (TENSION) (KIP) (KIP) (KSI) TENSION SHEAR 0.54 0.23 2.33 0.47 0.40 COMPUTER OUTPUT: MZKAJKW dt. 6125182 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (SPS-427) FIGURE 210A.59-2 ,.
- REF. AXES 0
N
*v y
' , 0, x
APPLIED LOADS: LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ'
- © 0
0.0 0.0 0.226 0~202
-0.189
-0.189
-5.161
-4.951 0.0 0.0 0.0 0.0
.RESULTS:
BASED ON F.S. OF 15 MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ALLOW. BOLT ALLOW. BOLT (TENSION) (KIP) (KIP) (KSI) TENSION SHEAR 0.38 0.07 6.42 0.69 0.64 COMPUTER OUTPUT: MZKAJ09 dt. 6/25/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (WM-2092-63) FIGURE 210A.59-3
Z' REF. AXES .~- - - . y
- R... %" x 11" x24%"
x N J ' 0 0 0
.~- -'~
5/8" </>PHILLIPS WEDGE
.~- -
N~
~
z 0 ~TS 4x4x3/8 - ~
© 0
'~
ANCHOR y
/ ,,
0 0 N
~
N
'f
'~
FORCE IN KIP MOMENT IN KIP-IN. N --
%" ~11 .. ~1 .
2" 4" ~ I. 8" 8" APPLIED LOADS: tx* LOCATION OF LOAD FX' FY' FZ' MX' MY; MZ'
- RESULTS:
0 0 0.0 0.0 0.023
-0.023 0.19 0.19
-0.209 0.209 0.0 0.0 0.0 0.0 BASED ON F.S. OF 15 MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ALLOW. BOLT ALLOW. BOLT (TENSION) (KIP) (KIP) (KSI) TENSION SHEAR 0.15 0.0 1.05 0.69 0.64 COMPUTER OUTPUT: MZKAJ4F dt. 6/29/82 AMENDMENT NO. 13 (2/83)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (CH-2081-685) FIGURE 210A.59*4
SL2-FSAR
- Question No.
210A.60 FP&L will determine for the cases where expansion anchors were used if SRSS summation of effects are appropriate and formally provide the results of this determination to the NRC.
Response
A review of calculations for both small-bore and large-bore support designs indicates that SRSS summation of biaxial effects was, in fact, not used. All such effects were summed directly *
- 210A.60-l Amendment No. 13, (2/83)
SL2-FSAR Question No
- 210A.61 FP&L will submit the results of the analysis of 10 worst case supports selected by Bergen-Paterson.
Response
The 10 Bergen-Paterson selected cases have been analyzed using the ANSYS finite element computer program. All resulted in a factor of safety of at least five for bolt tension and.shear. Nine additional cases, selected by Ebasco, were similarly analyzed. All but one resulted in a factor of safety of at least 6 .5. The one exception is a main steam restraint whose analysis resulted in a factor of safety of 2.4 for bolt tension (bolt shear is negligible). This restraint is unique in that the loads are very large. The pullout force is an order of magnitude greater than that of any other expansion-anchored restraint. The design of this restraint is being modified to achieve a* factor of safety of at least four. Analysis summary sheets for the 19 cases were transmitted to the NRC in FP&L letter*L-82-408 and are contained here as Figures 210A.61-l thru 210A.61-19 *
- 210A.61-1 Amendment No. 13, (2/83)
REF. AXES tX'
- y l/,xz 0 0 0 Cjoli Q
-- j R... 5/8" x 10" - 18" N
I© 0 r -TS3x3x% 0 1 CD I 0
~
Q 1/2" rp PHILLIPS WEDGE ANCHOR FORCE IN KIP
._ Cj-1 Q
MOMENT IN KIP-IN. Y' I* 0'-3*+* 0'-6 .. ,.. 0'-6 ..,l.,.o'-3..,, Z' APPLIED LOADS: LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0 0.15 -0.83 1.0 3.69 -7.12 0.0
- RESULTS:
0 0.15 . -0.83 1.06 3.47 -7.12 BASED ON F.S. OF 15 0.0 MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) {KIP) 1.49 0.282 5.87 7.1 6.0 COMPUTER OUTPUT: MZKAJJM dt. 7/29/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (SPS-427) FIGURE 210A.61-1
~
~u REF. AXES 0 0 - -
j ~ y R.. Y.z" x 9" x 9" N~ ~ r:- Q Y.z" <jJ PHILLIPS WEDGE ANCHOR FORCE IN KIP 1, MOMENT IN KIP-IN
- z
- j"-
r 0 . L.TS 4x4x3/8 0 - z*------------------------------------------------- -- ---..x* Ol1t1r I.. -1~ 0'-7¥.z %"
-~ .. 1 APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0 3.600 -1.000 -2.900 7.500 9.060 5.200 RESULTS: MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS REMARKS (TENSION) (KIP) (KIP) (KSI) (KSI) 1.16 6.8 sx = 5.24 (PRINCIPAL MAX. STRESS) SY= 5.18 NOTE: SX AND SY ARE STRESSES IN X AND Y DIRECTIONS RESPECTIVELY. UL TIM.ATE TENSION ULTIMATE SHEAR (K) (K) 7.1 6.0 COMPUTER OUTPUT: MZKAJLP dt. 7/29/82 AMENDMENT NO. 13 (2/831 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (CC-2074-44) FIGURE 210A.61-2
- 1-3/8'
-~
~
REF. AXES y R.. 1" x 15" x 14%"
.~
I 0 -...
%"</>PHILLIPS WEDGE ANCHOR L
FORCE IN KIP MOMENT IN KIP-IN. TS ll.x2x2x8
- --e- - ,__
Z' ....._-------------------~ -~----)Illa* X' -.
~
J 1%" I 11 3/8"
. 2 1/8" I 1 1 APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0.08 -0.66 0.50 0.90 -1.156 -0.60 RESULTS: MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.217 0.184 0.806 11.6 14.8 COMPUTER OUTPUT: MZKAJ5A dt. 8/10/82 AMENDMENT. NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAHT UNIT 2 ANCHOR PLATE ANALYSIS (CW-30008-8151) FIGURE 210A.61-3
- 0 W4x13
\
0 ll.. Y.z,, x 9" x 9"
.Co %" </>PHILLIPS WEDGE ANCHOR FORCE IN KIP MOMENT IN KIP-IN.
z ASSUMED Y' r.;"\ FIXED Z
- 0 X'
_..x, 1Y.z" 6" APPLIED LOADS: (IN X, Y, Z COORDINATES) LOCATION OF LOADS FX' FY' FZ' MX' MY' MZ'* 0 0.309 -0.515 0.309 19.930 0.0 19.930 RESULTS: (IN X', Y', Z' COORDINATES) MAX. BOLT LOAD MAX. BOLT SHEAR MAX. ll.. STRESS REMARKS (TENSION) (KIP) (KIP) (KSI) (KSI) 25.07 fbx = 12.94
. 1.375 0.002 (MAX. PRINCIPAL STRESS) fby = 24.95 REACTIONS: (IN X', Y', Z' COORDINATES) ULTIMATE ULTIMATE TENSION SHEAR (K) (K) 7.1 6.0 TOTAL REACTION FX' FY' FZ' MX' MY' MZ' FORCE 0
(ALONG FIXED EDGE)
-2.73x10- 2 0.434 -2.909 -3.318 -1.404 0.0 COMPUTER OUTPUT: . MZKAJFV dt. 8/12/82 AMENDMENT NO. 13 (2/83)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (CH-2081-147) FIGURE 210A.61-4
- y REF. AXES 0 0 R... %" x 9" x 1'-5
- x 0 TS 3x3x% 0
~
% <fa PHILLIPS WEDGE ANCHOR FORCE IN KIP z MOMENT IN KIP-IN.
0 0 .--
~
Z' ...._~~~~~~--' -=-- 0'-6 APPLIED LOADS:
- LOCATION OF LOAD 8
FX' 0.095 FY' 0.22 FZ'
-1.247 MX'
-4.604 MY' 3.259 MZ'
-0.059 0 0.095 0.28 0.367 -5.282 3.259 0.059 RESULTS:
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R... STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.41 0.09 6.78 7 .1 6.0 COMPUTER OUTPUT: MZKAJBZ dt. 7/19/82 AMENDMENT N0.13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYS.IS (CH-2081-14) FIGURE 210A.61-5
- y REF. AXES 0
~TS3x~x~
0
.~
j IL. 5/8" x 11" x 11 "
.Co 5/8" </>PHILLIPS WEDGE ANCHOR FORCE IN KIP MOMENT IN KIP-IN.
0 0 - ' ,.....
....~,
-- ,.. X' 8"
*l* w* *I 1
APPLIED LOADS: LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ'
-2.785 2.786 0.061 0.0 0.0 RESULTS:
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.701 0.690 5.022 10.3 9.6 COMPUTER OUTPUT: MZKAJBR dt. 7/21/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (FS-2138-37) FIGURE 210A.61-6
.c.,..
f _3 7/8" 0 I 0 -- REF. AXES z
~
TS 3x3x% R... 1" x 15%" x 25%"
~* [Q] %"</>PHILLIPS WEDGE ANCHOR FORCE IN KIP MOMENT IN KIP-IN.
y .- 0 0 ~- r; Z' ...... . . - - - - - - - - - - - - - - - - ' .:.. - ___. X' I'
- APPLIED LOADS: .... 4 3/8" 1 i.-* 9 3/8" 11%"1
- i. ......
LOCATION OF LOAD FX' FY' MX' MY' MZ'
-0.375 .019 0.545 0.317 -4.50 0.295
.0 -0.375 -0.769 0.705 2.876 -4.50 -0.295 RESULTS:
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 1.04 0.324 1.97 11.6 14.8 COMPUTER OUTPUT: MZKAJ54 dt. 7/19/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (CW-30008-187) FIGURE 210A.61-7
- REF. AXES
.---~~~~~~~~~~~~~~--.
0 0
.~-- ~
... f z y R... 3/8" x 8% x 8%
x~ 5/8" </J PHILLIPS WEDGE ANCHOR FORCE IN KIP MOMENT IN KIP-IN. LTS3x3x% 0 0 .
~
-- ~
-*~
- - --""'ll*)llla~ X' .
0'-6 ...... I I 1%" APPLIED LOADS:
- LOCATION OF LOAD FX' 0.. 0 FY'
-0.174 FZ' 0.260 MX' 1.820 MY'
-0.015 MZ' 0.0 RESULTS:*
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.203 0.044 3.81 10.3 9.6 COMPUTER OUTPUT: MZKAJ5C dt. 7/20/82 AMENDMENT NO. 13 (2/831 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (CC-2051~8) FIGURE 210A.61-8
tv* --
- j
@jMx2 ~.....
Ln REF. AXES y -e- ~ - R.. 1" x 13" x 16"
~
~
- 3/4" </>PHILLIPS WEDGE ANCHOR
-- x -e--- - ,...._
~
FORCE IN KIP
. MOMENT IN KIP-IN .
z ~ .....
~
. ...I NW Z' _.._ ,..._X, I~ 5%"
*1 .. 1* .. I 4 7/8" 2 APPLIED LOADS: "
5/8" LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' CD -0.073 0.0 0.49 0.0 -0.194 0.0 RESULTS: I I MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.308 0.035 1.001 11.6 14.8 COMPUTER OUTPUT: MZKAJ96 dt. 7/21/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (CC-2164-41) FIGURE 210A.61-9
r
~
...~
0 0 -'- REF. AXES y R.. 3/8" x 10" x 10" j ' 3/4" </J PHILLIPS z -- 8 - TS %x4x2 j... WEDGE ANCHOR FORCE IN KIP x MOMENT IN KIP-IN. 0 0
.~ -
-" l Z' -- "
- X' I. . 1}1" 2 I 1'
*I* *I 1
APPLIED LOADS:
- LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0 0.0 0.0 0.208 0.0 1.144 '0.0 RESULTS:
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. ST~ESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.245 0.001 2.60 11.6 14.8 COMPUTER OUTPUT: MZKAJOH dt. 7/27/82 AMENDMENT NO. 13 (:z/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (WM-2092-76) FIGURE 210A.61-10
-- ....------------------------------------ -r-
- ~
N 6" 0 I 0 --
, ~TS 3/8x88 REF. AXES R.. 1%" x 14" x 18" x
.' c.t -
J
-+--+----------- -
0 - - - t - -..........- 0
--t----'li---.----
1" </J PHILLIPS WEDGE ANCHOR
~ z .
FORCE IN KIP
.c,, -- ~o MOMENT IN KIP-IN
- I y La.. ,.. x 10" x 10"
- in I
~
Y' (.;"\ I
, z* T x*P 1
~., __ ,
I I I I* --.x* I 2" I r r " " . , 1/11111III11, r 11i/I I 11i/ Z I
... 5" I
...4( 7" ... 4( ...
APPLIED LOADS:
- LOCATION OF LOAD RESULTS:
0 FX' 0.0 FY' 1.68 FZ' 16.82 MX'. 41.33 MY'
. 0.0 MZ' 0.0 MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 9.23 0.002 13.17 21.8 23.6 REACTIONS:
TOTAL REACTION FX' FY' FZ' MX' MY' MZ' FORCE 0 (ALONG Fl XED EDGE) 0.0051 -1.677 -5.87 -26.97 -6.25 0.0 COMPUTER OUTPUT: MZKJT6S dt. 8/11/82 AMENDMENT NO. 13 (2/83)
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (MS-4100-16A)
FIGURE 210A.61-11
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0.278 -0.905 0.310 1.452 0.364 0.175 RESULTS: MAX. BOLT LOAD MAX. SHEAR LOAD MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.317 0.254 1.855 11.6 14.8 COMPUTER OUTPUT: MZKAJ8N dt. 8/5/82 AMENDMENT. NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (Sl-2405-8149) FIGURE 210A.61-12
- REF. AXES 0 0 -- ..
y \TS%x3x3 R... 3/4" x 15" x 15"
-z 3/4" </J PHILLIPS WEDGE ANCHOR
~
FORCE IN KIP MOMENT IN KIP-IN. 0 0 ---
- X' 0'-9 (MIN.)
.. I.. *I 0
'-3 APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0 0.642 0.051 1.398 0.496 2.622 -1.201 RESULTS: MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R... STRESS ULTIMATE TENSION ULTIMATE SHEAR * (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP)
*0.512 0.196 2.72 11.6 14.8 COMPUTER OUTPUT: MZKIT4W dt. 8/11/82 AMENDMENT NO. 13 (2/83)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (CH-71-R1) FIGURE 210A.61-13
't.
- REF. AXES 0
I
~...
C") 0
.~--
- -~
y x-.: z' " ~ R.. %" x 10" x 15"
%"</>PHILLIPS WEDGE ANCHOR R. ..3/8"x2~"x3%"
FORCE IN KIP MOMENT IN KIP-IN. 0 0 .~-~ z*--------------- ...-- -.x* ~
........1*
*1 1%" 0'-7
- APPLIED LOADS:
LOCATION OF LOAD FX' FY' MX' FZ' MY' MZ' 0 0.0 0.0 1.54 0.0 0.0 0.0 RESULTS: MAX. BOLT LOAD MAX. BOLT SHEAR MAX~ R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.661 - 9.41 7.1 6.0 COMPUTER OUTPUT: MZKIT8Z dt. 8/11/82 AMENDMENT*NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ANCHOR PLATE ANALYSIS (SPS-117) FIGURE 210A.61-14
3%"
.---------------------. -- j 0
0 REF. AXES R... %" x 13" x 16" v
%"</>PHILLIPS
. WEDGE ANCHOR
!st.
c;, FORCE IN KIP MOMENT IN KIP-IN. x
.~ '.
0 0 '~
- --.x*
.-,II z* '------------------..-----' --
.........I 1
1%" 0'-9 3/4"
.. I..1*... I APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ'
-2.089 -0.07 5.125 0.16 -0.444 -0.027 RESULTS:
MAX. BOLT LOAD MAX. BOLT SHEAR MAX R... STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 1.78 0.53 7.44 11.6 14.8 NOTE: X-AXIS IS 17° CLOCKWISE FROM NORTH COMPUTER OUTPUT: MZKAJHN dt. 6/23/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (Sl-4203-441) FIGURE 210A.61-15
- N REF. AXES y
~*
z - APPLIED LOADS:
- LOCATION OF LOAD
© FX'
-0.102 FY' 0.261 FZ' 0.102 MX'
-7.895 MY'
-3.086 MZ'
0.0 RESULTS
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.56 0.07 3.62 11.6 14.8 COMPUTER OUTPUT: MZKAJH6 dt. 6/25/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (D0-18-84) FIGURE 210A.61~16
- REF. AXES 0 0
=
-~
C'll
~
~
y R.. 5/8" x 15%" x 19%" 5/8" <j) PHILLIPS
- - N WEDGE ANCHOR FORCE IN KIP MOMENT IN KIP-IN.
0 0
=
-- j C'll
-- _..x, 11%"
~ 142*~,
- APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0.0 -0.034 0.46 0.16 0.0 0.0
~ESULTS:
MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE" TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.13 0.01 1.91 10.3 9.6 COMPUTER OUTPUT: MZKAJ2V dt. 6/30/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS (CC-2051-5) FIGURE 210A.61-17
- ~
.-:-n
-. - r - - - - - - - - - - - - - - - - - - - - - - - - -
+ + ....
~
1S4"4><11xO':'IL7 REF. AXES r---------""'-----. _,;.... R... 1** x 15" x 14Y.z...
. '~
y ~
...... , ....*'... %.. q,*PHILLIPS l/
"";" 0 WEDGE ANCHOR 0
0
- z -- FORCE IN KIP MOMENT IN KJP*IN.
N
*o*~
*I
.~
. + +I I- -
z*,.......I
,Y.i..
- 1'-% ....1*"
--, *x*
- APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' *.MY' MZ' 0 0.041 0.0 1.099 -13.338 -1.029 0.686 RESULTS: MAX. BOLT LOAD MAX. BOLT SHEAR
- MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) (KIP) (KSI) (KIP) (KIP) 0.86 0.014 2.45 11.6 14:8 COMPUTER OUTPUT: . MZKAJ47 dt. 6/30/82 AMENDMENT NO. 13 l2/83)
FLORIDA*POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS .
. (CW-30008-8129)
FIGURE 210A.61-18
REF. AXES
-- 1 R.. 3/4" x 1'-3 x 1'-3
-z 1" </>PHILLIPS
.s WEDGE ANCHOR FORCE IN KIP N MOMENT IN KIP-IN.
LTS3x3x% C? 0 z*---~~~~~~~~---~~~~~~~~~~--' -'- ', ~~~~x* ... I* o*-3 *... 1* 0'-9 (MIN.) ...,... 0'-3 ... ,
- APPLIED LOADS:
LOCATION OF LOAD FX' FY' FZ' MX' MY' MZ' 0.0 -1.343 1.114 17.474 0.0 0.0 RESULTS: MAX. BOLT LOAD MAX. BOLT SHEAR MAX. R.. STRESS ULTIMATE TENSION ULTIMATE SHEAR (TENSION) (KIP) . (KiP) (KSI) (KIP) (KIP) 1.50 0.34 7.66 21.8 23.6 COMPUTER OUTPUT: MZKAJKP dt. 6/25/82 AMENDMENT NO. 13 (2/83) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS
. (CH-85-R6)
FIGURE 210A.61-19
SL2*FSAR
- Question No
- 210A. 62
Response
FP&L will submit a description *or the ArJSYS model and the Ebasco preprocessor program u,sed in. the analysis. To account for the flexibilities of both the concrete expansion anchor and the baseplate, the "ANSYS" computer program was employed. This program utilizes the finite element method of analysis. TQ facilitate the use of the "ANSYS" program Ebasco has developed a preprocessor "EMBEDP". A brief description of the "EMBEDP" computer program is given ~n Table 210A.(>2;..l. The test problem presented in Section. 5.0 of "Summary Report! of Generic Response to USNRC IE Bulletin No. 79-02 sase Plate/Concrete Expansion Anchor Bolts by Teledyne Engin~ering Services, August 30, 1979", was used to verify the "&IBEDP" program. Subsection 5.5.3 shows the plate geometry and Subsection 5.6.l gives the bolt load. Input and output of the F.MBEDP program involving the test problem are presented in sample problems nos. 1, 2 and 3, copies of which are attached. The ANSYS input data results from these sample problems were introduced in~o the A~SYS program, the output of which was then compare.d with that of the Teledyne report. Table 210A. 62-1 compares the EMBEDP and Teledyne results for the bolt loads *
- Description Of Model The plate is divided into a finite number of elements (STIF63).
While dividing the plate into elements it is desirable to increase the number of elements in the region of expected maximum stresses. In other areas fewer elements may be used. To increase the convergence of the.results.it is common tQ have two rows .of elements between the edge of the plate and the bolt line. The concrete is replaced by compression-only sp~ings (STIFlO) derived from the half-space theory as given. by Barkan. . The total stiffness Kc of concrete subgrade is given as: G K = c (2.2) ,.JWL c 1-Vc Ge = shear modulus of* concrete Ve = Poisson's ratio of concrete W = width of base plate L = length of base plate
- 210A.62-l Amendment No. 13, (2/83)
Compression springs representing the concrete subgrade are attached t.o each nod~ of the mo.del
- Bolts are represented by tension springs {STIF10) in the longitudinal direction. The longitudinal stiffnesses of the bol,ts are obtained from tests performed at the jobsite by the bolt ..
manufacturer. , Shear stiff1.1ess~s of. the bolts (.S';['IF1.4) are also derived from test
- results. In this .at).alysis these .values were taken* from "Anchor Bolt Shear and Tension Stiffness", Teledyne Engineering Services, May *25, 1979. *:*Since the stiffness of the plate in th~ horizontal direction (in-pla:ne stiffness) is. rela:tivelylarge compared to the shear stiffness... of the bolt., the. she~r force distribµtion among the. bolts (all o-f the same 1:ype and size) is n.ot affected by the s"hE;!ar stiffness,of the .bolt. : For .ths reason, it is possible to distribute the t.otal shear force among the bolts without resorting to the ANSY.S analysis~ ; However, ill the analyses performed for St.
Lucie 2,. all loads~ pul_lotit aqd moment as well as shear, were applied in. th~ same .run iI). the knowledge tha~ the shear force .'taken by the*.bolts -w:oµld .affect neithei: the tension in the bolt nor the plate stress*. , . A part of the attachment is included in the model as plate elements. The load is applied to this attachment'. Study Of Model Mesh Size The baseplate f*or r*estraint CH-71-Rl was selected to
,effect of ele~ent sizes .on the.stresses and the boit tension.
s.tudy the
*This *restraint baseplate is typical of the. majority. <;>f expansion anchored restraint applica,tions. Pull9ut load was applied to the plate. A part of the* attachment was modeled as plate elements.
Two computer runs were performed with the 3/4" plate divided into 5 *x:*S.* elements ,and 8 x *8 el,eme~ts as. shown on Fi,gure 210A.62-1. Table* 210A.62-2 presents. the values of maximum bolt load and maximum plate. stress obtained. from the. two computer runs for two different mesh sizes. From these two cases it can be seen that the difference in bolt tension is small (0.06%) while the maximum stresses differ by 2.20%.
- The combination of small pullet force/thick plate results in a small prying contribution *. We* a.re proceeding to analyze another case with large pullout/relatively thin plate for comparison.
- 210A.*6.2-2 Amendment No. 13, (2/83)
- SL2-FSAR
- Ebasco Computer Program "EMBEDP" The "EMBEDP" computer program was developed by Ebasco as a preprocessor for the "ANSYS" finite element program for baseplate and anchorage nonlinear analysis. This program automatically generates the finite element model including the load data using a minimum number of input cards. The preprocessor minimizes engineering time and allows solution of a large number of baseplate problems economically. The program has been completed and verified *
. The program structure is sufficiently flexible to allow the user to exercise options in considering spe~ial features of different problems. The following special features ate included and can be handled l>Y the program:
(a) Selection of the type of element (bending only or membrane _ plus bending) for the baseplate - For the case with uplift force only, the bending type element can be used to reduce the computer cost. (b) Generation of the spring constants of the concrete subgrade - using the half-space formula developed by Barkan. (c) Consideration of the pretorque in the anchor bolts *
- (d) Consideration of the friction between the baseplate and concrete surfaces. If it is required to take into account the friction betweeµ the baseplate and the concrete, the friction element (STF52) may be included in the analysis.
When this element is selected, the baseplate is automatically represented.by a membrane plus bending element. (Please note that in the analyses performed for St. Lucie 2, the friction element was not used to carry shear loads). (e) Location and Number of bolts - Any random distribution, up to 20 bolts can be input. * (f) The attachments - any attachment having components parallel to the sides of the baseplate can be input.
"EMBEDP" together with "ANSYS" provides stresses in baseplates and forces in bolts on plate assemblies subject to various loadings *
- 210A.62-3
- Amendment No. 13, (2/83)
SL2.:.FsAR TABLE 210A.62-l BOLT LOAD COMPARISON (See Program Verification Reference) Load Bolt Bolt Load (1 b) Case
- Number EM BE DP Teledyne Case 1 1 2324 2350 2 2324 2350 (Axial Load) 3 2324 2350 4 2324 2350 Case.2 1 2272 2316
- (45° Shear/
Moment) 2 3 972 o.o 1024 0 4 972 1024 Case 3 1 1860 1942 2 1860 1942 (00 Shear/ Moment) 3 0 0 4 0 0 Note: Primary difference is due to different formulas used-for concrete spring *
- 210A.62-4 Amendment No. 13, (2/83)
SL2-FSAR TABLE 210A. 6 2-2 1/ 4 Pull Out Load Kip BOLT LOAD COMPARISON 5 x 5 Elements 1/ 4 Plate Stress in KSI Load in Kip 8 x 8 Elements Maximum Plate Maximum Bolt ~ximum Plate Maximum Bolt Stress in KS! Load in Kip 0.3495 2.19 0.349 8 2 .24 0.3496 210A.62-:5 Amendment No. 13, (2/83)
1---__,+-......,..--+-........-__,1---...,--t--- - -
-~
~
' ,f
'~
c., ~
* ---f-------"---+---+---+---+--- -
M
)(
BOLT .- >< BOLT :'~ M t----+----+------<.->------+--- ..._ M t---+---+----.,t---+--<:>---+---+---1 __
~~
~
)(
.~
*~
~
)( '
- h
~ .... .... .-t---+---+----"t---+---+---+---+-- -
en I- 1----+----+-----,1-----+----1
~
'""""____,.__ _-+---1----+----I - -
-i--
~
;/.... ~ \
en I-i--..........--1---+--+---+---+-~
~;--
.~ t---+---t--__,t---+-__,-+-__,-+-__,-+---1 ---
.... ::: ll'
~
ct - - ct.__......__.__ _,__ _.__......__....____....______, i=;lt- "-...
- I
-ti(
1%" 2 I 1 %" 2 I 1%" 1* 1%" I I 1%" . . . x I*"l*"I I *I I I I I 1"......1".... 1"....1".... 1 **...... 1".. . x ct (a) ct (b) PLATE THICKNESS= 3/4 INCH
- MODEL OF 1/4 PLATE 5 x 5 ELEMENTS 3/4" </J PHILLIPS WEDGE ANCHOR 1/4 PULL OUT LOAD = 0.3495 K MODEL OF 1/4 PLATE 8 x 8 ELEMENTS AMENDMENT NO. 13 (2/83)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 BASE PLATE ANALYSIS FIGURE 210A.62-1
SL2-FSAR
- Question No.
210A.63 F&L will perform. an analysis of a sample baseplate subjected to pullout load' and moment, considering these effects applied separately, and an analysis considering these effects applied simultaneously.
Response
Four individual load cases .were considered separately. The four load cases were (1) pullout load Fz, Mx and~ applied simultaneously. The plate selected was CH-71-Rl. Results from these cases are summarized below. It may be pointed out that the location of maximum stress is different for each load case. However, in the actual analysis using the "ANSYS" program for St. Lucie 2 critical combinations of individual loads were used. Single load applications were never considered in the analysis. Maximum Bolt Maximum Plate Applied Load Tension Load Stress Fz = 1.398 o. 350 . 2.19 Mx = 0.496 0.026 0.20 My = 2.622 0.138 1.06 Fz = 1.398 Mx 0.496 0.512 2. 7 2 My = 2.622 Load in KIP Moment in IN-KIP Stress in KSI
- 210A.63-l Amendment No. 13, (2/83)
SL2-FSAR Question No. 210A.64 FP&L will revise its expansion anchor design criteria to reflect a factor of safety of 15 across the board. Reference to prying calculations will be deleted.
Response
The criteria will be revised to incorporate the following statement:
"The use of a safety factor of 15 for all types of loading precludes the necessity of a prying calculation. Where the use of a safety factor of 15 is impractical and the presence of large loads results in a significant prying effect, baseplates shall be analyzed using the ANSYS finite element computer program."
- 210A. 64-1 Amendment No. 13, (2/83)
SL2-FSAR
- Question No.
210A.65
Response
Provide confirmation of piping analysis open items. An investigation was performed for all Safety Class 2 & 3 Systems (irrespective.of operating temperature) to demonstrate that the number of equivalent thermal cycles, as defined in ASME Subsection NC 3611.2, was sufficiently low to confirm the conservatism of the existing stress analyses. In accordance with the agreement reached at a meeting with the NRC and Florida Power & Light Company on October 14, 1982 an acceptance criteria of 1000 "Realistic" cycles was employed. In conducting this analysis, the following Safety Class 2 and 3 systems were reviewed: Reactor Coolant Component Cooling Water Charging *Letdown Safety Injection Auxiliary Feedwater Main.Steam Containment Spray Main Feedwater Intake Cooling Water A sample calculation specifying methodology and a summary of results is provided on Table 210A.65-1. Using realistic values of cycle frequencies, all systems were shown to exhibit approximately 700 equivalent cycles. Using all the thermal transients that appear in the Safety Class 1 specification (Refer to Tables II and III), which are conservative both in frequency and temperature variation, all systems were shown to have less than 1000 equivalent thermal cycles. Therefore, the above results confirm the conservatism of the existing stress analyses for Class 2 & 3 systems employing weldolets *
- 210A.65-l Amendment No. 13, (2/83)
SL2-FSAR
- Thermal Transient ATE TABLE 210A.65-l SAMPLE CALCULATIONS Charging ATI NI N 5% Ramp Up 455 70 15000 2 5% Ramp Down 455 208 15000 299 10% Step Up 455 47 2000 0 10% Step Down. 455 .145 2000 7 Loss of RCP Flow 455 204 40 1 Reactor Trip 455 187 400 5 Loss of Load 455 204 40 1 Loss of Sec SS 455 330 5 1 Normal Var. 455 6 106 1 Hydro 455 335 10 2 Leak Test 455 335 200 43 Loss of Chrgr 455 335 20 6 Loss of Letdown 455 400 50 26 Regen H-X Iso 455 349 120 32 Max Purification 455 128 9000 16 Max Dilution 455 126 8000 14 Low \TCT 455 45 2000 0
- Norm Start 455 455 500 500 Aux FW Inj TOTALS 955
- 1. Charging T max = 520°F
- 2. Ambient Temp (Tamb) 650F
- 3. aTE =;: T max - Tamb = 4550F
- 4. Number of cycles = NI
- 5. Temp change of transient = aTI
- 6. Equiv # of cycles at ATI5 N - NI ( ar.I/ ATE)
- 210A.65-2 Amendment No. 13, (2/83)
SL2-FSAR Question No *
- 220.1 Table 3.3-1 lists the Auxiliary Building as being designed for a tornado velocity of 300 mph while the other buildings are designed for 360 mph. The Containment Shield Building is omitted from this table. Subsection 3 .3 .2 .2 states the Shi.eld Building is designed for 300 mph. Correct Table 3.3-1 to reflect your actual design velocities. Justify in detail why the velocity of 300 mph is used instead of 360 mph. Submit your supporting calculations.
