ML20211G075
| ML20211G075 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 02/05/1987 |
| From: | Cheng T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8702250271 | |
| Download: ML20211G075 (157) | |
Text
, . February 5, 1987 Docket No.: 50-029 LICENSEE: Yankee Atomic Electric Company (YAEC)
FACILITY: Yankee Nuclear Power Station (Yankee Plant), Rowe, Mass.
SUBJECT:
MEETING
SUMMARY
ON SYSTEMATIC EVALUATION PROGRAM (SEP)
SEISMIC REVIEW (SEP TOPIC III-6)
Reference:
Systematic Evaluation Program Evaluation of Yankee Seismic Analysis (NUREG-0825, Section 4-11)
A meeting was held on January 6 through January 9, 1987 between members of the NRC staff, YAEC personnel, and their respective consultants at the Walnut Creek, California office of CYGNA Corporation (Consultant of YAEC). This meeting is the last technical review meeting held to resolve the Yankee Seismic evaluation of SEP Topic III-6. Any issues which could not be resolved will be considered as an open item in the staff's safety evaluation report (SER). A list of attendee's is contained in Enclosure A.
Enclosure B contains copies of handouts presented by YAEC and NRC consult-ants. Enclosure C summarizes the discussions and conclusions for the safety concerns listed in the question matrix. In Enclosure D, issues not incor-porated in the question matrix are discussed.
(original signed by)
Thomas M. Cheng Integrated Safety Assessment Project Directorate Division of PWR Licensing-B
Enclosures:
DISTRIBUTION As stated Central File NRC PDR Local PDR ISAP Reading File J. Partlow C. Grimes T. Cheng OELD C. Jordan B. Grimes ACRS (10)
NRC Participants See previous concurrence sheet
- ISAPD PWRA:PD1 D:ISAP TCheng:cm
- EMcKenna
Docket No.: 50-029 LICENSEE: Yankee Atomic Electric Company (YAEC)
FACILITY: Yankee Nuclear Power Station (Yankee Plant), Rowe, Mass.
SUBJECT:
MEETING
SUMMARY
ON SYSTEMATIC EVALUATION PROGRAM (SEP)
SEISMIC REVIEW (SEP TOPIC III-6)
Reference:
Systematic Evaluation Program Evaluation of Yankee Seismic Analysis (NUREG-0825, Section 4-11) l A meeting was held on January 6 through January 9, 1987 between members of the NRC staff, YAEC personnel, and their respective consultants at the Walnut Creek, California office of CYGNA Corporation (Consultant of YAEC). This meeting is the last technical review meeting held to resolve the Yankee Seismic evaluation of SEP Topic III-6. Any issues which could not be resolved will be considered as an open item in the staff's safety evaluation report (SER). A list of attendee's is contained in Enclosure A.
Enclosure B contains copies of handouts presented by YAEC and NRC consult-ants. Enclosure C summarizes the discussions and conclusions for the safety concerns listed in the question matrix. In Enclosure D, issues not incor-porated in the question matrix were discussed.
Thomas M. Cheng Integrated Safety Assessment Project Directorate Division of PWR Licensing-B
Enclosures:
DISTRIBUTION As stated Central File NRC PDR Local PDR ISAP Reading File J. Partlow C. Grimes T. Cheng OELD E. Jordan B. Grimes ACRS (10)
NRC Participants ISAPDh PWRAtFDif D:ISAPD CGrimes TCheng:cm EMcKenna-02/3 /87 02/ f//87 02/ /87
8 'o,, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION O
% :* WASHINGTON, D. C. 20666 February 5, 1987 s...../
Docket No.: 50-029 LICENSEE: Yankee Atomic Electric Company (YAEC)
FACILITY: Yankee Nuclear Power Station (Yankee Plant), Rowe, Mass. d
SUBJECT:
MEETING
SUMMARY
ON SYSTEMATIC EVALUATION PROGRAM (SEP)
SEISMIC REVIEW (SEP TOPIC III-6) -
Reference:
Systematic Evaluation Program Evaluation of Yankee Seismic Analysis (NUREG-0825, Section 4-11)
A meeting was held on January 6 through January 9, 1987 between members of the NRC staff, YAEC personnel, and their respective consultants at the Walnut Creek, California office of CYGNA Corporation (Consultant of YAEC). This meeting is the last technical review meeting held to resolve the Yankee Seismic evaluation of SEP Topic III-6. Any issues which could not be resolved will be considered as an open item in the staff's safety evaluation report (SER). A list of attendee's is contained in Enclosure A.
Enclosure B contains copies of handouts presented by*YAEC and NRC consult-ants. Enclosure C summarizes the discussions and conclusions for the safety concerns listed in the question matrix. In Enclosure D, issues not incor-porated in the question matrix are discu sed.
[4toir. I-homas ,M. Cheng Integrated Safety ssessment Project Directorate
, Division of PWR Licensing-B
Enclosures:
As stated
W
']
w ENCLOSURE A ATTENDEES <
NAME AFFILIATION BILGIN ATALAY CYGNA CORP.
THOMAS CHENG NUCLEAR REGULATORY COMMISSION (NRC)
JOHN 0. HASELTINE YANKEE ATOMIC ELECTRIC CO. (YAEC)
BRUCE HOLMGREN YAEC DONALD LEFRANC0IS YAEC
~
DARLENE K. LEONG CYGNA CORP.
GARY L. MOCHIZUKI NCT ENGINEERING, INC. (NCT)
MARK RUSSELL IDAHO NATIONAL ENGINEERING LABORATORY MICHAEL SCHULMAN CYGNA CORP.
. LONG C. SHIEH LLL LARRY E. SHIPLEY CYGNA CORP.
TOM TSAI NCT T. Y. WANG CYGNA CORP.
PEPE VALLENAS CYGNA CORP.
NANCY H. WILLIAMS CYGNA CORP.
EILEEN MCKENNA NRC DOMINICK FUCITO YAEC 1
I I
A-1 l
l
-y , -
t ENCLOSURE 8 MEETING HANDOUTS o
YANKEE ATOMIC ELECTRIC COMPANY U.S. NUCLEAR REGULATORY COMMISSION CYONA ENERGY SERVICES SEP TOPIC III-6 MEETING AGENDA JANUARY 6 - 9, 1987 WALNUT CREEK, CALIFORNIA s
TUESDAY. JANUARY 6 0 PAB CONFIRMATORY ANALYSIS QUESTIONS: A1, A2, A3, B4, C5A, CSF, CSM, C7, C88, 117, I2A 0 ULTIMATE BEARING CAPACITY OUESTIONS: 81, B1A C1C 0 MISCELLANEOUS STRUCTURAL ISSUES CUESTIONS: 817, C5CI, H12B, H12C, I1DI O NON-SAFE SHUTDOWN STRUCTURES QUESTIONS: C8A, C8C, C80 0 FIRE TANK CONFIRMATORY ANALYSIS ,
OUESTIONS: C6A, C6B, C6C, C6E, C6F WEDNESDAY. JANUARY 7 0 CEDIC ATED SAFE SHUTDOWN PIPING CONFIRM ATORY ANALYSIS OUESTIONS: D21, DEC, G5 0 EXAMPLE ANALYSES OF PIPING OUESTIONS: D27 D38 YAN: JANAGENDA
o MISCELLANEO.US PIPING ISSUES QUESTIONS: D2.1, D2.2 D20A, D20s, D20C. D260. D26G.
D28, D321 D34 0 MISCELLANEOUS COMPONENTS / EQUIPMENT ISSUES QUESTIONS: B16. D24. D24A, D248, 024C. E23. E24, F38,
. F4F, G8, G9III, HS THURSDAY. JANUARY 8 2
0 PRESSURIZER CONFIRMATORY ANALYSIS OUESTIONS: F3C, F5D 0 AUDITS OR INFORMAL MEETING TIME O PIPING /EOUIPMENT SECTIONS OF DC-1, REV. 3 0 STRUCTUR AL SECTION OF DC-1, REV. 3 FRIDAY. JANUARY 9 0 REVI'EW OF QUESTION MATRIX 0 REVIEW OF NRC'S NOVEMBER MEETING
SUMMARY
0 FORMAL SUBMITTAL OF DOCUMENTS LIST ,
YAN: JANAGENDA
h$'s $b[
SOIL STRUCTURE INTERACTION EFFECTS IN TURBINE BUILDING
SUMMARY
The potential effect of soil structure interaction effects in the Turbine Building at the Yankee Nuclear Power Station is estimated.
The assessment is made on the basis of dynamic characteristics of the Turbine R
Building and the underlying soils, and on information gathered in the analysis of the PAB.
~
The information reviewed indicates that expected soil structure interaction effects for the Turbine Building are not significant and will have no detrimental effect on the Safe Shutdown system.
REVIEW OF PAB ANALYSIS AND COMPARISON TO TURBINE BUILDING Differences in the dynamic characteristics of the PAB and the Turbine Building' and differences in the soil conditions under both buildings result in the Turbine Building being less sensitive to SSI effects.
The analysis of the PAB indicates that the effect of soil structure interaction is most significant in the frequency range above 7Hz. In this frequency range, ground spectra accelerations increase as the frequency decreases. The PAB structure has many significant modes in that frequency range.
On the other hand, the first five significant modes of the Turbine Building fall below 3.75H z. It is clear from this that the dynamic characteristics of the Turbine Building are different (sof ter) from the PAB, and the Turbine Building will not be affected by SSI to the degree the PAB appears to be.
The soil depth under the Turbine Building is 25% deeper than under the PAB. Deeper soil would result in higher radiation damping and a reduced response when considering SSI ef fects 1
LTR SOILmemo
ESTIMATED IMPACT OF SSI ON TURBINE BUILDING ANALYSIS -
An assessment of the effect of SSI on the Turbine Building is made using a simplistic analysis neglecting radiation damping.
In order to estimate the expected frequency shift and change in ARS, soil springs with no dampers are introduced under a simplified model of the Turbine Building (Figure 1). The simp!!fied model represents the modal masses and frequencies of the significant Turbine Building modes.
s a) Frequency Shif t
~
The results summarized in Table I show a shif t in frequency of less than 0.3%
for the first five horizontal modes.
For the sixth horizontal mode, the shif t is 5% and for the vertical modes of 18.6 Hz and 25.4Hz, the shif t is 2.3% and 9.1% respectively, b) Change in ARS Peak Values The change in ARS peak values is conservatively estimated neglecting radiation damping by reviewing the change in mode shapes and participation factors resulting because of soil structure interaction. The results on Table 2 show that the largest increase for horizontal modes is 7L The increase in vertical ARS for the 18.6Hz and 26.4Hz modes is 31% and 0.4% respectively.
It should be noted that the 18.6Hz mode corresponds to the vertical vibration of the Turbine Hall Roof. The roof has a large margin of safety for vertical loads. In addition, no Safe Shutdown equipment is supported nor potentially impacted by the Turbine Hall Roof.
The simplified procedure used for determining the increase in ARS does not account for the cases where a frequency shift results in a new frequency corresponding to a higher ground acceleration. This would be the case for modes 6 through 8. However, the frequency shif t in these modes is small, and a review of the ground response spectra shape shows an increase in acceleration values of at most 0.004g.
i l
i l
2 LTR: SOILmemo
CONCLUSION 7 The soil structure interaction effects for the Turbine Building coulci result in an increase in the horizontal ARS peaks of (at most) 7.4% This is not a large value.
In addition, it should be recognized that the original fixed base ARS were developed using artificial time histories with spectra that conservatively envelope the NRC spectra. If a procedure such as direct generation were to be used, it could possibly result in ARS about 20% lower. It should also be noted that the original NRC spectra have been recognized to be conservative (Memorandum from R.E. Jackson to W. Russell, April 5,1983).
The vertical ARS at the top of the Turbine Hall could increase by 31%; however, this value could be reduced significantly if refinements such as radiation damping and direct generation were to be introduced. As discussed previously, this possible increase would not jeopardize the overall Turbine Building structure or Safe Shutdown systems.
3 LTR: SdILmemo
r Figwe 1
~ '
Simplified Twbine Building Model 9
HORIZONTAL MASS o P 25 VERTICAL MASS 5o o
12 11 ll 90 99 99 9 / SOIL SPRINGS vnvr 99 e
a i
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P Table 1 ~
Frequency Shif t in Turbine ,.
Building Due to Soll Structure Interaction />
/
C .t .
- t t
Frequency (Hz) ~
Mode Fixed SSI No. Direction Base Model Shif t (%)
1 X (N-5) 2.57 2.569 0.04 '
2 Y (E-W) 2.57 2.569 0.04 3 Y 2.74 2.733 0.3 4 X 2.74 2.740 0 ,
5 Y 3.75 3.738 0.3 6 X 12.54 11.915 5.0 7 Z (vert) 18.60 18.178 2.3 8 2 26.40 24.009 9.1 i
5 LTR: SOILmemo
- f. ,. l 4C + / __
,. I [.
S
~
Table 2 .
Increase in ARS Peaks in Turbine Building Because of Soil Structure Interaction J-f
-c Mode (Participation Factor) * (Mode Shape) 2 No. Direction SSI/ Fixed-Base Increase in ARS
.~ -
~ '
7 1 X (N-S) l.009 +0.9 %
2 Y (E-W) 1.052 +5.2%
3 Y 1.005 +0.5%
4 X 0.999 -0.1%
5 Y 0.996 -0.4%
6 X 1.074 +7.4%
- , 7 Z (vert) 1.314 +31.4%
8 Z l.004 +0.4%
i 6
LTR: SOILmemo
\
ITEM CSC I ISSUE
SUMMARY
- VERIFICATION OF BATS COMPUTER PROGRAM RESOLUTION:
AT THE REQUEST OF THE NRC CONSULTANTS TWO ADDITIONAL VERIFICATION PROBLEMS WERE RUN FOR THE BATS COMPUTER PROGRAM. THE FOLLOWING IS A
SUMMARY
OF THE TWO ADDITIONAL VERIFICATION PROBLEMS:
e (1) IN AN EARLIER AUDIT. THE NRC CONSULTANTS HAD NOTED A DISCREPANCY BETWEEN BATS AND ETABS RESULTS FOR A SPECIFIC PROBLEM. THAT SAME PROBLEM WAS RERUN USING ANSYS. THE ANSYS RESULTS AGREE WITH BATS RESULTS.
(2) THE NRC CONSULTANTS ~ REQUESTED A VERIFICATION OF THE BATS FLEXIBLE LINK ELEMENT. A SAMPLE PROBLEM WAS CREATED AND THE PROBLEM WAS SOLVED BATS USING BATS AND SAPIV COMPUTER PROGRAMS.
AND SAPIV RESULTS AGREE.
THE CALCULATIONS AND RESULTS WERE AUDITED BY NRC CONSULTANTS ON DECEMBER 17, 19S6 AND WERE FOUND ACCEPIABLE.
CONCLUSION: THE BATS COMPUTER PROGRA!! IS A.'t.P71iLE.
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l DSSPmemwssmnmxm .. Q RESULTS OF THE FIRE WATER TAlWC EVALUATION SHOW THAT ALL ITEMS SEEM TO BE SATISFACTO EXCEPT STEEL SHELL BUCKLING l
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ITEMS
.LOM20. ALLOWABLE APPLICABLE. CODE i
TANK SLIDING 770 KIPS 20,600 K-FT 1256 KIPS [ = 0.4 TANK OVERTURNING 69,100 K-FT l SOIL BEARING 2.9 KSF 10.6 KSF
, RING FOOTING l SHEAR 13 PSI 93 PSI ACI 318 i MOMENT 44K-IN 114 K-IN SHELL BUCKLING AXIAL EOMPRESSION ONLY 2.2 KSI 4.3 KSI CODE CASE N-284 SHEAR ONLY 3.5 3.2 1
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M STRUCTURAL FREQUENCIES OF THE NORTH WALL MODE FREQUENCY DIRECTION 1 5.28 VERT (STEEL FRAME) 2 5.63 EW (STEEL FRAME) 3 11.26 NS (BLOCK WALL) 4 15.26 NS (BLOCK WALL) 5 39.50 VERT 6 63.50 EW 1
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g,nSSP m.= svsr.s s-w moc .. 6' IN THE EVENT OF 0.2G EARTHOUAKE, RESULTS OF THE SEISMIC EVALUATION OF THE NORTH WALL i SHOW THAT 1
O BLOCKWALL WILL DEVELOP CRACKING l
0 ALL OTHER STRUCTURAL MEMBERS, CONNECTIONS WILL REMAIN FUKTIONAL (CONNECTIONS REMAIN ELASTIC)
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ITEM C7 AND CBB ISSUE
SUMMARY
ITEM C7: PERFORM CONFIRMATORY ANALYSIS OF A MASONRY WALL USING NRC SPECTRA AND SEP CRITERIA.
ITEM C8s: INVESTIGATE THE POTENTIAL EFFECT OF COLLAPSE OF THE PAB NORTH WALL.
REFERENCES:
(1) YAEC DWGS. 9G99-FA-1SBA, FC-40AA. FS-19D AND FS-190A (2) CYGNA CALCULATION 860G4/G-F, SET F.
(3) "SEISHIC REEVALUATION AND RETROFIT CRITERIA FOR YNPS".
D00. No. DC-1, REV. 3, CYGNA.
RESPONSE
THE CONFIRMATORY ANALYSIS OF THE PAB NORTH WALL WITH THE MODIFICATIONS AS SHOWN IN REF. (1) WERE PERFORMED. THE 8" CONCRETE BLOCKWALLS (P2F1 & P2F2) IN THE SECOND STORY AND THE 12" CONCRETE BLOCKWALLS (P1Ei & PIE 2) IN THE FIRST STORY WERE ANALYZED SEPAdATELY. THESE WALLS WERE ANALYZED FOR DEAD LO,AD PLUS NRC LOADS. THE NRC LOADS ARE THE ENVELOPED N-S AND VERTICAL ARS (7% DAMPIN3) LOCATED AT THE COLUMN LINES G. 7 AND S OF THE PAB ROOF. THESE ARS WERE GENERATED BY THE PAB
SSI CONFIRMATORY ANALYSIS.
