ML20212M030
| ML20212M030 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Diablo Canyon |
| Issue date: | 12/01/1977 |
| From: | Allison D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20150F500 | List:
|
| References | |
| FOIA-86-391 NUDOCS 8608250339 | |
| Download: ML20212M030 (41) | |
Text
- -
-.=
QI i
DEC 011977
-P DOCKET NOS:
50-275 AllD 50-323 APPLICANT:
PACIFIC GAS AND ELECTRIC COMPNtY (PG&E)
FACILITY:
DIABLO C#1 YON NUCLEAR POWER STATION, UNITS 1 #1D 2 (DIABLO CANYON)
SUMARY OF MEETING HELD ON NOVEMBER 10, 1977, TO DISCUSS DIABLO CANYON SEISMIC DESIGN - REACTOR COOLANT SYSTEM FRACTURE MECHANICS ANALYSIS We met with the applicant on November'10,1977, in Bethesda, Maryland, to discuss a fracture mechanics study related to the Diablo Canyon seismic design.
A list of attendees is provided in Enclosure No.1.
PG&E was reanalyzing the plant's seismic design capabilities to detemine what modifications might be necessary to withstand an earthquake with a reference horizontal acceleration of 0.75g (as opposed to an original value of 0.4g).
In Amendnent 50 to the operating license application, PG&E had sub-mitted the results of an analysis indicating that the reactor coolant g"
system would meet the stress limits of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for earthquake loads combined with nomal operating loads. This provided a high degree of assurance that an earthquake at the design levels would not cause a loss-of-coolant accident and, accordingly, PG&E had believed that this would be sufficient for the seismic design reevaluation. We had indicated that, nevertheless, an analysis i
of the reactor coolant system should be perfomed combining the cal-culated earthquake loads with the calculated loads from a postulated loss-of-coolant accident (LOCA). This subject had also been dis-cussed at an Advisory Ccanittee on Reactor Safeguards (ACRS) Sub-committee meeting on June 21-23, 1977.
t Westinghouse had perfomed a fracture mechanics analysis to investi-i i
gate further the effects of earthquake loads on the reactor coolant l
system. Westinghouse described the analysis to us at the meeting.
The description is sumarized in the slides presented at the meeting (Enclosure 2). A more detailed report on the analysis was to be submitted for our review in Amendnent 56 on November 14. 1977.
I l
t ome s
- 82 9 860001 HOUGH 36-391 su==* = e >
-l-oave >
I Imc PORM 318. (9 76)- NRCM 0240
- vse.oovunamew '
~~
F1::EA & -J9/
. A -97 A -97
a
(
DEC 011977 t
The scope of the analysis was limited to the large reactor coolant system main loop piping where a postulated pipe rupture could be similar in effect to a design basis LOCA.
The results were cast in tenns of what it would take to cause a pipe break or rupture rather than a small leak which would have different effects than a design basis LOCA.
The fracture mechanics analysis had indicated that, in order to have an earthquake (at 0.4g or 0.759) cause such a pipe break, there would have to be a very large pre-existing flaw in the main loop piping at one of the most highly stressed locations. Alternately expressed, the probability of such an carthquake causing a pipe break was very small.
Earthquake loads had been combined with normal operating loads in the analysis. Due to conservatisms in the original i
design and the model frequencies involved, the original design earth-quake loads at 0.4g had been more limiting than the reevaluation earthquake loads at 0.759 for the purpose of the analysis.
The fracture mechanics analysis had quantified the applicant's belief that an earthquake at the design levels was so unlikely to cause a LOCA that earthquake loads and LOCA loads need not be combined in the reevaluation of the reactor coolant system. The analysis was also responsive to questions that had been asked at the ACRS Sub-committee meeting in June 1976. Nevertheless, the applicant was I
' planning to submit the results of a reactor coolant systen stress analjais with earthquake loads and LOCA loads combined as we had reques ted.
Oddn:KW pennis P ?$,
Dennis P. Allison, Project Manager Light Water Reactors Branch No.1 i
Division of Project Management
Enclosures:
i i
1.