Response
FSAR Table 3.3-1 has been revised to include the hurricane and tornado wind speeds used in the design of the Shield Building and Condensate Storage Tank Building. The external pressure coefficients for these dome-cylinder structures are given on Figures 3.3-1 and 3.3-2, respectively. The 300 mph average combined tangential velocity and translational velocity tornado design and wind speed specified for the Shield Building and Reactor Auxiliary Building was based upon consideration of the horizontal dimensions of those structures and _wind speed ban widths and conclusions presented in FSAR Section 3.3 Reference~ 3 and 4. The tornado design wind speeds specified for the structures are considered conservative since the tangential velocity versus
- height above ground is held constant at the maximum average values. Based upon data presented in References 3 and 4, the tangential velocity reduces substantially close to the ground and, in particular, in the height range of St Lucie Unit 2 structures.
See revised Section 3.3 and Table 3.3-1. -* 220.1-1 Amendment No. 5, (8/81)
SL2-FSAR Question No
- 220.2 Provide your basis and method of calculations to support the pressure differential ia the' Diesel Generator Building of 2.25 psi. If venting is considered in any other building provide method of computing differential pressures.
Response
The pressure differential value of 2.25 psi, used in the design of the Diesel Generator Building for St Lucie Unit 2, was established during St Lucie Unit 1 design and accepted by the NRC during Unit 1 operating license review. Based upon the large ventilation and cooling openings in the exterior walls, the reduced differential pressure was considered a conservative assumed design value; no calculations were perf9rmed to determine that value. Each of the two equipment housing compartments within the Diesel Generator Building contains approximately 49,200 cu ft of air volume (before deducting equipment volume) and each compartment has approximately 618 sq ft of available ventilation area after deducting wire screen area. Venting is not considered in any other structure. No FSAR change required *
- 220.2-1 Amendment No. 5, (8/81)
SL2-FSAR
- Question No.
220.3 You stated in Subsection 3.5.3.1.1 that the modified Petry Formula was used to evaluate the concrete for missile protection. The current requirement for computing required concrete wall thickneses is the NDRC Formula. Submit a table showing the required wall thicknesses compared with the actual wall and roof thicknesses for all Category I structures, ,based on the NDRC Formula.
Response
FSAR Table 3.5-11 lists roof and exterior wall thicknesses for all Category I concrete structures. The attached table summarizes design values obtained by use of the NDRC Formula for missile penetration (x), thickness required to prevent scabbing (s) and the maximum thickness of concrete which a missile will completely penetrate (e). Please note that the NDRC Formula is only applicable to hard missiles and not the soft missiles; i.e., the automobile and wood missiles. No FSAR change required *
- 220.3-1 Amendment No. 5, (8/81)
SL2-FSAR TABLE 220.3-1 TORNADO MISSILE IMPACTIVE ANALYSIS Penetration of Concrete for Design Base Spectrum of Tornado Missiles Using Modified National Defense Research Committee Formula (NRDC) Required Actual Concrete Minimum w A d Vo x x e s e s Thickness Concrete Missile 2 d d d 1.25 s Thickness No. Missile QE.L (in. (in.) (ft/sec) (in.) (in.) (in.) (in.) for Remarks 1 4" x 12" Plank Wall or 12' Long 200 48.0 322 Roof Slab Soft** of All Missile 2 l"diax3' 8.0 0.785 1.0 163 1.58 1.58 3.28 4.27 3. 28 4.27 5.34 Class I Steel Rod Structures (in.) 3 6" dia Sch 40 284.5 5.58 2. 67 116 4.69 1.76 3.50 4.51 9.35 12.05 15.06 x 15' Pipe 4 12" dia Sch 40 743.4 14.6 4.31 116 6.27 1.45 3.12 4.09 13.44 17.63 22.04 ?: 24" Critical x 15' Pipe Case N N p 5 13 *. 5" dia x 35' 1,497 143.0 153 Soft** long Wooden Missile N I Utility Pole 6 Automobile 4,000 2,880 84 Soft* Missile 7 2" x 4" Plank 27.8 8.0 403 Soft** 10' Long Missile 22.04"< 24" minimum wall or roof slab concrete thickness
- Missile does not penetrate based on ASCE Reference Page 6-28
** NRDC Formula is not applicable based on .AscE Reference Page 6-41 References - Design Based Spectrum of Tornado Missiles Table 3.5-10 SL2-FSAR Tornado Missile Concrete Barrier Minimum Thickness Table 3.5-11 SL2-FSAR ASCE "Manual of Standard Practices for Design of Nuclear Power Plant Facilities" Chapter-6
SL2-FSAR Question No . 220.4 In Subsection 3.4.1 you stated that polyvinyl chloride (PVC) water stops were used in the construction joints. When PVC is exposed to radiation it converts into hydrochloric acid which will attack the concrete. Show that the water stops will not be degraded by radiation.
Response
FSAR Subsection 3.4.1 addresses flood penetration provided for design flood water level. The flood penetration provided on and within exterior walls of seismic Category I structures with basements (Reactor Building and Reactor Auxiliary Building), consists of waterproofing membranes and polyvinyl chloride (PVC) water stops, respectively. Radiation levels at exterior walls of those buildings are well below the levels which PVC can tolerate without appreciable damage (5 x 105 rads) and the level at which corrosive gases could be liberated (approximately 106 107 rads based upon the behavior of similar compounds). The only area oh St Lucie No. 2 where radiation levels could be high enough to cause damage to PVC water stops was within the Shield Building Steel Containment Structure. Rubber.water stops were specified for use in construction joint~ below EL 23 within the Steel Containment Structure. Rubber has a threshold to damage above 2 x 106 rads and does not liberate gases until very high radiation levels are reached (above 108 rads) which is considerably above the maximum radiation levels predictable for the plant. No FSAR change required *
- 220.4-1 Amendment No. 5, (8/81)
SL2-FSAR
- Question No.
220.5 On Page 3.5-24 you stated the maximum thickness of concrete barriers is two feet. What is the maximum thickness provided and show that this is enough to stop the potential missile using the NDRC Formula.
Response
Page 3.5-24 incorrectly stated the maximum thickenss of concrete barriers is two feet. The page has been corrected to read "the minimum thickness ***** etc." Refer to response to Question 220.3 for discussion of required concrete wall thicknesses based upon the NDRC Formula.
- See revised page 3.5-24 *
- 220. 5-1 Amendment No. 5, (8/81)
SL2-FSAR Question No *
- 220.6 ResEonse Describe your procedure used to predict thicknesses which prevent spalling or scabbing of concrete barriers and generation of secondary missiles.
The procedure used to prevent the generation of secondary missiles by spalling was to provide a minimum concrete thickness of two feet, and to check that the calculated missile penetration depths calculated with the Modified Petry formula, were less than half the wall thickness. Deepest penetration for steel missile (l" dia x 3' steel rod) was calculated to be 3 * .13 in. and the deepest 10 penetration for a wood missile (2" x 4" plank 10' long) was 5.10 1 inches. *The FSAR references tests which show that wood missile splinter into pieces without causing any local damage for concrete barrier thicknesses of 12 inches or more. Therefore, the steel rod should penetrate the deepest of all the missiles. The ratio of 2 ft. wall thickness to maximum depth of penetration would be 7.67. 110 NOTE; The calculation results, presented on the table in response to Question 220.3, indicate that the 24 inch minimum thickness of concrete walls and roof slabs is more than the 1.25s minimum requirement to prevent spalling as calculated with the NDRC Formula *
- No FSAR change required *
- 220.6-1 Amendment No. 10, (6/82)
812-FSAR
- Question No.
220.7
Response
Provide label on ordinate of Figure 3.7-5. Figure 3.7-5 has been revised by adding references to obtain ordinate values. See revised Figure 3.7-5 *
- 220. 7-1 Amendment No. 5, (8/81)
SL2-FSAR Question No
- 220.8 State where the design time history is applied to the mathematical (3.7.1) model relative to the finished grade. If deconvolution procedures are used, describe these procedures and furnish the response spectra computed for the input to the math models shown on Figures 3.7-30 thru*3.7-51.
Response
The design time history is applied at the foundation level of Category I structures in the free field. Deconvolution procedures are not used. No FSAR change required *
- 220.8-1 Amendment No. 5, (8/81)
SL2-FSAR Question No. 220.9 Some of the response spectra points computed for the artificial (3.7.1) time histories fall below the design response spectra. Show that no more than 5 points fall more than 10 percent below the design spectra for each damping value.
Response
Comparis.ons between the response spectra points computed from the artificial time histories and the design response spectra suggested in R.G. 1.60 has been done. Both horizontal and l 5 vertical time histories are considered. The spectra values are generated at 1/2 percent, 2 percent, 5 percent, 7 percent and 10 percent damping, as suggested in R.G. 1.60. The frequency intervals used are that suggested in SRP Table 3.7.1-1. Results show that only for the 1/2 percent damping curve more than 5 points (out of 7 5 points) fall more than 10 percent below the design spectra curve. However, for St Lucie Unit 2, the lowest damping valve specifi.ed is 1 'percent (for steel piping) so the case of 1/2 percent damping has no effect on the seismic analysis. Moreover, the lowest frequency value* for St Lucie Unit 2 is 1.22 cycles per second (SL2-FSAR-Table 3.7-18, Reactor Building), therefore, points falling below design spectra for frequencies less than 1.22 HZ do not affect the results of seismic analysis. For the other points, the design spectra for the time histories show substantial higher values than the R.G. 1.60 design spectra. Thus the positive values should compensate the effects, if any, of the negative values and insure a conservative design. In summary, the 1/2 percent damping curve is not used for any design purpose on St Lucie Unit 2 and the remaining response spectra curves meet the criteria of no more than 5 points falling more* than 10 percent below the design spectra. No FSAR change required *
- 220.9-1 Amendment No .* 5, (8/81)
SL2-FSAR Question No *
- 220.10 (3.7.2)
Response
Your seismic models include provisions for structural torsion however, the soil springs do not include a torsional component. Describe how you account for the torsion at the foundation soil interface. Two dimensional seismic models are used for analyzing the Category I structures, since the Category I structures are supported independently and the geometrics of the structures are largely symmetrical. In the two dimensional models, torsional degrees of freedom of mass points are considered as fixed conditions. For the soil springs, the torsional component is also fixed but may be visualized to have a spring with very large torsional spring constant. Original analysis of Waterford No. 3 also utilized two dimensional models without torsional degrees of freedom. In response to NRC questions, a new three-dimensional model with torsional degrees of freedom and torsional soil spring was developed. The accelerations obtained from the new model (with torsional degrees of freedom) compared to those of the original two-dimensional model are smaller. No FSAR change required *
- 220.10-1 Amendment No. 5, (8/81)
*
- SL2-FSAR TABLE 220.10-1 WATERFORD NO. 3 NATURAL FREQUENCIES IN CYCLES PER SE~~D (CPS) I 12 E-W EARTHQUAKE N-S EARTHQUAKE Nlde G = 6,400 PSI G = 16,050 PSI G = 6,400 PSI G = 16, 050 PSI No.
Without With Ebasco Without With Eb as co Without With Eb as co Without With Eb as co Torsion Torsion Dynamic Torsion Torsion Dynamic Torsion Torsion Dynamic Torsion Torsion Dynamic 1 1.091 1.086 1.079 1.706 1.700 1.68 1.087 1.086 1.08 1. 702 1. 700 1.68 2 2.445 1.684 2.473 3.334 2.620 3.35 2.468 1. 815 2. 51 3.410 2.833 3.42 3 4. 562 2.450 4.540 5.248 3. 363 5. 22 4. 275 2.468 4.26 4.883 3.410 4.86 4 7 .535 4.545 7 .487 7 .571 4.684 7.51 7 .475 4.265 7 .43 7 .491 4.701 7 .42 5 10.936 4.678 9.183 10.965 5.184 9.10 10.254 4.680 8.40 10.284 4.860 8.01 6 11.97 5 6.529 11.149 11.982 6.587 11.12 10.807 6. 741 9.72 10.863 6.797 9.81 7 12.154 7.626 11.987 12.155 7. 696 11. 99 12.125 7.511 12.07 12.129 7. 539 12.07 8 14.874 11.464 14.241 15.046 11.471 14.37 14.914 10.054 14.16 14. 940 10.083 13.77 9 20.438 12.004 16.567 20.464 12.009 16. 46 19.270 10.826 15.16 19.303 10.877 14.74 10 21.640 13.113 17.924 21.640 13.176 17 .84 21.637 12.105 17 .24 21.638 12.108 17.20 N N 0
'i' N
*
- SL2-FSAR TABLE 220.10-2 WATERFORD NO. 3 OJMPARISON OF ACCELERATION OF DYNAMIC ANALYSIS I 12 WITH AND WITHOUT TORSIONAL .DEGREE OF FREEDOM SOIL SHEAR MODULUS G = 6400 PSI, SSE, SPECTRUM METHOD, 5% DAMPING E - W DIRECTION N - S DIRECTION STARDYNE - 3 EBASCO STARDYNE - 3 EBASCO DYNAMIC 203 7 DYNAMIC 2037 MASS CASE -I* CASE - II** DIFF CASE I* CASE I* CASE II** DIFF CASE l*
NO. % % 1 o. 2 78 0.272 -2.2 0.287 0.231 0.231 0 2 0.257 0.251 -2.3 0.263 0.218 0.217 -0.5 3 2.240 0.234 -2.5 0.246 0.208 0.207 -0.5 4 0.226 0.220 -2.7 0.230 0.199 0.198 -0.5 5 0.211 0.205 -2.8 0.214 0.190 0.189 -0.5 6 0.194 0.188 -3.l 0.195 0.179 0.179 0 7 0.178 0.172 -3.4 Q.178 0.169 0.169 0 N 8 0.167 0.161 -3.6 0.165 0.162 0.161 -0.6 N 9 0.156 0.150 -3.9 0.153 0.154 0.154 0
- => 10 0.148 0.143 -3.4 0.149 0.148 -0.7
..... 0.145 0 11 0.141 0.137 -2.8 0.139 0.144 0.144 0 I
12 0.234 0.228 -2.6 0.239 0.200 0.200 0 13 0. 223 0.217 -2.7 o. 226 0.194 0.193 -0.5 14 0.211 0.205 -2.8 *O. 214 0.187 0.186 -0.5 15 0.200 0.195 -2.5 0.202 0.180 0.180 0 16 0.190 0.184 -3.2 0.191 0.174 0.174 0 17 0.180 0.174 -3.3 0.180 0.168 0.168 0 18 0.170 0.165 -2.9 0.169 0.162 0.162 0 19 0.161 0.156 -3.1 0.160 0.157 0.156 -0.6 20 0.153 0.148 -3.3 0.151 0.152 0.151 -0.7 21 0.146 0.141 -3.4 0.143 0.147 0.146 -0.7
- Without torsional degree of freedom
- With torsional degree of freedom
~
p
"'9 ID ...p z ~
N CD CD N
SL2-FSAR TABLE 220.10-2 (Cont'd) E - W DIRECTION N - S DIRECTION STMDYNE - 3 El:IASCO STARDYNE - j EBASCO DYNAMIC 2037 DYNAMIC 2037
*MASS CASE -I* CASE - II** DIFF CASE I* CASE I* CASE II** DIFF CASE .l*
NO. % % 22 0.167 0.161 -3.6 0.166 0.160 0.159 -o.6 23 0.164 0.159 -3.1 0.163 0.158 0.158 0 24 0.167 0.156 -3*. 7 0.160 0.156 Q.156 0 25 0.157 0.152 -3.2 0.156 0.154 0.154 0 26 Q.153 0.148 -3.3 0.151 0.152 0.151 -o. 7 27 0.148 0.143 -3.4 0.146 0.148 0.148 0 28 0.145 0.140 -3.5 0.142 0.146 0.146 0 29 0.179 0.164 -8.4 0.179 0.169 0.169 0 30 0.161 0.147 -8.7 0.160 0.158 0.157 -o.6 31 0.153 0.154 +o.7 0.151 0.152 0.151 -o. 7 32 0.145 0.149 +2.8 0.144 0.147 0.148 +Q.7 35 0.171 0.180 +5.3 0.172 0.164 0.164 0
;:> 36 O.H3 0.170 +4.3 0.162 0.158 0.15~ +o.6 ..... 39 0.137 0.135 -1.5 0.134 0.14i. 0.140 -0.7 0
I
* >* Without torsional degree of freedom ffi ** With torsional degree of freedom J:l .
CJ. ffi J:l rt z 0
* *SL2-FSAR TABLE 220.10-3 WATERFORD NO. 3 ffiMPARISON OF ACCELERATION OF DYNAMIC ANALYSIS I 12 WITH AND WITHOUT TORSIONAL DEGREE OF FREEDOM SOIL SHEAR MODULUS G = 6400 PSI, SSE, SPECTRUM METHOD, 5% DAMPING E - W DIRECTION N - S DIRECTION STARDYNE - 3 EBASCO DYNAMIC 203 7 STARDYNE - 3 EBASCO DYNAMIC 2037 I
MASS CASE -I* CASE - II** DIFF CASE I* CASE I* CASE II** DIFF CASE 1* NO. % % 1 0.492 0.4 79 -2.6 0.430 0.429 -0.2 2 0.453 0.440 -2.9 0.401 0.399 -0.5 3 o. 423 0.411 -2.8 0.379 0.377 -0.5 4 0.395 0.384 -2.8 0.358 0.357 -0.3 5 o. 367 o. 356 -3.0 0.337 0.336 -0.3 6 0.333 0.322 -3.3 0.311 0.310 -0.3 7 o. 299 0.290 -3.0 0.286 0.285 -0.3 8 0.27 5 0.266 -3.3 0.268 0.266 -0.8 9 0.250 0.242 -3.2 0.249 0.248 -0.4 N N 10 0.232 0.224 -3.5 0.234 0.233 -0.4
;' 11 0.216 0.209 -3.2 0.222 0.221 -0.5 ,_. -0.3 0 12 0.356 0.344 -3.4 0.312 0.311 I
ln 13 0.341 0.330 -3.2 0.303 0.301 -0.7 14 0.325 0.315 -3.1 0.293 0.292 -0.3 15 0.310 0.300 .-3.2 0.284 0.282 -0.7 16 0.295 0.286 -3.1 0.274 0.273 -0.4 17 0.280 o. 271 -3.2 0.265 0.26/i -0.4 18 0.265 0.257 -3.0 0.255 0.254 -o.4 19 o. 251 0.243 -3.2 0.245 0.244 -0.4 20 0.237 d.229 -3.4 0.236 0.235 -0.4 21 0.223 o. 217 -2.7 o. 226 0.226 0
- Without torsional degree of freedom
- With torsional degree of freedom 8
ID
- l p.
~
- l rt z
~
N
~
00 00 N
- 512-FSAR TABLE 220.10-3 (Cont'd)
E - W DIRECTION N - S DIRECTION STARDYNE - 3 EB AS CO STARDYNE - 3 EBASCO DYNAMIC 2037 DYNAMic*2037 MASS CASE -I* CASE - II** DIFF. CASE I* CASE I* CASE II** DIFF CASE 1* NO. % % 22 0.259 0.250 -3.5 0.248 0.247 --0. 4 23 0.255 0.246 -3.5 0.246 0.245 --0.4 24 0.251 cr.243 -3.2 0.243 0.242 --0. 4 25 0.244 0.236 -3.3 0.239 0.238 --0.4 26 0.236 0.229 -3.0 0.235 0.234 --0. 4 27 0.228 0.220 -3.5 0.229 0.228 -0.4 28 0.221 0.214 -3.2 0.225 0.224 -0.4 29 0.277 0.254 -8.3 0.268 0.267 -0.4 30 o.251 0.229 -8.8 0.248 0.247 -o.4 31 0.238 0.239 +o. 4 0.238 0.236 -0.8 32 0.224 o.228 +1.8 0.228 0.229 -l-0.4 35 0.268 0.282 +5.2 0.259 0.262 +1.2 36 0.254 o.267 +5.1* 0.250 0.252 +o.8
"' 39 0.206 0.203 -1.5 o. 215 0.213 -0.9 O*
0 I
°'
- Without torsional degree of freedom
~** With torsional degree of freedom ~ ~ ...::l 0
z
~
co co c
ct, REACTOR
,__...,.__,_I~B-LDG. N CONTAINMENT VESSEL EL. 197.5 12 nZERO DISTANCE SHIELD BUILDING 1 EL. 198.3
<t.
REACTOR BLDG. 1---..::i::-----1 RAB PLAN EL. 176.0 13 2 EL. 172.4 't, REACTOR BLDG. EL. 154.0 14 3 EL. 150.7
/ RAB El.-35.0 4 EL. 131.0 ELEV.
STRUCTURAL LAYOUT 5 El. 111.0 FUEL HANDLING BUILDING EL.90.0 29 REACTOR AUXILIARY EL. 88.0 6 EL. 86.0 BUILDING BEAM ELEMENT 33,34,& 35 (TYP.) EL.69.0 EL. 66.0 J:L.61.5 22 7 EL. 61.0 EL.45.5 30 EL.55.0 23 36 EL.46.0 EL.46.0 24 19 COMBINED EL.44.0 8 STRUCTURE EL. 41.0 20 31 & 37 EL.21.0 131 L19.0 137 EL.8.0 CONNECTING MEMBER EL.-4.0 32&38 BETWEEN FHB& RAB 140 z RIGID UP LINK$ _ _ _ _ __ x MAT Y~---....- ROTATION CENTER NORTH AT GRID POINT 139. KH = SOIL SPRING, TRANSLATION, KIPS/FT
- Ke = SOIL SPRING, ROCKING, KIPS/RADIAN AMENDMENT NO. 12 (8/82)
FLORIDA POWER & LIGHT COMPANY
- ST. LUCIE PLANT UNIT 2 WATERFORD NO. 3 REACTOR BLDG.
' MATHEMATICAL MODEL (NO TORSIONAL EFFECT)
FIGURE 220.10-1
REACTOR BLDG .
. ~ ZERO DISTANCE CONTAINMENT SHIELD VESSEL BUILDING EL.197.5 12 1 EL.198.3 RAB PLAN EL.176.0 13 2 EL.172.4
<t_ REACTOR BLDG.
EL.154.0 14 3 EL.150.7 RAB EL.-35.0 EL.132.0 15 4 EL.131.0. ELEV. STRUCTURAL LAYOUT FUEL HANDLING EL.110.0 16 5 EL.111.0 BUILDING REACTOR AUXILIARY EL.90.0 29 BUILDING EL. 88.0 17 EL. 86.0 EL.61.5 22 30 EL.55.0 23 EL.46.0 24 EL.21.0 131 EL. 8.0 27 CONNECTING MEMBER BETWEEN EL.-4.0 28 FHB& RAB 140 z UP x v~-___..,..... ROTATION CENTER AT NORTH GRID POINT 139 K KH =SOIL SPRING, X-DIRECTION, KIPS/FT-z
&Hz KHy =SOIL SPRING, Y-DIRECTION, KIPS/FT ..
Kox Kz =SOIL SPRING, Z-DIRECTION, KIPS/FT 3 TRANSLATIONAL & Ke , Ke. =ROCKING SOIL SPRINGS 3 ROTATIONAL SOIL SPRINGS x y K<f> =TORSIONAL SOIL SPRING FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 WATERFORD NO. 3 REACTOR BUILDING MATHEMATICAL TORSION MODEL AMENDMENT NO. 12 (8/82) . FIGURE 220.10-2
SL2-FSAR Question No. 220.11 In Table 3.7-24 the maximtim moment at mass points 20, 22, 23 and 24 show smaller values for the time-history method than for the respbnse spectra method. Explain these differences since the time-history method is expected to yield a larger response than the response spectra method. This situation exists for other structures shown in other tables.
Response
Table 3.7-24 has been revised *to reflect correct values for maximum moments at mass points 20 and 22 for time-history method. These values are higher than response spectra method as anticipated. Maximum moments at mass *points 23 and 24 are negligibly higher for response spectra method than time-history method and could be attributed to the conservatism used in reading acceleration values £°rom ground response spectra and/or due to slightly different models used in two different methods. See revised Table 3.7-24
- 220.11-1 Amendment No. 5, (8/81)
SL2-FSAR Question No
- 220.12 The natural frequency of 70.36 shown in Table 3.7-18 for ES= 40 ksi, E-W direction, Mode 1, is in error. What is the correct frequency?
Response
The correct frequency is 1.36. Table 3.7-18 has been corrected. See revised Table 3.7-18 *
- 220.12-1 Amendment No 5, (8/81)
SL2-FSAR Question No
- 220.13 Outline the method used to account for differential structural moveme~t during an earthquake for piping that is supported by different seismic Category I structures.
Response
- 1. For piping that is supported by two buildings on a common mat or two structures within the same building, the relative seismic displacements between interface support/restraints are derived by taking the squa.re root of the sum of the squares (SRSS) of each relative seismic displacement towards the common reference. The common reference can either be the common mat or the structure base which is considered as anchorage in the analytical model of the structure.
- 2. For piping that is supported by two buildings on separate mats, the relative seismic displacements between interface support/restraints, in general, are derived from the combination .of co-directional maximum absolute seismic displacement of the two buiidings at the supporting elevation by square root of the sum of the square (SRSS) method. If the interface support/restraints at both buildings are located near the ground level or the two adjacent buildings have similar base response, the larger of the two maximum absolute
- 3.
seismic displacements may be considered as the maximum relative seismic displacement between the interface support/restraints. All the maximum relative seismic displacement are placed at the interface anchorage such as the penetration connections of the Reactor Steel Containment Vessel in the seismic displacement analysis. The seismic displacement analysis are at first performed for each of the three orthogonal directions independently. Then the result data are combined by SRSS method. No FSAR change required *
- 220.13-1 Amendment No. 5, (8/81)
812-FSAR
- Question No.
220.14
Response
Describe the method used to analyze the Turbine *Building for seismic motion. Although the Turbine Building is a nonseismic building, we have taken into account the equivalent static "g" loads in the stress analysis of the Turbine Building Framing. The seismic required design "g" values .for the Turbine Building structure evaluation were obtained from the dynamic seismic response analysis using a simplified model to represent the dynamic behavior of the Turbine Building structure. No FSAR change required *
- 220.14-1 Amendment No. 5, (8/81)
SL2-FSAR
- _Question No.
220.15
Response
Describe your criteria for system/subsystem decoupling. Review Plan 3.7.2 contains an acceptable criteria. Standard For the reactor building in particular, studies using seismic models with and without subsystem are made to ensure the coupling effect is minimal. Models with major equipment (such as steam generators and reactor vessel) and the supporting structure (i.~., the internal structure) modeled separately and modeled together are constructed and the Computer Code STARDYNE is employed. Dynamic responses such as frequencies, accelerations, and response spectra are compared. The differences-are found negligible. No FSAR change required *
- 220.15-1 Amendment No. 5, (8/81)
SL2-FSAR
- Question No.
220.16 (3.7.3) The comparison points in Table 3.7-38, sample problem 2 do not match. Show other point comparisons for each problems and show points where difference between methods is a maximum.
Response
Because of the difference in the consideration of load distribution between static and dynamic analyses, the maximum computer stress will not .always appear at the same point. However, with the consideration in the procedure delineated in Subsection 3. 7 .3.1.lb," the maximum computed stress will be higher .for the "Modified Equivalent Static Load" *method which is a frequency based static analysis. The original calculations for the three sample problems were performed on the basis of a flat response spectra of l.Og acceleration*. For an .actual response spectra generated fr<?m ordinary structure, the response value.s beyond *the resonant region usually decay rapidly. This is evidently demonstrated on the attached Figures 220.16-1 thru 220.16-3. The three sample problems have been recalculated for both flat response spectra and this more realistic response spectra using combination
*methods of NRC Regulatory Guide 1.92 Rev 1 and ASME Code Sullllll.er 1973 version for Code Class 2 & 3 piping
- The attached tables represent a comparison of pipe stresses computed by both Modified Equivalent Static Load Method" and the mode response spectra analysis. Table 220.16-1 is the
- result based on a flat response spectra of l.Og acceleration.
Table 220.16-2 is the re.sult based on 'the envelOped response on Figures 220.16-1 thru 220.16-3. In all cases, the maximum c~mputed stress is higher for the frequency based static analysis. As it is shown in Table 220.16-2 this is even more evidential in the comparison based on a real response s*pectra. It is worth to mention that while the sample problems 2 & 3 were arbitrarily picked from actual piping systems, the sample problem #1 does not reflect any normal restrained piping system. It was purposely modeled and restrained to exemplify the possible dynamic response of the piping. In general practice, the restraints are placed near the valves, corners and offsets as much as possible. No FSAR change is required *
- 220.16-1 Amendment No. 5, (8/81)
SL2-FSAR TABLE 220.16-2
- Sample Problem 1 (Fig. 3. 7-254)
Point No. HIGH STRESS COMPARISON based on floor response spectra Seismic Stress (PSI) Difference Static Dynamic 16 6853 1634 5219* 18 3645 1087 2558 1 3226 1936 1290 14 3152 1502 1650 15 2647 644 2003 4 2384 1524 860 30 711 1322 -611 29 709 1320 -611 Sample Problem 2 (Fig. 3. 7-255) (Fig. 3.7-256 Point No. Seismic ~tress (PSI) Difference Static Dynamic 31 3894 1315 2597 19 3673 1315 2358 17 3649 1010 2639* 21 3165 1236 1929 5 3042 1330 1712 13 2983 1101 1882 12 2784 1055 1729 30 2745 1005 1740 Sample Problem 3 (Fig. 3. 7-257) Point No. Seismic Stress (PSI) Difference Static Dynamic 27 3289 1134 2155* 1 3029 1324 1705 13 2804 1100 1704 2 4212 2518 1694 4 3162 1510 1652 20 2615 991 1624 8 3286 2055 1231 3 3139 2088 1051
- greatest difference
- 220.16-3 Amendment No. 5, (8/81)
0 M
- 0 N
ID 0 00 Ci) Ln CL
~
-(.)
(.) zw 0 MW a: LL z 0 1-(.) w a: ... c ID N ci (9) NOl.L'o'a31300'o' AMENDMENT NO. 5 (8/81) FLORIDA POWER & LIGHT COMPANY
- ST. LUCIE PLANT UNIT 2 RESPONSE SPECTRA USED FOR HIGH STRESS COMPARISON BETWEEN MODIFIED EQUIVALENT STATIC LOAD METHOD & MODE RESPONSE SPECTRA ANALYSIS FIGURE 220.16-1
0 M
- 0 N
0 C') co cg LC) enc..
->uu
'Id' 2 w
aw M c:: LL
<(
u l-a: w LC)
>I ci (9) NOl.L\1'=131300\1 AMENDMENT NO. 5 (8/81)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 RESPONSE SPECTRA USED FOR HIGH STRESS COMPARISON BETWEEN MODIFIED EQUIVALENT STATIC LOAD METHOD & MODE RESPONSE SPECTRA ANALYSIS FIGURE 220.16-2
0 C")
- 0 C"ll in 0
O'> 00 cc Ci) in Q..