MODELING AND EVALUATION OF THE BLOCKUALLS ARE DISCUSSED AS FOLLOWS.
(I) 8" CONCRETE BLOCKWALLs (P2F1 & P2F2) t w-
1
~
THE MATHEMATICAL MODEL OF THE 8" CONCRETE BLOCKWALLS AND THEIR SUPPORT FRAME IS SHOWN IN THE FOLLOWING FIGURES.
,DUE TO LARGE N-S RELATIVE DISPLACEMENT BETWEEN THE PAB ROOF AND THE FIRST FLOOR HORIZONTAL CRACKS BETWEEN ELEMENT LAYERS ACROSS THE ENTIRE LENGTH OF THE WALL WERE ASSUMED. ,
THE FREQUENCY AND PARTICIPATION FACTOR OF THE SIGNIFICANT MODES ARE SHOWN IN THE FOLLOWING TABLE. FOR h THE MODES CONSIDERED IN THE ANALYSIS ONLY A SMALL PORTION OF THE VERTICAL MASS IS INCLUDED. THEREFORE, -
THE EFFECT OF THE VERTICAL SEISMIC LOADS WERE CONSIDERED BY SUBTRACTING THE ZPA 0F THE VERTICAL ARS (0.23 G) FROM THE GRAVITY LOAD.
AS SHOWN IN THE FOLLOWING TABLES. THE STRESS RATIOS OF THE SUPPORT FRAME AND THE CONNECTION HARDWARES ARE LESS THAN THE ALLOWABLES PER REF. (3). HOWEVER. THE MASONRY STRESSES EXCEED THE ALLOWABLE BY A MAXIMUM OF 78%. THE OVERSTRESS IN THE MASONRY CAN BE ATTRIBUTED TO THE CONSERVATISM OF THE ENVELOPED RESPONSE SPECTRUM. AT THE FIRST FREQUENCY. 9.55 HZ. THE N-S ARS AT COLUMN LINE 8 HAS AN ACCELERATION OF 1.79 G. THE ARS AT COLUMN LINES 6 AND 7. HOWEVER. HAVE THE ACCELERATION OF 0.59 G AND 0.71 G. RESPECTIVELY. THE FIRST FLOOR ARS HAS A MAXI,M,UM ACCELERATION OF 0.54 G. THEREFORE. THE INPUT SPECTRUM IS TOO CONSERVATIVE FOR THE MASONRY AWAY FROM COLUMN LINE 8 0F THE PAB ROOF. SINCE ALL THE OVERSTRESSED MASONRY ELEMENT IS AT LEAST 10 FEET FROM COLUMN LINE 8 0F THE PAB ROOF. IT CAN BE CONCLUDED THAT REAL MASONRY STRESSES UNDER NRC LOADS ARE MUCH SMALLER THAN THOSE SHOWN IN THE FOLLOWING TABLE AND THE INTEGRITY OF THE WALL WILL NOT BE JEOPARDIZED.
5
MODAL CHARACTERISTICS SECOND STORY NORTH WALL (P2F1 & P2F2)
M005 FREQUENCY PARTICIPATION FACTORS No. (Hz) Y (N-S) Z (VERTICAL) i 9.55 8.716 0.017 2 12.05 1.363 0.024 3 13.98 0.080 0.040 1 4 19.46 1.730 0.087 5 22.23 3.861 0.015 G 23.78 1.650 0.052 7 25.19 0.408 0.052 i
STRESSES OF 8" CONCRETE BLOCKWALL ELEM. STRESS PERPENDICUL AR TO BED JOINT (PSI) STRESS
- h SRSS or Y&2 NRC DL-VERT. NRC Slui RATIO 10 48.0 -15.8 32.2 1.33 11 48.8 -15.8 33.0 1.36
' 49 32.7 -11.4 21.3 0.88 1 50 32.6 -11.4 21.2 0.88 69 36.8 -9.6 27.2 1.12 70 38.1 -9.6 28.5 1.18 89 42.1 -9.6 32.5 1.34 ,
90 42.6 -9.6 33.0 1.36 109 34.4 -5.9 28.5 1.18 110 32.4 -5.9 26.5 1.10 130 20.9 -3.9 17.0 0.70 153 44.8 -1.7 43.1 1.78 154 43.9 -1.7 42.2 1,74
' ALLOWABLE STRESS = 0.65,/Ho 24.2 PSI 4
1
STRESS OF 8" CONCRETE BLOCKWALL ELEM. STRESS PARALLEL TO BED JOINT (PSI) STRESS
- No. SRSS OF Y&Z NRC RATIO 28 62.0 1.11 30 78.1 1.40 31 67.7 1.22 4 48 56.8 1.02 52 57.6 1.03 56 74.6 1.34 62 57.3 1.03 63 57.4 1.03 76 61.0 1.10 80 60.6 1.09 81 72.3 1.30 88 64.7 1.16 89 67.7 1.22 92 81.3 1.46 96 65.6 1.18 108 67.0 1.20 109 61.6 1.11
- ALLOWABLE STRESS = 1.5 6 55.7 PSI.
6
EVALUATION OF SECOND STORY NORTH WALL SUPPORT FRAME AND CONNECTION HARDWARES MAX. STRESS RATIO STRESS DESCRIPTION AT ELEMENT / NODE No. RATIQ EXISTING W8X31 E308 0.36' NEW W8X28 COLUMN E170 0.37 NEW W8X28 BEAM E213 0.31 s
NEW W8X25 E211 0.15 NEW L3X3X3/8 BRACE E236 0.09 -
NEW C8X13.75 E286 0.35 VERTICAL RIB NEW C6X13 E253 0.41 EXIST. STR AP ANCHOR E307, 308 0.72 TA = 6000 LBS/ ANCHOR)
NEW 5/8" ANCHOR R N267, 507 0.61 (TA = 3256 LBS/ ROD)OD NEW 1/2" HILTI N108 0.91 fk! Bf0 NB!/HSA)
NEW 3/4" HILTI N31 0.68 KWIK-BOLT DETAIL H 9" EMBEDMENT NEW 3/4" HILTI N160, 280 0.33 ..
KWIK-BOLT DETAIL G, 5" EMBEDMENT
- INCLUDES THE STRESSES OBTAINED FROM THE PAB BUILDING ANALYSIS 4
MATHEMATICAL MODEL OF 8" WALL AND SUPPORT FRAME OVERALL VIEW s %
%>sN '
T
/ [t 's., l J's 5
@s N '
< <. a ..
. ~
s.
s 4 > '> k1 ayd'x ss m N s ,'s
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\
N >N'N N N s\
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), s%s N s
'N, N \s k N s \ ,
N k N'N s N
sN Ns PA0 NOUM W8LL P!'l & P!'I = N*C it tuvtttP!D AAS 6
MATHEMATICAL MODEL OF 8" WALL AND SUPPORT FRAME (CONT'D)
ELEMENT NUMBERS .
M "
,, L3X3X3/8 (NEW)
- ! 'Ib l61 N S b tl I?1 ,
/ ' M*
g
/ I /t1
/ ut
, gg 66 inu L2 ,
. x ... ,s, r ,n it, )
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177 lL 8X28 a to
" j W(NEW) t ra N9 .,
176
/
lug
/
S L 7*,
St P90 WORTM WALL P2ft 4 P2F2
- NRC 72 ENVfLOPED ARS s
i h
MATHEMATICAL MODEL OF 8" WALL Afl0 SUPPORT FRAME (CONT'0)
ELEMENT NUMBERS 8" CONCRETE BLOCKWALL ASSUMED CRACK b
l I I io li iu tu i .. is i.,in i.,i. , n.,n , na
@ @ as i ., a n, na 124 12 . 123 123 12' 1 21 12b 127 121 I?l 131 13 ; 13d 13 1 13'l IPi 1 31 13' 1 31 13'
=
son tu seu t it.1 ini 10<, in% ill iQih(('. i t i it. ' it i sti l ti<, sti tt 11e sti g g .2 6 '5
U (: 8 'i (" " " 'S @" "
60 61 62 63 6'l 65 66 67 60
@@ ?! 12 73 7's 75 h 17 78 71 41 42 93 49 15 46 41 99 99 50 41 52 53 lt 55 h 57 St 59 pt 77 79 Pu 79 Pn F7 g te 'g 'g it 31 De 19 % 37 94 99 en 1 2 3 4 3 6 7 8 9
@ { 12 83 14 15 16 17 IS 19 20 J
Ppl N0mfM WALL P2FI 4 P2F2 - temt it ENytLOPt0 mmt 6
l 1
MATHEMATICAL MODEL OF 8" WALL AND SUPPORT FRAME (c0NT'0)
NODE NUMBERS 8" CONCRETE BLOCKWALL 6 7 10 8 9
s .
l I I n
,u i sus ,su s seu -ius sua sus ium sue tsn 'ici ses ,ici seu , sce its act man ice sin 161 -
i , i, i it, it s i.e its ite i r r, its its nie i3m ist i33 m i3. i3e iss. ist i3a ise i.e i3g
.7
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- ist is, m iss ist isa ist ime ise inn ist int inn in n na t ans, ist man ime inn aq
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, tat its ist tan ist ika tat ita ime itn ist its its itu itt itk .ist its te tan igg
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-3 HINGED BOUNDARY CONDITION '
ASSUMED CRACK (TYP.)
i i
PA0 NORtM WALL P2Ft & P2F2 - NRC 71 (NvtLOP(0 ARS l 1 l
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w asxIc ts )IAa-( > f 111 3XIS1* M8XcT JWS N elle 4s1142J19 disI - MW2 2 3NA3*Od30 ehS
l (II) 12" CONCRETE BLOCKWALLS (P1El & PIE 2)
THE MATHEMATICAL MODEL OF THE 12" CONCRETE BLOCKWALLS AllD THEIR SUPPORT FR AME IS SHOWil IN THE FOLLOWIflG FIGURES.
THE N-S RELATIVE DISPLACEMENT BETWEEN THE PAB FIRST FLOOR AND GROUfl0 IS SMALLER THAN THAT BETWEEll THE ROOF AND THE FIRST FLOOR. THEREFORE. ONLY OllE HORIZONTAL CRACK ACROSS THE MID-HEIGHT OF THE WALL WAS ASSUMED.
- SINCE ONLY A SMALL PORTION OF THE VERTIC'L MASS CAN BE INCLUDED FOR THE MODES BELOW 50 zH . THE .ERTICAL RESPONSE SPECTRUM ANALYSIS WAS NOT PERFORi1ED. THE EFFECT OF THE VERTICAL SEISMIC LOADS WERE CONSIDERED BY SUBTRACTIllG THE ZPA 0F THE VERTICAL ARS (0.23G FOR THE SEC0llD STORY WALL AND 0.16G FOR THE FIRST STORY WALL)
FR0ll THE GRAVITY LOAD.
THE FREQUEllCY AND PARTICIPATION FACTORS OF THE SIGNIFICANT MODES AND EVALL'ATION OF THE STRUCTURE ARE SHOWil IN THE FOLL0tlING TABLES. ALL MASONRY STRESSES.
THE STRESS RATIOS OF THE SUPPORT FRAME. AND THE CONNECTI0fl HARDWARES ARE LESS THAN THE ALL0llABLES PER REF. (3). THEREFORE. THE 12" CONCRETE BLOCKWALLS ARE ACCEPTABLE. ..
6
c' J
. 1 MODAL CHARACTERISTICS FIRST STORY NORTH WALL (P1Ei & P1E2)
.. MOD E FREQUENCY PARTICIPATION 4 N~ o. (Hz) Y (N-S) 7 1 9.74 8.852
, 2 16.00 0.763 4 25.22 2.401 ,
G- 36.29 3.015 9 41.58 3.160 _
1 f
e e 9
- . - , . , , . - . . -- , , - - - . ~ , . . - . - . . _ . , - , , , - - - . . . , _ , . , _ . .
STRESSES OF 12" CONCRETE BLOCKWALL STRESS PERPENDICULAR TO BED JOINT (PSI)
ELEM.
NO. Y- NRC DL-VERT. NRC SUM ALLOllABLE 19 36.S -45.3 -8.7 20 31.1 -34.0 -2.9 2 34 44.6 -41.9 2.7 0.65 6o 37 32.2 -21.0 11.2 24.2 55 32.1 -19.2 12.0 73 20.0 -17.0 3.0 90 27.4 -15.1 12.3 91 30.9 -15.1 15.8 109 20.4 -13.4 7.0 STRESS PARALLEL TO BED JOINT (PSI)
ELEM.
No. Y-NRC ALLOWABLE 34 25.4 43 19.1 1.5 Eg 55.7 52 26.9 54 24.2 l 72 22.4 l
89 20.7
EVALUATION OF FIRST STORY NORTH WALL SUPPORT FRAME MAX. STRESS RATIO DESCRIPTION AT ELEMENT / NODE No. STRESS RATIO EXISTING W8x31 ELEM. 199 0.26' NEW W8X28 COLUl1N ELEN. 144 0.13 NEW W8X28 BEAM ELEM. 161 0.03
~
NEW C8X13.75 ELEM. 190 0.04 NEW C6X13 ELEM. 179 0.15 EXIST. STRAP ANCHOR NODE 171 0.40 (1/4" x 1 '/4") '
(TA = 6000 LBS/ ANCHOR)
NEW 5/8" ANCHOR R00 NODE 176 0.64 (TA = 3256 LBS/ ROD)
NEW 1/2" HILTI NODES 181 & 0.65 SLEEVE ANCHOR 281 (TA = 810 LBS/HSA)
INCLUDES THE STRESSES OBTAINED FROM THE PAB BUILDING ANALYSIS.
b e ,- - - --
MATHEMATICAL MODEL OF 12" WALL AND SUPPORT FRAME OVERALL VIEW 4
N N
Ns\ s\s\
N s N ss s N
- N N (x t s) g xs s '
N N N
> sN N ss ssN s sss N N
ss s 'x N's N
<xt)x_(>>
s N 2 :::3 x'x s N \9 N
hh s ms ( s NN s x\
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s ss N N 'N'N Ns N 'N N s N Ns
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N
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, m nearn nau.. nast stear - wa: n owinc ans in i-ninterian i
l l
MATHEMATICAL MODEL OF 12" WALL AND SUPPORT FRAME (CONT'D)
ELEMENT NUMBERS 12 l3b g s6 .
tse
< \ l?1
- l50
$7 ft 142 6 i:
,, A - hn its n .j; C6X13 C6X13 (NEW) lu (NEW) is 199 tem l15 35 l'#
7 W8X28 (NEW)
Le9 pag NORTH NALL. FIRST STORT - NRC 71 OmmPING RR$ IN T. DIRECTION i
l 6
- - - - - - - - - , . -- .- , , , - - - - - ~
MATHEMATICAL MODEL OF 12" WALL AND SUPPORT FRAME (CONT'D)
ELEMENT NUMBERS 12" CONCRETE BLOCKWALL 4
s:
1 y tta 119 IJti t Pr IP) IP l I Pi: IP! 1Pr IP1 IPA iM 13ll 13 IS ' sai 131 13%
100 IUt IL2 80s lut 104 lul i 104 tot lui !!U lit II<' II;l ll'i att IIt !!7 ut UJ um us 04 us ou cv wu MI 4d VJ 49 vs 96 vs 40 ww 64 65 64 67 60 69 70 7! 72 7J 79 73 76 77 70 79 80 08 F gn 47 m 49 50 Si sP sa sg sn as st sa so an at 67 na al 32 31 31 35 36 37 38 39 40 41 42 43 4t 45 16 17 to 14 Fu Pt PP Pg P4 P% Ph Pt PM P9 mi I 2 3 4 5 6 7 8 9 10 Il 12 13 15 IS LASSUMED CRACK PR9 NORTN WRLL. FIRST STORT - NRC 7% OfDtPINC Rfl5 IN T-0!MECTION i
l 9
4 1
I - - , .-- -.
~
MATHEMATICAL MODEL OF 12" WALL AND SUPPORT FRAME (CONT'D)
NODE NUMBERS, 12" CONCRETE BLOCKWALL T
8
@ @ o I
, -,,, ...,,,, , t ,m ,,, ,.a ,. , ,,n I,,,
,,, ,,, ., , ,,, I ,,, .,,, , , .
l,,,
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.lSt 159 iStigg l91 iSa i9e ttn 191 I49 rtt itu 94 it. It? its i}q in' ins int ina .i n<t . tin tii .iis int i,n t,e ii. let .tta .tsq inginq ,
4 ASSUMED CRACK INGED BOUNDARY CONDITION (TYP.)
PR8 NORTM NRLL. FIRST STORY - NRC 71 DRMP!NG RMS IN 1-DIRECTION l
t b
. .- n- ._.,n.- - , - ,-- . - , _ , - ,-
MATHEMATICAL MODEL OF 12" WALL AND SUPPORT FRAME (CONT'D)
NODE NO. (TYP. )
7 h
Phi ELEMENT NO.(TYP.)
ISI h " " SOS fil
-3 na O ma O iat O ins S ist O igg jg3 3g
,g < g C8X13.75 (TYP.) ,, ,,
3 191 3
l54 130 lli EXIST. W8X31 PRO NO4fN N4LL. FIR $f STORY - SIRC 72 DAMP!NC RRS IN 1-DIRECTION 4
ITEM C8A ISSUE
SUMMARY
INVESTIGATE THE POTENTIAL COLLAPSE OF THE
~
ACCUMULATOR TANK.
REFERENCE:
CYGNA CALCULATION BINDER 86064/ 7F. CALC.
SET D. DATED NOVEMBER 1986.
?
RESPONSE
THE ACCUMULATOR TANK HAS BEEN EVALUATED IN THE REFERENCE. THE EVALUATION WAS PERFORMED USING A FINITE ELEMENT ANALYSIS. THE GEOMETRY OF THE TANK AND THE FINITE ELEMENT MODEL UTILIZED ARE SHOWN ON THE FOLLOWING PAGES.
r i
l l
I
ACCUMULATOR TANK r .