List of Attendees 2.
Slides cc: See page 3 i
i t
LWR-gk LWR-h.
DPA1Mson:sjh JFStol j
o.ve >'
l.1/22/.17-ilh /27 -
NRC PORM 31s (9 74) NRCM'0240 -
W us e. oovunesassary psiesmose orric : s eye -enees4
m-u e.+ -
DE ? '01 1977
~
cc: Pacific Gas and Electric Company Mr. William P. Cornwell.
ATTN: Mr. John C. Morrissey P. O. Box 453 Vice President & General Counsel Morro Bay, California 93442 77 Beale Street San Francisco, California 94106 Mr. James 0. Schuyler. Nuclear Projects Engineer Philip A. Crane, Jr., Esq.
Pacific Gas and Electric Company Pacific Gas and Electric Company 77 Beale Street 77 Beale Street San Francisco,. California 94106 San Francisco, California 94106 Mr. W. C. Gangloff Janice E. Kerr, Esq.
Westinghouse Electric Corporation California Public Utilities Commission P. O. Box 355 350 McAllister Street Pittsburgh, Pennsylvania 15230 San Francisco, California 94102 Brent Rushforth, Esq.
Mr. Frederick Eissler, President Center for Law in the Public Scenic Shoreline Preservation interest Conference, Inc.
10203 Santa Monica Boulevard 4623 More Mesa Drive Los Angeles, California 90067 Santa Barbara, California.,93105 Arthur C. Gehr, Esq.
Ms. Elizabeth E. Apfelberg Snell & Wilmer 1415 Cazadero 3100 Valley Center San Luis Obispo, California 93401 Phoenix, Arizona 85073 Ms. Sandra A. Silver Bruce Norton, Esq.
425 Luneta Orive
,1216 North 3rd Street San Luis Obispo, Californ,ia 93401 Suite 202
)
Phoenix, Arizona 85012 Mr. Gordon A. Silver 425Luneta$ve Michael R. Klein, Esq.
San Luis Odaspo, California 93401 Wilmer, Cutler & Pickering 1666 K Street, N. W.
Paul C. Valentine, Esq.
Washington, D. C.
20006 400 Channing Avenue Palo Alto, California 94301 David F. Fleischaker, Esq.
1025 15th Street, N. W.
Yale I. Jones, Esq.
5th Floor 100 Van Ness Avenue Washington, D. C.
20005 19th Floor San Francisco, Califonria 94102 Mr. R. C. Martin, Geologist California Division of Mines and Ms. Raye Fleming Geology 1746 Chorro Street 107 South Broadway San Luis Obispo, California 93401 Room 1063 Los Anceles, California 900l?
- t. -
Y
. -........ = -..... -......
e,
)
j 4
ENCLOSURE NO.1 LIST OF ATTENDEES DIABLO CANYON MEETING NOVEMBER 10, 1977 NRC Staff D. Allison W. Hazelton R. Johnson P. Chen R. Bosnak K. Jabbour R. Mattu F. Litton PG&E H. Gonnly J. Hoch G. Blanc Westinghouse W. Bamford F. Witt T. Esselman W. Gangloff Fora- &-39I R -TI
~..
e,
.)
.)
v 1
ENCLOSURE 2 STlDY:
IffIEGRITY T TIE PRif%3Y PIPliB SYSlBiS T TlE DIABLO CNNON IAJClfR POWER PIMIS DURIIS POSTULATED SEISUC EVBfTS ORECTIVE: TO DBDISTRATE lllAT, FOR lliE DIABLO CAiNON !!UCLEAR PQER PIMTS, POSTULATED DESIGil SEISMIC EVBffS CAN,0T REAS0t! ABLY ifAD TO A LOSS T C00lMff ACCIDBIT, um E
. ~...