~
u o::t zw
- I 0
w C") a: LL Z* 0 t:= ..... u w a: in 0 ci x (fl) NOl.1':fl:l31300'1 AMENDMENT NO. 5 (8/81) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 RESPONSE SPECTRA USED FOR HIGH STRESS COMPARISON BETWEEN MODIFIED EQUIVALENT
*STATIC LOAD METHOD & MODE RESPONSE SPECTRA ANALYSIS FIGURE 220.16-3
SL2-FSAR Question No
- 220.17 In Subsection 3.7.3.9, Item C, you used the words "significant support displacement." Define the threshold of significant and the basis for this lower bound if a lower bound is used.
Response
All the relative. seismic displacement between restraint/support points which may contribute an estimated bending stress more than five percent of the code allowable stress limit are considered to be significant. No FSAR change is required *
- 220.17-1 Amendment No. 5, (8/81)
SL2-FSAR Question ~o *
- 220.18
Response
In Subsection 3.7.3.3 you stated sufficient mass points will be included in the. model.and sufficient dynamic modes computed. Define "sufficient" for number of mass points and dynamic modes. If the mass points ~nd corresponding dynamic degrees of freedom are distributed to provide for appropriate representatton of the dynamic characteristics of the subsystem, then it is considered to be sufficient. As indicated in Subsection 3.7.3.1.1.a, the maximum spacing of the mass points may not exceed one half the distance for which the frequency of a simply supported beam would be 20 Hz* The criterion for sufficiency in number of dynamic modes is that the inclusion of additional modes does not result in more than a 10 percent increase in responses. In general, this can be satisfied by including all the dynamic modes bel9w 33 Hz, if the highest mode calculated by 33 Hz has already fallen into the flat rigid response region of the corresponding response spectra, the effect of the remaining high modes are taken care of by adding the dynamic analysis result with an equivalent static solution in SRSS summation. No FSAR change is required *
- 220.18-1 Amendment No. 5, (8/81)
SL2-FSAR Question No *
- 220.19
Response
In Subsection 3.7.3.4 what is your criteria for moving the frequency of.the subsystems with respect to the supporting system? The subsystems in general, are designed or restrained to be in the rigid region to avoid resonance with the supporting system. If the first mode period of the piping is more than 70 percent of the first mode period of the structure, a multimode response analysis is performed. If the first mode period of the piping is 70 percent or less of the first mode period of the structures, the procedures as outlined in Subsection 3.7.3.1.1.b and care followed as an alternative of analysis approach. No FSAR change is required *
- 220.19-1 Amendment No. 5, (8/81)
SL2-FSAR Question No
- 220.20 Your modal response .combination procedure (Subsection 3.7.3.7) uses only SRSS and omit consideration of closely spaced modes.
Use Regulatory Guide 1.92 for combining modes that are closely spaced and correct the _appropriate loads.
Response
The consideration of closely spaced modes for subsystems are stated in Subsections 3.7.3.7.1, 3.7.3.7.2, 3.7.3.1.1.a,4 and 3.7.3.1.2.3.d.b following Regulatory Guide 1.92 for combining modes that are closely spaced. No FSAR change is required
- i'
- 2.20.20-1 Amendment No. 5, (8/81)
SL2-FSAR Question No . 220.21 Your procedure in Subsection 3.7.3.1.1.a,l(b) for determining piping support locations is unnecessarily complicated. State the frequency you intend to use for support spacing.
Response
Subsection 3.7.3.1.1.a.l.(b) addresses to mass spacing not restraint spacing. As it is stated in the response to Question 220.18, the mass points are distributed to provide for appropriate representation of the dynamic characteristic of the piping system. The maximum spacing between mass points are limited so as to provide fair mode shape for all the significant modes. No FSAR change required *
- 220.21-1 Amendment No" 5, (8/81)
SL2-FSAR Question No
- 220. 22 Your description of the method for analyzing -the. buried seismic Category _I piping and tunnels is very sketchy. Provide a more detailed procedure and copies of calculations including any ref ere need mate ria 1.
Response
The seismic Category I buried piping between mats and floor slabs are encased in either a sand or cement mixture or a concrete fill. The *seismic Category I buried piping outside the buildings are embedded in class-1 compact backfill soil of sufficient density so that liquefaction shall not take place and backfill will not lose its integrity during a SSE. The effect of the loads imposed by the adjacent non-buried portion piping on the buried piping are considered for a distance into the soil based on . Heteny's approach (Reference 4) to a semi-infinite beam on elastic foundation. The loads' are considered to have died out in the form of a damped wave along the buried pipe further into the soil. For the portion of buried piping in which the pipe is not fully restrained by the soil and can be influenced by the non-buried piping, the soil resistance is simulated by a system of equivalent lateral restraints. The stiffness *of soil restraints and spacing of these restraints are calculated by considering the buried pipe as semi-infinite beam on elastic foundation. The friction and
- slippage effect which permit relative movement between the soil and the buried pipe are considered in evaluating the pipe stress due to seismic wave motion (Reference. 1, 3 & 5). The portion of buried piping is coupled with the nonburied piping and analyzed as a whole by the usual method delineated in FSAR.Subsection 3.7.3.1.1. The stresses in buried portion of piping due to various load including the loading due to relative movement between the soil and buildings are inclu~ed in the analysis and the stress limits for various load combinations given in Subsection NC-3650 or ND-3650 or ASME Section III are satisfied.
For the portion of buried piping in which the.pipe is fully-restrained by the surrounding soil and can not' be influenced by the non-buried piping,. the "pipe moves with the ground. assumption is used in evaluating the pipe stresses. The maximum axial, bending and shear stresses due to compressional wave and shear wave are calculated separat*ely using the stresses formulas presented by Nemark (Reference 6) and Yeh (Reference 9). Since the maximum stresses due to various seismic waves do not occur simultaneously, the maximum combined stresses are calculated by SRSS method using principal stress formula. Finally the combined stresses due to seismic wave motion are .combined with the longitudinal pressure stress and compared with the yield stress of the pipe for normal, upset and emergency conditions. For faulted condition, 2 times the yield 220.22-1 Amendment No. 6, (9/81)
SL2-FSAR stresses of the pipe is considered as the allowable. For buried
. Pipe near the buried elbows, the friction and slippage effect are considered following technique presented by Goodling (Reference 1
& 2).
In addition to the seismic consideration, the buried piping are also evaluated for plant settiement and maximum thermal expansion stresses with proper consideration of slippage and friction effect (Reference 1 & 3). The following literatures are referenced in the above mentioned calculation procedure which is in general agreement with SRP 3.7.3 criteria. References
- 1. E. C. Goodling, Jr. "Flexibility Analysis of Buried Piping",
an A5ME Publication 78-pvp-82.
- 2. E. C. Goodling, Jr., "Seismic Stress in Buried Elbow".
preprint 3595, ASCE National Convention, Boston, April 1979.
- 3. E. C. Goolding, Jr., More Flexibility Analysis of Buried Pipe", A5.ME Publication 80-C2/pvp-67.
- 4. M. Hetenyi, "Beams on Elastic Foundation", The Univer~ity of Michigan Press (1946).
5. 6. M. A. Igbal and E. c. Goodling, "Seismic Design of Buried Piping", ASCE CDnfernence ( 197 5). N. M. Newmark, "Earthquake Response Analysis of Reactor
*Structures", Nuclear Engineering and Design, Vol. 20., PP*
303-322 (197 2).
- 7. R. J. Roark and W. c. Young, "Formulas for Stress and Strain",
McGraw-Hill ( 1975). 8.. G. R. Tschebotarioff "Soil Mechanics, Foundations and Earth Structures", McGraw-Hill (1951).
- 9. G. C. K. Yeh "Seismic Analysis of Buried Metal or Concrete Pipes", April 1974.
220.22-2 'Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
220.23
Response
What method is used to determine the composite damping for the reactor vessel and the primary loop system. As illustrated in Subsection 3.7.3.1.2.1, uniform modal damping is used. We note that one percent damping is utilized for OBE and two percent damping is utilized for SSE. No FSAR change is required *
- Amendment No. 5, (8/81)
SL2-FSAR 0824W-3 Question No. 220.24 Provide a comparison of results of Unit 1 seismic analysis to the results of Unit 2 analysis to support your conclusion this site is a "Multi-unit site" addressed by Regulatory Guid~ 1.12.
Response
The St. Lucie Unit 2 structural designs.are essentially the same as those of Unit 1, hence the seismic responses expected t9 be experienced in the St. Lucie Unit 2 plant are similiar to those of Unit 1 plant. Using identical seismic inputs, a s.eismic response analysis of the structures comprising St. Lucie Units 1 and 2. would demonstrate identical effects for both units. No FSAR change is required
- 220.24-1 Amendment Noo 5, (8/81)
SL2-FSAR Question No
- 220. 25 The ductility relationships for reinforced concrete beams and (3.5) slabs are larger than those acceptable to the staff. Re-evaluate your concrete beams and slabs for a ductility of
.05 ~10. The ductility ratios for steel members are larger p-p' than those acceptable to the staff. Re-evaluate the beams for a ductility ratio of 10 and columns with a kl/r ~ 20 for a ductility ratio of l* 7. For columns with a kl/r ) 20 use a ratio of 1.0.
Response
We have evaluated concrete slabs for ductility of
.05 ~ 10 and have found them acceptable. The concrete beams p-p' had already been designed for that valve. See Table 3.5-12 *
- 220.25-1 Amendment No. 5, (8/81)
SL2-FSAR. Question No
- 220. 26 Provide a comparison of your load combinations with the load combination equations in Standard Review Plan 3.8.2, 11~3 and address the effects of not meeting the load combinations, including any loads that are missing from your combinations.
Response
Load Combination St. Lucie 2 Remarks SRP 3. 8. 2 Table 3. 8-1 (1) D + L +.W Construction load case (1) is equivalent to SRP load combination (1) with Pt and Tt not applicaple. W = lateral wind load. (1) D + L + Pt + Tt (2) D+W+ Pt + Tt + L SL 2*1oad combinations (2) 16 or* (3) is equivalent to (3) D + E + pt + Tt SRP load combination (1)
+Ro + T0 + L except with W0 r E, R0 and 16 T0 added. *
-* (2) D + L + To + Ro ( 3) D + L + To + Ro
+ E (4, 5) D + L + T0
+ Ro + E + Pe SL2 load combination (4) or (5) is equivalent to SRP load combination (3)
(governs over SRP load combination (2)) except Pe was added to the load case. ( 4) D + L + Ta + Ra (6) D + Ta + Ra,+ Equivalent
+Pa+ E Pa + E + L j6
( 5) D + L + Te + Re (4,5) D + L + Te Equivalent
+Pe+ E +Re + Pe + E
- 220.26-1 Amendment No. 6, (9/81)
- SL2-FSAR Load Combinatfon SRP 3.8.2 (6) D + L + Ta + Ra
+ Pa + E' St Lucie 2 Table 3. 8-1 (7) D +Ta+ Ra
+Pa+ E'+ L Equivalent Remarks (3) D + L + T0 (8) D + L + T0 812 load combinations equivalent to SRP load
+Ro+ E + *Ra + E + Yr combination (3) except that Yr or Yj was added to (10) D + L + T0 the load case.
+Ro*+ E + yj (7) D + L +Te (12,13) D + L + Te + Equivalent 6
+ Re + Pe + E' Re+ Pe+ E' (8) D + L + Ta (9) D + L + T0 St~ Lude 2 load combination is equivalent to SRP load
+ Ro + Yr + E' combination.with Ym, Pa
*and Yj not applicable. With
+ Yj + Ym + E' Pa = O, Ra and Ta reduce to Ro and T0 respectively.
(8) D + L + Ta (11) D + Ta + Ra St. Lucie 2 load combination is equivalent to SRP load
+ Pa + Yj + *E' + L combination with Yr and 16 '
Ym not applicable. SRP load combination (9) was not compared because FL is not applicable *. With the elimination of this term, this SRP load combination is enveloped by SRP load combination (3). No E!SAR change .requir;e*d~ 220.26-2 Amendment No. 6, ( 9 I 81) __
SL2-FSAR
- Question No
- 220.27 Describe how the containment steel shell is anchored to the concrete foundation slab. Describe the procedures used to account for the shear stresses between the steel shell and the concrete on both sides of the ellipsoidal head for the loads which will produce these stresses.
Response
The steel containment ellipsoidal bottom head is completely embedded in concrete. The containment dead weight and any overturning moments due to 'seismic are assumed to be transferred to the concrete in bearing. Shear stresses are assumed to be zero. We believe that these are valid assumptions s~nce the head is not hemispherical. No FSAR change is required *
- 220.27-1 Amendment No. 5, (8/81)
SL2-FSAR Question No *
- 220.28 The methodology used to analyze the containment shell to guard against buckling is not completely described. Provide the following:
(1) Details of the assumptions and boundary conditions used in the analyses of the dome and the cylinder and justify why the analysis for each section was done separately. (2) All the load combinations which are considered critical (the limiting cases) for the buckling analysis. For each load combination state the most likely effected regions of the shell for this type of compressive loadings. (3) Compare and justify the methodology used in your analysis with the acceptable methods stated in the current version of the ASME Code subsubarticle NE-3100. Provide a discussion of the factor of safety for each service level. (4) Provide a copy of the referenced papers used in the buckling analysis of the containment shell~
Response
(1) The design of the containment is in accordance with ASME Code Section III, NE-3133 design rules with assumptions and boundary conditions inherent in the design rul~s* (2) The loading combinations are considered critical for buckling. a) Case 5 - Cold shutdown at ambient temperature. This case includes OBE seismic with external pressure. b) Case 9 - Condition with safe shutdown earthquake. This case includes SSE seismic with no internal pressure. The design most likely affected regions of the shell are the top head near the cylinder junction and the bottom tangent line on the cylinder. Buckling in the ellipsoidal head is not considered since it is embedded in concrete. No FSAR change is required *
- 220.28-1 Amendment No. 5, (8/81)
SL2-FSAR (3) As stated above, the design of the containment is in accordance with ASME Code Section III, NE-3100 design rules. External.pressure for cylinder and head are checked using design rules in NE-3133.3 and NE-3133.4. The cylinder is checked for axial compression using the design rules in NE-3133.6. Seismic and dead loads are considered to cause axial compression. For those areas with unequal biaxial compressive stresses, *the ASME rules
.as modified by WRC 69 have been used. Basically, the allowable stress for compression determined by NE-3133 remains the same for all "design conditions" except for SSE earthquake where the ASME Sect III; Winter 1972 Addenda NE-3131 C(2) allows a 20 percent increase. The St Lucie 2 .design did not use this 20 percent increase.
(4) Reference paper is WRC 69, June, 1961. 220.28-2 Amendment No. 5, (8/81)
SL2-FSAR
- Question No.
220.29 State the code used in the design of the steel structural supports for the reactor coolant system and show a comparison of the code used to the current version of the ASME Code Section III, Division I, Subsection NF. Also show a comparison of the ACI 349 Code to the ACI 318-71 Code you used for design of the Concrete Internal Structure.
Response
Part I The AISC code is used in the design of the structural steel supports for the Reactor Coolant Systems. Refer to Subsections 3.8.3.2.1 and 3.8.3.5.2. Part II The major difference between the ACI 318-71 and ACI 349 codes is in the area of design loading. The design loads specified in AC! 349 are supplemented by RG 1.142 which m:*akes the design loads consistent with those presented in SRP 3.8.4. Refer to the attached partial lineup of SRP 3.8.4 for comparison of load combinations specified therein and those used in the design of seismic Category I structures. The attached lineup of RG 1.142 gives supplemented requirements on design procedures to the AC! 349 Code, with statements of compliance, alternate compliance and remarks on impact of deviations. RG 1.142 also requires, that for concrete structures used to provide radiation shielding, the provisions of Sections 5,1 and 10 of ANSI Standard NlOl. 6-1972 "Concrete Radiation Shields" be followed. The provisions of those sections are followed with the following clarifications: ANSI NlOl. 6-72 Section Clarifications 12 5.1.2 No high density concrete is used. 5.1.3 No hydrous aggregate is used. 5.1.4 No boron containing aggregates are used. 5.1. 6 Coatings of clay, silt, gypsum, calcite or caliche on coarse aggregate total no more than three and one half percent of the total weight of the aggregate. Radiation attenuation calculations take this into account *
- 220.29-1 Amendment No. 12, (8/82)
SL2-FSAR 10.1. 2 10.1.3 Dimensional tolerances for hatches and openings as specified in ACI-347 are used rather tgan those given in Table 1 of ANSI NlOl.6-72. Minimum practicable joint clearances are specified. Service trenches are not used. 12 10.2.2 The weight of each block is indicated on the design drawing, not marked on the block. 10.2.3 Blocks are cured according to good construction practice, e.g., use of wet burlap or curing compound, but not necessarily in the absence of direct sunlight or heat. This sunlight or heat, however, does not result in the loss of shielding efficiency. 10.3.1 There are no present plans for penetrations through shielding plugs. However, if they are required, streaming is prevented by proper design of the penetration. 10.4 No movable or removable poured walls are used. 10.6 Precast shielding components are fabricated at the site. No FSAR change is required. 220.29 .. la Amendment No. 12, (8/82)
DOCUMENT: SRP J.8.4 TITLE: SL2-FSAR OTHER SEISMIC CATEGORY I STRUCTURES ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS Load Combinations for Concrete Structures For concrete structures, the load combinations are acceptable if found in accordance with the fol-lowing: a - For service load conditions, a - The strength design either the working stress method was used. design (WSD) method or the strength design method may be used. i - If the WSD method is used, i - Not applicable the following load combin-ations should be considered: (1) D + L (2) D + L +E 0 (3) D + L + W \0 I If thermal stresses due to T0
"' and R0 , are present, the fol-lowing combinations should be considered:
(la) D+L+T0 + Ro (2a) D+L+T0 + Ra +E (3a) D+L+T0 +Ra +w Both cases of L having its full value or being completely absent should be checked.- r t
- sl"t
.z 0
DOCUMENT: SRP 3.8.4 (Cont'd) SL2-FSAR ACCEPTANCE CRITERIA COHPLIANCE ALTERNATE COMPLIANCE REMARKS ii - If the strength design i i - St. Lucie Unit 2 design ii - The alternate load combinations Since the acceptance criteria method is used, the following complies with the loac! used are: load combinations have a multi-load combinations should be combinations listed, plication factor of 0.75, the considered: with the exception of (lb) 1.4 (B + D) + 1.3 combined loads used as ident-load combinations (1 b), (R0 + T0 ) + 1. 7 (L + H) ified in alternate compli-(1) 1.4 D + 1.7 L (2b) and (3b). (2b) 1.4 (B + D) + 1.3 ance for the St. Lucie Unit 2 (2) 1.4 D + 1.7 L + 1.9 E (Ro+ To) + 1.7 (L + H') design would be greater for all (3) 1.4 D + 1.7 L + 1.7 W + 1.9 E design cases. The load combi-(3b) 1.4 (B' + D) + 1.3 nations used on St. Lucie Unit 2 If thermal stresses due to (R0 + T0 ) + 1.7 (L + H) were based upon guidance pro-T0 and R0 are present, the + W) vided in AEC letter to FP&L following combinations Co., August 30, 1973, "Enclo-should also be considered: where sure 2 - Structural Design B = Buoyancy at normal Criteria for Category I (lb) (0.75) (1.4 D + 1.7 L groundwater level Structures Ou_t;e_ide the Con-1.7T0 +1.7 R0 ) tainment," in addition to those given in the.AC! 318-71 Code. (2b) (0.75) (1.4 D + 1.7 L B' Buoyancy at maximum
+ 1.9 E + 1.7 T0 + groundwater level re-0"' 1.7 Ro) resulting from a PMH.
I (3b) (0.75) (1.4 D + 1.7 L H Lateral earth* loads
"' +1. 7 W + 1. 7 T0 + under normal conditions 1.7 R0 H' Lateral earth loads Both cases of L having its under normal and earth-full value or being com- quake conditions pletely absent should be checked. In addition the following comhination should be considered:
(2b') 1.2 D + 1.9 E (3b') 1.2D+1.7 W Where soil and hydrostatic Soil and hydrostatic pres-pressures are present, in sures are included in the addition to an the above design and the requirements combinations where they have of ACI-318-71 Sections been included in L and D 9.3.4 and 9.3.5 were con- ,.... respectively, the require- sidered. 00 ...... ments of Sections 9.3.4 and
....00 9.3.5 of ACI-318-71 (Ref 1)
""' should also be satisfied.
SL2-FSAR DOCUMENT: SRP 3.8.4 (Cont'd) ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS b - For factored load conditions, b - St Lucie Unit 2 design complies which represent extreme environ- with the load combinations mental, abnormal, abnormal/severe listed. environmental and abnormal/ extreme environmental conditions, the strength design method should be used and the following load combinations should be considered. (4) D + L + T0 + R0 + E' (5) D + L + T0 + R0 + Wt (6) D + L + Ta + Ra + 1.5 Pa (7) D + L +Ta + Ra + 1.25 Pa + 1.0 (Yr + Yj + Ym) + 1.25 E (8) D + L + Ta + Ra + 1.0 Pa = 1.0 (Yr+ Yj + Ym) + LO E' In combinations (6), (7), and (8), the maximum values of Pa, Ta, Ra, Yj, Yr and Ym, includ-ing an appropriate dynamic load factor, should be used unless a time-history analysis is performed to justify otherwise. Combinations (5), (7), and (8) and the corre-sponding structural acceptance criteria of Section 11.5 of this plan should be satisfied first without the tornado missile load in (5) and without Yr, Yj, and Ym in (7) and (8). When consid-ering these concentrated loads, local section strength capacities may be exceeded provided there will be no loss of function of any safety-related system. Both cases of L having its full value or being completely absent should be checked.
SL2-FSAR DOCUMENT: SRP 3.8.4 (Cont'd) ANSI NlOl. 6-72 Section Clarifications 5.1. 2 No high density concrete is used. 5.1.3 No hydrous aggregate is used. 5.1. 4 No boron containing aggregates are used. 5.1.6 Coatings of clay, silt, gypsum, calcite or caliche on coarse aggregate total no more than three and one half percent of the total weight of the aggregate. Radiation attenuation calculations take this into account. 10.1. 2 Dimensional tolerances for hatches and openings as specified in ACI-347 are used rather than those given in Table 1 of ANSI NlOl.6-72. Minimum practicable joint clearances are specified. 10.1. 3 Service trenches are not used. 10.2.2 The weight of each block is indicated on the design drawing, not marked on the block. 10.2.3 Blocks are cured according to good construction practice, e.g.; use of wet burlap or curing compound, but not necessarily in the absence of direct sunlight or heat. This sunlight or heat, however, does not result in the loss of shielding efficiency. 10.3.1 There are no present plans for penetrations through shielding plugs. However, if they are required, streaming is prevented by proper design of the penetration. 10.4 No movable or removable poured walls are used. 10.6 Precast shielding components are fabricated at the site. 220.29-5 Amendment No. 5, (8/81)
SL2-FSAR DOCUMENT: RG 1.142 (REV. 0) TITLE: SAFETY-RELATED CONCRETE STRUCTURES FOR NUCLEAR POWER PLANTS (OTHER THAN REACTOR VESSELS AND CONTAINMENT) ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS The procedures and requirements described in AC! The design and analysis procedures Design and analysis of St. Lucie Standard 349-76, "Code Requirements for Nuclear Safety utilized for safety-related concrete Unit 2 started before AC! 349-76 Related Concrete Structures," are generally acceptable structures are in accordance with was issued. to the NRC staff and provide an adequate basis for the AC! 318-71 Code. complying with the Commission's regulations with regard to the design of safety-related concrete structures other than reactor vessels and containments, subject to the following:
- 1. The applicability of strength design methods 1. Not applicable to St Lucie -
to structures whose principal function is to Unit 2. provide a barrier to contain or retain pressure such as the divider barrier of the ice-condenser of the PWR containment is questionable. 'There-fore, for those structures, mere conformance with the requirements of AC! 349-76 is unaccep-table to the staff, who will continue to review the design of these structures on a case- by-case basis. N
- 2. When concrete structures are used to provide 2. Refer to response to Question N
0 radiation shielding, the provisions of 220.29 for compliance to N Sections 5.1 and 10 of ANSI Standard Sections 5.1 and 10 of ANSI NlOl.6-1972,2 "Concrete Radiation Shields," Standard NlOl,6-1972. "'I and those of ANSI Standard NlOl.4-1972,3 as
°' endorsed by Regulatory Guide 1.54, "Quality Assurance Requirements for Protective Coatings Applied to \later-Cooled Nuclear Power Plants,"
are applicable
- I 12 N
SL2-FSAR DOCUMENT: RG 1.142 (REV. 0) (Cont'd) ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS
- 3. ACI Standard 349-76 lacks specific require- 3. Appendix A of ACI 318 "Special ments to ensure ductility of framed struc- Provisions for Seismic Design" tures. Adherence to the requirements of is applicable when seismic Appendix A to ACI Standard 318-71 is loads are based on empirical acceptable. formulae such as those of the Unified Building Code. For the Category I structures, seismic loads are obtained from dynamic analysis of the structures based on SSE and OBE design response spectra.
Shear walls and bracing systems are designed to take the seismic forces calculated from such analysis. For these N reasons Ebasco feels that the N 0 N requirements of Appendix A of ACI 318 are not applicable to the nuclear plant structures I whose design is based on con-servative criteria and detailed seismic analysis.
- 4. Section S.1.2 permits depositing concrete 4. Water is removed before 4. Refer to Concrete Specifi-without the prior removal of.water from the concrete is deposited. cation FLO 2998.473.
place of deposit at the discretion of the owner. Since the presence of water in the place of deposit may seriously affect the strength properties of concrete, it is important that water be removed before con-crete is deposited unless a tremie is used.
- s. Section S.4.1 allows concrete that has parti- S. The placement of par- s. Refer to Concrete Specifi-
- <>- ally hardened or has been contaminated with tially hardened, contam- cation FLO 2998.473.
9 ro foreign materials or remixed after initial inated or retempered
- sp.
set to be reused at the discretion of the concrete is not permitted. ~ engineer. Such a material would be defective
- s rt and therefore should not be used.
z ? l/1 00 00
- DOCUMENT: RG 1.142 (REV. O) (Cont'd)
SL2-FSAR ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS
- 6. In addition to the requirements of Section 6. FP&L to indicate compliance 1.3.l of AC! Standard 349-76, the inspectors based on site inspection should have sufficient experience in rein- practices.
forced and prestressed concrete practice to interpret plans and specifications. The inspectors should be thoroughly familiar with the applicable AC! and ASTM Standards. AC! Standard 311-74,1 "Recommended Practice for Concrete Inspection," should be followed except where the requirements of Section 1.5 of ACI Standard 349-76 control.
- 7. The frequency of cylinder testing required by ANSI N45.2.S-1974 concrete Section 4.3.1 of AC! Standard 349-76 is not cylinder testing frequency consistent with generally accepted practice. was followed on St. Lucie A test frequency in conformance with ANSI Unit 2.
Standard N4S.2.S-1974,4 as endorsed by Regulatory Guide 1. 94, "Quality Ass.uran9e Requirements for Installation, Inspection, ..,.., and Testing of Structural Concrete and Structural Steel During the Construction 0 \C Phase of Nuclear Power Plants," is acceptable
- I 00
- 8. The minimum pressure-testing requirements 8. Not applicable. AC! Standard for embedded piping of AC! Standard 318-71 318-71, Section 6.3.2.4, indi-have been deleted from AC! Standard 349-76. cates that piping, with the In order ~o ensure that minimum pressure- exceptions of Section 6.3.2.S, testing requirements are met, the pressure is to be tested prior to con-tests of embedded pipes in Section 6.3.2.4 of creting. Section 6.* 3.2.S is AC! 349-76 should also satisfy the require- as follows:
ments of Subsection 6.3.2.4 of AC! 318-71.
"Drain pipes and other piping designed for pressures of not more than 1 psi above atmos-pheric pressure need not be tested as required in Section 6.3.2.4."
z 0 00 ~
....00
SL2-FSAR DOCUMENT: RG 1.142 (REV. 0) (Cont'd) ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS
- 9. More conservative load factors are appropriate 9, Load factors utilized are pre-in accounting for the effects of normal or sented in SRP 3.8.4 line-up.
shutdown thermal loads, postulated pipe break The noted load factor changes accidents, and an operating basis earthquake make the load combination con-(OBE) in combination with a postulated pipe sistant with those presented _break. The load factors used in Section 9.3.l in SRP 3.8.4. of ACI Standard 349-76 are acceptable to the staff except for the following:
- a. In load combinations (9), (10), and (11),
1.7 T0 should be used in place of 1.4 To.
- b. In load combination (6), 1.5 Pa should be used in place of 1.25 Pa*
- c. In load combination (7), 1.25 Pa and 1.25 E0 should be used in place of 1.15 Pa and 1.15 E0 , respectively.
0
- d. In load combinations (2) and (10), 1.9
\0 I E0 should be used in place of 1.7 E0 \0
- 10. Structures must be able to withstand the 10. Not applicable since each effects of differential settlement under safety-related concrete environmental loads as well as under abnormal structure is supported on an loads. Thus, in Section 9.3.2 of ACI 349-76, individual mat. Differential consideration of the effects of differential settlement within a building settlement should be included in load combi- was not expected to occur and nations (1) through (11). was not included as a design consideration.
- 11. The provisions of Section 9.3.3 of ACI Stand- 11. Load combinations used in design 11. The load combinations used on ard 349-76 to account for the effects of either useful dead and line loads St Lucie Unit 2 were based upon a>
CD transitory loads are not sufficiently general. or full dead and zero line loads. guidance provided in AEC letter E. Thus, in Section 9.3.3 of ACI Standard 349-76, to FP&L Co., 8-30-73, "Enclo-a CD when any load reduces the effects of other sure 2-Structural Design
- ! loads, the corresponding coefficient for that Criteria for Category I Struc-
'!<:"' load should be taken as 0.9 if it can be tures Outside the Containment," ;:> demonstrated that the load is always present in addition to those given in \.Jl or occurs simultaneously with the other loads. the ACI 318-71 Code.
Otherwise, the coefficient for that load a> should be taken as zero. CX>
* *SL2-FSAR DOCUMENT: RG 1.142 (REV. 0) (Cont'd)
ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS
- 11. (Continued)
Exeption is taken to the regulatory position which re-quires that a coefficient of 0.9 or zero be applied to any load which reduces the effects of other loads. Considering live load as having its full value or being completely absent satisfies the require-ment for setting a transitory load to zero. However, apply-ing the 0.9 coefficient to all other such mitigating loads, which are always present or occur simultaneously, would increase the number of load ...,..., combinations to an impractical 0 level with no demonstrated or ..., meaningful increase in the overall conservatism of the I governing load combinations
- 0 Our position is consistent with ACI 318-77 and ACI 349-76.