I*
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FINITE ELEMENT MODEL OF THE ACCUMULATOR TANK
~
u.. gg 20 l9 v LO ll l6 is l3 l ly O
L2 l
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RCCUMULATOR TANK 1
I i
LOAD COMBINATION:
ENVELOPE DL + P YCS
~
WHERE: P = HYDROSTATIC PRESSURE YCS GROUND SPECTRA ARE INPUT ALONG THREE DIRECTIONS.
FOR THE MORE CRITICAL ITEMS, YCS SEISMIC LOADS ARE MULTIPLIED BY TWO TO DEMONSTRATE ADEQUACY OF THE TANK UNDER NRC SPECTRUM 1 SEISMIC LOADS.
A DAMPING RATIO OF 3% IS USED.
THE TANK IS FULL OF WATER DURING NORMAL OPERATION.
i h
MODAL ANALYSIS RESULTS ARE AS FOLLOWS:
MODE FREo 2 /IN)
P ARTICIP ATION F ACTOR (LB-SEC No . (Hz) Lx Ly . LZ
- 1. 20.42 13.02 -- --
- 2. 20.42 --
13.02 --
- 3. 79.57 6.068 2.905 --
- 4. 79.57 2.905 6.068 --
- 5. 93.93 -- --
14.50 '
- 6. 152.55 0.3794 2.185 --
- 7. 152.55 2.185 0.3794 --
, 8. 189.70 0.1306 1.279 --
- 9. 189.70 1.279 0.1306 --
]~ L2
_ 221.35 221.35 210.25 NOTE: TOTAL MASS = 222.19 LBS-S'Ec2 f13, (O'86K) i e o
EVALUATION OF RESULTS:
- 1. ANCHOR BOLTS 3/4" O BOLTS
'" FORCE'S/ BOLT (LBS):
LOAD CASE AXIAL (+ = COMPR.) SHEAR DL 10732 0 YCS 9159 1983 DL + YCS 19891 1983 o DL - YCS 1572 1983 DL - 2YCS -7587 39GG UNDER DL-YCS:
BOLTS ARE NOT IN TENSION: 0.K.
[UNDER DL-2YCS i ( T DL-NRC)]
TENSION CAPACITYlBOLT = 8.84 K > MAX. TENSIONIBOLT =
7.59K.O.K.3
- 2. BUCKLING OF SKIRT MAX. COMPRESSIVE STRESS (a NODE 11) = 0.89 KSI < Sg =
13.9 KSI WHERE THE ALLOWABLE IS Sg PER ASME C00E PARA. NE-3133.G.
- 3. max. SKIRT TENSILE STRESS = 0.11 KSI: LOWS 0.K.
MAX. SKIRT SHEAR STRESS = 0.12 KSI: LOWS 0.K.
i l
- 4. TANK SHELL STRESSES (a NODE 14)
(A) COMPRESSIVE = 0.45 KSI: LOW 0 K.
(B) TENSILE =
0.12 KSI: LOW 0.K.
i
(C) SHEAR =
0.06 KSI: LOWS 0.K.
- 5. MAX. DISPLACEMENTS: HORIZONTAL: 0.0065 IN: SMALL: 0.K.
VERTICAL: 0.0012 IN: SMALL: 0.K.
- 6. SOIL BEARING STRESS (UNDER DL+2YCSl T DL+NRC)
=
3.46 KSF < 8 KSFi 0.K.
CONCLUSION: THE ACCUMULATOR TANK HEETS CODE ALLOWABLES UNDER THE POSTULATED LOADS AND WILL NOT COLLAPSE ON OTHER PLANT COMPONENTS.
't 4
ITEM C8c III ISSUE
SUMMARY
INVESTIGATE THE POTENTIAL COLLAPSE OF' THE ELEVATOR TOWER
REFERENCES:
[1] CYGNA C ALCULATION BINDER 86064/7F, CALC. SET C, DATED DECEMBER 1986.
s
[2] CYGNA C ALCUL ATION BINDER 83033/17F, CALC. SET E. DATED APRIL,1984.
RESPONSE
THE ELEVATOR TOWER HAS BEEN EVALUATED IN REFERENCE [1]. THE EVALUATION WAS PERFORMED USING A FINITE ELEMENT ANALYSIS. SEE THE FOLLOWING PAGES FOR THE GEOMETRY OF THE ELEVATOR TOWER AND THE FINITE ELEMENT MODEL UTILIZED.
- 9 l
l l
ELEVATOR TOWER
. .. +- *..- ,
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w BRACING OF ELEVATOR TOWER TO FUEL TRANSFER CRANE SUPPORT STRUCTURE - 6
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6
4 FINITE ELEMENT MODEL OF THE FUEL TRANSFER CRANE SUPPORT STRUCTURE AND ELEVATOR TOWER - s
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FINITE ELEMENT MODEL OF THE ELEVATOR TOWER
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E s[)
'j/
ELEVATOR TOWER t i
l LOAD COMBINATION ENVELOPE DL d: YCS
YCS ' GROUND SPECTRA ARE INPUT ALONG THREE DIRECTIONS'.
A DAMPING RATIO OF 5% IS USED FOR A STRUCTURE WITH RIVETED CONNECTIONS. EVALUATION OF RESULTS s
- 1. ANCHOR BOLTS AND BASE PLATES: LOAD CASE = DL - YCS ( d ) ._
BASE PL N MAX. PL. ANCHOR BOLTS FORCES (KIPS) COL. LINES STRESS. KSI TENSION SHEAR BOLT i BOLT 2 BOLTS 1&2 D1/2-2 3/4 !.56 1.59 1.46 2.24 D1/2-3 1/2 0.83 0 0 2.11 D1/2-3 3/4 1.23 1.10 1.07 2.29 D3/4-2 3/4 1.62 0 0 4.67 D3/4-3 1/2 '1.32 1.39 0.50 1.91 D3/4-3 3/4 0.86 0 0 0.06 MAX. PL. BENDING STRESS = 1.62 KSI < 0.6Fy = 21.6 KSI: 3, K2 , MAX. BOLT SHEAR STRESS = 5.95 KSI < Fy = 10.0 KSI: 0.K. MAX. BOLT TENSILE STRESS = 2.02 KSI < FT= 15.3 KSI: jld(u
- 2. COLUMNS HIGHEST STRESSED COLUMN IS AT COL. LINES D 3/4 3 2 3/4,
. ELEV. 1090'-5". ~
'~
~ AISC E0. (1.6 - 1A):
C " r b HY ' r
'A . MX F BX . IBY (1 }F gx (1 )F gy
. F 3 j{ j[g, F f EX EY s
, 6.38 . 0.85 (5.76) . 0.85 (9.77) 23.51 (1 _ 26/.92 )
6.38 33 (1 - 6.38 86.54) 29 - 0.73 < is 0.K. AISC Eo. (1.G - 18): p A . _ R1 . 'I BY , 6.38 . 5.76 9 77 g(0.6F) y gx F gy 26.4 33 9
=
0.75 < 1: 0.K. l l l l 1
, , , , , . - .- - . - - - , - - - - - . , . . - - - - - , - , . , - - , - . . . - --------------.--,.-,_a--
- - , . , - - - - . __m
ADEQUACY OF THE'SAME COLUMN ELEMENT UNDER NRC SPECTRUM SEISMIC LOADS (WITH NRC SPECTRUM SEISMIC LOADS = 2 X YCS LOADS): . u.. ,- e LOAD fA Mx i Mx 2 My i M2Y -
-8X hBY CASE KSI IN-K IN-K IN-K IN-K KSI KSI DL 0.964 4.26 18.56 4.63 19.16 0.677 2.071 YCS 5.415 17.18 139.51 12.86 71.20 5.087 7.697 s DL+2YCS 11.794 38.62 297.58 30.35 161.56 10.851 17.465 I X C3 -
[0.3( ) + 0.4 M + 0.33 1/2
- C l .6 KL/R = 41.54 < CC - 131.7 y FA 0.95 FCR - 29.791 KSI Fax - Fgy - 0.95Fy - 31.35 KSI THEN AISC Eo. (1.6 - 1A):
11.794 . 0.597 (10.851) . 0.621 (17.465) 29.791 (1 jgj 79 g] )(31.35) (1- 8 4
)(31.35)
= 1.01 T 1. : 0.K.
-= - - - - -
- 3. BEAMS HIGHEST STRESSED BEAM IS AT COL. LINE Di/2. ELEV.1112'-0" (g , = 13.3 KSI < -$- (0.66 Fy) = 29.0 KSI 0 . K .'
'-(NOTE: AXIAL STRESSES IN BEAMS ARE NEGLIGIBLE (<0.24 KSI)).
- 4. DIAGONAL BRACES OF ELEV. TOWER THE MAXIMUM SUM OF THE TENSILE STRESSES IN ANY ONE PAIR OF DIAGONAL BRACES IS: y 2.97 + 3.14 = 6.11 KSI < FT = 4/3 (0.6F y)=26.4 KSI: 0.K. -
( AT BR ACES ON COL. LINE 2 3/4. BETWEEN ELEVATIONS 1104'8 3/4" AND 1097' 5 1/2"). AS A BOUNDING EVALUATION OF THE DIAGONAL BRACES. A SECOND FINITE ELEMENT MODEL WAS USED. THIS SECOND MODEL IS IDENTICAL TO THE FIRST MODEL EXCEPT THAT ONLY HALF 0F THE DIAGONAL BRACES WERE KEPT IN THE MODEL: I.E., ONLY ONE OF A PAIR OF A DIAGONAL BRACES ARE MODELED. SEE ACCOMPANYING FIGURE. ACCORDING TO THE RESULTS USING THIS "SECOND" MODEL: MAX. BRACE TENSILE STRESS . (AT BRACE ON COL. LINE 31/2 BETWEEN ELEVS. 1104' 8 3'4"/ l AND 1097' 5 1/2"). l
=
5.4 KSI < 4/3 (0.6 Fy) - FT 26.4 KSI: 0.K. l ALL OTHER STRUCTURAL MEMBERS I.E., THE BEAMS. COLUMNS, BASE PLATES. ANCHOR BOLTS. ETC. ARE STILL QUALIFIED USING THE RESULTS OF THE "SECOND" MODEL: THE MODEL WITH ONLY HALF OF THE DIAGONAL BRACES.
- 5. BRACES CONNECTING THE ELEV. TOWER AND FUEL TRANSFER CRANE SUPPORT (FTCS) STRUCTURE.
IN COMPRESSION: -
~~ ,
F " MAX' A I 99 KSI < IA = 14.26 KSI: 0.K. IN TENSION: MAX T
=
1.65 KSI < FT = 4/3 (0.6Fy ) = 26.4 / KSI: 0.K.
- 6. NS DISPLACEMENT (AROUND VC EQUATOR) = 0.39"
. < NS CLEARANCE = 1.50".
- 7. FTCS STRUCTURE HAS BEEN EVALUATED AND FOUND ACCEPTABLE IN REFERENCE [23. THE SEISMIC LOADING CONSIDERED IN REFERENCE [2] IS THE NRC SPECTRUM SEISMIC LOAD. THE INTERACTION BETWEEN THE ELEVATOR TOWER AND THE FTCS STRUCTURE IS ACCOUNTED FOR IN THE ANALYSIS PERFORMED IN REFERENCE [23.
CONCLUSION: THE ELEVATOR TOWER MEETS CODE ALLOWABLES UNDER THE POSTULATED LOADS AND WILL NOT COLLAPSE ON OTHER PLANT COMPONENTS. l l l l
)
l i _ _ _ - - - . - , _ - . _ _ . - . . . - . . _ ~ _ _ .
~
ELEVATOR TOWER FINITE ELEMENT MODEL USED TO EVALUATE Tile DIAGONAL BRACES r e e **= =-=--e
? ,/
s
/ \
..J {
-)? p)
>- x
} $'< /
s
/ .)
x,
/
/
.- ,s N,
/ l
' ,/(
, )
>s ELEVATOR TOWER
~
~
ITEM D20A ISSUE
SUMMARY
PROVIDE AN EXPLANATION OF DIFFERENCES IN VERTICAL ARS ZPA'S USED IN THE ANALYSIS OF MAIN COOLANT LOOP (MCL) No. 4.
RESPONSE
THREE VERTICAL ARS EXIST AT THE MCL N0ZZLES OF THE RPV. SEE THE FOLLOWING TABLE FOR THEIR DESCRIPTION AND COMPARISON. REACTOR SUPPORT SEISMIC LOADING RING STIFFNESS ARS ZPA.G YCS ORIGINAL, DATED 5/32 0.360 NRC REVISED 6/8G 0.491 YCS REVISED G/86 0.239 HOTE THAT THE RATIO 0.491/0.239 APPROXIMATELY EQUALS 2 AS EXPECTED. SEE ALSO FIG. D20A.1 WHICH DEMONSTRATES THE EFFECT OF REVISION OF Tile RE ACTOR SUPPORT RING (RSR) STIFFUESS. NOTE THAT THE REVISED RSR STIFFNESS IS HIGHER THAN THE ORIGINAL ONE, AND THE RPV VERTICAL FREQUENCIES ARE 13.2 AND 17.2 HZ ( USING THE ORIGINAL AND REVISED RSR STIFFNESSES, RESPECTIVELY. WITH THE REVISED RSR STIFFNESS THE RSS AND RPV VERTICAL FREQUENCIES NO LONGER COINCIDE (12.3 HZ AND 17.2 HZ, RESPECTIVELY). l WITH REFERENCE TO THE ABOVE INFORMATION. Tile VERTICAL ZP10F j 0.491G'S USED IN THE LOOP 4 CONFIRl1ATION ANALYSIS IS ! JUSTIFIABLE.
i Y SPFCTRA
%, ~
P.O MON, 22 DEC 1986 - 16: 42: 2LI 6.00 FILts MCL.T.TCS.RS DestP]NQ 0.070. 0.030. 0.090 s..n - ---- Original (5/82 l RSR stiffness Revised (6/86) RSR stiffness *
;3 i !
I I U n.nn l I
~
l \ I I d
- l/ V
$ , !l
'=
\\
d 3.co i. y e )i ? s Y a ! ;- .1 lt (( i: i} G ., h, s.
*mY I }
h' ei i 1 i.= i$
, --;~'~
C.t 0.0 0.3 0.4 0.50.4 1.0 2.0 3.0 4. 5. 6. 10. 20. 30. 90. 53. e3. 10s. rntautwei mzi , CYGNA ENERGY SERVICES FIG. D20A.1 EFFECT OF RSR STIFFNESS ON VERTICAL ARS AT RPV MCL N0ZZLES: YCS LOADING
ITEM D20B ISSUE
SUMMARY
GENERATE YCS ARS USING THE "NEW" RPV h0 DEL WHICH WAS USED TO GENERATE SPECTRA F'OR THE LOOP 4 ANALYSIS. REFERE NCE: CYGN A CALCULATION BINDER 86064/3F CAL. SET M. DATED DECEMBER 1986. 1
RESPONSE
ARS AT THE MAIN COOL ANT LOOP (MCL) N0ZZLES OF ._ THE RPV ARE SHOWN ON THE FOLLOWING PAGES. THE SEISMIC LOADING IS THE YCS. THE RPV MODEL USED IS THE "NEW" RPV MODEL USED TO GENERATE THE ARS FOR THE LOOP H CONFIRMATORY AN A LY SI S. THIS "NEW" RPV MODEL HAS A HIGHER REACTOR SUPPORT RING STIFFNESS (RSR) THAN THE EARLIER RPV MODEL SHOWN IN MAJOR MECHANICAL EQUIPMENT QUALIFICATION REP 0F.T. REV. 1. THE "NEW" RPV MODEL VERTICAL FREQUENCY IS 17.2 Hz. NOTE THAT THE RSR STIFFNESS HAS VIRTUALLY NO EFFECT ON THE HORIZONTAL ARS. THE VERTICAL ARS. HOWEVER. IS AFFECTED AS SHOWN ON FIGURE D20A.1 SEE. ALSO RESPONSE TO ITEM D20A.
~
HORIZONTAL ARS AT RPV MCL N0ZZLES DUE TO YCS SPECTRA 2.0 FRI, 12 DEC 1983 09:54:52 [ m M $ OnnPING 0.020. 0.030. c.tpeo. 0.050 d l i l E .00 U . f S b -
'".i0 0.2 0.. O . , 0. 0. . i.0 2.0 .0 ,. . . . . iO. ... .a. ......0. is.
FREQUENCY (HZ1 CTGNA ENERGY SEFi ICES l l
)
, - . .-n -., .-- ,,.n.._
VERTICAL ARS AT RPV MCL N0ZZLES DUE TO YCS SPECTBA 2.O FRI, 12 DEC 1986. 09:54:52 sY K ILES DUE TO TCS e DMPING O.080. 0.030. 0.090. 4.050' O.UU d 1 t.% 0 E F u ' N vs j ...
- 5. 10. 40. 30. 90. 50. 60. 800.
0.1 0.2 0.3 0.9 0.50.6 1.0 2.0 3.0 9. 6. FREQUENCT IH2) CTGNR ENERGY SERVICES l l m
ITEM D20c l ISSUE
SUMMARY
INVESTIGATE SG UPLIFT POTENTIAL BASED' ON RESULTS OF CONFIRMATORY ANALYSIS OF LOOP 4. REFERE NCES: [1] CYGNA CALCULATION BINDER 86063/13F. CALC. SET A. DATED OCTOBER 1986. [2] CYGNA CALCULATION BINDER 86064/3F. CALC. SET L. DATED DECEMBER 1986. ._
RESPONSE
THE SG SUPPORT FRAME IS SHOWN ON THE FOLLOWING , PAGE. THE FRAME IS EVALUATED IN REFERENCE [2]. THE EVALUATION IS PERFORMED USING A FINITE ELEMENT ANALYSIS. THE FINITE ELEMENT N0 DEL UTILIZED IS SHOWN ON THE FOLLOWING PAGES. D
- - ---- , , .- m - ,,,_.__..