)
2 DETER 1IllISTIC STRlITURAL RELIABILITY AtRLYSIS LINES OF DEFENSE:
I QRLITYASSURANCEIN 1) MTERIAL PROCWB4ENT
- 2) FABRICATIm/ CONSTRUCTION
- 3) OPERATION CRACK GROWN SmLL*
l.EAK DETECTION i
MINIMLN CRITICAL I
TWOUGH WAu. FtAw IS lARGE*
1 LOCAIlllTIATEDBY EARTinRE IS VERY, VERY REMOTE v
I
- NEWINFORMATIONPRESENTEDINTHISSTUDY l
O
... - ~
. = -....
c.
1 3
3 PROBABILISTIC STRUClllRAL RELIABILITY NMLYSIS
~
NoFLAW SEISMIC Even FLAW PRESEtR PRESENT (LOWPROBABILITY)
(LOWPROBABILITY) 4 PR (FLAW INITIATE)
PR 61ALL PENETRATE)
- VERYVERYLOW
- C.P-QUITELW
~.
I PR (PIPE RUPTURE)
PR (PIPE RuPTuaE)
- VERYVERYLOW f
LOCA INITIATED BY EARTmuAKE ISVERY,VERYLOW C.P.: CONDITIONALPROBABILITY OF ALL PRIOR EVENTS OCCURRING
1
)
y 4
OtrruNE:
MATERIAL PROPERTIES FATIGUE CRACK GROWTH PROPERTIES
. STRESSES ELASTIC-PLASTIC STRAINS ElALITYASSURANCE SELECTION OF INITIAL FLAW SIZE FATIGUE CRACK GRodrH DURING SEISMIC EVErg
(
CRITICAL FtAW SIZE DETERMINATION PROBABIUSTIC STRUCTURAL RELIABILITY EVALUATION CONCLUSIONS O
i t
n
e COMPARISON OF MATERIAL PROPERTIES OF PIPING WITH CODE AND NSMH VALUES v1 a
Test Report Test Report Code Values NSMH Values Temperature (ASME (Minimum (From all Reports Examined)
(RVOutletPiping) i Sec. III, 1974)
Expected)
Average Minimum Average Minimum Yield Stress (ksi) b c
Ambient 30.0 30.0 40.9 32.5 42.4 38.9 d
650 'F 18.5 18.4 23.5 20.5 25.0' 23.6 1
Ultimate Strength (ksi)
D C
Ambi :nt 75.0 75.0 83.2 77.0 83.9 79.8 650 ':
71.8 65.7 67.9 52.1 65.7*'#
52.1 d
i a - Nuclear Systems Materials Handbook, TID-26666.
b - 43 specimens c - 12 specimens d - 15 specimens e - 5 specimens f - The average is 68.4 ksi with the minimum value omitted, the data are 63.6, 67.4, 68.6, 70.2, and 72.2 ksi.
B
TENSILE PROPERTIES FOR SUBASSEMBLY WELDS
- m b
No of Test Report Code Values Code Values NSl1H Tests Average for for (308/308L Yield Strength Piping Fittings Weld Metal) 3 63.9 30 30 35 v
r Ultimate Strength 3
87.6 75 70 73 a - Ali stresses in ksi; temperature is ambient, b - !!inimum expected data (interim), from Nuclear System Materials Handbook, Volune 1, Design Data, Section.1, Weld Metals, TID-26666.
3
- w. a.
- --.....~..
....----a..
.. -.. ~.-....
s
}-
e,
.)
7 II,870-17 6 I f.4.se ll 0 o n sess.u
=
t u==
o n ciow-o...n=
l o
- Dg o
/
u
=
I
{
I s
l
~
1%
1,,,,
,1 I
1 f
a t
f o
/
f
{-
f 8-
- 1
{
/
I If
{
==
- 1 o nt ot t1 aa.n D
l o r' te
/\\
- y. m f
t c.=we e..
f
[
c.'.'c "'.7
^
== -/
i
==
y 1
I i
i i
i i
i
- =
+=
- =
- =
- =
- =
(Due 0 atimb.Ofe asest.rsesle g,
,gg Fig. 3.1.41: Jlc Determination-Figure 3.1.4 2: J c Duermination -
l 0
Cast 316 Stainless Steel-750F Cast 316 Stainless Steel -- 6000F
'"=
[
e.o.n usse I
9 me.