- 12. The provision in Section 9.3.6 of ACI Standard 12. Design criteria used on 349-76 permitting local exceedance of section St Lucie Unit 2 does not permit strength under concentrated dynamic loads does local exceedance of section not ensure that the section can withstand the strength.
associated distributed loadings. ',rhus, if the provision of Section 9.3.6 of ACI 349-76 per-mitting exceedance of local section strengths is invoked, it should be demonstrated that section strengths are adequate to accommodate load combinations (7) and (8) without the dynamic loads Yj, Ym* and Yr* 00 00
SL2-FSAR DOCUMENT: RG 1.142 (REV. 0) (Cont'd) ACCEPTANCE CRITERIA COMPLIANCE ALTERNATE COMPLIANCE REMARKS
- 13. The NRC staff would accept the local excee- 13. Same as 12 above.
dance of section strength for concentrated tornado-generated"""111issile loading under load combination (5). However, an analysis should be performed to demonstrate that section strengths are adequate to accommodate load combination (5) without the dynamic load effect of tornado-generated missiles.
- 14. ACI Standard 349-76 does not address the 14. Provisions of Section subject of openings in slabs and footings. 11-12 of ACI 318-71 Provisions of Section 11.12 of ACI 318-71 are followed.
are acceptable for this purpose.
SL2-FSAR Question No *
- 220.30 You stated that the allowable stress for the factored load combinations was increased. These load combinations contain the earthquake loading. The staff does not allow any increase in the allowable for earthquake loads. Re-evaluate the structures without the increase in the allowable stresses and provide the results of your re-evaluation.
Response
Our load combinations and allowable stresses, comply with those in the *Standard Review Plan 3.8.3. SRP 3.8.3 II 5 does allow an increase in the allowables for earthquake loads in the factored load combinations. No FSAR change is required *
- 220.30-1 Amendment No. 12, (8/82)
SL2-FSAR
- Question No.
220.31 (3.8.3) (3.8.4) In several analysis you have used static loads to represent a dynamic loading. Provide your procedures for transforming the dyna~ic loads into static loads, reference pipe reactions and pressurization of shield wall.
Response
The St. Lucie 2 structural design utilizes equivalent static load methodology. The loading combination for all structural loads is provided in FSAR Subsection 3.8.4. The method of transforming the dynamic loads with equivalent static loads is as follows: (1) The dynamic piping loads (i.e., seismic, relief valve discharge) are transformed into static loads in the piping stress analysis by either utilizing the model response spectra method or by applying an amplification factor to the excitation load. These pipe loads on structures are considered in the design for both the positive and negative directions with the same magnitude. (2) The dynamic effects of pipe whip and jet impingement are discussed in FSAR Section 3.6. The pipe whip loads on the structural components (i.e., p~pe whip restraints) were calculated utilizing static methods while various confirmatory dynamic analysis were utilized to confirm these piping loads (refer to Appendix 3.6E). The jet impingement analysis utilized a Dynamic Load Factor (DLF) of two applied to the KPA load to determine the loads. on structural components '(refer to Subsection 3.6.2). (3) The dynamic effects associated with the containment subcompartment pressure analysis were considered in the structural design of the subcompartment walls. The equivalent static loads were obtained by applying a Dynamic Load Factor to the peak of each dynamic loads. The calculated peak pressure for each subcompartment and the corresponding design values (with calculated margin) will be provided in FSAR 12 Subsection 6.2.3. No FSAR change is required *
- 220.31-1 Amendment No. 12, (8/82)
SL2-FSAR Question No
- 220.32 You stated that the cable tray restraints were designed for a "minimum natural frequency within 16 hz". You further say the HVAC restraints are designed. with a minimum natural frequency of 15 hz. What provisions were made to ensure the first natural frequency was 15 or 16 hz and how do you account for higher modes in the systems? Also state how the restraints are anchored to the structure. FSAR Subsection 3.8.3.1.5 contains several restraint designs.
Response
The first natural frequency of the cable tray (HVAC) restraint is determined after.member selection to ensure that the minimum natural frequency of 16 hz (15 hz) is satisfied. Amplification
. factors are used to account f.or the participation of higher modes. The cable tray (HVAC) restraints are welded to steel embedments with a fillet weld all around.
No FSAR change is required *
- 220.32-1 Amendment No. 5, (8/81)
0823W-4 SL2-FSAR Question No
- 220.33 You stated you used only the passive earth pressure on the portion (3.8.5.5) of the structures to resist sliding. Discuss how you accounted for this load on the walls and provide a table showing the structure and the maximum earth pressure.
Response
FSAR Subsection 3.8.5.5 states that, if we discount the ability of the waterproofing membrane (beneath the Shield Building and Reactor Auxiliary Building) to resist shear forces, passive earth pressures would be sufficient to resist sliding. A design impact review of this assumption has been completed and results are as follows: Shield Building - A substantial portion of the Shield Building has structural concrete and fill concrete throughout the building cross-section below plant island grade elevation. Therefore, the soil passive pressure loads on the building are of no design significance. However, the design analysis of the building determined a seismic movement and resisting earth pressure less than the full passive earth pressure. This requires the waterproofing membrane to be able to transfer shear forces . Therefore, the design calculations for resistance to sliding are based upon the combination of resisting earth pressure and the membrane shear strength value documented by the manufacturer. See Figure 220,33-1 for design resisting soil pressures. 11 Reactor Auxiliary Building - The building general arrangement and resulting mat layout allows sliding to be resisted through internal shear resistance of the soil and i*esisting earth pressures, in the east, west and south directions. The ability of the waterproofing membrane to transfer shear forces is not required in those directions. In the north direction, the earth pressure required to resist sliding is greater than the design capacity of the building walls. Therefore, the design calculations for resistance to sliding in that direction are based upon the combination of resisting earth pressure and the membrane shear strength value. See Figure 220.33-2 for design resisting soil pressures. 11 See FSAR Subsection 3.8.5.5 for revised statement on the design consideration of waterproofing membrane in structure sliding analysis *
- 220.33-1 Amendment No. 11, (7/82)
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(.L::I) 30'Vl::IE> M0138 H.Ld30 AMENDMENT NO. 11 (7/82) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 FB-SOIL PRESSURE UNDER EARTHQUAKE FIGURE 220.33-1
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SL2-FSAR
- Question No.
220.34 (3.8.5.2) State the codes used and list any deviations to the codes used for foundation design. Compare the codes used to the present version of the codes showing deviations and the effect of these devia.tions.
Response
Standard Review Plan 3.8.5 "Foundations" refers to Standard Review Plan 3.8.3 "Internal Structures of Containments" for f.oundation design codes and ACI 318-71 is the only applicable code listed. 12 St. Lucie Unit 2 foundation design is in accordance with ACI 318-71 as supplemented by Standad Review Plan 3.8.4 "Other Seismic 12 Category I Structures." See response to Question 220.29 for comparison of SRP 3.8.4 load combination requirements and those used in the St. Lucie Unit 2 design. A review of the 1977 edition of ACI 318 Code has determined that the changes have insignificant effect on foundation design requirements. No FSAR change is required *
- 220 .34-1 Amendment No. 12, (8/82)
SL2-FSAR Question No
- 220.35 Provide a tabie showing the factor of* safety against sliding, overturning and flotation for the load combinations shown in Standard Review Plan 3.8.5.
Response
The attached tables show the factors of safety against sliding; overturning and flotation for the load combinations shown in Standard Review Plan 3.8.5 for the major and typical seismic Category I buildings. No change to the FSAR is required
- 220 .35-1 Amendment No. 5, (8/81)
SL2-FSAR TABLE 220.35-1 REACTOR BUILDING FACTOR OF SAFETY AGAINST LOAD COMBINATION SLIDING OVERTURNING FLOATATION D + H + E )1.5 )2.69 D+ H +W )8.7 )21.7 110 D + H + El 1.23 2.69 D + H + Wt 8.7 21. 7 D + pl 3.12 D = Dead Loads E = QBE El = DBE
- w Wt pl
= Hurricance Wind @ 194 mph
= Tornado Wind @ 300 mph
= Buoyancy, Max GWT EL + 21.00 H = Soil Pressure 220.35-2 Amendment No. 10, (6/82)
SL2-FSAR TABLE 220.35-2
- LOAD COMBINATION REACTOR AUXILIARY BUILDING SLIDING FACTOR OF SAFETY AGAINST OVERTURNING FLOATATION D + H + E 1.35 2.64 D + II + W 3.88 4 * .14 110 D + H + El .1.16 2.15 2.84 3 * .15 2.35 D = Dead Loads E = OBE El = DBE w = Hurricance Wind @ 194 mph Wt *- Tornado Wind @ 300 mph
. Fl Buoyancy,_ Max GWT EL + 17.00 H = Soil Pressure
- Amendment No. 10, (6/82) ..
220.35-3
SL2-FSAR TABLE 220.35-3
- LOAD COMBINATION
.CONDENSATE STORAGE TANK SLIDING FACTOR OF SAFETY AGAINST OVERTURNING FLOATATION D + H +E )2.24 )2.54
( 10 D+ H+ W )4.71 )3.93 D + H + El 2.24 2.54
*. D + H +Wt 4.71 3.93 D +.F 1 - 5.90 D = Dead Loads Note:
E = QBE Factors of safety for load combinations D+H+E and D+H+W El = DBE will be higher than for D+H+El and D+H+Wt respectively.
- w Wt Fl
= Hurricance Wind @ 194 mph
= Tornado Wind @ 360 mph
= Buoyancy, Max GWT EL+ 17.00 H = Soil Pressure
- Amendment No. 10, (6/82) 220.35-4
SL2-FSAR TABLE 220.35-4
- LOAD COMBINATION FUEL HANDLING BUILDING SLIDING FACTOR OF SAFETY AGAINST OVERTURNING FLOATATION
*D+:H*+E 2.11 2.33 D + H+W )l.50 )l.50 I 10' D + H + E;l 1.25 1.39 4.09 4.45 9.1 D = Dead Loads Note:
E = OBE Dyn soil pressure ) active soil pressure in calculations El = DBE H is neglected and this is con-servative. w Hurricance Wind @ 194 mph
=
Wt = Tornado Wind @ 360 mph Fl = Buoyancy, Max GWT EL + 17.00 H = Soil Pressure
- 220.35-5 Amendment No. 10, (6/82)
SL2-FSAR TABLE 220.35-5
- LOAD COMBINATION DIESEL GENERATOR BUILDING SLIDING FACTOR OF SAFETY AGAINST OVERTURNING FLOATATION D + H + E 2.98 9.68 D+ H +W )7.45 )23.92 110 D +*H +El l.55 4.71 7.45 23.92 6.44 D = Dead Loads E = OBE El =DBE w = Hurricance Wind @ 194 mph
- Wt Fl H
= Tornado Wind @ 360 mph
= Buoyancy, Max GWT EL+ 17.00
= Soil Pressure
- 220.35-6 Amendment No. 10, (6/82)
SL2-FSAR TABLE 220.35-6 COMPONENT COOLING FACTOR OF SAFETY AGAINST LOAD COMBINATION SLIDING OVERTURNING . FLOATATION D + H + E )1.52 )4.16 D+ H+ W 110
)3.07 )12.68 D + H + El l.52 4.16 D + H +Wt 3.07 12.68 D + Fl 3.ll D = Dead Loads E = OBE El = DBE
- w Wt Fl
= Hurricance Wind @ 194 mph
= Tornado Wind @ 360 mph
= Buoyancy, Max GWT EL+ 17.00 H = Soil Pressure 220.35-7 Amendment No. 10, (6/82)
SL2-FSAR TABLE 220.35-7 INTAKE STRUCTURE FACTOR OF SAFETY AGAINST LOAD COMBINATION SLIDING OVERTURNING FLOATATION D + H + E 1.64 1.52 D + H+ W 4.38 1.65 I 10 D + H + El 1.13 l.2J 3.83 1.61 2.31 D = Dead Loads E OBE El = DBE w = Hurricance Wind @ 194 mph
- Wt Fl H
= Tornado Wind @ 360 mph
= Buoyancy, Max GWT EL + 16.00
= Soil Pressure
- 220.35-8 Amendment No. 10, (6/82)
- SL2-FSAR question No
- 220.36 The load combinations listed in Subsection 3.8.4.3.2.l are not in accordance with Standard Review Plan 3.8.4. Compare your load combinations and discuss the effects of your deviations*
Response
0 Refer t8 response to Question 220.29 for comparison of load combinations specified in SRP 3.8.4 and those used in the design of seismic Category I structures, statements of compliance, alternate compliance and remarks on impact of deviation. No FSAR change is required
- 220.36-1 Amendment No. 5, (8/81)
SL2-FSAR Question No
- 220.37 Identify all masonry walls in.your facility which are in proximity to o.r have attachments from safety-related piping or equipment such.that wall failure could affect a safety-related system.
Describe .the systems and equipment, both safety and non-safety-related, associated with these masonry walls. Include in your review, masonry walls that are intended to resist impact or pressurization loads, such as missiles, pipe whip, pipe break, jet impingement, or tornado, and fire or water barr~ers, or shield
. walls. -Equipment to be considered as attachments or in proximity to be walls shall include, but is not limited to pumps, valves, motors, beat exchangers, cable trays, cable/conduit, HVAC ductwork, and electrical cabinets, instrumentation and controls.
Provide a re-evaluation of the desig~ adequacy of the walls, identified above, to determine whether the masonry walls will perform their intended function under all postulated loads and load combinations. Submit a written report upon completion of the re-evaluation program. Th~ report shall .include the foilowing information: 1 - Describe, in detail, the function of* the masonry walls, the configurations of these walls, the type and strengths of the
.material of which they are constructed (mortar, g"l!l&ut,
- *concrete and steel), and the reinforcement details (horizontal steel,. vertical steel, and masonry ties for multiple wythe construction). A wytbe is considered to be (as defined by AC!
Standard 531-1979) "each continuous vertical section of a
- wall, one masonry unit or groute4 space in thickness and 2 in.
minimum in thickness~" 2 - Describe the construction practices employed in the construction of these walls and, in particular, their adequacy in preventing.significant voids or other weaknesses in any mortar, grout, or concrete fill'. 3 - The re-evaluation report should include detailed justification for the criteria used. References to existing codes or test data may be used i f applicable for the plant conditions. The re-evaluation should specifically address the following: 220.37-1 Amendment No. 5, (8/81)
SL2-FSAR a - All postulated loads and load combinations should be evaluated against the corresponding re-evaluation acceptance criteria. The re-evaluation should consider the loads from safety and non-safety-related attachments, differential floor displacement and thermal effects (or detailed justification that these* can be *considered self limiting and cannot induce brittle failures)," and* the effects of ariy potential cracking under dynamic loads. Describe in detail the methods* used'to*account for these factors in the re-evaluation and the adequacy of the ac~eptance*criteria for both in-plane and. out-of-plane loads. b - The mechanism for load transfer. into the masonry.walls and postulated failure modes should be reviewed. For multiple wythe walls in which composite behavior is r*elied upon, describe the methods and acceptance criteria used to assure that these walls, esp~cially with regard to shear and tension transfer at the wythe interfac*es. With regard to local loadings such as piping and equipment support reactions' the acceptance criteria should assure that "the loads are ~dequately transfer~ed into the wall, such that any assumptions regarding the behavior of the walls are appropriate. Include the potential for *tensile stress transfer through bond at the wythe interfaces. Existing test data or conservative assumptions may be used to justify the re-evaluation acceptance criteria if .the* criteria are shown to be conservative and applicable for the actual plant conditio'ns. In the *absence of appropriate acceptance criteria a confirmatory masonry*wall test program is required by the NRC.in order to quantify the safety margins inherent in the re-evaluation criteria. Describe in detail the actions planned and their schedule to justify the re-evaluation criteria. *If a test program is necessary, provi4e your commitment for such a program and a schedule for submittal of a description of the test program and a schedule for completion of the program.* This test.program should address all appropriate loads (seismic, tornado, missile, etc). Submit the results of the te~t program upon its completion.
Response
This question is NRC Bulletin 80-11. A field inspection program and a design re-evaluation program were.instituted and completed on St. Lucie 1 which addressed the reqµirements of IE Bulletin 80-11. A program, simi!ar to St. Lucie 1,. of field inspections and re-evaluation of design adequacy has been implemented on St. 113 Lucie 2 as follows: 220. 37-2 Amendment No. 13, (2/83)
SL2-FSAR
- 1.
2. Perform field surveys of all masonry walls to identify all masonry walls which are in proximity to or have attachments from safety-related piping or equipment such that wall failure could affect a safety-related system. Masonry walls identified by the field inspection program as safety-related will be re-evaluated to demonstrate their capacity to withstand the postulated design loads. A re-evaluation report has been submitted to the NRC (FP&L letter 13 L-82-459). This report included the information as requested by NRC. Justification for the re-evaluation criteria was based on reference to effective codes and established standards of practice related to concrete and masonry design typically used throughout the industry. It has been justified appropriate and a test 113 program will not be necessary. Masonry walls on St. Lucie 2 are reinforced and intended to resist seismic forces *
- 220.37-3 Amendment No. 13, (2/83)
SL2-FSAR Question No. 220. 3 7A Question 220.37 is the SEB Interim Criteria for evaluating masonry walls. Document and explain any deviation to this criteria you used to evaluate the walls.
Response
A review of the SEB Interim Criteria for Safety-Related Masonry Wall Evaluation indicates the following deviations of the proposed St. Lucie 2 re-evaluation criteria. from the SEB criteria: Paragraph 3(c) We propose to use allowable tensile stresses normal to bed joints per ACI 531-79 for unreinforced walls. Although not an operating plant, St. Lucie 2 should not be treated as "new construction' since masonry wall construction is essentially complete and unreinforced walls, not permitted by SEB f'or new construction, have been used extensively. For reinforced walls, there is no permissible stress for tension normal to bed joints. However, the modulus of rupture is utilized in the determination of the cracking moment _in order to calculate the equivalent moment qf inertia for reinforced cracked walls in accordance with the formula I =(Mcz:\ 3 I + (l _ Mer )r e Ma J t Ma er f I where M = er uncrackett moment capacity (cracking moment) =--- Ma = applied moment on the wall It = moment of inertia of the transformed section Icr = moment of inertia of the cracked section fr =modulus of rupture (allowable tensile.stress for mortar) y = distance of tension face from neutral plane Paragraph 3 ( d) We propose to use the followl.ng allowable stress increse _factors for abnonnal, abnormal/severe environmental, extreme environmental and abnormai/extreme environmental conditions which exceed those specified in the SEB criteria: Masonry tension parallel to bed Joint 1.67 Masonry tension normal to bed joint (unreinforced) 1.67 Reinforcement tension or compression 2.25 No FSAR change is required. 220. 37A-l Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
220.-38 (3.8~2) The allowable stresses shown in Table 3.8-7 are not in accordance with the ASME Code for acceptance testing at ambient temperature. Revise the allowable stresses to confonn to the ASME Code allowables for the testing conditions your extent to use. In several of the load cases, it appears you are increasing the AISC allowable stresses and it is the staff's position that increases in the allowable stresses are only allowed for thermal loads and no others. Revise your allowable stresses to conform to this position. Re spouse: The allowable stresses for load case 2 (Acceptance Test at Ambient temperature) in Table 3.8-7 are in accordance with the ASME Code allowables. Refer to ASME Sect III Summer 1972 Addenda NE-6322. For load cases 7, 9 and 11, no increase in AISC allowables was used. No FSAR change is required *
- 220. 38-1 Amendment No. 6, (9/81)
SL2-FSAR
- Section 1.0 Title TWO DIGIT TABLE OF CONTENTS INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
1.2 GENERAL PLANT DESCRIPTION 1.3 COMPARISONS 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS
- 1. 5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.6 MATERIAL INCORPORATED BY REFERENCE
- 1. 7 DRAWINGS 1.8 NRG.REGULATORY GUIDES 1.9 OTHER CONCERNS AND COMMI 'l'MENTS 1.9A TMI RELATED REQUIREMENTS 1.9B INADEQUATE CORE COOLING ITEM II.F.2 OF NUREG-0737 2.0 SITE CHARACTERISTICS 2 .1 GEOGRAPHY AND DEMOGRAPHY 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 2.3 METEOROLOGY 2.4 HYDROLOGY 2.4A EROSION ESTIMATES 2.5 GEOLOGY, SEISMOLOGY AND GEOTECHNICAL ENGINEERING 2.5A BORING LOGS & ~NfA SUNMARIES 2.5B FLORIDA EARTHQUAKE OF OCTOBER 27, 1973 3 .O DESIGN CRITERIA-STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS 3 .1 CONFORMANCE WITH NRC GENERAL DESIGN CRI-:;:"ERIA
- 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTS I Amendment No. 5, (8/81)
SL2-FSAR Section 3.3 Title - TWO DIGIT TABLE OF CONTENTS (Cont'd) WIND AND TORNADO LOADINGS 3.4 WATER LEVEL (FLOOD) DESIGN 3.5 MISSILE PROTECTION 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE RUPTURE OF PIPING 3.6A HIGH ENERGY PIPE RUPTURE ANALYSIS INSIDE CONTAINMENT
- 3. 6B* HIGH ENERGY PIPE RUPTURE ANALYSIS' OUTSIDE CONTAINMENT 3.6C PIPE WHIP RESTRAINTS AND BREAK LOCATIONS 3.6D STRUCTURAL DETAILS OF PIPE WHIP RESTRAINTS 3.6E MAIN STEAM AND FEEDWATER ANALYSIS 3 3.6F MODERATE ENERGY PIPING ANALYSIS 3.7 3.8 3.9 SEISMIC DESIGN DESIGN OF CATEGORY*I STRUCTURES MECHANICAL SYSTEMS AND COMPONENTS 3.9A OPERABILITY CONSIDERATIONS FOR SEISMIC CATEGORY I ACTIVE PlJNPS AND VALVES 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.lOA CRITERIA FOR SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT AND THEIR SUPPORTS 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT 3.llA ENVIRONMENTAL QUALIFICATION FOR MAIN STEAM LINE BREAK 3.llB EQUIPMENT ENVIRONMENTAL QUALIFlCATION PROGRAM 4
3 .llC EQUIPMENT ENVIRONMENTAL QUALIFICATION GUIDEBOOK 4.0 REACTOR 4.1 .
SUMMARY
DESCRIPTION II Amendment No. 5, (8/81)
SL2-FSAR
- Section 4.2 Title TWO DIGIT TABLE OF CONTENTS (Cont'd)
FUEL SYSTEM DESIGN 4.3 NUCLEAR DESIGN 4.4 THERMAL AND HYDRAULIC DESIGN 4.5 REACTOR MATERIALS 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1
SUMMARY
DESCRIPTION 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY (RCPB) 5.2A OVERPRESSURE PROTECTION FOR ST LUCIE UNIT 2 - ~RESSURIZED WATER REACTOR 5.2B ANALYSIS OF NATURAL CIRCULATION COOLDOWN WI'l'llOUT UPPERHEAD VOIDING 5
- 5.2C 5.3 5.4 NATURAL CIRCULATION COOLDOWN REACTOR VESSEL COMPONENT AND SUBSYSTEM DESIGN
- 6. 0 ENGINEERED SAFETY FEATURES 6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.2 EBASCO MODIFICATIONS TO THE CONTEMPT-LT MOD 26 COMPUTER CODE 6.2B WATEMPT-A COMPUTER CODE TO CALCULATE THE SHIELD BUILDING ANNULUS TRANSIENT 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6 .5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.6 INSERVICE INSPECTION OF QUALITY GROUP B AND C CO~WONENTS 7 .O INSTRUMENTATION AND CONTROLS
7.1 INTRODUCTION
III Amendment No. 5, (8/81)
SL2-FSAR Section 7.2 Title TWO DIGIT TABLE OF CONTENTS (Cont'd) REACTOR PROTECTIVE SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHU'.::'DOWN 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION 7.5A SAFETY ASSESSMENT SYSTEM 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 8.0 ELECTRICAL SYSTEMS
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSITE POWER SYSTEM 9.0 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4* AIR.CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEM 9.5 OTHER AUXILIARY STEAMS
- 9. 5.A FIRE HAZARD ANALYSIS
- 10. 0 STEAM AND POWER CONVERSION SYSTEM 10.2 TURBINE GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF THE STEAM AND POWER CONVERSION SYSTEM 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS IV Amendment No. 5, (8/81)
SL2-FSAR
- Section Title
. TWO DIGIT TABLE OF CONTENTS (Cont'd)
I. 11.lA DERIVATION OF RESIDENCE TIMES LIQUID WASTE SYSTEM GASEOUS WASTE SYSTEM 11.4 SOLID WASTE MANAGEMENT .SYSTEM 12.0 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) . RADIATION SOURCES 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3A TM! SHIELDING STUDY
- 12. 4.
12 * .5 13.0 DOSE ASSESSMENT HEAL TH PHYSICS PROGRAM CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE* OF APPLICANT TRAINING 13.3 EMERGENCY PLANNING 13.4 REVIEW AND AUDIT 13.5 PLANT PROCEDURES 13.6 INDUSTRIAL SECURITY 14.0 INITIAL TEST PROGRAM 14.1 SPECIFIC INFORMATION TO BE INCLUDED IN PSAR 14.2 SPECIFIC INFORMATION TO BE INCLUDED IN FSAR 15.0 ORGANIZATION AND ME'P!ODOLOGY
- 15.1 INCREASED HEAT REMOVAL BY THE SECONDARY SYSTEM v Amendment No. 5, (8/81)
SL2-FSAR Section Title TWO DIGIT TABLE OF CONTENTS (Cont'd) 15.2 DECREAS.ED HEAT REMOVAL BY THE SECONDARY SYSTEM 15.3 DECREASE IN REACTOR COOLANT FLOW RATE 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.5 INCREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.7 RADIOACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 15.8 PRIMARY SYSTEM PRESSURE DEVIATION 15.9 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) 15A ANALYSIS OF LARGE STEAM LINE BREAKS 15B CONTAINMENT LEAKAGE AND DOSE CALCULATION MODEL 15C SUPPLEMENTARY INFORMATION 15D CE SEC 16.0 TECHNICAL SPECIFICATIONS 17.0 QUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE VI Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
240.1 (2.4.13) Subsection 1.9.5 of the FSAR states that hydrologic data for Hutchinson Island is being evaluated and that an amendment to Section 2.4 will be filled on or about March 1981. An amendment containing this data has not as yet been received. Please provide the promised information. Additionally, discuss potential effects on plant design basis and/or hydrologic evaluations.
Response
A revised Section 2.4 was included in Amendment 3, dated June 1, 1981. All pertinent hydrologic evaluations have been incorporated in.the new data *
- 240.1-1 Amendment No. 4, (6/81)
SL2-FSAR Question No. 240.2 The Safety Evaluation Report .(Construction Permit Stage) states (3.4) that stop log closures are to be provided to protect the entrances to s~fety related structures up to elevation +22 ft. MLW. Stop 12 log closures on the entrances of the Reactor Auxiliary Building and the Fuel Handling Building are shown in Figures 1.2-13 and 1.2-16 of the FSAR. However, they are not discussed or described in other sections of the FSAR dealing with flood protection. State whether the stop logs will be provided. If not, justify I 12 that they are not needed. Provide analyses of wave runup at all entrances below elevation +22 MLW. If the stop logs are to be installed, provide: 12 a) the bottom elevation of the door openings and the dimensions of the openings secured by the stop logs. 12 b) the time required to place the stop logs into position. c) the criteria to be used to determine when and if the stop logs are to be placed into position. d) a discussion of the relevant technical specifications *
- Response Based upon PMF high water elevation 17.2 ft mean low water (MLW) wave runup elevation 18.0 ft MLW and plant island elevation 18.5 ft HLW, flood protection stop logs at entrances (whose minimum elevation is at least + 19.5 feet) to safety-related buildings are I 12 not deemed necessary. Additional wave runup protection is
- provided to the entrances of the F\iel Handling Building and Reactor Auxiliary Building by the presence of adjacent buildings and structures (see Enlarged Site Plot Plan FSAR Figure 1.2-2). 12 Since no permanent structures are located on the south side of the Reactor Auxiliary Building, additional wave runup protection will be provided by installing stop logs in the entrance on the south 12 wall and the southern most entrance on the east wall. 7 A description of the stop logs found at St. Lucie Unit 2 is 12 contained in revised Subsection 3.4.1. 6 240.2-1 .Amendment No. 12, (8/82)
SL2-FSAR -~* Question No. 240.3 Discuss the hydrologic technical specification explaining (2.4.13) measures to be used to monitor the existing beach dunes and mangrove areas which are relied upon for hurricane protection. Provided the following information: (a) the means by which storm erosion will be measured and the frequency of such surveys. (b) the means by which the mangrove areas will be monitored to identify blighted areas and the frequency of such sur.veys.
Response
Beach dunes and mangroves areas are not relied upon to provide protection from hurricanes. As discussed in Subsection 2.4.S, the dunes are conservatively assumed to be eroded to an elevation of
+4 feet MLW prior to computing the maximum possible quantities of erosion from Highway AlA and the plant island. No credit is taken fo~ the energy dissipated or the time consumed in erosion of the dunes. Since. the "stable base plain" el~vations of +4 feet MLW to the east and south of the plant island and elevation +5 MLW on the north are the natural ground elevations of extensive flat areas,
*their viability is not depen~ent on the existence of the mangroves. In the wave analysis Subsection 2.4.5, no reduction of wave height or energy by the mangroves is assumed.
A visual inspection of the highway and dunes for breaching or erosion will be done periodically and aerial photographs of the beach and the dune areas will be taken once a year if the Unit I Technical 8pecification1 for mangroves are deleted. 7 No FSAR change is required
- 240. 3-1 Amendment .No. 7, (10/81)
SL2-FSAR
- Question No.
240.4 (2.4.13) Justify that the design groundwater level of elevation +16.2 MLW is conservative. Show that the groundwater level resulting from the PMH coincident with intense rainfall on the plant site will not be higher than your design level. Provide details of the analysis including: (a) the antecedent groundwater level assumed to exist before the PMH surge. (b) the contribution to this *level from the surge. (c) the contribution to *this level from wave runup and coincident precipitation.
Response
Plant island groundwater level is specified as follows: (a) The antecedent groundwater level assumed to exist before the PMH surge is Elevation 3.0 feet. This value was assumed based upon surface water levels observed on Hutchinson Island and information obtained from borings taken before start of plant island construction *
- (b) During a PMH design condition, the ocean surge elevation.was calculated to be Elevation 17.2 feet, per FSAR Subsection 2.4.5. FSAR Figure 2.4-12 shows maximum surge water levels during a PMH versus time duration. Civil design criteria assumes that the plant island ground water fluctuates at the same rate and level as results from the ocean surge. This is extremely conservative considering the distances that the water must pass through the plant island backfill material.