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-' . lt.. '
4 SA:.', f'
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,b - .y....j n - , . ;,
- . = .>
;. :, .. . -- -s - -
4, 5 T
- b. 1 r
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1 i 1
FINITE ELEMENT MODEL OF THE STEAM GENERATOR SUPPORT FRAME
~
MM 2'1 8 s'1 0
'D*' '.-
,J J J,~/ ,**m-- ****,
q
,.3
,. s'",'~'~...K~..~~
.p
- 33
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'1
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) (Call d N) b_- . ~..~.x 7 ',... l i 49.5* , . - 4 [1 *,*
%,. y i
1 l SG SUSFORT FRCME
LOADS: THE SG SUPPORT FRAME IS AN ALYZED UNDER LOADS CALCUL ATED IN THE LOOP 4 CONFIRMATORY ANALYSIS (REFERENCE [13). , LOAD COMBINATIONS CONSIDERED ARE: (1) (DL + TH) + SRSS (NRC, SAM) FOR VERTICAL LOAD, F (2) (DL + TH) - SRSS (NRC, SAM) y
+ SRSS (NRC, SAM) III
- DL + TH FOR BENDING MOMENTS M A , M g III THLG
- LOAD CASE Fy I2) K , Mg, K-IN Mg, K-IN (1) -208.5 11395 12610 (2) -246.2 11395 12610 NOTES: (1) SEE FINITE ELEMENT MODEL OF THE SG FRAME FOR ORIENTATION OF PIPING AXES A, B.
(2) (-)IS COMPRESSION. 6 r . , , ,~ av,. ,,-,,,,-,--,,---,,-ne v.- , - ,r_, , , - , , . , , . - , n. , - - - , - - - n,-------- , , , , - - ,
EVALUATION OF REsutTs
- 1. TENSION IN FRAME LEGS -
'0NLY LEG 1 (LEG CONTAINING NODE 1) IS IN TENSI'ON.
LOAD CASE (1) CONTROLS. TENSILE REACTION = 83.8K. TENSION / BOLT = "PT = 83.8 = 21.0 K. TENSILE C APACITIES: (1) A-307 BOLT, i 1/2"p: FT = 35.K >J_T (2) PULL-0UT: EMBEDMENT IS LARGE: BOLTS ARE WELDED TO PLATE WHICH IN TURN IS WELDED TO PIER CEBAR (3 # 11 REBARS/2 ANCHOR BOLTS): THEREFORE. PULL-0UT CAPACITY IS LARGE. (3) WELD BETWEEN PLATE AND ANCHOR BOLTS: Fw-57.3 KIPS >fT
- . L EG 1 0.K. IN TENSION 4
- 2. COLUMNS MOST CRITICALLY LOADED COLUMN ELEMENT IS EL. 15 (EL.
, BETWEEN N0 DES 53 AND 63) UNDER LOAD CASE (2)
- n. .
IKl )F y = 21.915 KSI F4 = 2/3 (1 - t FB = 4/3 (0.66) Fy = 29.04 KSI CH A . B , 8.68 . 0.85 (20.29) 21.915 F A g_ k'A , )p 8 (3 _ j 8h ) 29.04 F E
= 0.99 < is 0.K.
- 3. BEAMS MAX. BENDING STRESS [(AT EL. 38. LOAD CASE (2)]
=
8.93 KSI < Fg = 4/3 (0.66 Fy) = 29.0 KSI: 0.K. NOTE: COMPRESSIVE STRESSES IN BEAMS ARE NEGLIGIBLE.. CONCLUSION SG SUPPORT FRAME IS ADEQUATE UNDER LOOP 4 CONFIRMATORY ANALYSIS LOADS.
l QUESTION D21 1 CONFIRMATORY ANALYSIS OF THE DEDICATED SHUTDOWN PIPING IN THE PRIMARY AUXILIARY BUILDING Prepared for the U.S. Nuclear Regulatory Commission Washington D.C. 20555 Under 00E Contract No. DE-AC07-76I001570 j FIN No. A6808 i i 1 s
._ . . - _ . - . - - - , _ - _ _ . _ _ , _ . _ , , - - , . _ . , _ . , - , . ,,-...-,___...,_.,_______,,.,._.,vy-.,_._.,,,r.,.
o SCOPE PRIMARY SIDE DEDICATED SAFE SHUTDOWN PIPING IN THE PAB SECONDARY SIDE DEDICATED SAFE SHUTDOWN PIPING IN THE PAB ATTACHED PIPING SUFFICIENT FOR ADEQUATE BOUNDARY CONDITIONS d o 0BJECTIVES TO PROVIDE INDEPENDENT NRC SPECTRA /SEP GUIDELINES EVALUATION OF THE PIPING FOR COMPARISON WITH THE LICENSEE'S YCS EVALUATION I - TO DETERMINE IF YCS/ CODE ANALYSIS IS ADEQUATE IN DETERMINING NRC SPECTRA /SEP GUIDELINES ACCEPTABILITY (QUESTION D1) TO EVALUATE CONCRETE ANCHOR BOLT MARGIN FOR YCS LOADING (QUESTION DIS) TO EVALUATE THE ADEQUACY OF THE BRACKETING CRITERIA (QUESTION D22) j i 2
o APPROACH TWO MODELS WERE ANALYZED FOR THE SECONDARY SIDE ONE TO MATCH THE LICENSEE'S MODEL ONE WITH BOUNDARY CONDITIONS ACCORDING TO THE JUDGEMENT t 0F THE ANALYST THE TWO MODELS WILL BE COMPARED TO EVALUATE THE - BRACKETING CRITERIA; RESULTS FOR THE MODEL MATCHING THE LICENSEE'S WILL BE USED FOR OTHER EVALUATIONS ONE MODEL WAS ANALYZED FOR THE PRIMARY SIDE THE ANALYST AGREED WITH THE BOUNDARIES VALVE WEIGHTS SPECIFIED BY THE ANALYST DIFFER FROM THOSE USED BY THE LICENSEE SPECTRA USED: NRC IN THE PAB FOR THE SECONDARY SIDE ANALYSES NRC IN THE PAB ENVELOPED WITH THE VC PENETRATION YCS SPECTRA TIMES 2 FOR THE PRIMARY SIDE 3 _ _ - , _ - - . - . . - , . _ . - _ _ , _ _ . -- - - - - ,. . - . - . _ _ , . . , , . . . , , . , . . , , ,v,_ , - - . _ . _ , _ . ..- - , . , _ - ,m_-___ , . . .. . _ _ _
P 4 e k 86
=
og 888 1i bge pet *
,,, ' U p,,,, y,, e ,, ,
' Tn.
~
ase
/% u
,,~ ** /
i V C. een " r&e" .
. ,,~
~
J. " PRIMARY SIDE MODEL 4 I n._n,_- . , , - . n_, -.-,-...,-_,,_...__n__
9 9 40 Q,h I
%a. s y
- ' gf g.4 ?
- 4 y e'* >
it*7.pt 1:t +* V c y - ., . x ,
'S
- . .n*
d..fs
/s%giy c, **
ge
/
/
~.Q f s
/
/
/e "
,s
\\
/
.+
s# # .* 5 -. ,. - . , _ , _ . - . _ _ . , , . , _ . _ _ , _ , _ _ , . . . - - . , . . , . . . _ _ - . _ . . . . _ , - . , _ - . , . . _ . - , _ - - . _ . _ _ _ _ . . _ _ . , , . - - , _ , _ . , _ + . - _ , , , . _ _ _ . . _ _ _ _ _ , _ _ _ ,
l i l 1 1 o RESULTS - PRIMARY SIDE MODEL PIPING t Eq. 9 Eq. 9 Stress (ksi) Allowable (ksil Component Node 295 29.5 28.6 Socket weld at elbow 300 30.0 28.6 Socket wald at elbow 305 29.5 28.6 Socket weld at reducer 310 28.1 28.6 Reducer 359 19.6 28.6 Butt weld at elbow I i e t l l 6 I
o COMENTS ON THE PRIMARY SIDE ANALYSIS
- TWO ROD HANGERS DESIGNATED PRSH-26 ON DRAWING NO. 80023-P1-1063 WERE PREDICTED TO BUCKLE BASED ENTIRELY ON THE LOADING .
LARGE TOTAL DISPLACEMENTS (7 INCHES) WERE PREDICTED IN THE PIPE TUNNEL THE LIMITING STRAIN CRITERIA RATIO BASED ON ELASTIC STRESS IS 0.80 b 0 7
o R5SULTS-FULLSIZESECONDARYMODELPIPING Eq. 8 Eq. 8 Eq. 9 Eq. 9 Node Stress Allowable Stress Allowable Component 280 14.4 15.0 26.8 27.0 Tee 355 18.4 15.0 27.1 27.0 Str. Pipe 360 21.2 15.0 32.3 27.0 Socket Weld 365 17.0 15.0 26.4 27.0 Socket Weld 1045 28.2 15.0 38.9 28.6 Tee - o RESULTS - REDUCE 0 SIZE SECONDARY MODEL PIPING Eq. 8 Eq. 8 Eq. 9 Eq. 9 Node Stress Allowable Stress Allowable Component 555 14.1 15.0 25.9 27.0 Elbow / Soc. Wald 560 15.6 15.0 26.8 27.0 Elbow / Soc. Weld 640 16.6 15.0 29.8 27.0 Tee 690 15.7 .15.0 27.8 27.0 Elbow / Soc. Weld 695 17.5 15.0 29.3 27.0 Elbow / Soc. Weld 8 - . , . ,nn..- . - , , -,..--. . , - - . . - , - - , . - - - - . - - - , , . - , , , , _ , , - . , , . , . _ . - , _ , . . - _ . - - - . _ _ , - - - . , - - - - - . . - - - - , - . , , - -
A o COMMENTS ON THE SECONDARY SIDE ANALYSES VALVE WEIGHTS OF 90 POUNOS WERE USED IN CONTRAST TO THE LICENSEE'S WEIGHTS OF 25 POUNDS DEADWEIGHT OVERSTRESS WAS PREDICTED
- ATTRIBUTED PRIMARILY TO VALVE WEIGHTS THE MODELING OF THE LINES RUNNING TO THE PIPE TUNNEL WERE A SECONDARY FACTOR
- THE LIMITING STRAIN CRITERIA RATIO BASED ON ELASTIC STRESS IS 1.03 FOR THE FULL MODEL AND 0.80 FOR THE REDUCED MODEL
- USE OF THESE VALUES IS NOT ALLOWE0 FOR A PASS / FAIL DECISION UNLESS THE VALVE WEIGHT DISCREPANCY IS RESOLVED IN FAVOR OF THE LICENSEE BECAUSE DEADWEIGHT STRESSES CONTRIBUTED MORE THAN HALF OF THE EQN 9 RESULTS
- IF USED IN A DIRECT COMPARISON TO LICENSEE RESULTS, COMPARISON WILL BE CONSERVATIVE BECAUSE THE HEAVIER VALVES WOULO CAUSE RELATIVELY HIGHER SEISMIC STRESSES i
I I 9 i l
- , - - . . . . . . - , . .y_, - _ _ . _ . . _ _ _ - , . . , , , _ . . _ , . , , .._-.w.,_ __ _ . .. - . _ . _ _ _ _ _ , - . _ _. . . , . , , , . . ,_,, . - . m, -. ,.. , . _ __
i o RESULTS - PIPE SUPPORT EVALUATIONS LOADS FOR FULL AND RECUCEO SECONDARY SIDE MODEL WERE COMPARA8LE RESULTS FOR SECONDARY SIDE REDUCEO MODEL WERE USED LICENSEE'S SUPPORT COMPONENT STRESS RATIO STRESS RATIO SSS-H-8 3 X 3 X 1/4 ANGLE 0.47 SSS-H-9 3 X 3 X 1/4 ANGLE 0.80
. HILTI KWIK BOLTS 0.18 SSS-H-10 2 X 2 X 1/4 ANGLE 0.64 HILTI KWIK-BOLTS 1.35 I
l 1 10 ( 4 1
l
~
c l o CONCLUSIONS - GENERAL
- PRIMARY SIDE PIPING FOUND ACCEPTABLE, BUT TWO SUPPORTS ARE PREDICTED TO BE OVERLOADED SECONDARY SIDE PIPING OVERLOADED IN DEADWEIGHT AND SEISMIC IF w
VALVE WEIGHTS IN THE ANALYSIS ARE CORRECT 2 SECONDARY SIDE ROD HANGERS WERE PREDICTED TO BE OVELOADED BASED ON LOAD ONLY ONE OF THE THREE NEW SUPPORTS IS OVERLOADED (NEW SUPPORTS ARE COPMON TO BOTH SYSTEMS) o CONCLUSION - BRACKETING CRITERIA RESULTS FOR THE FULL AND REDUCED SECONDARY SIDE ANALYZES INDICATE THAT THE BRACKETING CRITERIA ARE ACCEPTABLE AND HAVE BEEN ADEQUATELY APPLIED 1 o CONCLUSIONS - CONCRETE ANCHOR BOLT CRITERIA INTERACTION EQUATION RESULTS FOR TWO SUPPORTS HAVE BEEN PRESENTED FOR COMPARISON TO THE LICENSEE'S RESULTS i 11
- - - , ,--,v .-,- -e -w,---.,,e--- --- -
, --~,-.--=----------r-- - - - - - - - - - . -e-- - - . - - . - -, -- - -- -- r-----e - --
O ITEM D24A Issue Summary Justify using full distance to the end of the boron tank when an intermediate support exists. Reference Calculation YRC 386: Qualification of Piping and Supports inside SSS Building, Rev. 3, dated October 31, 1986
Response
The intermediate lateral support was not considered in the nozzle stiffness calculations since it was judged to provide insignificant additionci stiffness .. at the nozzle locations. The support is not welded to the tank wall, and only makes point contact at fcur locations. In a(dition, review of the calculations for these nozzles indicated that the
' results are relatively insensitive to an intermediate support ut the given location. If it were assumed that the lateral support was significant, the follouin6 numbers would be calculated per WRC 297:
. Values of .A.
Existing analysis Assuming Support Effective 10 74 Node 1165 35 27 Node 2070 Review of figures 59 and 60 indicate that the values of stiffness used in the referenced calculation are still valid for the J( values tabulated above. I I I l
ITEM D24.b
. Issue Summary Justify using cylindrical geometry methodology to a spherical structure.
Reference (1) Calculation YRC-386, " Qualification of Piping and Supports Inside SSS Building," Revision 3 dated October 31, 1986.
Response
This question will be addressed in two parts: (1) Stiffness calculations, and (2) Stress calculations. R (1) Stiffness Calculations The choise of stiffness value for the nozzle at Node 3025 (underside of tank) wil'. affect the stress results for the attached piping and the tank shell. , For the attached piping, a conservatively low value chosen for stiffness will result in conservative stress values. The referenced calculation assumed the tank bottom was cylindrical rather than spherical in calculating stiffness. The calculation has been revised to include further calculations which indicate that the stiffness values used in Revision 3 were conservative. In addition, Revision 3 indicates that stress margins in this piping are greater than 7 to 1, so we conclude that the piping is adequately qualified. For the tank shell, a high value of stiffness will result in conservative tank shell stress values. The referenced calculation was revised to include more conservative stiffness calculations. These were input to the piping analysis to obtain conservative nozzle loads. (2) Nossle Stress Calculations ,, Stress calculations were made using the new nozzle loads discussed above and utilizing WRC Bulletin 107. Results indicate that the tank shell stresses are still well within allowables. 5194R. s
ITEM D.24.c Issue Summary l Will calculations be revised to remove use of WRC 297 for tank nossle evaluations? ,, Reference (1) Calculation YRC-386, " Qualification of Piping and Supports Inside SSS Building," Revision 3, dated October 31, 1986. . Response One of the three nossles in the referenced calculation has been re-evaluated using WRC Bulletin 107. The remaining two nossles do not fall within the Bulletin's parameters however, and so they remain qualified in accordance with ' WRC Bulletin 297. This calculation is available for audit in association with the use of this Bulletin. . i f i 5194R ,
1. ITEM H5 Issue Summary Demonstrate that the flexible hose used to connect the buried piping to the Fire Tank is adequate for NRC seismic loads. References (1) Cygna calculation 85005/1-F, SET CES 2 REV. 2
Response
The hose manufacturer was contacted to obtain the maximum allowable offsets.
- The following numbers were received:
~
- 4"% Hose 2i"% Hose i
47" Long 34" Long Allowable Axial Movement 2" 2" Allowable Transverse Movement 4" 4" 4 These numbers are a minimum of 5 times greater than the expected YCS anchor j
. motions (see Ref. 1). It is expected that NRC anchor motions would not exceed approximately 2 times the YCS values. Taerefore we conclude that the hose is adequate for YCS and NRC seianic loads.
t i o e 4 i e 1
0 32b AXIAL RESTRAINT FOR U-80LTS t Prepared for the U.S. Nuclear Regulatory Comission Washington D.C. 20555 Under 00E Contract No. OE-AC07-76ID01570 FIN No. A6808
o CREDITING U-BOLTS WITH CAPABILITY TO RESIST AXIAL LOADS IS NOT STANDARD INDUSTRY PRACTICE e o A BASIS HAS BEEN ESTABLISHED FOR EVALUATING THE PROPOSAL TO CREDIT U-BOLTS WITH AXIAL RESTRAINT CAPABILITY BASED ON THE FOLLOWING: LETTER FROM T. T. MARTIN (NRC) TO R. A. UDERITZ (PSE&GC) DATED APRIL 9, 1984;
SUBJECT:
INSPECTION REPORT NO. 50-272 AND 50-311/84-0E. LETTER FROM T. T. MARTIN (NRC) TO R. A. UDERITZ (PSE&GC) DATED JUNE 15, 1984;
SUBJECT:
COMBINED INSPECTION 50-272/64-05 AND 50-311/84-05. l l l l 2 l l l
l l l o A BASIS MUST BE PROVIDED FOR THE ALLOWABLE LOADS AND STIFFNESSES ESTABLISHED FOR AXIAL LOADING OF U-BOLTS THE BASIS MUST INCLUDE SUBSTANTIATING TEST RESULTS FROM A TEST PROGRAM CONOUCTED PER QA/QC REQUIREMENTS STANDARD IN THE INDUSTRY o LOCAL STRESSES DEVELOPED IN THE PIPE WALL UNDER THE U-BOLT MUST BE CALCULATED AND MUST MEET APPLICABLE ASME CODE REQUIREMENTS o A MEANS OF ENSURING THAT THE REQUISITE PRELOAD ON THE U-BOLT EXISTS FOR THE LIFE OF THE PLANT MUST BE ESTABLISHED 4 e 3
. s, y
ITEM D24A Issue Summary Justify using full distance to the end of the boron tank when an intermediate support exists.