/
g g
.I isuse I
l i
la l
/
O s
som U "N /
i I
i1 f
3
- 4 i
O
...n a
'l y
t'u s**n ** * ***s
.e l
C I
{
/
~
l r *"" -- (0 n
/
0 f
f a
t Do O
"* p
-*4 o es c e es.ao.s e.
O t' t' t.ari.
p
/ '.,,,
- ...* e s e in n. _
Ja, he en N
i f
m m
/
/
.f t
.L..... L.. _.
i t
i
- =
e
.~
......e..e......t e.u.....,.,....
J c., Determination -
Fig. 3.1.4 3: Jje Determination -
Fig. 3.1.4 4:
i Forged 304 Stainless Steel 6000F Stainless 304 Plate Material 6000F 5
The corresponcling 'C meus a0 W J
r_.
_-._r
.2
--.e----
l 8
)
- I s..ess.rt.s.re f act
(.p. ir.)
I i i lillij i i iIllii i i i l l11ll e.-._:
.. (. r.p
' g o
,/
1_
/
i i a i i t ilj i a i 1/iiij
=
g
=
2 y
n...u
. o o
/
,, e r
g
<*m...
s j
I g
- ' ". oa
/
Sl4.
E
)
... St;,
=
/
=
3 g
r g
Y
/
g.. -
s ti g
gg
/
=.
j
.I y
5
/
g
/
r 6
o%.,
2 a =%.
....n...
o m%.
.i. r
=
g t
=
a.
e.
- n..
- n..
e.
4
. m..
........ ~
't.
f f ! ! If'!
! f f II f f!!
f. f f f fl!
f I f fffIf I
f I IIIff
- g..
......c a
in u.i n..ri.s...,me....i.
4.s. em
- 8a. A Fig. 3.2-1: Fatigue Crack Growth Figure 3.2 2: Fatigue Crack Growth Results. Stainless Steels in Air Results Stainless Steels in Pressurited Environment.
Water Reactor Environment.
l
.,,,.. ii..t. [. 3 I I
a
~a
'r, e
..c fi 2
i e i siriit i i i si si;
~
= i i i iiiiig m i e i t ei[
i
...... 9 o
.... o e2 i
.o
=
oaa4 a.-
q o aaaa a '-
i, a na - a.-
1 e
l g
E ir.
I s
~
...n I
g r
it.
y
'/
e r
5 y
J g
_..,,., =
1 g
S m
.,4 5
.u wu....... a c.
=_
_ 3........ (......)..
,=
6 er
=
f f
,e.
,v'
=
sa.t.*
~
E 5
..L.
~
s
(
y'
.e '
{
i
~
~
I f
f ! i 1i11l t
I ! !Ifff d
l i
.,,,.., e... i. i [o..r. ]
- (* Al l
i l
Fig. 3.2 3: Use of Walker Model to Correlate Fig. 3.2 4: Effec' of Test Frequency I
Fatigue Crack Growth Results Stainless on Fatigue Crack Growth of Type Steels in Pressurized Water Reactor 316 Stainless 1.tect in PWR Environment.
Environment l
l l
1
',)
)
9 Reactor Pressure Vessel Outlet Nozzle Axial Stresses - Safe End inside Surface Outside Surface Load Type (ksi)
(ksi)
Internal
+2.8
+ 9.0 Pressure Thermal
+0.4
- 0.4 Discontinuity Normal Axial 0.0 0.0 (P + DW + Th)
Normal Bending
+15. 5
+18. 2
~
(P + DW + Th)
Seismic Axial I 4.9 i 4.9 Seismic Bending
!12.2 1]4.3 e
t t-
)
3 10 QUl\\LITY ASSURNICE,
- 1. ALL MATERIAL FOR THE PRIMARY PIPING WAS PURCHASED TO ASTM STANDARDS.