Also, the peak surge of Elevation 17.2 feet, only lasts for a duration of approximately one-hal.f hour. (c) During a PMH design condition, the wave runup elevation was calculated to be Elevation 18.0 feet, per FSAR Subsection 2.4.5. Civil Engineering assumes that the impact on plant island ground water level, during a wave runup and coincident precipitation, is minimal *and therefore the already conservative value of Elevation 17.2 feet should not be increased. Ebasco Civil Design criteria for Category I plant island structures specifies the following ground water levels: 240.4-1 Amendment No. 4, (6/81)
SL2-FSAR Normal groundwater table - Elevation + 3.00 feet PMH groundwater table - Elevation+ 17.00 feet for all structures, except Elevation
+ 21.00 feet at the Reactor Building.
No FSAR changes required as a result of above responses. 240.4-2 Amendment No. 4, (6/81)
SL2-FSAR --* Question No. 240.5 (2.4.13) Provide the groundwater levels assumed to exist coincident with other extreme events (natural and man induced) for calculation of combined loads on safety relat~d structures. Justify your selection of these levels considering the causative events and the frequency of occurrence.
Response
Groundwater level during all other design conditions (including tornado and earthquake) is assumed at normal groundwater Elevation
+ 3.00*feet. There is no event which would cause the groundwater elevation to rise .beyond normal conditions.
See .FSAR Subsection 2.4.13.5
- 240.5-1 Amendment No. 4, (6/81)
SL2-FSAR
- 240.6 (2.4.2)
Describe the water level and wave climate at the site as result of the close passage of Hurricane David in' September 1979. Describe any wave erosion resu~ting from the storm and its implication as to existing design basis. Include in your discussion, the source of the data, the accuracy of the data, and other pertinent details.
Response
Hurricane David struck the St Lucie plant site on September 3, 1979. The eye of the storm skirted the site and provided a momentary period of calm. A top wind speed of 85 mph was recorded at one point on the meteorological tower at 190 feet above grade elevation. The prevailing wind direction during the storm's passage was from the Southeast. Rainfall at the site measured about 6.20 inches over a 54-hour period, starting late on September 3 and extending through September 6. No visible evidence of rainfall or flood damage to plant facilities was reported. *The plant's drainage system functioned properly and easily handled this amount of rainfall. Tides were reported to be generally 3 to 5 feet above normal (Ref. 1). Highwater marks from possible flooding were not visible; hence, flooding can be characterized as light. No records of wave heights along the shore or at the plant site were made. Since damage resulting from wave attack was minimal, it can be concluded that the wave climate was light to moderate in height. Observations of this damage included: (a) A one foot drop at the* base of the dunes along the eastern shore due to wave action. No dune breaching occurred. (b) Canal damage was limited to only one location; the elbow region of the intake canal on the inside* fa~e of the southern-most dike. A six foot vertical drop was reported to have o*ccurred, starting at Elevation + 13 feet MLW down to Elevation+ 7 feet MLV. Reference 1 NOAA.- National Weather Service, "PreliminaJ;"y Report Hurricane David, August 25 - September 7, 1979. ii 240.6-1 Amendment No. 4, (6/81)
SL2*FSAR (c) No visible damage to mangroves. No FSAR change required due to above responses. 240.6-2 Amendment No. 4, (6/81)
- .* Question No.
240.7 SL2-FSAR Describe the method of protecting the sheetpile groins and ( 3 .4 ) bulkheads from weakening due to corrosion over the life of the plant. If this protection involves periodic inspections, discuss a technical specification stating the frequency of inspection and criteria for determining the need for corrective maintenance/replacement. *specifically, state whether the sheetpile bulkhead installed for erosion protection at the nose of
'the Discharge Canal is covered with a concrete pile cap. If it is, provide a drawing of the bulkhead as built, showing the elevations and thickness of the pile cap.
Response
The hurricane prot~ction groins and bulkheads are protected from corrosion by impressed current systems. Florida Power & Light Company's Preventive Maintenance Standard No. ES-2.1 requires a monthly check of groups and/or all individual anode circuits. Any anode circuit found nonfunctional is reported and a work order for repair is issued. The sheetpile bulkhead at the nose of the Discharge Canal is covered with a concrete cap from one foot below grade (grade elevation is Elevation 18.0 to Elevation 22.0 feet. The concrete cap is detailed on Figure 240.7~1. No FSAR change required *
- 240.7-1 Amendment No. 4, (6/81)
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AMENDMENT NO. 4 (6/81) FLORI DA POWER & LIGHT COMPANY SECT A-AtFii DET "C't*8l ST. LUCIE PLANT UNIT 2
~'*f*O 'L*l;o CIRCULATING -WATER SYSTEM DISCHARGE CANAL NOSE PROTECTION REF DWG: 8770-G-719 (REV.3) FIGURE 240.7-1
0778W-4 SL2-FSAR
- Question No.
240.4 (2.4.13) Justify that the design groundwater level of elevation +16.2 MLW is conservative. Show that the groundwater level resulting from the PMH coincident with intense rainfall on the plant site will not be 'higher than your design level. Provide details of the analysis including: (a) the antecedent groundwater level assumed to exist before the PMH surge. (b) the contribution to this *level from the surge. ( c) the contribution to this level from wave runup and coincident precipitation.
Response
Plant island groundwater level is specified as follows: (a) The antecedent groundwater level assumed to exist before the PMH surge is Elevation 3.0 feet. This value was assumed based upon surface water levels observed on Hutchinson Island and information obtained from borings taken before start of plant island construction. (b) During .a PMH design condition, the ocean surge elevation was calculated to be Elevation 17.2 feet, per FSAR Subsection 2.4.S. FSAR Figure 2.4-12 shows maximum surge water levels during a. PMH versus time duration. Civil design criteria assumes that the plant island ground water fluctuates at the same rate and level as results from the ocean surge. This is extremely conservative considering the distances that the water must pass through the plant island backfill material. Also, the peak surge of Elevation 17.2 feet, only lasts for a duration of approximately one-half hour. (c) During a PMH design condition, the wave runup elevation was calculated to be Elevation 18.0 feet, per FSAR Subsection 2.4.S. Civil Engineering assumes that the impact on plant island ground water level, during a wave runup and coincident precipitation, is minimal and therefore the already conservative value of Elevation 17.2 feet should not be increased. Ebasco Civil Design criteria for Category I plant island structures specifies the following ground water levels: / 240.4-1 Amendment No. 4, (6/81)
0778W-5 812-FSAR Normal groundwater table - Elevation + 3.00 feet PMH groundwater table - Elevation+ 17.00 feet for all structures, except Elevation*
+ 21.00 feet at the Reactor Building.
The above design criteria were based upon PMH maximum ocean surge level documented witltln the FSAR, up to and including Amendment 2, at elevation 16.3 feet. The revised PMH maximum ocean surge level, incorporated into Amendment 3 of the FSAR, at elevation 17.2 feet and the assumed coincident equal groundwater table is 0.2 feet greater than the design criteria noted above. This small difference has a negligible impact on the design of the noted 11 structures. See FSAR Su~section 3.4.2 240. 4-2 Amendment No. 11, (7 / 8 2)
0778W-9 812-FSAR
- Question No.
240. 7 (3 .4) Describe the method of protecting the sheetpile groins and bulkheads from weakening due to corrosion over the life of the plant. If this protection involves periodic inspections, discuss a technical specification stating the frequency of inspection and criteria for determining the need for corrective maintenance/replacement. Specifically, state whether the sheetpile bulkhead installed for erosion protection at the nose of the Discharge Canal is covered with a concrete pile cap. If it is, provide a drawing of the bulkhead as built, showing the elevations and thickness of the pile cap.
Response
The hurricane protection groins and bulkheads are protected from corrosion by impressed current systems. Florida Power & Light Company's Preventive Maintenance Standard No. ES-2.1 requires a monthly check of groups and/or all individual anode circuits. Any anode circuit found nonfunctional is reported and a work order for repair is issued. The sheetpile bulkhead at the nose of the Discharge Canal is covered with a concrete cap from one foot below grade (grade elevation is Elevation 18.0) to Elevation 22.0 feet. The concrete 111 cap is detailed on Figure 240.7-1
- No FSAR change required *
- 240.7-1 Amendment No. 11, (7/82)
SL2-FSAR
- Question No.
241.1 (2.S.4.2.6) a) Clarify your statement, From this parameteric study, the numerically largest elastic modulus which was associated with the maximum acceleration was conservatively selected to determine the spring constants for the structural dynamic I analysis." b) What does the above mentioned maximum acceleration pertain to? c) Does the above p*rocedure give conservative results under all conditions, and for all affected plant elements? d) Compare the "largest elastic modulus" determined by the above procedure to the realistic soil modulus at the site that was determined by tests, and comment on the degree of conservatism achieved with respect to plant response parameters.
Response
a) Throughout the design and construction phases of the St. Lucie Plants, advances were made in the state-of-the-art for methods to determine elastic properties of soils. Originally elastic modulus values were calculated from compression wave
- velocities obtained from field refraction surveys using the following relationship:
E = P Vc2 (1 + u ) (1 - 2u )/(1 - u ) in which, E = elastic modulus p = mass density u =Poisson's ratio The value of elastic modulus calculated from the above formula varied from 10,000 psi to 250,000 psi. The highest elastic modulus value occurred at very low strains (io-6 in/in) and for deeper soils. The lower values represented large strains (10-l in/in) and soils closer to the surface. *Values of shear modulus were calculated for .these strain ranges using the. following equation: G = E 2 ( 1 + u) in which G = shear modulus E = elastic modulus u =Poisson's ratio
- 241.1-1 Amendment No. 6*, (9/81)
SL2-FSAR Figure 2.5-55 in the FSAR presents the shape of the shear modulus versus shear strain used in the original analyses. These values were later confirmed by field and laboratory tests. The data shown on Figure 2.5-55 is representative of the soils near the foundation level * . This data was used in a one-dimensional discrete mass analytical technique to analyze the response of the horizontal soil profile to earthquake motion (as discussed in Soil Behavior Under Earthquake Loading Conditions:
"State-of-the-Art Evaluation of Soil Characteristics for Seismic Response Analyses" by. Shannon & Wilson, Inc. and Agbabian-Jacobsen Associates). In this model, the soil profile is idealized by a series of discrete masses and springs with linear viscous dampers. The nonlinear and hysteritic stress-strain characteristics of the soil are introduced by using an equivalent snear modulus and an equivalent viscous damping factor. An iterative procedure is used with this model to introduce the proper values for the equivalent shear modulus and damping factor. Thus, the strain distribution in the soil profile is first estimated and the moduli and damping factors selected based upon the assumed strain distribution. The response of the soil profile (discrete masses) to the base motion with the assumed characteristic is then computed and the average strains developed within the soil profile are determined. These values are then compared with the initial assumptions and new values of moduli and damping selected for a second solution.
This iterative procedure is repeated until strains are computed which agree with those assumed for the model. The soil response studies for the higher strain level commensurate with the SSE of O.lg yielded a strain of 1.5 x 10-4 in/in, which corresponds to a shear modulus of 14SOOO psi. The lower strain level on the order of 5.0 x 10-in/in was obtained from soil response studies commensurate with the OBE of 0.05g using a higher shear modulus of 16,700 psi. Therefore, the parametric study refers to the response analyses and the numerically largest shear modulus for the SSE acceleration of O.lg was 14,000 psi and the largest shear modulus for the OBE acceleration of 0.05g was 16,700 psi. The shear moduli were used to calculate the spring constants for the structural dynamic analysis. The equation used to calculate the spring constants and details of the structural dynamic analyses is presented in FSAR Subsections 3.7.l and 3.1.2. 241.1-2 Amendment No. 6, (9/81)
SL2-FSAR
- b) c)
The maximum acceleration for the SSE is O.lg and the maximum acceleration for the OBE is 0.05g. The procedure yielded conservative results under all conditions and for all affected plant elements. In fact, additional conservative was allowed for by varying the maximum shear modulus + 20 percent. In general a higher shear modulus for the soil p;ofile and building model yielded higher responses. These higher responses were used in the design of the buildings and equipment. Subsection 2.5.4.11 of the FSAR describes a generic study which was performed which compared an available finite element analysis for a nuclear plant (Site X) to an available lumped mass cantilever model for Site X. This study, comparing the foundation level spectra for the two methods indicates that the lumped mass cantilever model results generally eveloped the spectra based on the finite element models. This means that additional convervatism was used in the dynamic analysis utilized for the St. Lucie Plant.
- d) Figure 2.5-55 in the FSAR shows the shear modulus vs shear strain curve developed from field and laboratory tests. The "largest elastic modulus" determined from the above procedures (converted to shear modulus) are:
- 1) 2)
G G
= 14,000
= 16,700 psi, psi,
=
=
1.5 x lo-4 in/in, a
. 1*in, a 5
- 0 x 10 -5 in The largest shear modulus value at 5 x 10-6 in/in strain is
= O.lg
= 0.5g 16,000 psi to 30,000 psi as determined from field crosshole velocity measurements. At the depth of the main plant structures 2
125 (900) Gmax= = 32.2 x 144
= 21,800 psi for shear strain = 5 x 10-6 in/in. The values obtained from the above procedure plots on the curve for their respective shear strains. The values obtained were.varied+ 20 percent to allow a degree of additional conservatism with respect to plant response parameters.
No FS1\R change is required *
- 241.1-3 Amendment No. 6, (9/81)
SL2-FSAR r*
- Question No.
241.2 (2~5.4.2.6) Published literature shows an extreme range of shear strains in soils subjected to most earthquakes between approximately 9 x 10-4 and 4 x lo-1 percent. Your calculations are reported to show a range of strains from 5 x lo-3 to 1.5 x lo-2 percent. a) Please explain why even the larger calculated value is less than the lower limit (approximately 2.5 x lo-2) of the published data for strong motion earthquakes. b) How do the shear moduli corresponding to these strain ranges compare with those that can be derived from the numerically largest elastic modulers referred' to in 241.1?
Response
a) An analysis of the elastic response of the site soils to a strong motion earthquake is presented in Subsection 2.5.2.5.2 This analysis was based on the work of Seed and Idriss. It is important to note that no strong motion earthquates have ever been instrumentally recorded in the southeastern United States. The attached Figure (241.2~1) shows the earthquake shear strain range for the site plotted with the ranges shown in the published literature. The limiting values for the published data were interpolated logrithmically. Aithough the site range upper limit is les*s than the strong motion range lower limit, the site data is in the *center of the extreme range for most earthquakes. Since the plant is designed for a maximum ground surface acceleration of O.lOg in order to comply with the minimum accepted acceleration as stipulated by lOCRFlOO, Appendix A, it can be expected that the St. Lucie range will fall below the maximum values of the extreme range. b) The answer to this question is contained in the response to 241.1. No FSAR change is required *
- 241.2.-1 Amendment No. 6, (9/81)
, .. STRONG MOTION(1)
EARTHQUAKE 1.9x10-2 % 2.3x10-1%
, EXTREME RANGE FOR MOST EARTHQUAKES( 1)
*I s.sx10-43 2.3x10- 1%
ST. LUCIE RANGE(2-) 1-- 5.0x10-3%
* *I 1.5x10-2%
(OBE) (SSE) I I I I I
)>
- S:*
m 10-4 10-3 10-2 10,...1 1 m ""Tl r z
)>
0 c
;;o
-i ;:o :s:m SHEAR STRAIN(%).
I (II -
-t 0 z D * )> -I c z (1) "SOIL BEHAVIOR UNDER EARTHQUAKE LOADING CONDITIONS" PREPARED FOR U.S. AEC BY A JOINT VENTURE "11 )> r- ""CJ p OF SHANNON AND WILSON, INC. (SEATTLE) AND AGBABIAN-JACOBSEN ASSOCIATES (LOS ANGELES), 1972
- A c: 0 Cll C'>m 0 ::e c: (/l mm iii (2) FLORIDA POWER AND LIGHT COMPANY, FINAL SAFETY ANALYSIS REPORT, ST. LUCIE PLANT UNIT NO. 2, VOL. 3
- 0 I ;:o ca mm N )>
"'D
,.!?'>
r-
=
,I:>.. ;:o )>
- """ (/l
- z -
N -t -t G'> I ;:o C: I
-~
z --tz -t0 ()
;:o
)>
""CJ z )> .
G'> m z
SL2-FSAR r* Question No
- 241.3 What were the "residual confining pressures" developed during (2.5.4.4) compaction of fill (referred in paragraph 2.5~4.4) and how were they measured.
Response
This statement refers to Figure 2.5-78 in the FSAR. It is meant to refer to the fact that plant grade was raised from approximately Elevation +5.0 to Elevation +18.0 by excavation and backfi~l. The confining pressure at the base of the Reactor Building is* calcu1ated as follows: Confining pressure EL + 18 to EL + 2 = 16ft x 115 lb/ft3 = 1840 psf Confining pressure EL + 2 to EL - 25 = 27ft x 60 lb/ft 3 = 1620 psf Total = 3460 psf The confining pressure of Elevation - 25 is 3.46 KSF which was approximated by performing tests at a confining pressure of 3.5 KSF. This statement was not meant to refer to "residual confining pressures" developed when compaction is performed adjacent to building walls. The effect of compaction is included in the at-rest earth pressure coefficient, Ko, as described in the response to
- Question 241.9. Since the dynamic earth pressures controlled structural design, the at-rest earth pressures were not measured.
See revised FSAR Subsection 2.5.4.4 *
- . 241.3-1 Amendment No. 6, (9/81)
SL2-FSAR Question No. 241.5 a) Explain how 98% modified Proctor density was selected to be (2.5.4.5.3) equivalent to a relative density of 85%. b) It is stated in Subsection 2.5.4.5.3 that any material which failed to meet the specified minimum density was recompacted and retested or removed. However, Figures* 2.5-61 and 2.5-62 show an actual sample failure rate of 7.6%. Is there a discrepancy between the information given in Subsection 2.5.4.5.3 and that in the above figures? c) Justify the selection of a sample of 277 tests instead of using the total tests of 4700 to arrive at the actual failure rate of 7.6 percent. (What was the basis of selecting the sample of 277 tests out of the total tests of 4700)?
Response
a) Modified Proctor compaction tests (ASTM Dl557) and relative density tests (ASTM D2049) were performed on representative samples of backfill material at the inception of Unit 2 backfill operations. Based on these test results and the results of Unit l's density correlation studies, it was determined that a field density equal to 98 % of the maximum Mod-ified Proctor density would yield an 85 % relative density. This correlation was periodically re-evaluated as backfilling progressed, using actual backfill test data. Figures 2.5-61 and 2.5-62 present the .results of the latest evaluation completed at the time of the FSAR sumittal. A subsequent evaluation was completed in August 1979. A total of twenty~three ASTM D2049 tests were used to establish in the relative density - Proctor density correlation. These twenty-three tests included the original seven tests used in the evaluation appearing in the FSAR, plus subsequent tests that had been performed. A sample of 349 field.density tests from a population of approximately 5400 tests was statistically an.alyzed. The results indicated a sample mean relative density of 97.63%, with 6.03% failures (See Figure 241.5-1.) Another evaluation is currently in progress, utilizing a total of 45 ASTM D2049 tests, including the previous 23 tests. This evaluation includes the results of backfill testing through March 1981. The correlation portion of the study has been completed. Consistent with the previous two studies, the results indicate that 98% compaction yields a 90% relative density. To achieve 85% relative density, only 97% compaction is required. However, our field
- 241.5-1 Amendment No. 6, (9/81)
- 812-FSAR control has not been relaxed: 98% compaction is still required. This is why the mean relative density for the site is much higher than the required minimum of 85% (See Figure 241.5-2).
b) -In the field, any material that failed to meet the specified minmum density was recompacted and retested. Therefore, ideally, no failures occurred. However, in the process of recording test data for the statistical study, retest density values were initially excluded, and the failing density values were used. Thus, the 7.6% failure rate shown on Figures 2.5-61 and 2.5-62 does not reflect the true failure rate. It is appropriate to call the 7.6% a gross failure rate; the net failure rate, incorporating retest data, would be much less, although it would not reach zero. This is because in certain instances the site Soils Engineer accepted marginal failures (for example, a relative density of 84.5%) in areas where such acceptance did not appreciably affect' the overall integrity and intent of the backfill. It should be noted that the acceptance of failing tests was a rare occurrence; the backfill material specified was well-suited for high degrees of compaction. Based on our experience at the site, we
. estimate that less than five precent actual failures occurred. This is substantiated somewhat by the backfill evaluation completed in August 1979, where the sample failure rate dropped to 6.03%. Included in this study were the retest values for failing tests rather than the initial values, for all testing performed since the previous study. We did not upgrade the failing tests used in the previous study to the retest values, as this would have involved extensive effort in
.searching through the Quality Control Documentation. Since the sample failure rate was well below the allowable failure rate of 16%, and since upgrading the results of the. previous study would have reduced the sample failure rate even further below the allowable, it did-not appear warranted to expend this effort. .
Note: The NRC, upon reading this answer (241.Sb) from FP&L, requested a more detailed discussion of the response, below is that material. The frequency distribution attached to the original response showed that 6.0 percent of the 349 tests fell below 85 percent relative density. Table 241.5-1 contains a listing of these 21 tests, along with their associated elevations, field densities, control and confirmatory proctor densities, percent relative densities, assigned groups for the 8/79 study, and, 241.5-2 Amendment No. 6, (9/81)
- SL2-FSA,R r
- where available as of this writing, the sample locations. The site Quality Control organization is presently determining the exact location of each of these tests. Th~ percent relative density was determined from the correlation curves developed for the 8/79 study, a copy of which is in the mail. Included in the 8/79 study were all test results used in the previous study {described in the FSAR), and the results of all testing done since the previous study. If a failing test was retested, the retest value was used rather than the initial, failing value, for all testing done since the previous study.
Thus, the study is cummulative, incorporating the results of all testing done at the site. An examination of the attached table shows that field density tests 770 and 3060 are at 85 percent relative density. Furthermore, field density test 6490 was retested (test 6993) and is* at 92 percent relative d,ensity. So only 18 of the 349 tests, or 5.2 percent show relative densities less than 85 percent. Only 2.6 percent are below 80 percent relative density, and only 0.3 percent 'are below 72 percent reiative density. No tests fall below 67 percent relative density. It should be noted that soils with relative densities greater than about 55 percent to 60 percent are sufficiently dense to preclude liquefaction. The site Quality Control organization is currently in the process of examining the failing tests to determine the present status of the soils in those locations. For example, the soil at the location and elevation of field density test 6510 has been excavated for a manhole. They are also inspecting for retest values of failing tests from the previous study. The results of this investigation will be used to upgrade the results of the backfill study. As stated in the original response, we expect that less than 5% actual failures occurred. As des.cribed in the response to Question 241.5 (c), the backfill was compacted to 98 percent of the control proctor density. Recompaction and retesting were done on tests falling below 98 percent of the control proctor density. Some failing tests occurred when the confirmatory proctor density was more than several pounds per cubic foot larger than the control proctor density. Since field compaction was based on the control proctor density, .these failing tests were not immediately evident. When the results of the confirmatory proctor tests were available, it was usually after several additional backfill lifts or concrete for ductbacks, etc. had
- 241.5-3 Amendment No. 6, (9/81)
SL2-FSAR been placed over the test locations which subsequently yielded failures. Since it was not possible to immediately identify all failures, and because of the magnitude of the backfill operations a certain amount of failures were unavoidable, a limit was placed on the percent of failing tests that would be acceptable. This limit was set*at 16 percent allowable, and therefore assumed that the mean relative density would be equal to 85 percent relative density plus one standard deviation. This maximum allowable of 16 percent below 85 percent.relative density also assumed that the frequency distribution of the density data fit a normal distribution curve exactly, and implied that failing tests would not be corrected. However, since a certain number of failing tests were to be corrected, the actual frequency distribution of the density data was slightly peaked and mode~ately skewed to the right than th~ normal curve. For the normal curve, the value of kurtosis, or degree of peakedness, is equal to three, and the value for skewness is equal to zero. The distribution of the 8/79 study data has a kurtosis equal of 3.76, which indicates a more peaked distribution, and has a skewness equal to 0.22. The positive value of skewness indicates that the sample data has a longer tail to the*right of the mean. For this type of slightly peaked and moderately skewed distribution, the area under the curve below one standard deviation less than the mean is slightly les.s than 16% (for the normal curve); liowever, 16% is a close approximation. c) Field control of backfill placement generally involved in the performance of one Modified Proctor Compaction Test (ASTM Dl557) for every ten in-place density tests. A control Proctor density was established as representative of the material in any particular lift, and was used to* determine the percent compaction of that lift until, at the location of the tenth in-place density test, a confirmatory Proctor density was determined. This confirmatory Proctor density then became the control Proctor. density for the next ten in-place density tests, and so on. Thus, only every tenth in-place density was associated with an actual Proctor density (the confirmatory Proctor), for which an actual percent compaction could be calculated. The intervening nine in-place densities were based on the control Proctor density and, since specification compliance was required, were all compacted to at least 98% of the control Proctor density. Inclusion of these intervening in-place density tests in the backfill study would have resulted in a 241.5-4 Amendment No. 6, (9/81)
SL2-FSAR decreased rat~ of failure, since they were all passing tests
- Therefore, it was conservative to use only in-place density tests for which an actual Proctor density was established in order to calculate the percent failures. Following this system, 4700 in-place fie~d density tests should have resulted in 470 Proctor density tests, rather than the 277 tests used in the study. However, many of the initial Proctor density tests were rejected from inclusion in the backfill study
- because the Proctor tests were not performed on material taken from the same exact location as the tenth in-place test. The Proctor test material was taken from random areas of the just-placed lift and, as such, could not be associated with a single in-place test for; the purpose of the study. When the correct procedure was established for taking Proctor test samples, the subsequent test data was incorporated into the backfill study. The study shows that the sample size of 277 tests is satisfactory to permit statistical inferences to be made on the population of 4700 tests.
No FSAR change is required *
- '241.5-5 Amendment No. 6, (9/81)
SL2-FSAR TABLE 241.5-1
~.AILING DENSITY TEST RESULTS Test Elev. ip (PCF) ~p (PCF) ¥p (PCF) Dr Assigned Group Sample No. (Ft.) Field Control Confirmatory % for 8/79 Studl Location Remarks 730 -0.5 116.4 118.7 120.3 84 80-85 RCB RCB MAT EL-15.42 to EL-25.92 770 -4.0 111.1 116.5 114.8 85 80-85 TGB 920 -9.4 112.4 114.4 116.-8 . 83 80-85 RAB RAB MAT EL-9.50 to EL-19.00 1910 -7.0 114.7 116.5 121.4 73 70-75 RAB 1960 +10.5 112.9 115.1 118.0 79 70-75 FHB FHB MAT EL-9.00 1970 +11.5 111.5 113.9 119.5 67 65-70 FHB 3030 +6.5 113.8 116.1 118.5 81 80-85 3060 +6.5 112.5 116.4 116.3 85 80-85 3420 +3.5 110.8 112.9 116.9 75 75-80 3430 5.0 110.8 112.9 117.4 73 . 70-75 N
.3750 +1.4 110.4 112.6 114.5 83 80-85 TGB
..... 4250
~ +8.0 107.5 111.4 112.6 80 75-80
.I.JI 4270 +9.5 108.1 114.0 114.7 73 70-75 0\ I 4330 +14.9 110.6 °112.3 114.6 84 80-85 4930 +7.6 110.8 112.2 115.9 80 75-80 TGB 5134 +i2*.4 109.8 112.1 116.4 73 . 70-75 TGB 5750 +17.0 106.3 110.3 110.3 84 80.85 O.F. O.F. = Outlying Facilities 5980 +8.7 113.5 112.1 117.6 84 80.85 6160 +10.6 110.4 112.1 117.1 72 70.75 6510 +13.7 118.1 120.2 123. 7 77 75.80 O.F.
*6490 +6.5 113.3* 120.2 120.2 71* 70.75 OGB/CCB *Retest 6993: 0 (Field) r = 117.9 PCF; therefore, Dr = 92%
t
- sr1'
.!2l 0
..0\
\0
- 00
SL2-FSAR
- . GROUP TABLE 241.5-1 (Cont'd)
NUMBER OF TESTS PRECENT OF TOTAL (349) TESTS
<85% 18 5.2 <84% 14 4.0* <83% 12 3.4 <82% 12 3.4 <81% 11 3.2 <80% 9 2.6 <79% 8 2.3 <78% .8 2.3 <77% !7 2.0 <76% 7 2.0 <75% 6 1.7 <74% 6 1.7 <73% 2 o.6 <72% 1 0.3 <71% 1 0.3 <70% 1 0.3 <69% 1 o.3 <68% 1 0.3 <67% 0 0
- 241.5-7 Amendment No. 6, (9/81)
- .28-
.2521
~
NOTES: NUMBER OF TESTS IN SAMPLE= 349 SAMPLE MEAN RELATIVE DENSITY= 97.63% SAMPLE STANDARD DEVIATION= 9.92 SAMPLE COEFFICIENT OF VARIATION= 10.2%
.24 - PERCENT <85% DR = (.0029 + .0201 + .0115 + .0258)
= 0.0603
= 6.0%
ALLOWABLE FAILURES"" 16% FR(X) =SAMPLE RELATIVE FREQUENCY SAMPLE POPULATION~ 5400
.20 -
.1777
.*1691 (X) .16 ,__
.1490
.12 -
.0888
.08 -
.0487
.04 - .0344
.0258
.0201 0 -- .0029 I r.om I .0086 I
.0115 I
JJ 65 70 75 80 85 90 95 100 105 110 115 120 125 130 (DR(%) AMENDMENT NO. 6 (9/81) FLORIDA POWER & LIGHT COMPANY . ST. LUCIE PLANT UNIT 2 RESULTS OF AUGUST 1979 BACKFILL STUDY FIGURE 241.5-1
140 DR= 130%
- 135 DR= 125%
DR= 120% DR= 115% DR= 110% DR= 105% 130 DR= 100% DR=95% 125 DR=75% DR=70%
- u. DR=65%
<:S I
120
~
(.) e:. (ii z 115 w c a: c c..J w 110 u:: 105 100 95 100 105 110 115 120 125 130 135 MAX. PROCTOR DRY DENSITY (PCF) - exp M'IENDMENT NO. 6 (9/81) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 t 1 RESULTS OF APRIL 1981 CORRELATION STUDY FIGURE 241.5-2
SL2-FSAR
- Question No.
241.6 a) In Subsection 2.5.4.7, reference is made to Section 3.7 for the subsurface soil response to dynamic loading. In Subsection 3.7.2.4 the values of the Young's modulus, E, used for the soil-structure interaction analysis of different siesmic Category I structures are shown to range from 17,400 psi to 40,000 psi. Explain the basis for this large variation in moduli values. b) Please indicate the actual values of Poisson's ratio used in the analysis of the different seismic Category I structures and explain the apparent discrepancy between the values used in Subsections 2.5.4.4 and 3.7.2.4. In Subsection 2.5.4.4 (b) the Posson's ratio was assumed to be 0.45 for saturated sand for calculation the dynamic modulus of elasticity, E. The Poisson's ratio for soil is assumed as 0.25 in Subsection 3.7.2.4 and it has been varied by+ 20 percent to account for any uncertainties in the soil properties. Therefore, the upper limit of the Poisson's ratio in the parametric studies is only 0.30 which appears low for the type of sandy soil that supports these structures.
Response
- a) Young's modulus, E, varies with the square root of the confining pressure. The heavier loaded structures have bearing pressures of 10 KSF and use a modulus of 40,000 psi.