/ Reference Calculation YRC 386: Qualification of Piping and Supports inside SSS Building, Rev. 3, dated October 31, 1986 g
Response
The intermediate lateral support was not considered in the nozzle stiffness calculations since it was judged to provide insignificant additionci stiffness at the nozzle locations. The support is not welded to the tank wall, and
,only makes point contact at four locations.
In addition, review of the calculations for these nozzles indicated that the results are relatively insensitive to an intermediate support at the given location. If it were assumed that the lateral support was significant, the following numbers would be calculated per WRC 297:
. Values of A.
Existing analysis Assuming Support Effective Node 1165 10 74 35 27 Node 2070 Review of figures 59 and 60 indicate that the values of stiffness used in the referenced calculation are still valid for the J( values tabulated above. G
ITEM D24.b l Issue Summary Justify using cylindrical geometry methodology to a spherical structure. Reference (1) Calculation YRC-386, " Qualification of Piping and Supports Inside SSS Building," Revision 3, dated October 31, 1986.
Response
This question will be addressed in two parts: (1) Stiffness calculations, and (2) Stress calculations. 1 (1) Stiffness Calculations The choice of stiffness value for the nozzle at Node 3025 (underside of tank) will affect the stress results for the attached piping and the tank shell. For the attached piping, a conservatively low value chosen for stiffness will result in conservative strece values. The referenced calculation assumed the tank bottom was cylindrical rather than spherical in calculating stiffness. The calculation has been revised to include further calculations which indicate that the stiffness values used in Revision 3 were conservative. In addition, Revision 3 indicates that stress margins in this piping are greater than 7 to 1, so we conclude that the piping is adequately qualified. i For the tank shell, a high value of stiffness will result in conservative tank shell stress values. The referenced calculation was revised to include more conservative stiffness calculations. These were input to the piping analysis to obtain conservative nozzle loads. (2) Nozzle Stress Calculations Stress calculations were made using the new nozzle loads discussed above and utilizing WRC Bulletin 107. Results indicate that the tank shell stresses are still well within allowables. K 5194R t
ITEM D.24.c Issue Summary Will calculations be revised to remove use of WRC 297 for tank nozzle evaluations? Reference
'1) Calculation YRC-386, " Qualification of Piping and Supports Inside SSS Building," Revision 3 dated October 31, 1986.
Response
One of the three nozzles in the referenced calculation has been re-evaluated using WRC Bulletin 107. The remaining two nozzles do not fall within the y Bulletin's parameters however, ari so they remain qualified in accordance with WRC Bulletin 297. This calculation is available for audit in association with the use of this Bulletin.
*e f
l l l r 1 5194R , f I
i i l l ITEM D32I ISSUE
SUMMARY
- IDENTIFY MEASURES TAKEN TO INSURE THAT U-BOLTS PROVIDING AXIAL RESTRAINT WILL NOT LOOSEN.
RESPONSE
ALL U-BOLTS WHICH ARE REQUIRED TO PROVIDE AXIAL RESTRAINT ARE TIGHTENED AS REQUIRED TO PROVIDE THE RESTRAINT. THE U-BOLTS ARE DOUBLE NUTTED TO PREVENT , LOOSENING. TO PROVIDE FURTHER ASSURANCE THAT AXIAL RESTRAINT WILL BE MAINTAINED. SMALL BORE SUPPORTS WHICH HAVE U-BOLTS PROVIDING AXIAL RESTRAINT WILL BE PLACED IN THE YNPS IN-SERVICE INSPECTION PROGRAM. CONCLUSION: ASSURANCE THAT U-BOLTS REQUIRED TO PROVIDE AXIAL RESTRAINT WILL MAINTAIN THAT RESTRAINT WILL BE PROVIDED THROUGH PERIODIC IN-SERVICE INSPECTIONS. G b
- .,y . . .- , .- _.,__. ._. _. - - _ .. . __ _ ,.__ m ,_ _.__ _ _
QUESTION 038
~~
EXAMPLE PIPING ANALYSIS RESULTS Prepared for the U.S. Nuclear Regulatory Conunission Washington D.C. 20555 Under 00E Contract No. DE-AC07-761001570 FIN No. A6808 O
l o SCOPE PROBLEM NO. 1 FEED AND BLEED HX, CHARGING AND SI ORAINS (SP-FB-5) PROBLEM NO. 10 STEAM GENERATOR E-7-3 BLOWDOWN (SP-SG-11) 1 o GENERAL OBJECTIVE: TO PROVIDE INDEPENDENT NRC SPECTRA /SEP GUIDELINES EVALUATION OF THE PIPING FOR COMPARISON WITH LICENSEE'S ANALYSES TO DETERMINE ACCEPTABILITY OF EXISTING YCS + PVRC OAMPING CALCULATIONS o SPECIFIC OBJECTIVES: TO EVALUATE THE FOLLOWING QUESTIONS YCS + PVRC OAMPING (QUESTION 038)
- 60*. SAM (QUESTION 027)
FACTOR FOR YCS RESULTS TO OBTAIN LOADS AND DEFLECTIONS FOR EVALUATION OF UNIDIRECTIONAL SUPPORTS (QUESTION 03a) INTERACTION EQUATION CRITERIA - LINEAR INTERACTION WITH YCS (QUESTION DIS) 5/3 POWER INTERACTION WITH YCS (QUESTION E25) USE OF FAULTED SUPPORT ALLOWABLES 2 l l
~
o ISSUES RAISED DURING THE ANALYSES USE OF FLOOR SPECTRA AT BRANCH POINTS (DECOUPLING CRITERIA) ACCURACY OF FIELD DATA USED IN ANALYSIS ACCURACY OF LOADING DATA DIRECTION OF SPECTRA (N0 UNIFORM PLANT COORDINATE SYSTEM) MIXED RELATIVE AND ABSOLUTE SAM AND TAM VALUES LOADS FOR GANG SUPPORT ANALYSIS REDUCED ALLOWABLES FOR 1/2" HILTI KWIK-BOLTS ADEQUATE AREA 0F REINFORCEMENT FOR FORGED BRANCH FITTINGS 3 l i
o - APPROACH INDEPENDENT ANALYSIS OF BOTH PROBLEMS s MODEL BOUNDARIES INDEPENDENTLY DEFINED VALVE WEIGHTS ESTIMATED INDEPENDENTLY 2 5ANGSUPPORTLOADMETHODOLOGYUSEDWITHOUTENDORSEMENT BOTH MODEL BOUNDARIES ADJUSTED TO FIT LICENSEE'S MODELS TO ALLOW EVALUATION OF DECOUPLING QUESTION ADDITIONAL RUN PIPING REMOVED FULL MODEL ANALYSIS OF BOTH PROBLEMS COMPLETED FOR GENERAL COMPARISON REDUCED MODEL ANALYSIS OF PROBLEM NO. 1 COMPLETED FOR SPECIFIC QUESTIONS DONE TO REMOVE EFFECT OF DIFFERENT MODELS FROM COMPARISONS FULL MODEL ANALYSIS OF PROBLEM NO. 10 USED FOR SPECIFIC l QUESTIONS USAGE LIMITED TO RESULTS FROM AREAS NOT AFFECTED BY OIFFERENCE IN BOUNDARY CONDITIONS 4
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o LIMITING RESULTS: WORST CASE NUMBERS d
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EQN 9 STRESS LIMITING PROBLEM (KSI) PARAMETER RATIO PROBLEM NO 1. FULL MODEL 97.4 FATIGUE: STRESS 1.9 FATIGUE: CYCLES 25.7 { REDUCED MODEL 63.1 FATIGUE: STRESS 1.3 FATIGUE: CYCLES 3.7 PROBLEM NO. 10 873.4 STRAIN 16.4 i i I 7
- , .. , , _ - - , - - - _ _ . , _ _ . - ,y-. . , , _ _ _ , - , . - . - , _ - - . , , . .. _
l l l o RESULTS: GENERAL COMPARISON FULL MODEL NRC SPECTRA /SEP GUIDELINES RESULTS WERE COMPARED TO LICENSEE'S REDUCED MODEL YCS/ CODE RESULTS PROBLEM NO. 1 2 POINTS ON RUN PIPING ANO 10 POINTS ON BRANCH PIPING s EXCEEDED STRAIN CRITERIA LIMITS 1 R00 HANGER ON RUN PIPING AND 1 ON BRANCH PIPING HAD , BUCKLING LIMITS EXCEEDED
- LICENSEE'S ANALYSIS INDICATED ACCEPTABILITY, WITH MAXIMUM PIPING STRESS RATIOS OF 0.9 AND 0.7 FOR ANSI B31.1-1977 EQUATIONS 12 AND 14, RESPECTIVELY, AND SUPPORTS ACCEPTABLE PROBLEM NO. 10 6 POINTS ON RUN PIPING AND 9 POINTS ON BRANCH PIPING EXCEEDED STRAIN CRITERIA LIMITS 7 SUPPORTS ON BRANCH PIPING EXCEEDED SEP GUIDELINES (RUN PIPING SUPPORTS NOT ANALYZED)
- LICENSEE'S ANALYSIS INDICATED ACCEPTABILITY, VITH MAXIMUM STRESS RATIOS OF 0.96 AND 1.02 FOR ANSI B31.1-1977 EQUATIONS 12 AND 14, RESPECTIVELY, AND SUPPORTS ACCEPTABLE (WITH USE OF FAULTED ALLOWABLES) 8
o RESULTS: ACCEPTABILITY OF YCS + PVRC DAMPING AND 60% SAM REDUCED PROBLEM NO. 1 MODEL NRC SPECTRA /SEP GUIDELINES RESULTS WERE COMPARED TO LICENSEE'S YCS/ CODE RESULTS 3 POINTS EXCEEDED STRAIN CRITERIA LIMITS DUE TO COMBINED INERTIA AND SAM LOADING t
- LICENSEE'S ANALYSIS INDICATED ACCEPTABILITY, WITH MAXIMUM STRESS RATIOS OF 0.9 AND 0.7 FOR ANSI B31.1-1977 EQUATIONS 12 AND 14, RESPECTIVELY, AND SUPPORTS ACCEPTABLE APPLICABLE PORTIONS OF PROBLEM NO. 10 MODEL NRC SPECTRA /SEP GUIDELINES RESULTS WERE COMPARED TO LICENSEE'S YCS/ CODE RESULTS AT LEAST 2 POINTS EXCEEDED STRAIN CRITERIA LIMITS DUE TO SEISMIC INERTIA LOAD ALONE AT LEAST AN ADDITIONAL 8 POINTS EXCEEDED STRAIN CRITERIA i WHEN SAM LOADS WERE INCLUDED
- LICENSEE'S ANALYSIS INDICATED ACCEPTABILITY, WITH MAXIMUM STRESS RATIOS OF 0.96 AND 1.02 FOR ANSI B31.1-1977 EQUATIONS 12 AND 14, RESPECTIVELY, AND SUPPORTS ACCEPTABLE (WITH USE OF FAULTED ALLOWABLES) l l
9
o RESULTS: FACTOR FOR UNIDIRECTIONAL SUPPORTS (QUESTION 03a) APPLICABLE PORTIONS OF PROBLEM NO. 10 MODEL NRC SPECTRA /SEP GUIDELINES RESULTS WERE USE0 TOTAL DEFLECTION IN INCHES PERPENDICULAR TO THE RESTRAINT DIRECTION WAS CALCULATED USING STANDARD LOAD COMBINATION METHODOLOGY 9 LICENSEE'S SUPPORT ID DEFLECTION DEFLECTION RATIO WGB-H-62 16.2 WGB-H-63 13.0 WGB-A-3 11.8 WGB-A-4 10.5 WGB-H-64 2.3 10
o RESULTS: INTERACTION EQUATION CRITERIA (QUESTIONS 015 AND E25) SINCE NO APPLICABLE RESULTS WERE AVAILABLE FOR COMPARISON TO LICENSEE'S RESULTS, OUPNY YCS LOADS WERE OBTAINED BY FACTORING PROBLEM 10 RESULTS BY 1/2 5 SUPPORT NO. SP-SG-11-HS SP-SG-11-H8 NRC SPECTRA LOADS TENSILE (K) 1992.00 1044.00 SHEAR (K) 33.00 292.00 YCS LOADS TENSILE (K) 996.00 522.00 SHEAR (K) 16.50 146.00 1/2" STAR SLUGIN ALLOWABLE LOADS TENSILE (K) 2000.00 2000.00 SHEAR (K) 2000.00 2000.00 INTERACTION EQUATIONS NRC SPECTRA LOADS POWER OF 2 1.00 0.54 YCS LOADS LINEAR 0.51 0.33 5/3 POWER 0.31 0.18 RATIO - NRC/YCS LINEAR 1.97 1.62 5/3 POWER 3.18 2.97 11 ( b
o RESULTS: FAULTED SUPPORT ALLOWABLES AT LEAST 7 0F THE SUPPORTS WERE SHOWN OVERLOADED BY THE APPLICABLE PROBLEM 10 RESULTS, ONE WITH A STRESS RATIO OF - 3.6, THE OTHERS WITH A RATIO OF AT LEAST 1.1 THE LICENSEE'S ANALYSIS, WHICH INCORPORATED FAULTED SUPPORT ALLOWABLES, SHOWED THEM ACCEPTABLE e e 12
, -,,,--,,a. --- , - . - . - . - . - , _ - . - - . -
i o RESULTS: DECOUPLING ISSUE TOP FIVE SEISMIC INERTIA STRESS RATIOS WITH AND WITHOUT RUN PIPING ARE PRESENTED FOR 80TH MODELS EQN 9 STRES3 (KSI) PROBLEM NODE COMPONENT WITH RUN PIPING NO. NO. TYPE _I,N , OUT RATIO 1 501 SOCKET WELD 56.8 10.6 5.3 500 SOCKET WELD 52.9 13.6 3.9 305 TEE JUNCTION 35.2 11.1 3.2 304 SOCKET WELD 55.6 17.7 3.2 303 SOCKET WELD 56.0 17.8 3.2 10 560 SOCKET WELD 34.2 7.1 4.8 550 SOCKET WELD 33.9 7.6 4.5 545 SOCKET WELD 33.6 8.1 4.1 535 SOCKET WELD 32.4 8.8 4.1 565 SOCK 0LET 38.3 12.1 3.2 l l l 13 i s
- - .n. - - , - ,, , , , - . . , , . ~ . . . . , . , - - , . ,,--,n. - . --
o CONCLUSIONS FOR EXISTING YCS + PVRC DAMPING CALCULATIONS: CALCULATIONS WIT'H MAXIMUM STRESS RATIO LESS THAN 0.5 ARE ACCEPTABLE CALCULATIONS WITH STRESS RATIO LESS THAN 0.7 ARE ACCEPTABLE, PROVIDED THAT: A REVIEW OF THE CALCULATION BY THE LICENSEE HAS ESTABLISHED THAT SPECTRA AT BRANCH POINT MODEL CUTS AND VALVE WEIGHTS ARE ACCURATE A CASE-BY-CASE REVIEW BY THE NRC STAFF FINDS THE CALCULATION ACCEPTABLE CALCULATIONS WITH MAXIMUM STRESS RATIO AT AND ABOVE 0.7 ARE NOT ACCEPTABLE NONE OF THE OTHER PROPOSED CRITERIA IDENTIFIED IN THE SPECIFIC OBJECTIVES ARE ACCEPTABLE 14
~
1 o CONCLUSIONS (continued) CURRENT DECOUPLING CRITERIA ARE NOT ACCEPTABLE SUITABILITY OF APPLICATION OF GROUND SPECTRA AT BRANCH POINTS MUST BE CONSIDERED IF SPECTRA ARE NOT SUITABLE, THE MODEL MUST BE EXTENDED THE CONVENTIONAL NUMBER OF SUPPORTS AND FLOOR SPECTRA APPLIED TO THOSE SUPPORTS THIS CONCLUSION IS APPLICABLE TO ALL SSS PIPING OTHER ISSUES RAISED IN THE EXAMPLE ANALYSES MUST BE RESOLVED i 15
e 1 9 I e
' SAFETY INJECTION EXAMPLE ANALYSIS .