- 3. WESTINGHOUSE SPECIFICATIONS WERE A-PLIED Tifl00GHOUT MANUFACTURE OF TE PRIMARY PIPING.
II. TE SUPPLIERS WERE REQUIRED TO OPERATE TO DETAILED MANLFACTURItG PLANS AND TO DOCLNENT ALL.EXNilNATIONS AND TESTS.
- 5. WESTIfEHOUSE REVIEWED AND APPRWED MANUFACTURING DRAWINGS, PROCESSES, AND PROCEDURES TO ASSURE THAT CORRECT PRACTICES AND METHODS WERE EM-PLOYED.
- 6. THE FINISHED PIPING ASSEMBLI5S MET THE ASPE CODE REQUIREMENTS AS CERTIFIED BY T E AUT}ORIZED INPSECTION AGalCY.
- 7. THE PIPING SYSTEMS MET TIE ASFE CODE APPLICABLE FOR INSTALLATION,
- 8. PRIOR TO SERVICE TIE REACTOR VESSEL SAFE EtlD YELDS ERE INSPECTED 3
~
TIES BY ULTRAS 0flIC EXAMINATION AND ONCE BY DYE PENETRN4T WITH fl0 RECORDABLE INDICATIONS.
10.
IN-SERVICE INSPECTION WLL.1NCUJDE ULTPASONIC EXAMIfMTION AND DYE PENETRANT EXAMINATION PER INE CODE.
- 11. REDUNMNT MEANS ARE PROVIDED FOR DETECTING AND IDENTIFYING THE LOCATION OF TW SOURCE OF REACTOR C00LNIT LEAKAGE.
r A
n Estimates of the Elastic-Plastic Strains O
e 1
i o,ff = g /(oj - 02)#
- I'2 - "3 '
- I'1 ~ '3)#
I l
i l
b hg.
1 i
11
'eff "
7 4 6
'2)' + (c2 c3N + (ci c g 3
I
+
e l
i i
.-__-_n
)
)
.n ll.870-9 28 24 YlELD l
20 e
,e(PSI) l (f.)
n PROPORTIONAL LlHIT P.L. 17025 0.068 18000 0.0 73 O
16 20000 0.I06
{
22000 0.202 e 23045 0.290 y
t y
24000 0.399 5
12 25000 0.545 4
26000 0.729 28000 1.229 30000 1.92 9 8
32000 2.862 34000, 3.495
{
i i
i I
i l
l I
I I
I I
I O
0.00.1 0.20.30.4 0.5 0.60.70.80.9 1.0 1.1 1.2 1.3 STRAIN (PERCENT)
I l
Avarey et* -- %
,....;.......,. J I U '. : al6b000
.)'
.)
~
13 11.870-1 o
u)
.,.3 E
Q?
=s a%
~
t 2
n-fun
-a
\\w225 m>
a o
==
=
in g
\\
l h.
d b
\\
wa s
f*
\\
\\ \\
es s
ne
\\
\\
l ur
=
sc
\\
\\\\t e=
sE
=
\\
s i wD Mm-e
\\(
d.
o g = '0
\\
\\}
C'E O
$ =o E
's E
\\
Wts
~
m
-No
\\
\\
g
- 02 i
\\
\\
E E5 E=
\\
s.E m
w
.o = -
%uE
\\
e a> v.
O m
\\
o c
w m m u
\\
ts.
O a
k C Co ohm W
- C.O
- s mm ce anw E E M o
$)
o8 5
s
>um i
o m
u es ;
7-
,E%E E
=22 m
1 W
g eo o
- n. <t m m
1 E
i i
y 2
No a85 m =z h
I I
I o
h 3
O e
(IS' ) SShj S.1All'qJJJ Eft:.1011Hid-;,i10V'3 A
- ..=-.
. ~.... _..