The intake structure is a lightly loaded structure with a bearing value of approximately 1.9 KSF. Therefore proportioning and using the square root of confining pressure
<i.9)1/2 40,000 x = 17.400 (10)1/2 The values shown in FSAR Subsection 3.7.2.4 are proportionate to their confining pressure.
b) Poisson's ratio used in the analysis of the different seismic Category I structures was 0.25. There is no discrepancy between the different values used for Poisson's ratio since it varies with strain. For low ranges of strains on the order of 10-6 in/in Poisson's ratio is approximately 0.5. *For the data obtained from cross hole velocity measurements we calculated a Poisson's ratio of 0.45 which is consistent with this low strain measurement. Poisson's ratio decreases with
- 241.6-1 Amendment No. 6, (9/81)
SL2-FSAR increasing strain. It is difficult to make an exact evaluation of the value of Poisson's ratio has a small effect upon engineering calculations and evaluations. The value of 0.25 was originally selected for use in analyses consistent with typical values used at that time. The variation of + 20% allows for some uncertainties. No FSAR changes is required. 241.6-2 Amendment No. 6, (9/Sl)
SL2-FSAR
- Question No.
241. 7 (2.5.4.8.1) While discussing the simplified procedure for evaluating soil liquefaction potential, you have implied that, even under simultaneous occurance of hurricane or flood with the SSE, the water level in the soil beneath the plant will not use noticeably because of the soil permeability. Please substantiate the above contention by suitable calc~lation for
*I (a) the assumed simultaneous occurrence of SSE and 25 yr precipitation, and (b) OBE and standard project storm and (c)
OBE and standard project hurricane.
Response
In accordance with standard design requirements we have assumed that the water level at the site for liquefaction calculations is El. +2, which is tµe mean high water level. This level is a conservative water level to use in the liquefaction analysis. Any water associated with flooding or rainfall from storms or hurrianes, could not noticeably raise the ground water level in the soil beneath the plant because of the permeability of the soil (i.e. 5 x 10-3 cm/sec). For the purpose of indicating extreme conservatism the lowest
- safety factors shown on Figure's 2.5-85a and 2.5-85b of the FSAR have been recalculated to show the minimum safety factors assuming ground water levels equal to plant grade of El. +18.
These* c*alculations are much more conservative than assuming any of the conditio~s requested in (a), (b) and (c) above. The minimum safety factor against liquefaction thus calculated assuming loose soil conditions and a ground water level of Elev. +18 for amax = O.lg and 10 cycles of strong motion is 1.78. Also assuming loose soil conditions and a ground water level of Elev. +18 for amax = 0.05g and 30 cycles of strong motion, :the minimum safety factor against liquefaction was calculated to be 3.06. Comparing the data for average soil conditions, which is more realistic, the minimum safety factors are 2.70 and 4.65. These high safety factors again demonstrate that liquefaction would not occur at the site under the postulated condition even for a ground water level equal to plant grade. No FSAR change is required *
- 241. 7-1 Amendment No. 6, (9/81)
SL2-FSAR.
- Question No.
241.8 (2.S.4.10.2) Why were no *post-construction settlement readings recorded? Furnish settlement data to compare measured settlement with estimated settlement of all Category I structures. What is the basis for the statement that differential settlements are expected to be within the design limits? What -are the assumed design limits? What are the measured differential settlements?
Response
Settlement readings for the following structures were begun on the indicated dates: Reactor Containment Building - 3/18/78;* Reactor Auxiliary Building - 6/24/78; Turbine Building 12/4/77; Main Steam Trestle - 4/4/79; Circulating Water System Intake Structure - 12/16/78. Reading were begun immediately after mat placement, and the settlement markers were raised as backfill progressed. The measured settlements, although larger than the estimated settlements, are neither unreasonable nor intolerable. The largest average settlement (the Reactor Containment Building)
- is less than 1.5 inches. The trend of the settlement data indicates that settlement has essentially ceased. This is consistent with the FSAR prediction that all settlement would
.occur during the construction period. (See Table 241.8-1) b) . The design limit for differential settlement is one-half inch between structures that contain mechanical and electrical interconnections. Actual differential settlements are expected to be within the design limit because all settlement is expected to occur during the construction phase of the plant. No additional settlement is expected during the plant's operating phase, and therefore differential settlement will not occur. This is substantiated by the settlement data. Most of the settlement for each structure occurred prior to late *1979; the trend since then indicates that settlement has essentially. ceased. Measured differential settlements are not entirely revealing, since, as mentioned above, most of the settlements occurred early in the construction period, before any mechanical or electrical interconnections were made *
- 241. 8-1 Amendment No. 6, (9/81)
SL2-FSAR The April 26, 1980 settlement data for the Reactor Containment Building, the Reactor Auxiliary Building and the Main ~team Trestle is apparently in error, and is probably the result of a survey bust or equipment malfunction. The trend of the data does not lend validity to these values. Additionally; an inspection of the dta reveals an upswing in the values, with the structures reaching maximum settlement between late June and mid-July, 1980. This upswing is the result of elastic rebound caused by the removal from service of a perimeter and several local deep watering systems during mid-1980, thereby decreasing the effective stresses at the foundation levels. Post January, 1980 settlement data is currently being prepared for presentation, and the trend of the data is e~pected to show the cessation of all foundation movement. There is no dewatering system in operation on site at present, and none is planned as all subaqueous excavations have been completed. No FSAR change is required. 241.8-2 Amendment No. 6, (9/81)
- SL2-FSAR TABLE 241.8-1 ST. LUCIE NO. 2 SETTLEMENT DATA Value shown are cummulative total settlements in inches for the indicated dates and structures.
All values are negative unless noted. 1977 1978 1979 1980 1981 12-4 3-18 6-24 9-30 12-16 3-3 4-4 4-24 6-9 7-14 9-8 10-20 12-11 1-29 3-17 4-26 5-31 6-28 7-19 8-16 12-17 1-30 TURBINE GENERATOR BUILDING 0 0.44 0.11 0.49 0.72 0.96 1.12 0.83 1.24 1.17 0.98 1.17 1.24 1.27 1.16 1.22 1.47 1.46 1.69 1.22 1.16 1.10 REACTOR CONTAINMENT BUILDING- 0 +0.20 0.18 0.32 0.60 0.12 o.33 0.87 0.94 0.91 0.93 1.30 1.17 1.27 2.17 1. 36 1.68 1.51 1. 54 1.39 1.42 REACTOR AUXILIARY BUILDING - 0 0.20 0.20 0.47 o.ss 0.20 0.70 0.63 0.57 o.so 0.87 o. 75 o.76 1.65 o. 71 0.95 0.87 0.69 0.35 0.35 CWS INTAKE STRUCTURE 0 o.s2 o.68 0.28 o.ss 0.31 0.20 0.06 0.22 a.so 0.27 0.48 0.46 o.48 0.56 0.23 0.07 +o.1s MAIN STEAM TRESTLE 0 +o.25 0.28 0.33 0.24 0.21 0.59 a.so 0.47. 1.08 .. o. 74 0.83 0.61 0.65 0.39 0.31
SL2-FSAR
- Question No.
241.9 Please justify the use of 0.5 as the value for the at rest (2.5.4.10.3)earth pressure coefficient, K9 , for coml[lected granular backfill.
Response
There are several ways to compute the at-rest earth pressure coefficient~ K0
- Since the K0 state is a condition of no lateral strain, it can be related to Poisson's ratio, u, by:
K0 = u/(1-u) For the plant backfill, Poisson's ratio was taken to be 0.25. This yields a K0 equal to 0.33. This value was increased by fifty percent to 0.5 to account for the effects of compaction~ Another approach is to use the soil friction angle, ~' which is equal to forty degrees for_ the plant backfill. In this case, using K0 = 1 - sin ~ the at-rest pressure coefficient is* equal to 0.36. It is apparent that the K0 values agree reasonably well (0.33 and 0.36) when computed by two unrelated methods. Although the value of Ka = 0.5 may appear somewhat
- low, it should be noted that the, standard site procedure for compacting around a structure was for compaction of each lift to proceed in a direction away from the structure, thus reducing the effects of locked-in stresses against the structure. In any event, the static earth pressure case did not control structural design; the dynamic earth pressure was the governing case by a wide margin. One final note: Figure 2.5-87 contains an error.
The earth pressure value shown is for the active case, whereas it should be for the at-rest case. This figure has been revised to reflect the correct earth pressure coefficient. See revised FSAR Figure 2.5-87 *
- 241.9-1 Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
241.10 Explain, in detail, the procedures/specifications adopted to verify that the compaction piles installed along the slopes were effective in producing the desired densification of the soils. Were the soil densities determined before and after the installation of the compaction piles. b) Please furnish the compaction pile driving data and heave plate measurements. c) Describe the provisions made for soil drainage during driving of the compaction piles and the performance of these drai"nage features.
Response
a) The proc~dures/specifications adopted to verify that the compaction piles were effective are contained in:
- 1) Ebasco Specification - Ultimate Heat Sink Compaction Piles
.No. FLO 9560.069.
- 2) Ebasco Report-Compaction Pile Soil Stabilization Program-dated January 1976.
- 3) Ebasco Report-Soils Foundation of the Emergency .Barrier Wall-dated May 1976.
The soil densities were determined by in-place density tests reported in the Ebasco report dated May 1976. The results indicated the relative densities greater than 85% were obtained as a result of driving the compaction pile~. b) This data is conta.ined in the Ebasco reports refellenced in the response to (a) above. The heave plate readings were generally negative indicating that the insitu material was densified. c) Prior to driving compaction piles at least five pile locations were preaugered to ~llow disappation of pore pressures during driving. Visual observations made at the time of pile driving indicated little or no movement of ground water due to pile driving. No FSAR change is required *
- 241.10-1 Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
251.2 Provide da~a on the quaiifications of the personnel performing the
.fracture toughness tests that indicates the personnel wer~
qualified by training and experience and had demonstrated competency to perform the tests in accord with written procedures of the component manufacturer. If the above.information cannot be provided or i f the information provided does not comply with.the requirements for training of personnel, state why the information cannot be provided and identify why the methods used for training are equivalent to those required in paragraph III.B.4 of Appendix G.
Response
The personnel performing impact testing were qualified in accordance with the ASME Boiler and Pressure Vessel Code. Compliance with the ASME B&PV Code is verified by Combustion Engineering quality assurance procedures, a~d reviewed by ASME and NRC audits. As required by the Code, the personnel performing impact t:esting are certified by qualified supervisory personnel. Records of this certification are maintained in accordance with NA-4900, "Records and Data Reports" and are available for review at Combustion Engineering's Chattanooga facility *. Combustion Engineering training methods comply with the revision of 10CFR50 Appendix G published November 14; 1980 "for comment". No FSAR change ls required *
- 251.2-1 Amendment Nci. 6, (9/81)
812-FSAR
- 1ueston No.
251.1 Identtfy the heat treatment given the test specimens for the welds in the corebelt line region. Indicate whether this heat treatment gives equivalent metallurgical effects as the actual core ~elt line welds. Resp0nse The test specimens for core. beltline welds are made with the same pre, interpass, and post-weld temperature requirements as the reactor 11essel 111elds. The specimens receive a ~O hour, ll.50 + 25 F stress relief treatment equivalent to the stress relief given to the vessel. The test specimens are in the same metallurgical condition as are the welds in the core beltline region. No FSAR change is required *
- 251. 3-1 Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
251.4 For Welds Inside the RV Beltline Region
- 1. Provide CVN impact curves and the NDT temperature for each beltline weld.
- 2. If sample material cannot be provided to the requirements of paragraph III.C.2 of Appendix G, provide:
- a. CVN impact test curves f.rom samples which were produced using the same process, post weld heat treatment, electrode type, flux type and weld manufacturer as the SL-2 RV beltline welds to demonstrate the welds would meet the fracture toughness requirements of paragraphs IV.A.l and IVB of Appendix G, 10CFRPart 50. Sufficent CVN impact test data must be provided to assure the data is a
*conservative representation of the RV beltline weld.
- b. CVN impact test data for RV beltline weld materials from each weld metal test per paragraph NB-2400 of the ASME Co*de.
Response
-* Impact test data for all weld materials used in the beltline region of St Lucie 2 is given in Table 251.4-1. This information was taken from the weld metal certification tests of NB-2400. The highest RTNDT for the beltline region weld materials is -40 F. In all cases RTNDT is fixed by the NDT temperature. No FSAR change is required *
- 251.4-1 Amendment No. 12, (8/82)
SL2-FSAR TABLE ~51.4-1 (Cont'd) CVN Data Absorbed Wire lleat Ele~trode or MDT Temp. Energy Shear Lat. Exp. Sealll* No. No. Plux Lot No.
- -10 ft-lbs 114 60 Mils.
73 83 40 51 116 60 69
+60 145 80 80 148 80 *so 163 100 83
+160 167 100 84 161 100 88 153 100 87
&8018-HABJC -10 -10 127 ND 72 129 ND 75
-10 132 HD 76
+10 158 78 167 82 162 85 88018-PAAPC -.60 0 107 68
- +10 103 124 121 118 117 66 75 75 69 71
!8018-JAGB -40. +10 180 90 146 83 157 86
+20 196 87 210 88 177 81 '
101-1248 83642 3636 see (Tyff 0091) sea11 101-124A E8018 llABJC sqe seam 101-124A E8018 GACJC -80 -30 66 44 68 43 53 37
+10 109 78 118 81 112 75
- 251.4*3 Amendment No. 6, (9/81)
'() _,
SL2..,F~AR
- Wlre Heat TABLE 251.4-1 (Cont'd)
Electrode or NDT .Temp. CVN Data Absorbed Energy Shear Lat. Exp. Seam No. ~o. : Flux Lot No. F F ft - lbs % Mils. 101-1,24C '83642 3536 see (Type 0091) seam 101-124A E8018 IAOCE see seam 101-124A BABEF see
~eam 101-124A HABJC see seam 101-124A FAAFC see seam 10l,.-124A HAGB see seam 101-124A
- 33637 1122 (Type 0091) 1!!8018 JAHB
-so
. -40
+10
+10 1S3 131 125 16S ND ND ND ND 85 81 77 86 138 ND 82 139 ND 76
+20 1S7 ND 86 137 ND 82 157 ND 85 E8018 BAOED -so +10 156 ND 84 149 ND 79 145 ND 77 E8018 AACJD ..;.60 0 119 ND 73 0 101 ND 62 Qi 152 ND 90
+10 126 ND 74
+10 149 ND 91
+10 127 ND 77 101-142A 83637 . 1122 see (Type 0091) seam 101-124C
- 251.4-4 Amendment No. 6, (9/81)
SL2-FSAR
- *Wire Heat TABLE 251.4-1 (Cont'd)
Electrode or NDT Temp. CVN Absorbed Energy Data Shear Lat. Exp.
- Seam No. No, Flu~ Lot No. F F ft lbs % Mils.
101-142B 83637 1122 see (Type 0091) seam 101-124C E8018 FAOED -60 ~so 9 0 4 19 5 10 16 ,5 7
-40 76 40 52 70 ' 40 44 89 50 55 0 98 60 60 102 60 61 86 50 52
+100 157 100 90 152 100 86 153 100 86 Ego1a EACAE -80 -100 21 5 12 19 0 11 27 5 16
-20 ' 83 40 52 89 50 55 85 40 54
+100 123 100 78 124. '100 81 136 100 91 101-14'?B E80l8 GABID -50 +10 104 ND 55 105 64 101 64
+100 153 89 156 91 136 86 101-1.42C 83637 1122 see (Type 00~1) seain 101-124C 101-171 83637 0951 Type 124 -70 -10 88 60 61 99 80 72 94 70 66
+100 112 100 83 116 100 84 116 100 85
- 251.4-S Amendment No. 6, (9/81)
SL2-rFSA~ TABLE 251,4-1 (Cont'd) CVN Data Absorbed Wire T:leat Electrode or ND'l! Temp. *Energy .Shear Lat Exp Seam No, No* Flux Lot* No. .F F Ft-lbs % Mils 3P73l7 0951 Type 124 .,.so ... 20 51 30 41 50 30 39 52 30 41
+100 94 100 64 93 100 63 100 100 70 E8Ql8 lt\QCg s~~
. seam No. 101-124A 68018 F.\AFF -70 -10 66 30 43 52 25 37 74 35 48
+160 127 100 81 132 100 80 122 100 80
- E.8018 KABIF -~o +10 97 50 62 93 50 57 96 50 60
+160 132 100 75
~ 139 100 79 156 100 85
- .251.-4-6 Amendment No. 6, (9/81)
*- 7
SI.Z-FSAR
- Question No.
251.5 For e~ch weld adjacent to a nozzle, a flange and a shell region near a geometric discontinuity in the RV indicate the RTNDT and the method of detet'lllintng the RTNDT* If the method for de~er.111tn:J:ng the RTNDT ls different from that required by paragraph lV*A~l of Appendi~ G~ provide data to demonstrate that the method ls conserv-.,tive.
Response
IE Bullettn*No. 78-12, dated September 29, 1978, required the review and sub11lisJlon of all reactor vessel welding materials test d~ta tQ th~NRC. CE produced '"Information Requested by IE Bulletin 78-12, 'Atypical Material in Reactor Pressure Vessel W~ld9", A,E. Scherer to H.D. Thornburg, LD-79-036, dated June 8, J979, in response to this request. Review of the data in the IE Bul1,ettn respon~e shows tha~ no .~lding material used in CE vess~ls*has aq RTNDT high~r than -ior. Therefore, the St. Lucie II wel~ments tn. ~he RV have RTNDT ~ -lOF. This ts less than the RTNDT of the adjacent base material; and therefore these
~eldments are not limiting for operation.
No FSAR change ts required *
- 251.5-1 Amendment No. 6, (9/81)
SL2rFSAR
- Question No.
251. 6 Fo~ RCPa*welds outside the RV, indicate the RTNDT of the welds that are limiting for operation. Indicate the method for determining the RTNDT* If the method for determining the RTNDT is dt(ferent than that required by paragraph IV.A.l of
.Appendl~ G, provide daea to demonstrate that the method is cons~rvative.
- Regponse The controlltng items for ~he system P-T Uniit curves are the belt line region of the RV, the reactow vessel flange, and the RV flange to upper shel\ transition region. The belt Une region has added to its initial RTNI>T the shift in RTNDT over _the first ten yeus of plant operation. This produces a much highe_r RTNDT than ts a llQwed anywhere else in the RCPB (maximum RTNDT = 60F).
Therefore, no *w~lds OQtside the RV are limiting for operation. No FSAR change is req~ired *
- *251.6~1 Amendment No. 6, (9/81)
SL2-FSAR
- Question No.
251.7 For Spray Nozzles (Code Nos. M-9213-1 and M~9123-2), Letdown Drain Noz~les (Code Nos. ~-9214~1, M9214-2, M-9214-3, and M-9214-4) and Drain No~z1e (Code No. M-r9215-l) piping:
- l. Provide all CVN impact and,drop weight test data from the m<it~rtal.
- 2. If there were insuff~cient tests performed to meet the fracture toughness requirements of NB~2332 of the Summer 72 Addeqda to ~he i971 ASME Code. provide. CVN impact data and drop ~eight test data from o~her material which h~s been procured to the same material specification and heat treated to a metallurgf.cal:J..y equivalent condition as the a~ove Identified piping. The applicant must analyze the data to conservatively demonstrate that the SL-2 piping would have met the Summer 72 Addenda fracture toughness requirements. The applicant must alsp provide heat treatment information from the SL-2 piping and the additional piping which 4emonstrates the piping were heat treated to a metallurigcal equivalent condition.
Response
These components were prq~ured ~o the requirements of NB-2332.a, material with no1Dinal wall thickness of*2..,.l/2" or less. Drop weight testing was not required, and Charpy V-notch testing (Table 251.7-1) was performed to establish a lowest service temperature of 40 F. No FSAR change is required *
- 251. 7-1
- Amendment No. 6, (9/81)
SL2-FSAR
- TABLE 251.7-1 CHARPY V-NOTCH IMPACT DATA FOR NB-2332.a COMPONENTS CVN DATA Component Heat Temp Energy Absorbed Lat. Exp.
Code No. No. F (ft-lb) (Mils) Shear (%) 1 2 3 1 2 3 1 2 3 M-9213-1 -100 2 3 2 2 4 1 1 l 1 M-9213-2 -1.. 0 40 12 11 44 11 9 1 1 1 M-9214-1 A422QT 0 21 30 33 20 30 30 10 20 20
-2 +20 240 240 65 90 88 46 100 100 *30
-3 +40 133 240 240 88 -86 90 80 100 100
-4 +100 240 240 240 81 88 88 100 100 100 M-9215-1 All seven nozzles were manufactured from three forgings taken from the same heat *
- 251. 7-2 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 251.8 Provide a conservative demonstration for Pressurizer Manway Nuts (Number C-5364) and Pressurizer Manway Studs (Number C-5365) that the material when tested at 40 F or lower will meet or exceed 25 mils lateral expansion. Lower bound CVN curves for*SA-193 Gr. B-7 and SA-540 Gr. B-24 materials are considered acceptable methods for extrapolating the CVN impact data from the test temperature to 40 F. In addition, demonstrate that the metallurgical condition of the materials used to generate the lower bound curves for SA-193 Gr. B-7 and SA-540 Gr. B-24 materials are equivalent to the metallurgical condition of the SL-2 material. This can be accomplished by providing the heat treatment information for the material used to generate the lower bound curves and for SL-2 Pressurizer Manway Studs and Nuts.
Response
A. CVN data for SA-193 Gr. B-7 are given in Table 251.8-1. Since full curves are not required for this material, testing over a range of temperatures is not normally done. Results for the Pressurizer Manway Nuts, code no. C-5384, were: Temp F Ft-lbs % Shear Mils Lat. Exp.
+10 53 80 33
+10 25 40 18
+10 41 60 27 At 10 F, two specimens met the mils lateral expansion requirement, 25 mils, while one did not. Since two specimens exhibited over 50 percent shear, this is the center of the temperature range in which the toughness increases rapidly with temperature. By testing at a temperature 30 F higher, 40 F, CE expects that all specimens would exhibit 75-100 percent shear. From the data presented in Table 251.8-1, CE expects an excess of 25 mils lateral expansion at 75 percent shear.
Heat treatment for code no. 5364 is given below; heat treatment data for Table 251.8-1 is given in Table 251.8-2. CE feels that each of these heat treatments produce similar metallurgical structures in this alloy. Austenitized 1550 F, oil quenched Tempered 1000 F Stress Relieved The preload temperature for these fasteners, equivalent to
- their lowest service temperature, is 70 F.
251.8-1 Amendment No. 6, (9/81)
SL2-FSAR The SA 540 Gr. B-24 material used for St. Lucie Unit 2 pressurizer manway studs (code No. C-5365) exhibited between 54 and 58 ft-lbs absorbed energy at 10 F. NRC Bulletin 175, PVRC Recommendations on Toughness Requirements for Ferritic Materials", August 1972, contains energy (ft-lbs) versus lateral expansion (mils) data for bolting steels of the 4340 (SA-540 Gr. B-23 and B-24) composition. Based on this data, CE could expect that no less than 30 mils lateral expansion was developed, but recorded. This would meet the 25 mil lat-eral expansion requirement at 10 F. No FSAR change is required. 251.8-2 Amendment No. 6, (9/81)
SL2-FSAR
- Heat Piece TABLE 251.8-1 IMPACT DATA FOR SA-193 GR. B-7 MATERIAL Absorbed Energy Shear Lateral Exp *.
No. *No. Temp. (ft-lbs) (%)* (Mils.) 1 *2 3 1 2 3 1 2 3 85589 876 +10 69 68 69 100 100 100 45 45 56 876-1 +10 69 69 70 100 100 100 48 48 46 831 +10 65 65 65 100 100 100 44 46 43 831-1 +10 66 68 68 100 100 100 43 43 45 832 +10 60 60 59 100 100 100 41 38 37 832-1 +10 60 60 60 *100 100 100 39 38 38 907 +10 59 60 61 100 100 100 36 38 38 907-1 +10 64 66 65 100 100 100 43 45 45 42540 +10 65 66.5 53 62 64 53 47 50 37 215272 +40 61 74 64 100 100 100 38 44 45 63 59 61 100 100 100 43 33 36 59 68 49 100 100 73 31 30 34 216276 +40 60 63 61 100 100 100 42 37 39 215272 +40 75 90 86 100 100 86 40 73 54 215272 +40 64 55 54 79 66 69 39 32 30
- 251.8-3 Amendment No. 6, (9/81)
SL2-FSAR
- Heat No.
TABLE *251.8-2 HEAT TREATMENT FOR SA-193 GR. B-7 MATERIALS. 85689 Austenltized .1550 F, 1-1/2 hrs at heat, water.quenched Tempered 1100 F, 8 hrs. air cooled Stress Relieved 1100 F, 5 hrs. ' air cooled 42540 Austenltized 1500 F, 2-1/2 hrs. oil quenched Temp~red 1100 F, 5 hrs. ' oil quenched
*215272 Austenitized 1580 F, 8 hrs. ,. oil quenched Tempered 1120 F, 6 hrs.
215276 Austenltized 1570 F, 6 hrs.* Tempered 1140 F, 7 hrs. water quenched 215272 Austenitized 1560 F, 5 hrs. , oil quenched Tempered 1130 F, 7 hrs. 215272 Austenltized 1580 F, 8 hrs. oil quenched Tempered 1170 F, 8 hrs
- 251.8-4 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 251.9 RCPB Heat-Affected Zones 1 ... Provide CVN impact data for all RCPB heat-affected zones a) in the RV which were not fabricated using submerged arc or covered electrode weld processes; and b) outside the RV which were not fabricated using submerged arc or covered electrodes weld processes and are limiting for RV operation.
2 *. If the CVN impact data in B.l cannot be provided, estimate the RTNDT from samples which duplicate the RCPB heat-affected zones. The samples' base material must be fabricated from material that has been procured to the same specification as the SL-2 material and heat treated to the same requirements as the SL-2 material *. The sample's weld metal shall be produced using the same process, post-weld heat treatment, electrode type, flux type and weld manufacturer as the-SL-2 weld.
Response
Since all ferritic RCPB welds for the St. Lucie 2 Unit were made using submerged arc or covered electrode weld processes, no further information regarding CVN impact data is required in
- response to this question.
No FSAR change is required *
- 251.9-1 Amendment No. 6, (9/81)
SL2-FSAR
- Quest.ion No.
251.10 *The materials surveillance program uses six specimen capsules that should contain reactor vess*e1 steel specimens of the limiting base material, weld metal and heat-affected-zone materi*at. To demonstrate compliance with Appendix H, 10CFR Part 50, provide a table that includes the following information for each specimen:
- 1. Actual surveillance material
- 2. Origin of each surveillance specimen (base metal: heat number, plate identification number; weld metal: weld wire, heat of filler material, production welding conditions, and plate material used to make weld specimen);
- 3. Test specimen and types;
- 4. Chemical composition of each test specimen.
Provide the location, lead factor and withdrawal time for each specimen capsule calculated with respect to the vessel inner wall.
Response
Table 251.10-1 'lists the requested information. The weld and HAZ specimens are produced using the same weld procedure as is used to weld the vessel. The HAZ specimens are 1/2 weld metal, *and 1/2 limiting base metal plate, M-605-1. The weld metal: specimens are produced by welding plates M-605-2 and M-605-3 together. The surveillance capsule withdrawal schedule for St. Lucie Unit 2 was established in accordance with 10CFR50, Appendix H, paragraph II.C.3(b). The first capsule is scheduled for withdrawal when the encapsulated base metal material is conservatively estimated to exhibit a reference temperature shift of 50 F. This is predicted to occur after approximately one effective full power year (EFPY) which corresponds to a neutron fluence of about 1.3 x lol8n/cm2 (E lMeV). The second and third capsules are scheduled for withdrawal after 12 and 24 EFPY, respectively. A stgnificant advantage will result from withdrawal of the first surveillance capsule after 1 EFPY, because it will provide an early indication of the validity of the reactor vessel fluence and reference temperature shift predictions used to set the vessel operating limits. Actual dosimetry and shift measurements will then be available for projecting radiation induced changes in the toughness properties of the vessel beltline materials *
- 251.10-1 Amendment No. 6, (9/81)
SL2-FSAR This withdrawal schedule is consistent with the objectives of ASTM El85-79 (Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels) and 10CFR50, Appendix H: - " * *
- to verify the initial predictions of the surveillance material response to the actual radiation environment
. . . . . and " ** to determine the conditions under which the vessel can be operated with adequate margins of safety against fracture. No FSAR change is required. 2~1.10-2 Amendment No. 6, (9/81)
- SL2- FSAR
- TABLE 251.10-1 ST. LUCIE UNIT 2 SURVEILLANCE PROGRAM Withdrawal Schedule Capsule Azimuthal EFPY (Target Lead Surveillance Specimen Type Chemical*
No. Location Fluence) n/cm2 Factor Materials No. & Orientation Composition 1 83° 1.15 1. Plate M-605-1 12 CVN-,L 0.11 Cu 12 CVN-T .008 p 3 Tensile
- 2. Weld Metal* 12 CVN 0.07 Cu Linde 124 Lot No. 09 51 3 Tensile ,009 p Mil B-4 Wire Heat No. 8363 7
- 3. HAZ material 12 CVN-T 0.11 Cu Plate M-605-1 3 Tensile .008 2 97° 24 1.15 1. Plate M-605-1 12 CVN-L 0.11 Cu (2.7x1019) 12 CVN-T .008 p
- 2. Weld Metal* 12 CVN 0.07 Cu N Linde 124 Flux 3 Tensile ,009 p V>
Lot No. 7951 0 Mil B-4 Weld Wire I Heat No. 83637
- 3. HAZ material 1.2 CVN-T 0.11 Cu Plate M-605-1 3 Tensil .008 p 3 Standby 1.15 1. Plate M-605-1 12 CVN-T 0.11 Cu 3 Tensile .008 p
- 2. Weld Metal* 12 CVN 0.07 Cu 124 Flux, Lot 0951 3 Tensile .009 p Mil B-4 Weld Wire Heat No. 83637
- 3. HAZ material 12 CVN 0.11 Cu Plate M-605-1 3 Tensife .008 p z
- 4. SRM Material, .HSST 12 CVN-L Ref. c 0 Plate 01
- Weld Metal specimens are fabricated from the same heat of wire and lot of flux as was used in the weld seam No. 101-171 (closing girth seam weld)
SL2- FSAR
'I' ..\BLF. 251.10-1 (Cont'd)
Wi thdrawa 1 Schedule Capsule Azimuthal EFPY (Target Lead Surveillance Specimen Type Chemical No. Location Fluence) n/cm2 Factor Materials No. & Orientation Composition 4 263° 1.15 1. Plate M-605-1 12 CVN-L O.ll Cu 12 CVN-T .008 p 3 Tensile
- 2. Weld Metal* 12 CVN 0.07 Cu Linde 124 Flux 3 Tensile .009 p Lot No. 0951 Mil. B-4 Wire Heat No. 38637
- 3. HAZ Material 12 CVN O.ll Cu Plate M-605-1 3 Tensile .008 p 5 277° Standby 1.15 1. Plate M-605-1 12 CVN-T 0.11 Cu 3 Tensile .008 p U1
,..... 2. Weld Metal* 12 CVN 0.07 Cu ,..... Linde 124 Lot 0951 3 Tensile .009 p 0 I Mil. B-R Heat 83637 .i:-
- 3. HAZ Material 12 CVN O.ll Cu Plate M-605-1 3 Tensile .008 p
- 4. SRM 12 CVN Ref. c HSST Plate 01 6 284° Standby 1.15 1. Plate M-605-1 12 CVN-L O.ll Cu 12 CVN-T .008 p 3 Tensile
- 2. Weld Metal* 12 CVN 0.07 Cu Linde 124 Lot 0951 3 Tensile .009 p Mil B-4 Heat 8363 7
- 3. HAZ Material 12 CVN O.ll Cu Plate M-605-1 3 Tensile .008 p z
0
*Weld Metal specimens are fabricated from the same heat of wire and lot of flux as was used in the weld seam No. 101-171 (closing girth seam weld)
SL2-FSAR Question No *
- 252.1 We have reviewed Sub sec ti on 10. 2*. 3 of the Final Safety Analysis Report submitted by the applicant. Our evaluation cannot be completed without additional information from the applicant relating to the design, assembly and operating conditions of the low pressure turbiµe discs. Past-experience with similar equipment in the United Kingdom and more recently with Westinghouse turbines in the United States has revealed a propensity for stress corrosion cracking in discs ~hich was not predictable. In order for the staff *to assess the potential for stress corrosion cracking in the applicant's plant, the following informa~ion will be required.
a) What lubricant was used in the hub area of the discs for assembly. b) What are the similarities/differences between the discs in the St. Lucie Unit No. 2 turbines and those used by Westinghouse. c) What are the operating temperatures in the bore area of the discs. d) Which disc or discs are exposed to a moisture level during operation that approximates the level of moisture present in cases of cracking *
- e) f)
'What are the calculated critical crack sizes and what is the method used to calculate that size.