O e 9 9 3 l l t
SYSTEM DESCRIPTION 0 CONSOLIDATION OF L ARGE BORE PROBLEMS #201 AND #207 s. O INCLUDES PIPING OF THE FOLLOWING NOMINAL SIZES: 8", 6", 4", 3". 2" O ANCHORED AT 4 MAIN COOLANT LOOPS 3" THERMAL SLEEVE AT MAIN COOLANT LOOPS e INCREASING TO 4" PIPING AFTER MOVS 0 EXITS THE VC AT 3 PENETRATIONS 8" PIPING REDUCING TO A 6" PENETRATION 3" PENETRATION 2" PENETRATION 7 ALL PENETRATIONS ARE REINFORCED PAD WITH PIPING WELDED AT INSIDE AND OUTSIDE SURFACES 0 INCLUDES 2" CROSS-0VER LINES BETWEEN LARGER LINES l 9 s
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ANALYSIS 0 ANALYZED TO NRC SPECTRA AND SAMS
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0 '2 ANALYSES WERE PERFORMED ORIGIN AL CONFIGUR ATION ( AS-BUILT) SEISHIC OK WITH STRAIN CRITERIA SAM OVERSTRESS IN 4 AREAS (2 VC PENETRATIONS. I MAIN )
~
COOLANT LOOP CONNECTION, 1 2" CROSS-CONNECT) REVISED CONFIGURATION 4 SUPPORTS REMOVED SEISMIC OK WITH STRAIN CRITERIA SAM OVERSTRESS AT 1 VC PENETRATION l l I 1 l 1 I
ANALYSIS RESULTS 0 ORIGINAL CONFIGURATION E 6" VC PENETRATION, SIF = 2.1
, EON 12 STRESS = 34677 PSI (2.4SH = 41820 PSI EON 14 STRESS = 73676 PSI (W/ SAM) (S4 + Sy = 46625 PSI) e E 3" VC PENETRATION, SIF = 2.1 EON 12 STRESS = 9459 PSI (2.4SH = 41820 PSI)
EON 14 STRESS = 60414 PSI (W/ SAM) SA*8H = 46625 PSI) E MAIN COOLANT LOOP #4 N0ZZLE, SIF = 2.0 EON 12 STRESS = 34372 PSI (1.8Sg = 31365 PSI) EON 14 STRESS = 67443 PSI (W/ SAM) (S4 + Sg = 46625 PSI)
- E 2" CROSS-CONNECT, SIF = 2.1 EON 12 STRESS = 21297 PSI (2.4Sg = 41820 PSI)
EON 14 STRESS = 56048 PSI (W/ SAM) (S4 + Sg = 46625 PSI) 0 REVISED CONFIGURATION E 6" VC PENETRATION, SIF = 2.1 i EON 12 STRESS = 40609 PSI (2.4SH = 41820 PSI) EON 14 STRESS = 71789 PSI (w/ SAM) (S4 + Sg = 46625 PSI) ALL OTHER POINTS BELOW ALLOWABLES t m n
i
)
f: : ANALYSIS OF THE MAIN STEAM AND FEEDWATER PIPING AND SUPPORT STRUCTLRE OUTSIDE THE VAPOR CONTAINER
1.0 INTRODUCTION
The main steam /feedwater piping outside the VC is supported on a series of light structural frames. The mass and stiffness of the support structure and piping runs are of similar magnitude; therefore, the rigid support assumption generally used in the majority of the piping analyses is not valid here. Additionally, a number of pipes may be supported by the same frame, promoting interaction between the pipes. For the above reasons, the main steam /feedwater piping and the support structure shall be analyzed together in one problem or they can be analyzed separately with interaction explicitly considered. This document outlines the methods and criteria which shall be used in the analysis and qualification of the main steam /feeddater piping outside the VC. Both seismic and non-seismic portions of the piping shall be analyzed. The extent of the amount of non-seismic piping shall be determined based on its effect on the response of the seismic piping and the support structure. Certain modifications ney be designed to limit the areas of interaction between the seismic and non-seismic portions. Specific analysis details and evaluation criteria are discussed in the following sections. 2.0 MAIN STEAM /FEEDWATER PIPING ANALYSIS 2.1 Geometry and Computer Modeling The combined main steam /feedwater piping and support structure system is large and complex, making it prohibitively costly and time-consuming to analyze a detailed model of the entire system. To
- minimize the size of the actual seismic analysis model, i
substructuring methods shali de used to model the support structure frames to limit the degrees of freedom to a manageable number. Each frame shall first be modeled in detail, considering all structural members, connections, and other attached piping and equipment. Consideration shall be given to the supporting scheme on each frame to determine whether the main steam and feedwater pipes will interact on that particular frame. If the supporting scheme allows interaction, the attachment points of the piping shall be retained, and the detailed frame model shall then be condensed to . 1 l YAN: 0599 ( l l _ . , _
retain only the important meters and equivalent masses and stiffnesses of the remaining med ers. This simplified model shall then be used for the system analysis. The above procedure shall be used for every support frame.' The analysis model shall consist of the eight main steam and -feedwater lines with equivalent support frames connecting them as appropriate. After the system analysis is completed, the reactions at the support points shall be used to analyze the detailed support frames for qualification. If modifications to the supporting scheme or the support frames are required, the frame models shall be reevaluated to assess the interaction and to determine the necessary degrees of freedom for the revised configuration. The revised ' support frame models shall then be regenerated and recondensed. This iterative process shall be repeated until an acceptable supporting scheme is determined. The support frames and the main steam and feedwater lines are connected to a nu m er of structures. For the PAB and turbine building, the walls and structural penetrations are stiff compared the support frames and piping, and anchors shall be taken at those attachment points. The piping penetrations for the VC, however, are relatively flexible by comparison. For those penetrations, equivalent stiffnesses shall be calculated. In all cases, the size and thickness of the reinforcing pad on the VC shell shall be considered when calculating the equivalent stiffnesses. The relative stiffness of the piping section exiting the VC and the reinforcing pad shall be considered in determining the local stiffness of the VC shell. The connection of the main steam and feedwater piping to the VC reseeles a nozzle penetrating a vessel. The most commonly used c method to determine the stiffness of the nozzle attachment is WRC ! Bulletin 297. This method is based on empirical data for cylindrical shells. In this analysis, the VC is.a sphere rather than a cylinder. To determine a compatible stiffness for the penetrations WRC 297 shall first be used, considering the circumferential . stiffness of a nozzle on a large cylinder (comparable to the VC , . curvature). To verify and bound this stiffness, the longitudinal stiffness data of nRC 297 shall be used to determine an upper bound stiffness of the equivalent cylinder, and theoretical means shall be l used to determine a lower bound stiffness in a flat plate. These l three values shall be used to determine a reasonable value to be used at the VC boundary. l 2 YAN: 0599
l The remaining nodeling aspects of the main steam /feedwater piping and support structure model shall be as specified in the Seismic Retrofit Criteria document, DC-1, Revision 3. 2.2 Loading Conditions 7 The main steam /feedwater piping and support structure shall be analyzed for deadweight, thermal, and site specific spectral loads, including the required anchor motions associated with each of the three load cases. The intent of this evaluation is to ensure the adequacy of the system to function during and after a seismic event to attain a safe shutdown. For this evaluation, the thermal conditions for normal operation shall be used in combination with the ' applicable deadweight and seismic _ loads for evaluation of the piping and support structures. Thermal anchor motions of the anchor points shall be incluCed as appropriate. Ground and amplified response spectra shall be generated for each of the anchor points in the system model for the site specific spectra. The amplified response spectra shall be generated for appropriate elevations on the VC, PAB, and turbine building. PVRC damping (ASME Code Case N-411) shall be used in the development of the spectra. Seismic response spectrum analyses shall be performed for the system. The seismic load shall be input as independent support motions at the appropriate anchor points (multi-level response spectrum method). As suggested in NUREG-1061, the response between the support levels shall be conbined by absolute summation, and combination within a level shall be performed by square-root-sum-of-the-squares. The Reg. Guide 1.9210% grouping method shall be used to combine modal effects. Seismic anchor motions shall be considered for each building as appropriate and as specified in the Seismic Retrofit Criteria document DC-1, Revision 3. In addition to the building motions, relative seismic anchor motions of adjacent support frames shall be considered due to seismic wave propagation. To . determine the proper differential motion to be applied to adjacent supports. the seismic wave shall be propagated along the surface. Using the predicted wave length, frequency, magnitude of displacement, and the spacing of the support frames, a maximum relative displacement between any two anchor points can be calculated, as well as the distribution of the anchor motions between a series of supports. A series of maximum l 3 YAN: 0599
, , - - - -- - - - +, . . _ - - , . , - , . , . - - - - . -
,,,-.,,.,.---,-,,-n,
displacements between supports, to maximize the pipe stresses and ) support loads, shall be used in the analysis unless a reduced set of , displacements can be justified. Response from the seismic wave load ' case shall be combined with the standard seismic anchor motion ! response using the SRSS method. -. 2.3 Acceptance Criteria Acceptance criteria for the main steam /feedwater piping shall be as specified in the Seismic Retrofit Criteria document DC-1, Revision 3. The basic ASME/ ANSI B31.1 equations shall first be used in the qualification. In the initial evaluation, the piping shall be compared to the SEP criteria as follows: DW < Sh Eqn 11 DW + Site Specific Spectra < 2.4Sh Eqn 12 .. Thermal + TAM + SAM < Sa Eqn 13 DW + Thermal + TAM + SAM < Sa + Sh Eqn 14 If Equations 13 and 14 cannot be qualified with the addition of the seismic anchor motion stresses, the SAM stresses can be included in Equation 12. If the Equation 12 allowables are exceeded, the strain-based criteria and its associated requirements, as specified in the Seismic Retrofit Criteria document DC-1, Revision 3, may be used. These strain criteria and requirements apply to the seismic portions of the main steam and feedwater piping and to the non-seismic portions which have an effect on the response of the seismic piping or the frames which support the seismic piping. 3.0 MAIN STEAM /FEEDWATER PIPING SUPP(RT FRAME ANALYSIS i 3.1 Geometry and Computer Modeling The piping shall not be included in the detailed model for the evaluation of the MS/FW support frame. Tne frame model shall include all the merrbers effective in transmitting the seismic load, witn other members considered as lumped masses. The boundary of this frame is connected to the VC, PAB and Turbine Building. All these structures shall be modeled with equivalent stiffness. 4 YAN: 0599 w
i 1 3.2 Loading and Methodology The MS/FW support frames shall be analyzed for dead load , seismic load and pipe reactions. The wind load does not govern for these frames. I The seismic inertia force due to the frame weight shall be considered using the equivalent static method. It is considered as an accelera-tion equal to the peak of sne ground spectra multiplied by the mass of the frame. Since this frame is made of bolted steel mencers, and the maximum stress is expected to be close to the yield stress, the 7% damping spectra shall be used. The load conbinations shall be as follows: Case 1: DL + RT + (RX + FX) + (Ry + Fy) + (RZ + FZ) i Case 2: DL+R T + (R X + F X) + (R y + F y) - (R Z + FZ ) Case 3: DL + RT + (R X + F X) - (R y + F y) + (RZ + FZ) Case 4: DL + R T + (RX + FX) - (Ry + Fy) - (RZ & FZ) Case 5: DL + RT - (RX + FX) + (RV + Fy) + (RZ + FZ) Case 6: DL + RT - (RX + FX) + (Ry + Fy) - (RZ + FZ) s v . Case 7: DL + RT - (R X + FX ) - (Ry + Fy) + (RZ+F) Z , Case 8: DL + R T - (R X + F X) - (R y + F y) - (R Z + FZ ) Where DL = dead load due to frame weight and pipe weight RT " Pi Pe reactions due to pipe thermal load - i RX, Ry and RZ " P i Pe seismic reactions applied in the X, y l and Z directions I l FX, FY and FZ = frame seismic inertia loads in the X, y and I directions . 5 YAN: 0599 l 4
- . - - - ~ - -
\
t 3.3 Acceptance Criteria Acceptance criteria for the MS/FW support frame shall be as specified in the Seismic Retrofit Criteria document DC-1, Rev. 3, S.ection
~
'5.4. The members and connections shall be evaluated for the most critical load case as discussed in Section 3.3 of this document.
4.0 NON-RETLRN VALVE ENCLOSLRE ANALYSIS The enclosure structure around the non-return valves on the main steam lines outside the VC will be analyzed using the Seismic Reevaluation and Retrofit Criteria, DC-1, Rev. 3. , Loads and load combination will be per Sections 5.2 and 5.4.1 of the Criteria. The response spectra to be used as input loading will be
- calculated using the mode shapes and participation factors from the main steam /feedwater piping analyses.
Analysis methodology and acceptance criteria will be per Sections 5.3.1 and 5.4 of the Criteria. However, the non-return valve enclosure structure is not part of the Safe Shutdown System, and the sole purpose of the analysis is to demonstrate overall integrity of the enclosure structure such that it does not jeopardize the adequacy of the non-return valves on the main steam piping. Therefore, a nonlinear analysis method
. such as the one described in Section 6.0 of the Criteria and/or a more liberal acceptance criteria may be used to demonstrate the integrity of the enclosure structure if the allowable stresses of Section 5.4.2 are exceeded.
b i - 6 YAN: 0599
\
s( s
- ITEM F3B.III i
ISSUE
SUMMARY
PROVIDE THE FOLLOWING INFORMATION:
? :'
III. ALLOWABLES FOR THE BOLTS IF BOLTS WERE USED
REFERENCES:
(A) DC-1, REV. 3 (B) SEP GUIDELINES (C) IMPELL REPORT No. 02-0570-1204, SEISMIC EVALUATION OF THE YANKEE NUCLEAR POWER STATION REACTION INTERNALS
RESPONSE
f
, THE REACTOR INTERNALS WERE EVALUATED BY IMPELL FOR YCS SEISMIC LOADINGS. UNDER YCS LOADINGS, THE BOLTS, CONNECTING THE UPPER AND LOWER CYLINDERS OF THE LOWER CORE SUPPORT ASSEMBLY ARE AT 87% OF THE ALLOWABLE.
To CALCULATE THE BOLT STRE,SS UNDER NRC SPECTRA SEISMIC LOADS THE STRESSES UNDER YCS WERE FACTORED. THE YCS STRESSES WERE FACTORED BY THE RATIO 0F NRC SPECTR AL ACCELERATIONS TO YCS SPECTRAL ACCELERATIONS AT THE . INTERNALS NATURAL FREQUENCIES. y e
~
FREQUENCY NRC SPECTRAL YCS SPECTRAL RATIO ACCEL. (5% DAMP) ACCEL. (3% DAMP) NRC/YCS 3.95 HZ (HORIZ) 0.26G 0.16G 1.63 5.63 HZ (HORIZ) 0.35G 0.14G 2.5 6.71 HZ (HORIZ) 0.36G 0.13G 2.7 13.09 HZ (VERT) 0.95G 0.79G 1.2 NOTES: YCS WITH 3% DAMPING WAS USED BY IMPELL FOR THEIR SEISMIC EVALUATION.
~
NRC WITH 5% DAMPING WAS USED FOR THIS EV ALUATION. THE USE OF 5% DAMPING IS CONSISTENT WITH DC-1, REV. 3 AND SEP GUIDELINES. F = SRSS [(F T HORIZ-) X - )* FT ERT. INER.) TNRC x 1.2) + YCSTygg (F (VERT. IMPACT) X C)U F Tggg " 10 7 KSI
~
FT ALLOW CONCLUSION: THE BOLTS CONNECTING THE UPPER AND LOWER CYLINDERS OF THE LOWER CORE SUPPORT ASSEMBLY ARE ADEQUATE FOR-NRC SPECTRA LOADS. l l l 1
QUESTION F5d PRESSURIZER SUPPORT CONFIRMATORY ANALYSIS RESULTS t Prepared for the U.S. Nuclear Regulatory Comission Washington D.C. 20555 under DOE Contract No. DE-AC07-76IO01570 FIN No. A6808 1
PURPOSE e o TO RESOLVE QUESTION OF ADEQUACY OF 10WF49 BEAMS IN UPPER SUPPORT
' RAISED DURING AUDIT OF LICENSEE CALCULATION -
o TO CONFIRM LICENSEE'S CONCLUSION OF ADEQUACY OF SAM ELIMINATOR BASED ON CONFIRMATORY ANALYSIS t 2
ANALYSIS DETAILS o 3-D FINITE ELEMENT MODEL o LOADS DEADWEIGHT SEISMIC 3 SPECTRA WITH R.G. 1.92 COMBINATION (10% RULE) 4% DAMPING USED YCS VERTICAL SPECTRUM
- 2 (EQUIVALENT NRC SPECTRA LEVEL)
NRC HORIZONTAL SPECTRUM USED o CRITERIA SEP GUIDELINES PERTINENT DETAILS BUCKLING INTERACTION EQUATION INCLUDED BENDING ALLOWABLES INCREASED BY 1.33 WITH NO AXIAL ALLOWABLE INCREASE PER ASME CODE
- ANCHOR BOLT 4.0 FACTOR OF SAFETY WAS NOT WAIVED BECAUSE:
SUPPORTS WERE UNDER PURE TENSILE LOADING LESS THAN 4 BOLTS PER SUPPORT LOCATION OF SUPPORTS DID NOT LESSEN PROBABILITY OF CRACK FORMATION IN SUPPORTING CONCRETE 3
. . _ , y _ _ . , . _ , . . . - . - - . , --,,4. ,--. r---, ,- , . ..., -----,
THE MODEL 25 24 N2s
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ELEMENTS 4
HORIZONTAL SPECTRA RSS ELEV.1101' YC5
- 2. 4% DAMPtNG 3.2 - . . ~ . . . . . _ . _ .
3- , s i 2.s - . 2.6 - . 2.4 -
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g 1.6 d i.4 a u l 12 ; 4 1i t 0.8 -l I 0.6 - ' 1 ~- O.4 q 0.2 d 0 , , , . , , , r , , , , r-- r - T- 7 , , , , 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 PERiCD (5) VERTICAL SPECTRA RSS ELEV.1056.5' YCS
- 2. 45 DAMPtNG 1.5 1.4 - q i 1.3 - l i.2 J :
1 i . 1.1 ]' g w i -, , i l g 0.9 9 c i ' g 0.8 4.3 d 0.7 U
- V 0.6 - i 0.5 1
0.4 - j N 0.2 - 0.1 3- - 3-- T , , - 1 ~ ~ - ;- -- 1 -- - h 0 0.2 0.4 0.6 0.8 1 PERT 00 (5) 5
RESULTS f o 10WF49 RESULTS MAXIMUM INTERACTION EQUATION IS 0.7 o SAM ELIMINATOR RESULTS BUCKLING CHECK SUPPORT INTERACTION EQUATION PZR-R-1 0.97
-2 0.90
-3 0.56
-4 0.42 ANCHOR BOLT EVALUATION SUPPORT INTERACTION EQUATION PZR-R-1 0.86
-2 0.77
-3 0.49 l -4 0.29 i
6 l I a .