..i:
)
j
- 11.870-2 g
~
2.0 l.5 l.0
.u 0.5' l
0.0 0.0 1.0 2.0 3.0 11. 0 5.0 I
'EFF 'y i
I Inelastic Strain Correction Factor Ior Cylinders Under Thormal Loading as a Function of the Ratio of Elastically Calr.uiatect Ef fective Stru',s Owe Yield Stress 9
- + - - - -
- w.,
.i..~..
.-.-...u.
)
-)
- e Ii.870-1I g
32 NORMAL OPERATING + DDE (OUTSIDE) 28 NORMAL OPERATION (0UTSIDE) 24 YlELD NORMAL OPER ATING 20
+DDE (INSIDE) 3r
- NORMAL OPERATING
^Y:
(INSIDE) 3 16 u
w y
>t e
g 12 u
/(
0
~~
N
~-
NORMAL OPERATION -DDE (OU TS IDE )
i O
NORt'.it OPER AllNG -DDI (INSIDE) lgl l
l I
l l
l l
I I
I 0.00.1 0.2
- 0. 3 0. II 0.50.6 0.7 0.8 0.9 I.0 I. I I.7 I.3 STRAIN (PERCEHT) 4.
Ave..'oe E f fet tive Sta"s - I'fect v St. ' Cur v:' for 316 S' M 6t20:'F :rdicat.:..; C :.is..e: ':n ' OLJ f 1/I s
.y Ol"
- ~
._.m._
1 l
11.870-i0 16 32 NORMAL OPERATlHG
+DDE (0UTSIDE)
~
28 NORMAL OPERATING (OUTSIDE) 24 YlELD NORMAL OPERATING 20
+ DDE (INSIDE)
(
h y
NORMAL OPERATING (INSIDE)
(
16 W
e
- (
l 1 r 12
> L 1 (r 8
I NORMAL OPERATING -DDE (00iSIDE )
14 i
NDRMAL OPERATING -DDE (INSIDE) i l
l l
l l
l l
l l
0 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.70.80.9 1.01.1 1.2 1.3 l
l STRAIN (PERCEHT) i Average Elfcctive Stress - Ef fective Strain Curve for 316 SS at 6500F Indicating Behavior for - DDE Followett by + DDE 9
1
~
s f
M Surface Calc.
Stresses Rs0 i-inside Effective Bending Strain t
Load o-outside Axial Circum.
Radial Stress (ksi) Membrane
(+ outside)
M j
1
-17.1 0
0 Piping)-
(-DDE o
-19.2 0
0 Maximum i
9.24 l
Strain (+DDE) o 0.74 Maximu:$
i 1.64 8.00
-2.2 8.92
{9.93 1.0) 0.14 Strain (-DDE) o 8.59 9.60
-2.2 10.93 0.64 Strain Range i
0.10 (after first o
0.10 i
cycle) l w
L First Cycle Strain Range Itasimum i
0.14 l
Strain Range o
0.52
(+DDE then
-DDE)
Maximum 1
0.20 f
Strain Range o
0.59 i
t
(-DDE then
+DDE) 0
.e y
.. - ~.
..... =
4 18 EuSTic FATIGbE QUCK GROWT}i ANALYSIS S lj x 10-12 l' MAX u-R)0.5;4.48 EE DN EusTic-PuSTIc FATIGUE CRACK GRWTil ANALYSIS A) J-INTEGRAL p = 0.0419 [J, G-R)]2.24 B) ELASTIC-PLASTIC K EVALUATION p = 5.4 x 10-12[14, G-R) 0.5 4.48 3
4 P
h
19 SMYW'T FATIGUE CPACK GRGfill DUE TO SE1911C EVEfff
- PaCTOR VESSEL OtntIT ib2ZLE SAFE ENo -
(OKSIDE SLRFACE)
~
CRACK GROWTH MILS) 50 CYCLES 100 CYCLES 150 CYCLES 20 CYCLES i
ELASTICANALYSIS 29 fA 104 152 j
ELASTIC-PLASTIC ANALYSIS b
KAPPROACH (26)*34 43 51 60 JAPPROACH (26)*32 37 43 50
- (
) - FIRST CYCLE CRACK GROWTH b
IS155KSidii WELUSI0th FATIGUE CRACK GROWTH IS VERY SMALL DURING SEISMIC EVENT
(< 0.1 IN) l
...s-
. >. :.i...:..,.