What capability for volumetric inspection of the disc hub areas is available to St. Lucie Unit No. 2.
Response
a) The Lubricant used in the hub area of the discs for assembly is Molybdenum Disulfide *or a Graphite Lubricant.* b) St. Lucie Unit No. 2 L.P. turbine discs are.similar in design to other Westinghouse units. Discs are shrunk on and keyed by means of three keys to the shaft. c) The operating temperatures and moisture leve 1 for the various discs in the bore area are as follows:
- 252.1-1 Amendment No. 6, (9/81)
SL2-FSAR Inlet Bore Exit Bore Metal Temp F Metal Temp F Inlet Steam Outlet Di SC iF 1 2 3 4 (Note 1) 438 280 228 191 (Note 2) 324 263 209 181 Moisture % 0 1.8 6.2 8.5 Steam Moisture 1.8 6.2 8.5 9.8
- 11. 9 5 186 188 9.8 Note (1) 2" from Inlet disc face Note (2) Mid key location or disc outlet edge D. See table above for moisture level of various discs during operation.
E. Critical crack size is calculated to be as follows: Critical Crack Size Critical Crack Size L. P. No. Disc No. Bore (in l.nches) in Keyway (in inches) L.P. No. 1 Disc 1 5.179 2 .665 Governor end L.P. No. 1 Disc 2 2 .696 1.207 Governor end L.P. No. 1 Disc 3 4.243 2 .116 Governor end L. P. No. 1 Disc 4 4.961 2 .537 Governor end L. P. No. 1 Disc 5 3.973 1.957 Governor end L.P. No. 1 Disc 1 2.573 1.135 Gen. end L. P. No. 1 Disc 2 2.220 0 .928 Gen. end L.P. No. 1 Disc 3 3.534 1.699 Gen. end L. P. No. 1 Disc 4 7.179 3 .839 Gen. end L.P. No. 1 Disc 5 8.204 4.440 Gen. end L.P. No. 2 Disc 1 2. 709 1. 215 Governor end L. P. No. 2 Disc 2 2 .196 0.914 Governor end 252.1-2 Amendment No. 6, (9/81)
SL2-FSAR Critical Crack Size Critical Crack Size L.P
- No. Disc No. Bore (in inches) in Ke~al *(in inches)
- L.P. No. 2 L.P. No. 2 L.P. No. 2 Disc 3 Governor end Disc 4 Governor end Disc 5 Governor end 3.697 6.483 7.813 1.795 3.430 4.211 L.P. No. 2 Disc 1 2.660 1.186 Gen .* end L.P. No. 2 Disc 2 2.231 0.934 Gen. end L.P. No. 2 Disc 3 3.648 1. 766 Gen. end L.P. No. 2 Disc 4 4. 749 2.412 Gen. end L.P. No. 2 Disc 5 8.237 4.460 Gen. end The method of calculation is defined in Westinghouse Report "Criteria for L. P.
Nuclear Turbine Disc Inspection" submitted to NRC in June, 1981. Discs 1 and 2 approximates the leve 1 of moisture pre sen_t in cases of cracking. F. The Bores and Keyways ca~ be inspected by ultrasonic inspection techniques without removing the disc from the shaft. St. Lucie Unit No. 2 has been subjected to this inspection at the site *
- 252.1-3 Amendment No. 6, (9/81) p
SL2-FSAR Question No *
- 260.1 .
Response
Provide a statement in Section 17.2 that the QA Program described in the latest revision of "Topical Quality Assurance Report", FPLTQAR l-76A, will be-followed for the operations phase of St. Lucie, Unit 2. See revised Section 17.2 *
- 260.1-1 Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 260.2
Response
.Correct Figure 17.2-1 to agree with FP&L's current organizational structure.
See revised Figure 17. 2-1.
/
- 260.2-1 Amendnent No. 6, (9/81)
SL2-FSAR Question No *
- 260. 3
Response
Provide a statement in Section 17.2 that the safety-related items covered by the QA Program are listed in Table 3.2-1 of the FSAR and supplement Table 3.2-1 accordingly. See revised Section 17.2 and Table 3.2-1 *
- 260.3-1 .Amendment No. 6, (9/81)
SL2-FSAR Question No *
- 260.4 .In Table 1.8-1, update the revisions of the regulatory*guides relating to QA to be consistent with those committed to in the Topical Report FPLTQAR l-76A (new Revision 5 following baseline review of new regulatory guidance).
Response
See revised Table 1.8-1 *
- 260.4-1 Amendment No. 6, (9/81)
SL2-FSAR 1uestion No *
- 260.S Subsection 17.1.2.2 of the standard format (Regulatory Guide 1.70) requires the identifi,cation of safety-related structures, systems, and components controlled by the QA program. You are requested to supplement and clarify Table 3.2-1 of the St. Lucie 2 FSAR in accordance with the following:
- a. The following items do not appear on FSAR Table 3.2-1. Add the appropriate items to the table and provide a commitment that the remaining items are subject to the pertinent requirements of the FSAR operational quality assu*rance program or justify not doing so.
- 1. Safety-related masonry walls (see IE Bulletin No. 80-11).
- 2. Biological shielding within reactor auxiliary building and fuel handling building.
- 3. Missile barriers within reactor auxiliary building, intake structure, fuel handling building, and component cooling area structure, as applicable.
- 4. Spent fuel pool and liner.
- 5. Spent and new fuel storage racks.
- 6. Spent fuel cask.
- 7. Spent fuel handling machine.
- 8. Refueling machine.
- 9. CEA change mechanism.
- 10. Cask handling crane.
- 11. Fuel transfer system.
- 12. Spent fuel handling system.
- 13. Steam trestles for support of main steam piping, feedwater piping, auxiliary feedwater piping and pumps, including associated missile barriers.
- 14. Radiation monitoring (fixed and portable).
- 15. Radioactivity monitoring (fixed and portable).
- 16. Radioactivity sampling (air, surfaces, liquids) *
- 260.5-1 Amendment No. 7, (10/81)
SL2-FSAR 17. 18. 19. Radioactive contamination measurement and analysis. Personnel monitoring. internal (e.g., whole body counter) and external (e.g., TLD system). Instrument storage, calibration, and maintenance.
- 20. Decontamination (facilities, personnel, and equipmen.t).
- 21. Respiratory protection, including* testing.
22.. Contamination control.
- 23. Radiation shielding (permanently installed).
- 24. Accident-related meteorological data collection equipment.
- 25. Expendable and consumable it'ems necessary for the functional performance of safety:...related structures, systems, and components (i.e., weld rod, fuel oil, boric acid, snubber oil, etc.).
- 26. Containment vacuum relief system piping and valves.
- 27. Flood and erosion protection structures such as the sheetpile groins protecting the UHS diversion channel, the concrete bulkhead in the nose of the discharge canal, and flood protection stoplogs.
2~. Roofs of safety-related structures.
- 29. Site drainage system - including drains, parapets, grading, culverts, and channels.
- 30. Intake canal slope.
- 31. Emergency cooling water canal slope (including roadway and retaining walls).
- 32. Int.ake pipes.
- 33. Class I conduits, cooling water lines, and manholes for class I components.
- 34. Class I backfill around safety-related structures.
- b. The following items from the FSAR Table 3.2-1 need expansion and/or clarification as noted. Revise the list as indicated*
or.justify not doing so. 260. 5:...2 Amendment No. 7, (10/81)
SL2-FSAR
- 1. Identify the safety-related instrumentation and control systems to the same scope and level of.detail as provided in Chapter 7 of the FSAR (this can be done by footnote).
Vertfy that this covers I&C for a) Combustible gas control system (item 0). b) Post-accident monitoring including the containment pressure monitor, the containment water level monitor, and the containment hydrogen concentration monitor. c) Containment vacuum relief system.
- 2. Clarify that the iodine removal system (item S) includes the hydrazine tank cover gas system.
- 3. The shield building ventilation system (item P) lists fans, filters, ducting and damper~, and instrumentation.
Either add the following safety-related components or verify that they fall under the headings alread~ listed: a) Fan motors. b) Demisters. c) Charcoal adsorbers
- 4.
d) e) Filter housings. Heaters. Expand the list of items under the emergency power system (item M) to include the following safety-related components or verify that they fall under the headings already listed: a) Raceways and their supports - Those with Class IE cables and those whose failure could damage other safety-related items. b) Diesel generator packages including auxiliaries (e.g., governor, voltage regulator, excitation-system). c) .cab le splices, connectors, and terminal blocks. d) Valve operators. e) Protective relays and control panels. f) 480 volt load centers *
- 260.5-3 Amendment No. 7, (10/81)
812-FSAR g) Battery chargers and 125 volt D.C. distribution equipment. h) 120 volt A.C. vital distribution equipment. i) Safety-related underground cable system.
- c. Enclosure 2 of NUREG-0737, "Clarification of TMI Action Plan Requirements" (November 1980) identified numerous items that are safety-related and appropriate for OL application and therefore should be on Table 3.2-1. These items are listed below. Add the appropriate items to Table 3.2-1 and provide a commitment that the remaining items are subject to the pertinent requirements of the FSAR operational quality assurance program or justify not doing so.
NUREG-0737 Clarification Item
- 1. Plant-safety-parameter display console. I.D.2
- 2. Reactor coolant system vents. II. B. l
- 3. Plant shielding. II.B. 2
- 4. Post accident sampling capabilities. II.B.3 5.
6. 7. Valve position indication. Auxiliary feedwater system. Auxiliary feedwater system initiation II.D.3 II.E.1.1 II.E.1.2 and flow.
- 8. Emergency power for pressurizer heaters. II.E.3.1
- 9. Dedicated hydrogen penetrations. II.E.4.1
- 10. Containment isolatin dependability. II.E.4.2
- 11. Accident monitoring instrumentation. II.F.l
- 12. Instrumentation for detection of II.F.2 inadequate core-cooling.
- 13. Power supplies for pressurizer level II.G.l indicators.
- 14. Automatic trip of reactor coolant pumps. II.K.3(5) 260.5-4 Amendment No. 7, (10/81)
SL2-FSAR
- 15.
16. 17. Power on pump seals. Emergency plans (and related equipment). Equipment and other items associated II.K.3(25) III.A.1.1/III.A.2 III.A.1.2 with the emergency support facilities.
- 18. Inplant I2 radiation monitoring. III.D.3.3
- 19. Control-room habitability. III. D.3. 4
Response
260.5a l~ The analysis and re-analysis (design calculations) 112 re quired by IE Bulletin 80-11 on the safety-related concrete block walls will comply with the pertinent requirements of the FSAR operational quality assurance program. The inspection program and the re-evaluation programs imposed by IE Bulletin 80-11 will erisure the safety-related concrete block walls are installed properly. The concrete block walls are considered as part of the RAB which is already indicated in Table 3.2-1.
- 2. See revised FSAR Table 3.2-1, structure
- 3.
4. 5. See revised FSAR Table 3.2-1, structure See revised FSAR Table 3.2-1, structure See revised FSAR Table 3.2-1, structure
- 6. See revised FSAR Table 3.2-1, structure
- 7. Se~ revised FSAR Table 3.2-l(T).
- 8. See revised FSAR Table 3.2~l(T).
- 9. See revised FSAR Table 3.2-l(T).
- 10. Item (alO) FHB Cask Handling Crane and item (a27) RAB Steel Stop Logs, sheetpile groins, and concrete bulkheads are not to be added to Table 3.2-1. These items include safety related design features to assure that no failure occurs that could significantly reduce the functional capability of associated safety related structures or equipment, these items are subject to the pertinent requirements of the FP&L quality assurance program *
- 260.5-5 Amendment No. 12, (8/82)
SL2-FSAR 11. 12. See revised FSAR Table 3.2-l(T) The spent fuel handling system is described in items a(5) t hroug h a ( 11) , above.
- 13. See revised FSAR Table 3.2-1, structure
- 14. See revised FSAR Table 3.2-l(U).
- 15. Item numbers 15 through 22 and 24 are important to the public health and safety and therefore are controlled by procedures which implementation will be audited under the 112
*FP&L quality assurance program. These items have testing, inspection, surveillance and audit requirements specified in the Tech. Specs. which provides assurance that these items and activities will perform their intended functions.
- 16. See response to 260.5a.15
- 17. See response to 260.5a.15
- 18. See response to 260.5a.15
- 19. See response to 260.5a.15 (Note: All instruments are to be stored in accordance with ANSI N45 *. 2.2 Class B.
- 20. See response to 260.5a.15
- 21. See response to 260.5a.15
- 22. See response to 260.5a.15
- 23. See revised FSAR Table 3.2-1, structure for shielding.
- 24. See response to 260.5a.15
- 25. Item number 25 has been identified as important to safety and is covered by the FP&L QA Program.
- 26. See revised FSAR Table 3.2-l(V)
- 27. See response to item 260.5a.10
- 28. See revised FSAR Table 3.2-1, structure
- 29. These items are not added to Table 3.2-1. In the event of a flood, these items will help alleviate the excess water conditions found on site, however they serve no safety purpose. FP&L procedures provide that proposed changes to plant systems are reviewed for their potential impact on plant safety.
260.5-6 .Amendment No. 12, (8/82)
SL2-FSAR
- 30. Intake canal slope is not seismically designed. In the case of a seismic event the intake canal slope will sluff off. However, even with this reduced area, the canal will be able to provide sufficient water to the plant to preclude an event from occurring. This will not be added to Table 3.2-1, although it is subject to the pertinent surveillance requirements of the Tech. Specs. The Tech *
. Spec. requirements will be audited under the FP&L quality assurance program.
- 31. The Emergency Cooling Water Canal Slope is seismic. The sides of this canal is* reinforced with piles to preclude excess sand from slumping into the canal and disrupting the flow of water. See revised Table 3.2-1.
3,2. Intake pipes are non-seismic. These pipes are not essential to the shutdown of the plant and will not be added to Table 3.2-1. This response is consistent with response a(30) above. *
- 33. See revised FSAR Table 3.2-l(m) for manholes. Class IE conduits are considered as a raceway and are covered under the Emergency Power Systems (M) No. 12. The raceway system is seismically analyzed. Cooling water lines are not used on St. Lucie Unit 2 and are not included on this table *
- 34. See revised FSAR Table 3. 2-1.
260.5b.l Safety. related instrumentation and control systems covered in FSAR Sections 7.2 through 7.6 are included in the quality assurance progr~m in accordance with note 12, Table 3.2-1. a) The combustible gas control system is completely described in FSAR Table 3.2-1(0). All instrumentation and control systems for the components of the systems delineated in this table meet the criteria for said system. b) Refer to revised FSAR Table 3.2-l(y). c) See revised Table 3.2-l(v) 260.5-7 Amendment No. 7, (10/81)
SL2-FSAR 260.5b.2 A 5% solution of Hydrazine is stored in a horizontal tank in -.5 elevation of the Reactor Auxiliary Building. A nitrogen cover gas at a .nominal pressure of 10 psig is provided by the bulk gas storage system. The storage tank is isolated from the non-safety N2 supply line by two seismically qualified Safety Class 2 check valves. The cover gas is not required for pump operability but is provided
*only to prevent the breakdown of the Hydrazine solution. A safety grade pressure alarm is provided in the control room to alert operators to a drop in tank pressure.
260.5.b.3a a) Fan Motors - They are described completely under the heading in Ite~ P (Fans) contained in Table 3.2-1. (b, c, d, e) Demisters, Charcoal Absorbers, Filter Housings, and Heaters are all described in Table 3.2-1 item P by the description of filters. 260.5b.4 a) Raceways and their supports ~ This is covered under Item M ~o.
- 12. The raceways with Class IE and non-Class IE cables are seismically supported except in areas where it can be demonstrated that their failure would not damage any safety-related systems.
b) This is covered under Diesel Generator Control Boards (6) and Diesel Generator Sets (1). c) Cable splices and connectors that are used as interface material were purchased as Class IE material, non-seismic. The terminal.blocks were purchased as Class IE seismic Category I. d) Valve Operators are handled under Diesel Start System (15), Diesel Oil day tanks, Diesel Generator Sets. e) Protective relays are covered under Diesel Generator Control Boards ( 6). f) This is handled under safety-related 480V Switchgear, 125V de, . 120V ac panels, transformers, motor control centers. 260.5-8 .Amendment No. 12, (8/82)
SL2-FSAR g) This is covered under.plant emergency batteries and. invertors
- i)
(8) and safety related 480V switchgear, 125V de, 120V ac
,panels, transformers, motor control.centers.
See Table 3.2-1 Item M.12. Cables by themselves are not seismic Category I. The underground raceway system is designed as seismic Category I. For tornado wind note b is correct. For fluid group one must add* a new note: all.cables are qualified to operate in a wet/dry environment *
.260.5c
- 1. Item I.D. 2, Plant Safety-Parameter Display Control - The* plant safety paramete.r displaY, system is required by NUREG-0695 "Fune tional Criteria for Emergency Response Facilities'",
-February 1981 is a non-seismic and non-Class IE system (Ref.
Appendix 7.5A). Hence it is not added to.Table 3.2-1, although it is subject to the pertinent requirements of the FP&L quality assurance program.
- 2. See revised FSAR Table 3.2-l(W).
- 3. Item I.I.B.2, Plant Shielding, Appendix 12.3A describe's the shielding study performed using TM! source terms. No. safety-re lated. structures or systems were added that needs to be incorporated into Table 3.2-1 *
- 4.
S. See revised FSAR Table 3. 2-1.(X). Ite'lll II.D.3, valve position indication - As .described in. Subsection 7.6.3.8 acoustic .valve flow monitors are provided for pressurizer safety and power operated relief valve position indication. This valve position indication is designed per NUREG 0737 item II.D.3 requirements and is not safety grade. Hence it is not included in Table 3.2-1, although it is subject to the pertinent requirements of the FP&L quality assurance program.
- 6. This is contained in FSAR Table 3.2-l(L).
- 7. This is contained inFSARTabl;e.3.,2.l(L) No. 4.
- 8. Item .II.E.3.1 Fmergency Power for Pressurizer Heaters - As described in App~ndix 1. 9A item II. E. 3.* 1, no safety *g~ade systems components, or structures were added because of this NUREG 0737 item hence no revisions for Table* 3. 2-1 *is necessary, al though* i.t is subject to the pertinent requirements of the FP&L quality assurance program *
- 260.5-9 Amendment No. 7, (10/81)
SL2-FSAR
- 9. Item II.E.4.1, De.dicated Hydrogen Penetration - As described*
in Appendix l.9A item II.E.4.1, no change to the existing design is required hence no revision is required to be made for Table 3.2-1.
- 10. Item Il*E.4.2, Containment .. isolation. Dependability - no safety grade system, components or structure* i's required to be added because of this NUREG* 0737 'requirements (Appendix 1. 9A), hence no change to Tabl*e 3. 2-1 is to be made.
- 11. Item II.F.l, Accident Monitoring Instrumentation - The list of instruments for this item have been added to Table 3.2-1 item Y.
- 12. A description of the l.nstrumerttation for. inadequate core cooli:ng (ICC) is provided *in Appendix l.9B. ICC instrumentation will meet the requirements of NUREG 0737 and will be added to Table 3.*2-1 when the design is finalized.
- 13. Item II.G.l, Power Supplies for Pressurizer Level Indicators -
same as above item 10.
- 14. The *St. Lucie Unit 2 reactor . penetration system does not include automatic trip of the RCP. Rather the pumps are tripp~d manually in accordance with emergency procedures established as part of the CE owners group post-TM! activities *
- 15. A safety grade supply of Component Cooling Water to Reactor Coolant Pumps Seals is not required since tests (see the response to Question 410.19) have shown that the pump seals
.can withstand a Loss of Cool*ing Water for more than 50 hours without losing integrity.
- 16. *This is .not* identified :as safety-related and is subject to the same requirements as outlined for items 15 et al. of Question 260.5a above.
I
- 17. Item III.A.1.2 Emergency Support Facilities - NUREG 0696 does not require safety grade instrumentation data display systems and*connnunication systems be provided for the Jµnergency Support Facilities. St'ructures and ventilation systems are not required to be seismic Category I (NUREG 0696), hence not included"' in Table 3. 2-1, al though they are subject to the pertinent.requirements of the FP&L quality assurance program.
- 18. The response for item 16 is applicable to. item 18, including the requirements noted in FSAR Table 3.2-1 (21.4)
- 19. Item III.D.3.4 Control Room Habitability - see item 10 above.
260. 5-10 Amendment No. 7, (10/81)
SL2-FSAR Question No *
- 271.1 In accordance with the requirements of GDC 2 and 4 all safety-related equipment is required to be designed to withstand the effects of earthquakes and dynamic loads from normal operation, maintenance, testing and postulated accident conditions. GDC 2 further requires that such equipment be designed to withstand appropriate combinations of the effects of normal and accident conditions with the effects of earthquake loads.
- The criteria to be used by the staff to determine the acceptability of your equipment qualification program for seismic and dynamic loads are IEEE Std. 344-1975 as supplemented by Regulatory Guides 1.100 and 1.92, and Standard Review Plan Sectlons 3.9.2 and 3.10. State the extent to which the equipment in your plant meets these requirements and the above requirements to combine seismic and dynamic loads. For equipment that does not
- meet these requirements provide justification for the use of ot.her criteria.
Response
Seismic qualification of equipment was performed for the effects of normal, accident and seismic load$ as described in FSAR Sections 3.9 and 3.10. The analysis and testing methods employed are consistent with applicable regulatory requirements and are described in FSAR
- See FSAR Section 3.7, 3.9 and 3.10 *
- 271.1-1 Amendment No. 9, (3/82)
SL2-FSAR Question No.
- 271.2 Provide a list of all safety-ielated systems together with a list of all safety-related equipment and support structures associated with each system. The equipment lists should indicate whether the equipment is NSSS supplied or BOP supplied. These lists should include all safety-related mechanical components, electrical, instrumentation, and control equipment, including valve actuators and other appurtenances of active pumps and valves.
Response
The list of all safety-related systems together with a list of all safety-related equipment have been supplied in letter L-81-507, dated December 2, 1981. No FSAR change is required *
- 271. 2-1 Amendment No. 9, (3/82)
SL2-FSAR Question No *
- 271.3 For each safety-related equipment item, the following information should be provided:
(a) Method of qualification used:
- 1) Analysis or test (indicate the company that prepared the report, the reference report number and date of the publication).
- 2) If by test, describe whether it was a single er multi-frequency test and whether input was single axis or multi-axis.
- 3) If by analysis, describe whether static or dynamic, single or multiple-axis analysis was used.
- 4) Provide natural frequency (or frequencies) of equipment.
(b) Indicate whether the equipment has met the qualification requirements. (c) Indicate whether the equipment is required for:
- 1) hot standby
- 2) 3)
4) cold shutdown both neither (d) Location of equipment, i.e., building, elevation (e) Availability for inspection (Is the equipment already installed a~ the plant site?) (f) A compilation of the required response spectra (or time history) and corresponding damping for each seismic and dynamic load specified for the equipment together with all other loads considered in the qualification and the method of combining all loads.
Response
The information requested above for all safety-related equipment are presented via filling out appropriate sections of the Qualification Summary of Equipment sheets. These sheets have been supplied letter L-81-507, dated December 2, 1981. It is to be noted that item (e), installation status of each equipment, is not presented in *these sheets. This data will be available by February 1982 and will be submitted under a separate cover
- No FSAR change is required.
271. 3-1 Amendment No. 9, (J/82)
SL2-FSAR Question No *
- 271.4
Response
Identify all equipment that may be effected by vibration fatigue cycle effects and describe .the methods and criteria used to qualify this equipment for such loading conditions. Where applicable, vibration and fatigue effects are included in the analysis or testing performed to deconstrate the qualification of equipment. A review of methods of analysis or test used to account for vibration and fatigue can be reviewed in detail for the equipment reviewed during the SQRT site audit. No FSAR change is required *
- 271.4-1 Amendment lfo,*. 9, (3/82)
.SL2-FSAR Question No *
- 271.5
Response
Describe the results of any in-plant tests, such as i.n situ impedance tests, and any plans for operational tests which will be used to confirm the qualification of any item of equipment. No specific in situ tests have been or will b~ conducted to confirm the q11alification of any equipment. The only in plant testing which is planned and is of a similar nature is Pre-operational Vibration, Thermal Expansion and Dynamic testing on piping systems. This* testing is described in FSAR Subsection 3.9.2.1. No FSAR change is required *
- 271.5-1 Amendment No, 9, (3/82)
SL2-FSAR Question No *
- 271.6 To confirm the extent to which the safety-related equipment meets the requirements of General Design Criterion 2 and 4, the Seismic Qualification Review Team (SQRT) will conduct a plant site review. For selected equipment, SQRT will review the combined required response spectra (RRS) or the combined dynamic response, examine the equipment configuration and mounting, and then determine whether the test or analysis which has been conducted demonstrates compliance with the RRS if the equipment was qualified by test, or the acceptable analytical criteria if qualified by analysis.
The staff requi.res that a "Qual:lfication Summary of Equipment" as shown on the attached pages be prepared for each selected piece of equipment and submitted to the staff two weeks prior to the plant. site visit. The applicant should make available at the plant site for SQRT review all the pertinent documents and reports of the qualification for the selected equipment. After.the visit, the applicant should be prepared to submit certain selected documents and reports for further staff review.
Response
The "Qualification Summary of Equipment" has been partially filled-out for all equipment and has been supplied as an attachment *. Upon the receipt of the list of selected equipment for SQRT site audit, "Qualification Summary of Equipment" sheets will be filled out completely and will be transmitted to NRC SQRT team two weeks prior to the scheduled plant site review date. No FSAR change is required *
- 271.6-1 Amendment No. 9, (3/82)
812-FSAR Question No *
- . 280.1 You state in your Fire Hazards Analysis how various safety-related cable trays, conduit and equipment are separated by distance from its redundant counterpart, and the criteria that were used to establish barriers between these redundant trains. However, your fire hazards analysis does not provide adequate protection for the effects of postulated expcsure fires involving permanent and/or transient combcstibles (exposure fires) on systems, circuit cable trays or equipment required for safe plant cold shutdown which are separated only by distance (e.g., no fire barriers between redundant trains 20 ft. or less from each other, as listed in page 9A-21 of the FHA). It is our position that as a minimum, redundant trains within 20 ft. of
.I each other should be protected by a one hour fire rated barrier as well as area automatic sprinklers. In some instances, such as the reactor coolant makeup and injection systems, redundant trains separated by more than 20 ft. may require additional protection. In the fire hazards analysis, you need to demonstrate that, assuming failure of the primary suppression system, a fire on installed or transient combustibles will not result in the loss of capability to achieve safe shutdown.
Where this cannot be demonstrated, an alternate means of assuming safe plant shutdown (cold shutdown) should be provided. Alternate shutdown will most likely be required for areas such as the cable spreading rooms, and the control room *
- Demonstrate:
(a) Safe shutdown from the main control room where a fire disables any remote shutdown panel, or any safe shutdown equipment including conduit/cable trays controlled from remote locations. (b) Safe shutdown from remote locations when the main control room is uninhabitable due to a fire or when fire disables safe shutdown equipment or cables in the cable spreading areas. Alternate shutdown capability without fire damage need only be provided for the essential instrumentation, controls and equipment necessary to bring the plant to a hot standby condition. Fire damage to systems necessary to achieve and maintain cold shutdown should be limited so that repairs can be raade and cold shutdown condition achieved within 72 hours. Attached (Enclosure 2) are our guidelines for alternate shutdown syste111s *
- 280.1-1 Amendment No. 9, (3/82)
SL2-FSAR
Response
Safety related cable trays, conduit and equipment which are required for safe shutdown (as defined in FSAR Appendix 9.SA, Section 2.0) are located, protected, and/or routed with regard to their redundant trains as follows: a) separate fire areas which are bounded by three hour barriers, b) in commonly designated fire areas separation by a 20 ft distance with no intervening combustibles, or c) limited fire barriers for congested areas with additional suppression features, or d) alternate shutdown equipment to maintain the plant safe in the event of disabling fire. Isolation is provided such that a disabling fire in any ar~a (including alternate shutdown locations) will not impair safe shutdown capability. In general, redundant trains of equipment, power and control cables are protected in accordance with the requirements of Section III G of Appendix R to 10CFRSO. For deta{ls see fire area by fire area analysis (Attachment 1) to letter L-82-20 January 19, 1982. 280.1-2 Amendment No. 9, (3/82)
.SL2-FSAR Question No *
- 280.2 Substantiate the fire resistance capability of the barriers used to separate safety-related areas or high hazard areas by verifying that their construction is in accordance with a particular design that has been fire tested. Describe the design, the test method used and the acceptance criteria.
*Provide information for the following components:
(a) Rated fire barriers, including floor and ceiling construction and the support for barriers that are not floors or ceilings; {b) Fire dampers and fire doors, including a. description of how they are installed in the ventilation ducts that penetrate rated fire barriers of safety-related areas; and
.(c) Fire barrier penetration seals around cuts, pipes, cables, cable trays and in other openings (e.g. concrete joints sealers and fillers) including verification that all seals are of the thickness specified in the tests, and that cables and cable trays are supported in a manner similar to supported in a manner similar to supporting arrangements used in any tests.