4 1
\
CONCLUSIONS o THE 10WF49 BEAMS ARE ADEQUATE a o THE SAM ELIMINATOR IS ADEQUATE o THE PRESSURIZER SUPPORTS ARE ADEQUATE o ALL MAJOR MECHANICAL EQUIPMENT IS ADEQUATE 7 i
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ITEM G8 ISSUE
SUMMARY
- VERITY THAT USING 40% OF SHEET METAL WEIGHT FOR APPROXIMATE WEIGHT OF INTERNALS IS ACCEPTABLE RESPONSE: THE 40% APPROXIMATION WAS USED FOR 3 CABINETS.
o NRV/CIS CABINET o WALL MOUNTED MCCS 0 MCC No. 1. BUSSES 1 AND 2 THE TOTAL WEIGHT OF THE NRV/CIS CABINET (SHEET METAL + INTERNALS) IS 450 LB/ CUBICLE. A WEIGHT OF 515 LB/ CUBICLE WAS USED IN THE ANCHORAGE EVALUATION. THE WEIGHT OF A WALL MOUNTED MCC IS 25 LBS. A WEIGHT OF 24 LBS. WAS USED IN THE ANCHORAGE EVALUATION. MCC No. 1 BUSSES 1 AND 2. WEIGHS 4532 LBS. THE WEIGHT USED IN THE ANCHORAGE EVALUATION IS 3000 LBS. ANCHOR BOLT INTERACTION FOR THE 3000 LB. WEIGHT WITH NRC SPECTRA IS 0.20. ANCHOR BOLT INTERACTION USING 4532 LBS. WITH NRC SPECTRA IS 0.30 ANCHOR BOLTS FOR MCC-No. 1 ARE 1/2" DIA. STAR-SLUG INS. A FACTOR OF SAFETY OF 5 WAS APPLIED IN C ALCUL ATING ANCHOR BOLT INTER ACTION. ,s CONCLUSION: THE 40% APPROXIMATION IS VALID FOR THE NRV/CIS CABINET AND THE WALL MOUNTED MCCS. THE ANCHORAGE FOR MCC NO.1, BUSSES i AND 2 WAS EVALUATED USING THE ACTUAL WEIGHT AND WAS FOUND TO BE ADEQUATE FOR NRC SPECTRA LOADS. t
ITEM G9III: ISSUE
SUMMARY
- VERIFY THE DESIGN ADEQUACY OF THE FOLLOWING COMPONENTS TO NRC LOADS:
III) COMPONENT COOLING SURGE TANK
REFERENCE:
YAEC CALCULATION YRC-529 RESPONSE: , THE COMPONENT COOLING SURGE TANK WAS ANALYZED USING THE NRC SPECTRA. THE ANALYSIS OF THE EXISTING CONDITION SHOWED THAT THE TANK LEGS WOULD APPROACH YIELD UNDER NRC SPECTRA LOADS. IN LIEU OF PERFORMING A MORE SOPHISTICATED ANALYSIS (I.E., TIME HISTORY. NON-LINEAR ANLYSIS. ETC.) WHICH COULD QUALIFY THE TANK YAEC WILL INSTALL L3 x 3 x 1/4 DIAGONAL BRACES (SEE NExT PAGE). CONCLUSION: WITH DIAGONAL BRACES INSTALLED. THE COMPONENT C00LLING SURGE TANK WILL MEET DC-1, REV. 3 CRITERIA WHEN SUBJECTED TO NRC SPECTRA LOADS. e
C0WONENT C0OLING SURGE TANK MODIFICATION our se4CEMYM56106(
.. )
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,5 % "dia. boIt w/nutc wa. sher (mfe 2) A325 /4 t
NOTES
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goces.
- 2. the /ntde'nedsfee/was/er. Bo/l ho/e.s lo k '%"
o'ta. Fie/ddri//k//kles in 23x3. J. (Ase E70, egga$orither eiectrcde 16ra// ne Mmg. 4 Rowok I'mik e/earance &dween d2nk s/x//and braces. . 5 Seve/ a // corners of 45x 3s. 4
ITEM H5 Issue Summary . Demonstrate that the flexible hose used to connect the buried piping to the Fire Tank is adequate for NRC seismic loads. References (1) Cygna calculation 85005/1-F, SET CES 2 REV.'2
Response
The hose manufacturer was contacted to obtain the maximum allowable offsets.
- The following numbers were received:
, 4"$ Hose 21"$ Hose 47" Long 34" Long Allowable Axial Movement 2" 2" Allowable Transverse Movement 4" 4" These numbers are a minimum of 5 times greater than the expected YCS anchor motions (see Ref.1). It is expected that NRC anchor motions would not exceed approximately 2 times the YCS valuec. Therefore we conclude that the hose is adequate for YCS and NRC seismic loads.
ITEM H12B ISSUE
SUMMARY
- AUDIT OF THE BURIED PIPING CALCULATI0]lS RAISED A QUESTION CONCERNING THE VALIDITY OF THE FRICTION EQUATION USED ON PAGE 22 (REF.
[1]). THE EQUATION SHOULD BE JUSTIFIED.
REFERENCES:
[1] CYGNA CALCULATION BINDER 85005/1F, CALC. SET No. CES-2, e REV. 2. DATED SEPTEMBER 1985. [2] CYGNA CALCULATION BINDER 85005/1F, CALC. SET NO. CES-2, REV. 4. DATED DECEMBER 1986. [3] "SEP TOPIC II-2-4.D. SLOPE STABILITY YNPS", LETTER FROM R. CARUS 0 (NRC) To J.A. KAY (YAEC). DOCKET NO. 50-29. LS05-82-09-009. SEPTEMBER 2, 1982. [4] YAEC SPECIFICATION No. YRS-005, SECTION 02225.
RESPONSE
A VALUE OF ju= 0.8 AND A" EQUATION VALIO ONLY FOR COHESIONLESS SOILS WERE USED ON PAGE 22 0F REF. [1] TO CALCULATE THE FRICTION FORCE BETilEEN SOIL AND PIPE. NRC CONSULTANTS REQUESTED A JUSTIFICATION OF THE VALUE OF 4( AND A VERIFICATION OF THE COHESION 0,F THE BACKFILL SOIL. THE BACKFILL SOIL IS COHESIONLESS PER REF. [4] AND THEREFORE THE USE OF THE EQUATION IS APPROPRIATE. i i n
. - - . -- , - .,,.___~
THE ACTUAL VALUE OF IS OBTAINED FROM [ = T AN ( - ())
WHERE 0.76 BETWEEN DRY SAND AND ROUGH. STEEL f). (0.80 BETWEEN SATURATED SAND AND ROUGH STEEL,
()= ANGLE OF INTERNAL FRICTION
=
400 FOR THE YNPS SITE (SEE REF. [33). SINCE THE SOIL CAN BE DRY OR SATURATED s jdd= TAN (0.76) (400 ) . 0.59 THIS VALUE WAS USED AND THE CALCULATIONS WERE REVISED IN REF. [23. SOME SAMPLE RESULTS OF THE REVISED (REF. [23) CALCULATIONS ARE SUMMARIZED AND COMPARED WITH PREVIOUS (REF. [13) RESULTS AS FOLLOWS: (1) PIPE STRESSES AT ELBOWS (ELBOWS GOVERN) STRESS. PSI PIPE PREVIOUS (REF. [il) REVISED (REF. [2]) ALLOUABLE 2" STAINLESS STEEL 28490 29673 33660 1 1/4" STAINLESS STEEL 29168 30535 33660 2" CARBON STEEL 27171 24853 27000 2 1/2" CARBON STEEL 19366 20300 .27000 1" CARBON STEEL 24231 25710 27000 4" CARBON STEEL 25907 26990 27000 (2) FLEX. HOSE DESIGNS ARE STILL 0.K. SINCE REVISED (REF. [23) AXIAL MOVEMENTS ARE LESS THAN DESIGN AXIAL MOVEMENTS.
~
(3) ANCHOR LOADS AT BUILDING PENETRATIONS: AXIAL SEISMIC LOADS, KIPS
, NPS. IN PREVI0 tis (REF. [11) REVISED (REF. [21) 1 3.6 2.7 1 1/4 4.5 3.3 ,
2 6.5 4.8 4 4.9 3.6 e iba3;: ALL OF THE ABOVE RESULTS (REF. [1] AND REF [23) ARE BASED ON CONSERVATIVE MAXIMUM GROUND ACCELERATION AND VELOCITY VALUES OF Ag = 0.274G AND Vn = 13.152 IN/SEC, RESPECTIVELY. q THE ACTUAL VALUES ARE Ag = 0.1G AND Vn :l 4.8 IN/SEC FOR THE j YCS. RESPONSE IS ALMOST DIRECTLY PROPORTIONAL TO VM , AND THEREFORE ABOVE RESULTS ARE CONSERVATIVE BY A FACTOR OF ABOUT 2.74. CONCLUSION: REVISING THE COEFFICENT OF FRICTION FROM 0.80 TO 0.59 DOES NOT AFFECT THE QUALIFICATION / DESIGN OF BURIED PIPING, FLEX. HOSES, BUILDING PENETRATIONS. THE PREVIOUS CONCLUSIONS / DESIGN (OF REF. [13) ARE STILL VALID AND THE BURIED PIPING AND ASSOCIATED HARDWARE ARE ADEQUATE. s
% 0 4
ITEM H12c ISSUE-Si!MMARY: VERIFY THE AXI AL FORCE S. C ALCUL ATION (ON PAGE 29 0F REF. [13). -
~.
REFERENCES:
[1] CYGNA CALCULATION BINDER 85005/1F CALC. SET No. CES-2 REV. 2. DATED SEPTEMBER 1985. 2 [2] CYGNA CALCULATION BINDER 85005/iF. CALC. SET NO. CES-2. REV. 4. DATED DECEMBER 198G. RESPONSE:AN ERROR APPEARS ON PAGE 24 0F REF. [1] IN THE CALCULATION OF THE AXIAL FORCE S. THIS ERROR WAS CORRECTED In REF. [2]. THE CORRECTION HAS NO ADVERSE EFFECTS ON RESULTS. FOR A
SUMMARY
OF REF. [2] RESULTS. SEE RESPONSE TO ITEM H128. l i l l
fW hWl'b'MI ISSUE
SUMMARY
Provide assurance that adequate arec of reinforcement exists for branch piping. REFERENCES
- 1. YS-497, Specification for Piping for Yankee Atomic Electric Plant, by s
Stone and Webster Engineering Corporation, July 1959.
- 2. YS-4652, Specification for Shop and Field Fabricated Nuclear Piping for Revised Arrangement of Safety Injection and Shield Tank Cavity Water Piping for Yankee Atomic Electric Company, August 1970.
RESPONSE
All piping at YNPS has been designed in accordance with the above specifications, which reference ANSI B31.1-1955, and ANSI B31.7-1969 for design and fabrication requirements. Specific limitations for different pipe sizes and classes meeting these codes are contained within the individual sections of the Specifications. i f a.--. _ . - _ . - - --- . - - . - - - - - , , - . , , - , . -
/V{W uffbiW ISSUE
SUMMARY
Loads on gang supports from nonseismic safety-related piping should be justified.
RESPONSE
The follo' wing supports fall into the scope of this question. Gang Supports In Scope Support Number Line Numbers Vertical Supports Added?
- 1. SP-SG-7-H6 3/4" VRH-2502-3 No
- 2. SP-SG-12-H7 2" WBG-601-7 Yes, one 3
- 3. SP-BD-1-H1 2" WGB-601-5 No 2" WGB-601-3 No -
2" WGB-601-1 No
- 4. SP-BD-1-H2 2" WGB-601-1 No 2" WGB-601-3 No 2" WGB-601-5 No
- 5. SP-BD-1-H3 2" WGB-601-1 No 2" WGB-601-3 No 2" WGB-601-5 No
- 6. SP-BD-1-H4 2" WGB-601-3 No 2" WGB-601-5 No
- 7. SP-BD-1-H5 2" WGB-601-3 No 2" WGB-601-5 No
- 8. SP-BD-1-H6 2" WGB-601-5 No 2" WGB-601-3 No
- 9. 9699-MP-144-AHS 2" CRBH-2502-2 No Piece MK No. CRBH-RH4 1-1/2" CRBR-2502-5 No SP-FB-2b 1-1/2" CRBH-2502-3 No
- 10. 9699-MP-144-AH6 2" CRBH-2502-2 No Piece MK No. CRBH-RHS 2" CRBH-2502-3 No SP-FB-2b 2" CRBH-2502-5 No The methodology used to estimate loads on gang supports due to nonseismic piping was to use the loads reported for the seismically analyzed lines on that support and ratio the load in proportion to the relative pipe weights.
This methodology is based on the assumption that the piping was hung in a similar fashion. Since both seismic and nonseismic lines were originally supported by these gang hangers for dead weight load only, the difference between the two is the addition of new supports for seismic loads. A review of the new supports for the seismic piping in this scope indicates that, with
. one exception *, all of the new supports were added to provide lateral restraint. We conclude then that for these lines the vertical support of the seismic and nonseismic lines are similar, and the estimation of the vertical forces imposed on the support by the nonseismic lines is adequate. No lateral loads were imposed on the supports by the nonseismic lines since, without exception, they support the lines for dead weight only.
*The one line which added a vertical restraint is 2" EBF-601-25 which is supported in part by Gang Hanger SP-SG-12-H7. However, the new support replaces a nearby existing vertical support which is being removed as part of the seismic upgrade. Therefore, the vertical support for this line is virtually unchanged.
s 9 e l l l L
I
~
ENCLOSURE C
SUMMARY
OF ISSUE DISCUSSIONS The followino is a summary of the results of issue discussions concerning the Yankee Atomic Electric Company's seismic upgrade of the Yankee Rowe Nuclear Plant. The summary has been organized to make use of the licensee's question matrix. Topics which have not been included in that matrix are discussed in Enclosure D. A. GENERAL QUESTIONS , A1, A2 The successful completion of the PAB confirmatory analysis, combined with a review of related structures for the one short-coming identified in the analysis (see discussion under 81 below), ' has led to a conclusion that the YCS analyses performed for the plant structures are acceptable. CLOSED. B. EVALUATION CRITERIA B1, Bla, C1c The question concerning ultimate soil bearing capacity was essentially resolved during a telephone conference involving the licensee and NRC geotechnical staff. This question will be resolved when the licensco has docketed the analysis results discussed during the telephone conference. 84.111,B8.1, C5m In connection with the confirmatory analysis of the PAB, the licensee has committed to redesigning the details of the steel to concrete connection at the second floor elevation of column line 8G. This action is anticipated to be incomplete at the end of the review. There are no other outstanding questions concerning the PAB confirma-tory analysis. B4.iv, C5f The licensee has established, based on a review of the turbine building design, that it included no Hilti anchor bolts. Because of this, the shortcoming identified in the PAB confinnatory analysis generated no concern for the turbine building design. The licensee provided justification to demonstrate that the soil-structure inter-action effects are negligible to the response of the turbine building under the earthquake loading and it is acceptable to the staff. .. Therefore, the original seismic evaluation results are adequate. CLOSED. 816 The one application of ASME Code Case N-284 to a building-founded tank, the boron injection tank, was subjected to a case-by-case review and found acceptable. CLOSED. Structures C5a Based on the positive PAB confirmatory analysis results, the licensee's use of 7% damping in the YCS analysis of the PAB was found acceptable. However, the adequacy of amplified response spectra generated besed on the YCS analysis will be confirmed when the safe shutdown piping confirmatory analysis is completed. i e
. . , - . - . , . , , , , . - - - - - , - . __ - . . , , . - , . , . ~ , -- _ , . - - - - - , . , - . . . - , ,- , . , , - . - - - - -
C6a Confirmatory analysis of the fire water tank has established that the tank wall lacks sufficient margin to resist shear buckling failure due to seismic inertia loads of the tank alone. Nozzle loads were not considered, but will be small because of the bellows installed in the attached piping. The tank base anchor bolts and freeboard space are acceptable. The inadequacy of the tank wall under shear induced buckling will be identified in the
, Safety Evaluation Report (SER).