- .... us.
)
Pipe with a Through Wall Crack in Bending 1
4 a
, a rb
-n
)
fk
/
a f
e 1-
+
(
I F
l l U
\\
s I
/
\\
~
l f
3 l
b = 4 (n-n)2 R'2 t ~-
R P
M c
-n m
g
[R 2 (2 cos s - sin a)]
g 2 (n-a)2 R t7 m
i l
~..
I 1
N 1.0 0.8 go %
O 0.6 732 0.4 NOTE:
,,~5 u* c' c= y w
to O EXPERIPENTAL DATA' FROM REFERENCE 4 0.2 i
[
n:
i-4 p*
- 2a j
p = INTEREAL FRESSURE l
l R = HEAN R AD IUS 0.8 I
2 4
6 8
10 A = a/[RT I
i Failure Pressure for Circumferential Through Wall Flaws l
i i
i l
i
.a l
.....~...x
. ~.....
. o..... -
0 j
~
.v i
9
~'
/
~
-7 7
(sooo O (o psi)
PSI) 6 FLAW GEOMETRY mT 5
- o 5
.5 x
p E
4 bx U
PREDICTION (O PSI) x3 3,
2
[ PREDICT 10tl*
(11000 PSI )
~
'N N
I I
i 1
I I
I I si i
i i
O O
20 tio 60 80 100 120 1110 160 180 200 220 2t10 260 280 2a (DEGREES)
Comparison of Limit Moment Prectictions with Experimental flesults - A10GB h: Woo (!;eh r,-nce d)
-..===..uc.=,.=..=_.=,.=.-
.s...= ~
= -.
=..
.)
3 i i.8,0 8 406 b,/
(R+f)
~
, NY O PSI FLAW GE0ETRY 200 me M
Z O6050 PSI 100
~
l PREDICTION (1050 PSI) t PREDICTION (2500 PSI) o_l 1
I I
I I
I I
I I
I I
I I
I 0
40 80 Iq0 160 200 240 280 320 2. (DEGREES)
Ccmparison on Limit Moment Predictions with Experiment.1!
Albl 3C 3 ping (R icrence 12) 4' P.? w it! -
--*---------e-
RESULTS OF C'UTICAL FLAW SIZE DETERMINATION -
y
!)
CIRCUMFERENTIAL THROUGH WALL FLAW WITH IMPOSED BENDING MOMENTa E
Applied Moment c
Critical Flaw Percent of Mean b
5 Location (x 10, in-lb)
Size (in.)
Circumference d
Reacto-Vessel Outlet Nozzle (safe end) 59.28 25.5 26 p
il P,eactor Vessel Inlet Nozzle (safe end) 15.26 62.5 65 8
Steam ^enerator Outlet Nozzle 22.69 69.0 64 y
Stean Cenerator Inlet Nozzle 32.32 62.0 57 i
Prinar. Pump Outlet Nozzle 24.40 45.0
.48 j
Primary Pump Inlet Nozzle 32.81 61.5 57 9
)
a - Internal Pressure was taken as 2500 psi, b - All locations are at safe ends, c - At mean radius.
CGICLUSIGh CRITICAL Tm0 UGH HALL CIRCUMFERENTIAL FALL IS LARGE (> 2 FEET IN LENGTH) c ooooo me m,, o o.
_.....__. ~
.....m
..._._m_
_._.....m___
z_
.z-.
)
25 PROPABILITY CALCUIATI0iG 1..
PROBABluTY & OCCURRENCE OF f.ARTHQUAKE AND PIPE RuerunE ASSlNING EVENT INDEPE?mCE
~
- 2. PROBABIUTY T CRACK INITIATION FROM AN UNFLAWED STATE DUE TO FATIGUE
- 3. PROBABILITY OF THROUGH WAu. PENETRATION DURING SEISMIC. EVENT (FLAW INITIALLYPRESENT) 11. PROBABIUTY & bSS T C00UWT ACCIDENT IN THE PRESENCE OF THROUGH WA FLAWS Y
(
~
l f
i l
m
.v.
w...