Response
2a) Barriers used to separate safety-related areas and high hazard areas are reinforced concrete construction of 18 inch to over 36 inch thickness. In some few particular instances a thickness of eight inches is used. Such walls and floor-ceiling designs have been established as having fire resistance rating of over four hours, as verified by N.F.P.A. 251, Standard Methods of Fire Tests of Building Construction and Materials (this test is synonymous with ASTM E 119 and UL 263). In few particular instances, eight inch and 12 inch concrete block wall are used as noted in the FHA. In these cases, UL equivalent block, rated for three hour fire resistance has been used *
- 280.2-1 Amendment No~ 9, (3/82)
SL2-FSAR 2b) The fire dampers provided are 3 hours fire rated and UL listed. They are installed in accordance with National F'ire Protection Code 90A. A maximum clearance of 1/2 in. is provided for expansion. The thickness sleeve is equal to the gage of the connecting ductwork. Mounting angles bolted to the sleeve are overlapping the wall by a minimum of 1-1/2 inches. Removable gasketed access doors* are installed in the ductwork for service, maintenance and resetting of the fire dampers. Fire doors, and acces;,;ori.es are specified UL listed or labeled. Fire resistance* rating of 1-1/2 hours and 3 hours are per UL 555 and lOB; "Fire. Test of Door Assembly". Installation is in accordance with NFPA 80-Standard for the
*Installation of Fire Doors and Windows, and in *accordance wi~h manufacturer's installation instructions.
2c) Fire barrier penetration seals around pipe, ductwork and cables essential to safe shutdown will be sealed. The fire rating of sealing media will be consistent with the fire barrier rating. 280.2-2 Amendment No. 9, (3/82)
812-FSAR
- Question No.
280.3 It is our position that you comply with Se.ction D.5(a) of Appendix A to BTP ASB 9.5-1, in that a fixed emergency lighting system consisting of sealed beamed units with individual (8-hour minimum) battery power supplies should be installed in all areas required for safe shutdown operations, including access and egress routes. Verify that you will comply with our position.
Response
The design is presently based upon the use of Multi-Lamp Units with a four (4) hour capacity battery supply. The design has been upgraded to assure compliance with eight (8) hour minimum requirement per BTB ASB 9.5-1 of Appendix A*
- 280.3-1 Amendment No. 12, (8/82)
/
.SL2-FSAR Quest ion No *
- 280.4 Demonstrate that primary and secondary power for the alarm system can be maintained by:
a) Using normal offsite power as the primary supply, with a 4"'.'hour battery supply as the secondary source.
- b) Having the capability for manual connection to the Class IE emergency power bus within 4 hours of loss offsite power.
Response
The fire detection system is supplied by an Uninterruptible Power Supply (UPS). . This UPS is normally connected to Motor Contra 1 Center (MCC) 2AB. This MCC is aligned with one of the plant class IE emergency Diesel Generators. The UPS is also connected to station battery 2C. Upon loss of normal offsite power the UPS will automatically be supplied by the station battery until they are loaded onto the emergency diesel generator. In addiiion to this highly reliable power supply, the Main Fire Detection Panel will be provided with a self.contained four hour battery *
- 280.4-1 Amendment No. 9, (3/82)
SL2-FSAR Question No *
- 280.5 Verify that the fire pumps can provide, in accordance with BTP ASB 9.5-1 Section 5.b.5, the largest firewater flow and pressure
{based on 500 gpm for manual hose streams plus the largest design demand of any sprinkler or deluge system as determined in accordance with NFPA 13 or NFPA 15) with the largest fire pump out of service. Also, indicate how the minimum firewater supply is reserved in each storage tank. *
Response
A 70 sq. ft. sprinkler spacing permits .20 gpm/sq. ft for the end sprinkler head at an end head pressure of 6.25 psi. This requires 2,256 gpm at 96 psi at the system interface (6 x 11 flange at elevation 23 ft). This requirement is available from either* fire pump after accounting for yard main friction loss and deduction of 500 gpm for hose streams. overall, the system will provide an average density of .226 gpm/sq. ft over 10,000 sq. ft. 2260 gpm plus 500 gpm for hose is the maximum flow requirement. On this basis, total flow required for a two hour duration is 331,000 gallons. 'nle plant has two 500,000 gallon tanks, each of which have a dedicated volume equivalent to two hour system operation for fire protection at full capacity *
- 280.5-1 Amendment No. 9, (3/82) *
- 0762W-6 SL2-FSAR
- Question No.
280.6 Verify that the plant fire brigade:
- 1) Consists of five members during every shift.
- 2) Confirm that the fire brigade provide for local fire department participation in drills at least annually.
Response
- 1) The site fire brigade consists of five members present during every shift. 11
- 2) Local Fire Department Supervision and Paramedic/Ambulance personnel presently participates in at least one drill annually, Supervision and personnel from the two nearest stations are given yearly familiarization tours of the site. Selected (large) Fire Department equipment has been brought to the St. Lucie Unit 1 site to check maneuverability and clearances .
- 280.6-1 Amendment* No. 11, (7 /82)
ST LUCIE FSAR Question No
- 280.7 Provide a plot plan of the site showing all fire protection water mains to include pipe sizes, valves, location of hydrants, and fire pump connections. Also include details on the fire pumps and suction and discharge headers. It is our position that the piping be arranged and valved such that no single break will cause the loss of both fire plUUps or shut off all fire protection water to any area of the plant. Also, it is our position that you provide standpipe and hose stations throughout the plant in accordance with NFPA 14, including permanent hose stations inside containment to meet Section C.3.d of Appendix A to BTP ASB 9.5-1 *
. Response Plot plan, flow diagrams and pipe and valve layouts demonsrate that no single break will cause loss of both fire pumps or shut off all fire protection water to any area of the plant (see Figures 1.2-2, 9.2-5, 9.2-6 and M-708 sheets (1 and 5a, FHA).
Standpipe and hose stations are provided throughout the plant in accordance with NFPA 14 (see FHA, Section 4.2.2 and revised Table 4-2
- Hose stations will be provided inside the Reactor Containment Building at strategic locations, primary water will be used in these basic stations *
- 280. 7-1 Amendment No. 9, (3/82)
SL2-FSAR Question No *
- 280.8 Verify that the fire pumps and their controllers are UL listed and installed in accordance with NFPA 20 requirements. 'lhe fire pumps start-up setpoints should be adjusted such that both fire pumps do not start simultaneously (at least a 5 to 10 second delay between pump start-ups i_s required by NFPA 20).
Response
Florida Power and Light Company has responded to the above question per a letter, L-81-4e, dated February 11, 1981 *
- 280.8-1 Amendment No. 9, (3/82)
SL2-FSAR Question No
- 280.9 You indicate that the normal ventilation systems will be used to provide venting of smoke and other combustion products. Verify that the ventilation systems are designed to handle the high temperature gases which may be expected. Also provide a description of the operating mode of all such equipment serving each area in the event a fire were to occur in that ar~a.
Indicate the operating modes of the ventilation systems in areas adjacent to the fire area.
Response
Recognizing the need to maintain ~he flow of cooling air to essential operating equipment the potential for inadvertent operation and the consequential blocking of air flows, is avoided by not providing fire dampers in all fire area ductwork penetrations of safety-related equipment areas. Fire dampers could be susceptible to inadvertent failure apart from fire considerations. Due to the limitation on the amount of exposed combustibles subject to a fire, and preaction automatic sprinkler systems where concentration of combustibles is heavier and to the continuation of ventilation flows to remove smoke and heat and to supply cooler makeup air, high temperatures are not expected to develop *
- Although localized high off gas temperatures can be expected in the immediate vicinity of combustion, air mixture temperatures in the area are not expected to reach high temperature. For limited durations of exposure, we believe that the standard ventilating system equipment could be expected to remain operational. in addition, excessive expansion deformation or heat failure of ductwork is not anticipated. The ventilation system consists of independent supply and exhaust systems except for specific areas which require separate exhaust for radiation reasons. i.e., ECCS Areas.
Fire Dampers will be installed in non-safety related ductwork penetrating barriers between redundant essential equipment. A complete summary of all fire dampers to be provided is shown .in Attachment .5 of letter L-82-20 dated January 19, 1982
- 280. 9-1 Amendment No. 9, (3/82)
SL2-FSAR Question No
- 280.10 It is our position that you provide portable radio communications between the fire brigade and any control center for fire emergencies.
Response
At least 2 portable radios for emergency use now provided in St Lucie Unit 1 Qlntrol Room, as is a base station. Normally the fire team would use a portable radio to communicate with the Qlntrol Room base station. If it were necessary to use a fire control center other than the Control Room, portable radios would be available and used *
- 280.10-1 Amendment No. 9, (3/82)
SL2-FSAR Question No
- 280.11 Provide the name and qualifications of the fire protection engineer assigned the responsibilit*y for the Fire Hazards Analysis at St Lucie 2.
Response
Resumes of Fire Protection Engineers responsible for Fire Hazard Analysis for St lucie Unit 2 were provided under separate cover letter L-81-369 dated August 25, 1981 *
- 280.11-1 Amendment No. 9, (3/82)
SL2-FSAR
- Question No.
280.12 Provide a list of all interior finish insulation, sound proofing, etc. that are other than noncombustible. Indicate the flame spread and smoke contributed ratings of each material where available. Indicate where their materials are use.d (by fire area) and in what quantities.
Response
As outlined in Section 6 of the FHA, 'li~alls and structural materials are non-combustible. An area by area comp~lation of combustible loads is presented in the FSAR, App. 9.SA, FHA
- 280.12-1 Amendment No. 9, (3/82)
812-FSAR Question No
- 280.13 It is our position that you provide an automatic suppression system for the diesel generator rooms and for the diesel generator fuel oil storage tanks to meet the guidelines of Section D.9 of Appendix 9 to BTP ASB 9. 5-1.
Response
The diesel generator rooms will be provided with an automatic sprinkler system. 'Th.e two 41,500 gallon diesel oil storage tanks are in separate enclosures of at least 3 hour fire resistance. Fire detectors for early warning, portable. fire extinguishers and available yard fire hydrants for manual first-aid fire fighting, and portable foam generating equipment combine to provide adequate fire protection for the oil storage tanks. For a detailed discussion for FHA of diesel fuel oil storage area see FSAR App 9.5A *
- 280.13-1 Amendment No. 9, (3/82)
SL2-FSAR
- Question No.
280.14 It is our position that you provide pressurized water type portable fire extinguishers at strategic locations throughout the plant, including areas such as the control room and cable spreading room to meet the guidelines of Section C.6 of Appendix A to BTP ASB 9.5-1.
Response
Portable pressurized water type fire extinguishers are provided in the Control Room, Cable Spreading Room, and other strategic I 12 locations throughout the plant *
- 280.14-1 Amendment No. 12, (8/82)
SL2-FSAR
- Question No.
280.15 It is our position that all valves in the fire protection water system be electronically supervised with alarm and annunciation in the *control room to meet the guidelines of Section C. 3. b of Appendix A to BTP ASB 9.5-1.
Response
All valves in the automatic fire prqtection water system will be I12 electronically supervised with alarm and annunciation in the control room as per BTP ASB 9.5-1, Section C.3.b. Valves in the fire water main are manual post indicator type and are under administrative controls
- 280.15-1 .Amendment No. 12, (8/82)
SL2-FSAR
- Question No.
280.16 Page 9.SA.7-22 of your analysis is not included in your submittal and needs to be provided to complete your FHA.
Response
Copies of pages 9.SA.7-22 were provided to the NRC during a meeting on May 5, 1981 *
- 280.16-1 Amendment No. 12, (8/82)
812-FSAR
- Question No.
280.17 It is our position that you provide loss of ventilation air flow alarms for all battery rooms with alarm and annunciation in the Control Room to meet the guidelines of Section D.7 of Appendix A to BTP ASB 9.5-1.
Response
Flow switch with alarms in the Control Room are provided to 112 annu~icate loss of ventilation in the battery rooms *
- 280.17-1 Amendment No. 12, (8/82)
SL2-FSAR Question
- 280.18 Fire Area 14 (Primary our position that you collection system for comply with Section D and Secondary Shield Wall location): It is provide an engineered oil containment and the reactor coolant pump lube oil system to of Appendix R to 10 CFR Part 50.
Response
The Reactor Coolant Pump Lube Oil Collection System will be provided to meet the following criteria: a) Capable of collecting lube oil from all potential pressurized and unpressurized leak.age sites in the reactor coolant pumps' lube oil systems, b) Capable of draining lube oil from the collection system at the pumps to a safe location at a rate in excess of the largest anticipated leak .in the lube oil systems, c) Seismically analyzed to insure the system will remain on the Reactor Coolant Pump Motors during Design Basis Earthquake conditions, and d) Capable of collecting 225 gallons of lube oil. This quantity is in excess of the quantity which would require unit shutdown to investigate (Approximately 15 gallons for each pump). Also the capacity is in excess of the entire capacity of lube oil system of a reactor coolant pump of 190 gallons *
- 280.18-1 Amendment No. 9, (3/82)
.SL2-FSAR Question No .
280.19 Fire Areas 22 and 23 (Electrical Penetration Area): Automatic suppression systems should be provided for each of the electrical penetration rooms.
Response
An automatic preaction suppression system will be providPd in P.ach electrical cable penetration room *
- 280.19-1 Amendment.No. 9, (3/82)
0771W-13 812-FSAR
- Question No.
280.20 Fire Area 29 (RAB Drumming Storage Area): An automatic water extinguishing system should be provided for this area.
Response
The RAB Drumming Storage Area is a separate fire zone, enclosed by I 11 walls and floors which have a grea'ter than 3 hour fire rating. The entrance doorway is a three Lour rated fire door. Based on the FHA combustible load, fire confinement, availability of manual fire extinguishing equipment, and the absence of any safety-related equipment or cable in the area, the existing fire suppression features a re considered more than adequate for the hazard involved. Area smoke detectors will be installed which will alarm in the control room .
- 280.20-1 Amendment No. 11, (7/82)
SL2-FSAR Question No *
- 280.21
Response
Drawing SK-2998-M-708, Sheet SA, was omitted from your submittal and should be provided to complete your FHA. A print of drawing SK-2998-M-708, sheet SA was provided at a May S, 1981 meeting with the NRC *
- 280.21-1 Amendment No. 9, (3/82)
0771W-15 SL2-FSAR
- Question No.
280.22 Fire Areas 32 and 33 (RAB Decontamination Room and Storage Area): An automatic water suppression system should be provided for both of these areas.
Response
As noted in Section 5.4 of the Fire Hazard Analysis, Fire Zones 32 I 11 and 33 (RAB Decontamination Room and Maintenance Storage Area and RAB Repair Shop and Storage Area, respectively) are each segregated from any adjacent fire zones, as well as from each I 11 other by walls, floors, and ceilings having fire resistance ratings of greater than 3 hours. The present FHA is based on a combustible loading of Class A materials such that the addition of combustibles (e.g. a small amount of lube oil) from the recent installations of PASS and RAD monitoring skids do not negate the conclusions of the analysis therein. These areas .contain no safe shutdown equipment or cable (other than the cable in the enclosed chase) and the smoke detectors with alarm signals to the control room will provide early warning of a fire condition. Effective manual fire. fighting, utilizing fire extinguishers and hose stations, will permit adequate suppression activities without compromising the plant's ability for safe slrutdown and without the need for an automatic suppression system in the area *
- 280.22-1 Amendment No. 11, (7/82)
SL2-FSAR Question No *
- 280. 23
Response
Fire Area 42 (Control Room): It is our position that area smoke detection be provided for the control room to meet Section D.2 of Appendix A to BTP ASB 9.5-1. Smoke detectors are installed in the hung ceiling above the control room, over cable trays, inside the RTG Boards, and in the kitchen area. *The control r-qom is manned on a 24 hour per day basis. St Lucie Unit No. 2 complies with Appendix A to BTP ASB 9.5-1, Section D~2 *
- 280.23-1 Amendment No. 9, (3/82)
812-FSAR Question No
- 280.24 Appendix R to 10 CFR Part 50 will also be used as guidance for our review of your fire protection program. Your compliance with the requirements set forth in Appendix R as modified by accepted exceptions your program takes to the requirements of Appendix R as well as BTP ASB 9.5-1, and describe your alternative for providing an equivalent level of fire protection.
Response
The evaluation of each Fire Area was provided under separate cover letter L-82-20 dated January 19, 1982 *
- 280.24-1 .Amendment No. 9, (3/82)
SL2-FSAR Question No
- 280. 25 In accordance with Section 9. 5.1, Branch Technical Po sit ion ASB 9.5-1, position C.4.a.(l) of NRC Standard Review Plan and Section III.G of new Appendix R to 10 CFR Part SO, it is the staff's position that cabling for redundant safe shutdown systems should be separated by walls having a three-hour fire rating or equivalent protection (see Section III.G.2 of Appendix R). That is, cabling required for o.r associated with the primary method of shutdown, should be physically separated by the equivalent of a three-hour rated fire barrier from cabling required for or associated with the redundant or alternate method of shutdown. We need assurance that redundant sh*1tdown cable systems and all other cable systems that are associated with the shutdown cable systems are separated from each other so that both are not suhject to damage from a single fire hazard. Complete responses to the following list of requests for information should provide the information for each system needed to bring the plant to a safe shutdown.
Response
An analysis of all essential equipment and cables has been performed in order to insure that no single fire can prevent St. Lucie Unit 2 from achieving a safe cold shutdown. In this regard a list of all equipment and power sources required to bring the plant to cold shutdown has been compiled (see Attachment 2 of letter L-82-20 dated January 19, 1982). This list differentiates between equipment required to maintain hot standby and that required for cold shutdown only. All cables required for the operation of these components and power sources are listed numerically with their routing by fire area in Attachment 4 of letter L-82-20 dated January 19, 1982. In addition these cables are shown on a set of drawings prepared specifically for use in this analysis. To facilitate this review a computer program has been prepared which lists all essential equipment and cables that appear in a fire area (Attachment 3 of letter L-82-20 dated January 19, 1982). This report is analyzed by fire area, in conjunction with the physical drawings, to determine if the operation of redundant essential components could be impaired by a single fire in this area. No credit is taken for repair of cold shutdown equipment. The results of this analysis are presented by fire area in Attachment I of letter L-82-20 dated January 19, 1982 along with a list of the equipment and cables which must be protected or relocated. This analysis demonstrates that with the plant changes listed in Attachment 1 of letter L-80-20 dated January 19, 1982 along with selected manual actions no singl~ fire could prevent the plant from being brought safely to cold shutdown *
- 280.25-1 Amendment No. 9, (3/82)
SL2-FSAR Question No
- 280.26 Provide a table that lists all equipment including instrumentation and vital support system equipment required to achieve and maintain hot and/or cold shutdown. For each equipment listed:
a) Differentiate between equipment required to achieve and maintain hot shutdown and equipment required to achieve and maintain cold shutdown, b) Define each equipment's location by fire area, c) Define each equipment's redundant counterpart. d) Identify each equipment's essential cabling (instrumentation, control, and power). For each cable identified: (1) Describe the cable routing (by fire area) from source to termination, and (2) Identify each fire area location where the cables are separated by less than a wall having a three-hour fire rating from cables for a redundant shutdown system, and e) List any problem areas identified by item 280.26d(2) above that will be corrected in accordance with Section III.G.3 of Appendix R (i.e., alternate or dedicated shutdown capability).
Response
- Attachment 2 of letter L-82-20 dated January 19, 1982 to this report is a list of all equipment and their power sources required to bring the plant to cold shutdown. This list differentiates between equipment required for hot standby and cold shutdown, defines component location by fire area, and lists the redundant components of an essential system.
As outlined in the response to Question 280.25 all essential cables are listed by fire area and numerically with routing by fire area from source to temination in Attachments 3 and 4 of letter L-82-20 dated January 19, 1982 respectively. All cables are shown on the drawings in Attachment 6 of letter L-82-20 dated January 19, 1982. Attachment 1 lists those areas where the function of essential redundant components are threatened by a single fire. In these cases, as specified, certain equipment and cables will be relocated or isolated to meet the intent of Appendix R *
- 280.26-1 Amendment No. 9, (3/82)
SL2-FSAR
- Question No.
280. 27 Provide a table that lists Class IE and Non-Class IE cables that are associated with the essential safe shutdown systems identified in item 1 above. For each cable listed: (See *Note*) a) Define* the cables' association to the safe shutdown system (common power source, common raceway, separation less than IEEE Standard-384 guidelines, cables for equipment whose spurious operation will adversely affect shutdown systems, etc). b) Describe each associated cable routing (by fire area) from
*source to termination, and c) Identify each location where the associated cables are separated by less than a wall having a three-hour fire rating from cables required for or associated with any redundant shutdown system.
Response
Associated cables are defined (per IEEE 384-1974) as non-class lE circuits that share power supplies, enclosures or raceways with
.Class lE circuits or are not physically separated from class lE circuits or equipment *
- Associated circuits comply with one of the following:
- 1) 'Ibey are uniquely identified as associated and remain with or are separated from the same as those class lE circuits with which they are associated. 'Ibey are subject to all the requirements placed on class IE circuits, such as cable derating, environmental qualification, flame resistance and raceway fill unless the absence of such requirements could not significantly reduce the availability of class lE circuits.
- 2) They are in accordance with (1) above from the class IE equipment up to and inc~uding an isolation device.
Isolation devices are devices in a circuit which prevent malfunctions in one section of a circuit from causing unacceptable influences in the other sections of the circuit or other circuits. Class IE qualified circuit interrupting devices actuated by fault and overlaod current are considered to be isolation devices. Beyond an isolation device a circuit is considered a non-class IE circuit provided it does not again become associated with a Class IE circuit *
- 280.27-1 Amendment No. 9, (3/82)
SL2-FSAR In addition to the above all power and control circuits are. provided with an interrupting device (ie fuses,, circuit breakers)
- Low level instrumentation cables do not require interrupting devices because the energy released because of an electrical fault is low and within the cables continuous energy-carrying capability.
Based on the above failure of an associated cable due to hot shorts, open circuits or shorts to ground will not prevent operation or cause mal-operation of redundant trains of systems necessary to achieve and maintain cold shutdown. 280.27-2 Amendment No. 9, (3/82)
.SL2-FSAR
*question No.
280.28 Provide one of the following for each* of the circuits identified in item 280.27c above: a) The results Qf an analysis that demonstrate that failure caused by open, grqund*, or hot short of cables will not affect i.t 's associ~ted shutdown sys.tem, *Note*
~
b) Identify each circuit requiring, a solution in accordance with Section III.G.3 of Appendix R, or c) Identify eac~ circuit meeting or tha.t will be modified to meet the requiremb'nts of Section III.G.2 of Appendix R (i.e., three-hour wall, 20 feet of clear space with automatic fire suppression, or one-hour barrier with automatic fire suppression) *
Response
See response to Question 280.27. I
"*l:i:-
- 280.28-1 Amendment No .. 9, (3/82)
't '
.SL2-FSAR *Question No~ .*
- 280. 29 To assure compliance with GDC 19, we require the following information be provided for the control room. If credit is to be taken for an alternate or dedicated shutdown method for other fire areas (as identified by item 280~26e or 280.28b above) in accordance with Section III.G. 3 of new ~ppendix R to 10CFR Part 50, the following information will also be required for each of these plant areas.
a) A tabl~ that lists all equipment including instrumentation and vital support system *equipment that are required by the primary method of achieving and maintaining hot and/or cold shutdown. b) A table that lists all equipment including instrumentation and vital support system equipment that are require_d by the alternate, dedicated, or remote metho~ of achieving and maintaining hot and/or cold shutdown.
- c) Identify each alternate shutdown equipment listed in item 280.29b, above, with essential cables (instrumentation, control and power) that are located in the fire area containing the 'primary shutdown equipment. For each equipment listed provide one of the following:
- (1) Detailed electrical schematic drawings that show the essential cables that are duplicated elsewhere and are electrically isolated from the subject fire areas, or (2) The results of an analysis that demonstrates that failure (open, ground, or hot short) of each cable identified will not affect the capability to achieve and maintain hot or cold shutdown.
d) Provide a table that lists Class IE and Non-Class IE cables that are associated with the alternate, dedicated, or remote method of shutdown. *For each item listed, identify each associated cable located in the fire area containing the primary shutdowu equipment. For each cable so identified provide the results of an analysis that demonstrates that failure (open, ground, or hot short) of the associated cable will not adversely affect the alterriate, dedicated, or remote method of shutdown *
- 280.29-1 Amendment No. 9, (3/82)
.SL2-FSAR
Response
The Hot Shutdown Control Panel (HSCP) provides a means of shutting the plant down independently. of the Control Room. The HSCP, which is fully described in FSAR Subsection 7.4.1.5, is located ih the B-Switchgear Room (fire area 34) on the 43'elevation of the Reactor Auxiliary Building. The HSCP would be utilized as the alternate means of plant shutdown in the unlikely event of a serious fire "in the Control Room or the Cable Spreading Room (fire areas 42 and 52 respectively) *. All cables required for the operation of the instrumentation and the controls of the HSCP (see FSAR Table 7.4-2) are listed in Attachments 3 and 4 and are shown on the drawings in Attachment 6. As indi cate"d in Attachment 1 of letter L-82-20 dated January 19, 1982 those cables and components required for plant shutdoWn from the HSCP that are located in FA-52 will be either isolated or relocated
- 280.29-2 Amendment No. 9, (3/82)
SL2-FSAR Question No *
- 280. 30 The residual heat removal system is generally a low pressure system that interfaces with the high pressure primary coolant system. To preclude a LOCA through this interface, we require compliance with the recommendations of Branch Technical Position RSB 5-1. Thus, this interface most likely consists of two redundant and independent motor operated valves with diverse interlocks in accordance with Branch Technical Position ICSB 3.
These two motor operated valves and their associated cable may be subject to a single fire hazard.* It is our concern that this single fire could cause the two valves to open resulting in a fire-initiated LOCA through the subject high-low pressure system interface. To assure that this interface and other high-low pressure interfaces are adequately protected from the effects of a single fire, we require the following information. a)* Identify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves} to isolate or preclude rupture of any primary
.coolant boundary *
- b) Identify each device's essential cabling (power and control) and describe the cable routing (by fire area) from source to termination.
c) Identify each location where the identified cables are separated by less than a wall having a three-hour fire rating from cables for the redundant device.
Response
The following Reactor Coolant System high-low pressure interfaces rely on redundant electrically controlled devices to maintain primary system integrity:
- 1) Shutdown C.Ooling Isolation (Figure 280.30-1)
- 2) Power Operated Relief System (Figure 280.30-2)
- 3) Letdown Isolation (Figure 280.30-3)
- 4) Primary Sampling (Figure 280.30-4)
- 5) Reactor C.Oolant Gas Vent System (Figure 280.30-5)
The Reactor Coolant System is isolated from the Low Pressure Safety Injection System by means of redundant locked-closed motor operated valves. Valves V-3481 and 3652 are powered by the SA-Train while valves V-3480 and 3651 are powered by the SB-Train. Each of these valves is controlled by independent interlocks which prevent valve opening when RCS pressure is 275 psig or above. The cross-tie valve (V-3545) is powered by the SAB-Train. In addition to *these pressure interlocks Low Pressure Safety Injection is protected by relief valves (see FSAR Subsection 6.3.2.2.6.1) to prevent damage to this essential system. All cables essential to the proper operation of these valves are listed in Attachments 3 and 4 and are shown on the 280.30-1 Amendment No. 9, (3/82)
SL2-FSAR
Response
Cont Id) drawings that form a part of Attachment 6 of letter L-82-20. In all cases where the cables for the redundant isolation valves are located in the same fire area one train of isolation valves will be protected within that area or relocated. This eliminates the possibility of a fire induced LOCA. The pressurizer is provided with redundant Power Operated Relief Valves. These are solenoid operated valves which are designed to fail closed on loss of power to preveni: uncontrolled steam dump. For additional protection each relief valve has a motor operated valves located upstream for isolation.
- For diversity the relief valves are de powered while the isolation yalves are ac powered.
In all cµses where the cables for a relief valve and its isolation valve appear in the same fire area (see Attachments 3, 4 and 6 of letter L-82-20) the relief valve cables will be protected within that area or relocated. This eliminates the possibility of a fire induced LOCA. The Letdown Line is provided with four pneumatic valves in series each of which is capable of isolating the Reactor Coolant System. All valves fail closed on loss of air or power. An analysis of the routing of all cables essential the operation of these valves (see Attachments 3, 4 and 6 of letter L-82-20), demonstrates that no one fire can prevent Letdown Line isolation. The Reactor Coolant System is provided with three independent sampling paths. Each line has redundant pneumatic valves which fail closed on loss of air or power. The valves located inside containment are powered from the A d-c bus while those outside are powered from the B d-c bus. In addition, each line has a restriction orifice designed to limit flow to less than the makeup capability of one charging pump in the event of any failure in the downstream line. Thus, these passive devices eliminate the possibility of a LOCA through these lines. In compliance with the positions outlined in NUREG Guide 0660 a Reactor Coolant Gas Vent System is being provided. Each vent path is provided with redundant solenoid operated valves which fail closed on loss of power. Both vent lines have restriction orifices which are designed to limit flow to less than the makeup capability of one charging pump in the event of any failure in the downstream lines. Thus, as stated for the sample lines, these passive devices eliminate the possibility of a LOCA through these vent paths. 280.30-2 Amendment No. 9, (3/82)
- SL2-FSAR TABLE 280. 30-1 FIRE HAZARD ANALYSIS RCS HIGH/LOW PRESSURE INTERFACE FIRE FIRE VALVE AREA POWER SOURCE AREA V-1460 14 125V de B 34 (RCS Vent Iso Valve)
V-1461 14 125V de A 34 (RCS Vent Iso Valve) v-1462 14 125V de A 34 (RCS Vent Iso Valve) V-1463 14 125V de B 34 (RCS Vent Iso Valve) v-1464 14 125V de B 34 (RCS Vent Iso Valve)
~
00 0 V-1465 14 125V de A 34 I.I> (RCS Vent Iso Valve) 0 I I.I> V-1466 14 125V de B 34 (RCS Vent Iso Valve) V-2515 14 125V de Bus 2B 34 (RCS Vent Iso Valve) V-2516 14 125V de Bus 2A 34 (RCS Vent Iso Valve) V-2522 24 125V de.Bus 2B 34 (RCS Vent Iso Valve)
- Non-safety alternate to LCV-2UOP & 2110Q
- 24 120V ae PP-208 51 letdown Isolation Valves. (letdown Control) r V-1474 14 125V de Bus 2B 34 t
rt z (RJRV) 0
- SL2-FSAR TABLE 280.30-1 (Cont'd)
FIRE FIRE VALVE AREA POWER SOURCE AREA V-14 75 14 125V de Bus 2A 34 (PORV) V-14 76 14 MCC 2A5 37 (PORV Iso Valve) V-1477 14 HCC 2B5 34 (POF.V Iso Valve) V-5200 through 5202 14 120V de Bus 2A 34 (Primary Sample Iso) V-5203 through 5205 24 125V de Bus 2B 34 (Primary Sample Iso) N> 00 ? w 0 I .i:-
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- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SHUTDOWN COOLING ISOLATION FIGURE 280.30-1
V-1474 FC V-1476 V-1475 FC V-1477 QUENCH TANK PRESSURIZER AMENDMENT NO. 9 (3/82) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 POWER OPERATED RELi EF VAL YES FIGURE 280.30-2
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