C7, C8b, 12a Confirmatory analysis of the PA8 north masonry wall established two action items:
- 1. The strap anchor located at the 1039 ft elevation is acceptable only if the licensee can establish that the anchor has been grouted. An acceptable means of accomplishing this must also be developed. t
- 2. Inclusion of PAB seismic anchor motion effects into the wall calculation resulted in a prediction of both horizontal and vertical cracking of the masonry. Because of this, the
- 1 licensee must establish acceptable criteria for evaluating fully cracked masonry walls. Such criteria would then be used 2
to establish acceptability of the wall. C8a The licensee's evaluation of the accumulator tank established that it will not collapse and therefore could not adversely affect nearby SSS structures or components. CLOSED. I C8c.iii The licensee's evaluation of the elevator tower established that i it will not collapse and therefore could not adversely affect ! nearby SSS structures or components. CLOSED. C8d Review of the licensee's proposal for evaluating the MS/FW piping and support structure has established that the evaluation of the 4 NRV enclosure will be included in that body of work. Since the evaluation of the MS/FW piping and support structure has been postponed until after the issuance of the SER related to this j review, this item will appear in the SER as an open item. i pipino Systems p .. D2.1 An adequate description of the criteria proposed for evaluation of pipe supports has been included in the draft of the DC-1 criteria document. CLOSED. D2.2 A sample of the document that will record the list of pipe support I calculations has been reviewed and approved by the NRC staff. This question will be resolved when the licensee has docketed the final document. t l 2 i l
03a The derivation of an acceptable factor for increasing YCS lateral displacements in evaluating rod hangers cannot be completed until the example analyses are completed (see question D38). In addition, the licensee must provide displacements from the YCS calculations for supports WGB-H-62, -63, -64, WGB-A-3, and -4 (piping problem No. 10) for comparison to the NRC spectra values. D15, E25 Evaluation of proposed criteria for pipe support anchor bolt evaluation has been deferred until issues raised in the NUREG-1030 review concerning acceptability of interaction equations have been resolved. This is not expected to occur until after the SER is issued. t D20a, D20b The licensee has resolved questions concerning the adequacy of spectra used in the loop 4 piping confirmatory analysis. This included presentation of YCS spectra generated using the refined reactor support ring model. However, since the design adequacy of the reactor support against sliding is still uncertain and will be reevaluated during the next refueling outage, the adequacy of the amplified responsa spectra used for the loop 4 analysis will be confirmed when the reevaluation is completed. D20c The licensee's evaluation of the potential of the steam generators to uplift yielded a prediction that minor uplift will occur. With one exception, the licensee's evaluation established that the sup-port structure is adequate under the new loadings. The exception is that a dynamic load factor associated with the impact predicted to occur was not considered in the evaluation. This needs to be done. D21, D22 Discussion of preliminary results of the confirmatory analysis of SSS piping in the PAB established that they are based on questionable data. Specifically, the confirmatory analysis models contain valve weights, based on a survey of vendor data, that are significantly larger than those used by the licensee. The licensee has provided valve weights based on plant specific information (weights of spare valves and vendor supplied information for specific valve models) which will be included in a revised model. In addition, discrepancies in boundary conditions will be resolved based on additional information , provided by the licensee. A discussion of the issues raised concer-ning data applied in confirmatory and example analyses is included in Enclosure D of this document. The final conclusions of the con-firmatory analysis will be based on the results of the revised model. 024 Resolution of questions D24a and 024b represents a successful re-solution of issues raised by the case-by-case review of the use of WRC-297 nozzle flexibility in analyzing piping attached to the boron injection tank. This application of WRC-297 is acceptable. CLOSED. i 3
024a, 024b The licensee established that the change in nozzle stiffness j associated with consideration of intermediate supports was small (approximately 16%), and demonstrated a margin against exceeding allowable stresses of 8.0 to accommodate this effect and the potential error associated with applying a cylindrical methodology to a spherical geometry. This was found acceptable by the NRC staff. Acceptance was based on the fact that the difference in stiffness between the thin walled tank and the short, stiff connecting piping was such that the stiffness of the nozzle could increase substantially without affecting the results. The piping controls the rotational motion of the nozzle. This, in combination with the demonstration of large margin, was the basis of acceptance. CLOSED. 024c The licensee attempted to evaluate nozzles on the side of the boron injection tank using the WRC-107 methodology, but found that i geometrical parameters of the nozzles exceeded WRC-107 limits. 4 Because of this, the licensee requested a case-by-case review of the applications. The NRC staff agreed to do this.,The 4 case-by-case review remains to be done. 1 l 026d The licensee has agreed to docket the basis for use of strain criteria in piping evaluation. This will resolve the question. 026g Previous work has established that 1983 ASME criteria application j with YCS assures that ANSI-B31.1 requirements would be met. This places the piping calculations done to this criteria with the population of calculations whose acceptability is to be determined by confirmatory analysis. Therefore resolution of the confirmatory 4 analysis questions will assure that these calculations will be properly handled. This question is resolved. CLOSED. 032.i The licensee has connitted to placing pipe supports with u-bolts credited with axial restraint in the population of plant components whose field condition is periodically verified. This will assure that the preload necessary to maintain axial restraint capability will exist for the life of the plant. CLOSED. 034 An audit of licensee calculations has established that asymmetric bending of pipe support members has been properly considered in , i all past calculations. A revision to the DC-1 criteria document > l assures that all future calculations will properly consider
- principal axes of asymmetric members. CLOSED.
037 The audit activity discussed under question 034 above also provided assurance that small-bore pipe supports do not include welds to the pressure boundary. This resolves the concern about the criteria for such welds found in the procedure that governed ! the small-bore pipe support calculations. CLOSED. 1 i i 4 i
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D38 Discussion of preliminary results of the example analyses l established that they are based on questionable data. Significant i differences were found in the estimates of valve weights which ! lacked definitive information. Also, portions of the models reflected as-built conditions, while the licensee's models included additional supports necessary to achieve acceptance. The licensee has provided valve weights based on plant specific information (weights of spare valves and vendor supplied information for specific valve models) which are being reviewed prior to inclusion in revised models. In addition, the support configurations used in the licensee's models have been provided for the revisions. A discussion of the issues raised concerning data applied in confirmatory and example analyses is included in Enclosure D of this document. The final conclusions of the example analyses will be based on the results of the revised models. 9 MS/FW piping and Support Structure E23 An audit of Cygna calculation No. 86063/21/f was performed to -
- evaluate the licensee's basis for u-bolt allowable loads. Two questions were raised
- 1. The calculation made use of a stiffness based on the assumption that point contact exists between the u-bolt and the pipe, so that the curved portion of the u-bolt contributes to its flexibility. This assumption was questioned, primarily because the calculation also included credit for axial restraint via friction resistance. The licensee was requested to demonstrate that the results did not significantly change with consideration of the potential for the existence of a much larger contact zone than postulated, such a zone being the result of the preload necessary to establish the frictional resistance necessary for axial restraint.
- 2. A review of the docket was performed to establish an acceptable basis for evaluating the licensee's proposal to credit u-bolts with axial restraint. Such a basis was derived, based on documents generated by an inspection of the i Salem Nuclear Generating Station (see correspondence related i to IE inspection reports 50-272/84-05 and 50-311/84-05 dated
; April 9, 1984, and June 15,1984). Evaluation of the licensee's document using this. basis established the following shortcomings:
2.1 An acceptable basis for demonstrating the adequacy of criteria allowing axial restraint credit for u-bolts must include correlation to applicable test results. No such results were included in the licensee's document. Such correlation is necessary to establish the margin maintained by the criteria.
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2.2 Criteria allowing axial restraint credit for u-bolts must preclude the occurrence of unacceptable local stresses in the associated pipe wall. Acceptable pipe wall stresses are defined according to the applicable s piping code (ANSI-831.1-1977 or ASME Section III, s Winter, 1980 Addenda). - ~ s The final requirement of the review, that a means of ensuring that the requisite preload on the u-bolts exists for the life ' of the plant, has been met by a areventative maintenance commitment made by the licensee. This is discussed under 4 question 8311. E24b The licensee's proposal to allow factors of safety for pipe support expansion anchor bolts of between 2 and 4 was accepted for interim operation based on the provisions of IE Sulletin 79-02. For long term operation, the full factor of safety of 4 is expected to be met. CLOSED. n Major Mechanical Components F3b.iii The licensee has demonstrated the adequacy of reactor internals bolts under NRC spectra loads. CLOSED. ' F4f, F8c Scheduled audit of the steam generator and feed and bleed internals calculations were cancelled to allow more time to pursue issues raised by the piping example analyses. The cancellation does not adversely impact the review for two reasons. First, the presentation of results of these calculations made during the November meeting contained a fine level of detail and indicated large margin in the internals designs. Second, case-by-case
- reviews have already been performed for all the major mechanical components to establish acceptability of using SEP Guidelines criteria with YCS (see question Fib). The methodology used in the 3
internals calculations has been adequately reviewed in conjunction 1 with the case-by-case reviews. CLOSED. > F5d \ Review of the results of the pressurizer confirmatory analysis i raised a question concerning the validity of horizontal spectra used in the analysis. Shortly after the meeting, the spectra were reviewed and found to be in error. Essentially, NRC horizontal *
- spectra were doubled, a process applied to YCS to obtain an acceptable equivalent for NRC spectra. This resulted in excessive conservatism in the calculation. Stress results were revised to reflect correct horizontal spectra, and the conclusion drawn that the pressurizer supports are acceptable under NRC spectra loadings. This is in agreement with the licensee's confirmatory analysis results. The handouts contained in Enclosure B of this i
l document have been revised to reflect the correct spectra. CLOSED. 6
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Electrical and other Mechanical Equipment' l
, G8 The licensee has established the actual weights of the contents of all cabinets originally qualified using a content weight based on 40% of the sheet metal in the cabinet. In all cases but one, the estimate was conservative. For the exception, the licensee reevaluated the cabinet using the actual weight. It was shown to be acceptable. CLOSED.
< . G9.iii The licensee's NRC spectra evaluation of the component cooling surge tank indicated that the tank supports are not acceptable. -
Rather than performing more detailed analysis to remove . conservatism, the licensee opted to modify the tank supports. The preliminary design was reviewed at the meeting and appeared acceptable. Installation of the modification is anticipated to be incomplete at the time of publication of the SER, so that ensuring acceptable installation of the modification will be an open item there. s ( G3.iv Audit of Yankee anchorage calculation No. YRC-405 established that the YCS evaluations of the anchorages for the D. C buss and the scram breaker cabinet did not establish sufficient margin to allow l a conclusion of adequacy under NRC spectra loading. In response to , s this concern, the licensee reviewed the YCS calculations and s established that there was sufficient conservatism in the -
' calculations to ensure acceptability under NRC spectra loading.
CLOSED. Dedicated Safe Shutdown System (DSSS) H5 The licensee has demonstrated that the flexible hose connecting DSSS piping to the fire water tank is adequate under YCS loading with sufficient margin to assure acceptability under NRC spectra loading. CLOSED. H12b The licensee responded to a concern raised about the validity of the friction coefficient used in buried pipe calculations by revising the calculations to a lower coefficient which the staff found acceptable. The revision demonstrated acceptability with a margin of 2.7. CLOSED. , H12c The typographical error found in the axial force calculation for l buried pipe was corrected during the revision discussed under l question H12b above. CLOSED. i i Masonry Walls '
, Ild.1 The licensee proposed to justify the 0.001 in-plane strain limit for confined masonry walls by calculating the stress associated with this level of strain and comparing to allowable stresses established using the Standard Review Plan (NUREG-0800). This was
- acceptable to the NRC staff, with the understanding that the t
stress calculations would require review and approval before the
,^l Justification was accepted.
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ENCLOSURE O ITEMS NOT RELATED TO EXISTING ISSUES The following items, discussed during the meeting, have not been incorporated into the question matrix.
- 1. The'following items were generated during the course of the example analyses.
1.1 Adequacy of Decoupling Criteria Review of the isometrics provided for the example analyses of g problem Nos. 1 and 10 revealed a potential flaw in the basis for accepting the decoupling criteria proposed by the licensee for the analysis of piping. This criteria allows application of floor spectra at the branch points in the branch piping analysis without consideration of the rigidity of the run piping through which floor motion is transmitted to the branch points. Review of the isometrics indicated that the implicit assumption of rigidity of the run piping may be questionable. This concern was addressed by the example analyses, where flexibility of the run piping was considered in defining the model boundaries. Results of these analyses were compared with the results of analyses that had boundary conditions matching those used by the licensee. The comparison established that nonconservative results can be obtained, and were obtained in the licensee's piping problem No.
- 1. Stresses from the full size model (including run piping) ranged as high as nine times the reduced (branch piping only) model. This had a significant impact on the level of overstress predicted. The full size model had a maximum low cycle fatigue usage factor cf 18.3, versus a usage factor of 1.7 for the reduced model. TM hmparison for problem No. 10 had full size model stresses from ..e to 2.4 times those of the reduced model.
However, the differences in results were isolated to the piping adjacent to the branch point, and did not effect the maximum stresses, which occurred elsewhere. Due to the limited time available to finish this review, this topic will be an open item. in the SER. 1.2 Adequacy of As-Built Drawings Review of three as-built isometric drawings (Nos. 9699-FP-260,
-26E, and 44E, Revision 1) generated several cases of erroneous or missing data. This was an error rate sufficiently high to generate concern about the adequacy of all as-built drawings. The licensee was requested to address this concern. An initial review identified two sources of error. The first source was the process of transferring the field data (8-1/2x11 sheets) to full size drawings. The licensee has made a commitment to review the full size drawings to correct any errors found. This work will not be completed before the end of the review. The licensee found the remainder of the discrepancies to exist on the field drawings, but found them to be insignificant, all being within the 1
tolerances specified for the walkdowns that generated the 8-1/2 x 11 drawings. The NRC staff will compare the drawings associated with the example analyses to corroborate the licensee's findings. Further, the verification of the field drawings did not include corroborative walkdowns, a result of the fact that the subject piping is in containment, and not available for inspection. All dimensional discrepancies resulted from a comparison of ANSI Standard component dimensions with field measurements, so the discrepancies found would necessarily be small. The concern that larger dimensions could be in error can not be verified without a walkdown. This will not be done prior to the finish of the review, so it will be identified in the SER as an open item. 1.3 Loads for Gang Support Analysis In a November 17, 1986 telephone conference, the licensee indicated that loads on gang supports from non-safety related piping are estimated based on the loads from safety-related piping. Tile licensee verified that this is normal practice, and was requested to provide justification for its use. This will not be completed prior to the end of the review, and will appear in the SER as an open item. Future documentation of this topic will be recorded under question D39. 1.4 Reduced Allowable Loads for 1/2" Hilti Kwik-Bolts As discussed in Information Notice 86-94, Allowable loads for 1/2" Hilti Kwik-Bolts have been revised. The new allowable loads are lower than the original loads for some embedment values. The NRC staff's example analyses will incorporate the new allowables. The licensee was requested to review the Notice and respond to the concern that the lower allowables may make some of the Yankee Rowe baseplate designs unacceptable. The licensee's response was that all such bulletins are reviewed as a matter of course and corrective action will be initiated by the review if necessary. Since this review is not anticipated to be completed soon, this item will appear in the SER as an open item. Future documentation of this topic will be recorded under question 040.
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1.5 Assurance of Adequate Area of Reinforcement for Branch Fittings ^ Based on a similar issue raised in the Diablo Canyon review, the licensee was requested to identify the measures taken to ensure l that branch fittings have adequate area of reinforcement. The licensee responded by stating that the plant had been constructed to ANSI B31.1-1955, which contains requirements for area of reinforcement identical to those in current standards. As further evidence, the licensee agreed to verify that a sample fitting does have adequate area of reinforcement, specifically, the fitting which generated the question, a Sockolet at the branch point of 3/4"-DRH-2502-17 from 2"-CRCH-2502-3. This is scheduled for completion before the SER is issued. Future documentation of this topic will be recorded under question 041. 2
1.6 Estimation of Weights for Valves Lacking Definitive Information During the course of the. example and confirmatory analyses, several valves were attributed weights less than expected. All 2" valves in the example analyses had weights of 25 pounds specified. A check of vendor literature for 2" valves of the specified pressure ratings yielded weights of 30 to 80 pounds. Valves VD-V-1157, VD-TV-401A, -4018, -401C, AND -4010 in the confirmatory analysis of the OSSS piping in the PA8 also had weights of 25 pounds specified. Based on a similar review of vendor literature for the specified pressure rating, a weight of 90 pounds was estimated. The licensee stated that weight specifications for all of these valves were based on estimates, since definitive information was not available. The preliminary 4 NRC staff deadweight analysis of the SSS piping in the PAB indicated that the support spacing is not adequate for deadweight. The licensee's analysis indicated that the spacing is adequats. This difference in results was attributed primarily to the difference in valve weights used. Because of this, the licensee was requested to verify the valve weights for which estimates had been used. The licensee proposed to contact the appropriate valve vendors to obtain weights for given sizes and model Nos. of valves where possible, and to weigh spare valves where possible to determine if the concern is justified. This effort is complicated by the fact that the example analysis varives are in containment, s6 that manufacturer /model No. information was not available. It was necessary to defer action on this concern until entry to containment becomes available. This item will appear in the SER as an open item. 1.7 Use of Correct Spectra and SAMS i The example analyses established two shortcomings in the piping analysis methodology. First, there was apparently no plant coordinate system established for the analysis of buildings or piping. Second, SAM values were either specified in absolute or relative terms. Both of these shortcomings increased the i likelihood of errors in obtaining the correct data for analysis. To address this concern, an audit of piping calculations was . conducted during the January meeting. Three calculations were ! audited, TW-042, No. 6, and No. 11. These calculations were chosen to provide a variety of boundary conditions, including attachment to the RSS, attachment to the VC, and attachment to other piping. The audit established that no errors occurred in specifying spectral direction or in SAM application due to relative / absolute interpretation. Two factors were identified that contributed to the generation of the issue. First, althou e each analyst had the option of arbitrarily defining coordinate systems, the plant spectra were all correlated to absolute , directions: North, South, and up. The one exception to this was the spectra associated with the main coolant loop (MCL). Only one l of the loops was analyzed, and it was oblique to the N-S-up coordinates. There was no commentary included with the spectra to aid the analyst in defining directions for the other three loops. The licensee was askoc about this potential source of error and l .
l I 4 I responded that the standard practice was to envelope the two MCL ) horizontal spectra, and to apply the envelope to the two horizontal directions in the analysis. Two of the three calculations audited had this feature. The second factor arose from the fact that all SAMs were relative to ground except those associated with the MCL spectra. This was not identified as a serious problem, because a mistake by the analyst in applying MCL SAMs would result in conservative results in all cases except for piping that wa:: routed directly from the MCL (including major mechanical equipment) to the VC. In this case, the RSS SAMs would have to be combined with either the MCL or the VC SAMs for correct results. However, all MCL piping and major mechanical equipment is surrounded by radiation shielding, which consists of monolithic concrete structures contiguous with the RSS. This m makes it unlikely that any piping would have such a routing. In the one occurrence of an applicatio'n of MCL SAMs found in the audits, they were correctly applied. CLOSED. . 1.8 Acceleration / Damping Mismatch in Spectra The example analyses established several instances where the accelerations specified at a particular frequency for a lower damping were smaller than accelerations for a higher damping at that frequency. Although the discrepancy was in the fourth and fifth digit of the numbers, the licensee was requested to explain how it could occur, to preclude the possibility of significant occurrences of the discrepancy in unaudited spectra. The licensee's response (letter, M. Schulman (Cygna) to M. Russell (INEL), subject: Response Spectrum Generation Algorithm, Yankee Rowe SEP, Job No. 86864, dated December 23, 1986) identified error as a result of the discretized algorithm used to generate spectra. The error occurs because of the fact that the actual maxima of a continuous acceleration time history will typically not fall on one of the discrete points usad to characterize 4 spectra in digital computers. The algorithm estimates the actual maxima based on discrete values in the near vicinity. This estimation is the error source. Because of this, the error is i dependent on the interval used to define the discretized spectra. The interval used by the licensee was very small (0.005 sec), so that the error can also be expected to be small. This was supported by the fact that the noted discrepancies were small, and that they all occurred in zpa areas of the spectra, where spectra for different damping values are nearly identical. Review of the licensee's response established that the claim that the error is negligible is correct. CLOSED.
- 2. The staff provided cormients to the final draft of Yankee Plant Seismic Reevaluation and Retrofit Criteria, DC-1. The licensee will incorporate the staff's comments and femally submit them to the staff when revision of DC-1 becomes available.
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