--- --- =:.: = a = = - - = = = =
.~ -....
.i 26 1
PROBABILITY T OC0JRRENCE T EARTH 9UAKE
- 0 PIE RUPTURE ASSililtE EVBIT IiOEPBDENCE
~
P h m m e s) = 10 % m g
4 P (LOCA OCCURS) = 10 HEAR t
I CRmCAL TIME, I (EARTH 0lnKE) = 30 SEC (9,5 x 10-7 YR)
E 5SEC(1.6x10-7 YR)
=
L P = P P T + P P T = 1 x IOS g t g g t o e
h
,..-........w,..-.
,. =.. ~... - -..
. -. - - -. -........~.-,.:..
s
- ?]
1 Sketches Outlining the Overlap Method of Reliability Assessement i
i S
f(s) f(S)
(a) POPULATION DENSITY' FUNCTIONS FOR NORMAL DISTRIBUTIONS SHOW f(v)
Y FAILURE DENSITY 0
(b)DIFFERENCEDISTRIBUTION I
i l
2=
15 - sl T
JS sl 0
l
/o32+o2 0
3 3
/o 2 + o 2 3
3 l
v.
7, Isl 1-s/gl l
s /l+(o /o3F s
I
=
102 8
Z LEGEND:
6 4
CURVE. t, SED IN ANALYSIS 2
I 10 HOLD TIME CORRECTED CURVE w
8 N
y 6
2
%'N g
750 F y
4 g
g 2
[ DATA BASE CURVE.
5 800 F 0
b l00 8
6 CURVE FROM CC1592-7
~
4 1
j 2
i 10-I I
'I'
100 2
4 6 8103 2
4 6 8102 2
4 68103 2
4 6 8104 5
2 4
6810 i
CYCLES TO FAILURE Data Based Strain Versus Cycles to Failure Curve Used in Probabilistic Structural Reliability Analyses - 304SS and 316SS at 7500F a,
l a
~
i 1*
I
,3 j
~
l CALCULATION OF PROBABILITY OF FLAW INITIATION FOR 50, 100, 150 and 200 CYCLES l
REACTOR VESSEL OLITLET N0ZZLE SAFE END Applied Conditions Failure Conditions Probability Results l
Strain Standard No. of Strain Standard Prob. of Flaw l
8 Range (%)
Deviation (%)
Cycles Range (%)
Deviation (%)
Z, Initiation j
0.7 0.175 50 5.0 0.525 7.77 3.9 x 10-15 g
0.7 0.175 100 4.2 0.441 7.38 7.9 x 10-I4 0.7 0.175 150 3.0 0.315 6.38 8.9 x 10'II I
0.7 0.175 200 2.5 0.263 5.70 6.0 x 10-9 a - 10.5% of mean.
i CONCLUSIQi: PROBABILITY OF FLAW INITIATItG DURING A SEISMIC EVENT IS VERY LOW.
i I
j
n.,.-..
u.a_.
- u..
- a.==,.=-
= :
?.a..
.. =.
)
)
,, 7v Probability of Through Wall Penetration During Seismic Event (Flaw Ini Present)
K C.__
mean = 300 ksi {TF standard deviation = 20.4 ksi (in, (6.8% at the mean)
Applied K mean = 155.2 ksi (iii standard deviation = 38.8 ksi /iii(25% of the mean) i Mean of squared distribution =
X2+a2 Standard deviation.of squared distribution =
(2 a (o2 + 21 )l/2 2
I For s T = 90416 For applied K c.
0 3 = 12254 s = 25592 v
- s " 1223 l
l 2 = 3.74 N =-
9 x 10-5 l
l l
t 5
i t
PROBABILITY OF WALL PENETRATION DUE TO THE WORST SITUATIONS EXISTING AT THE VARIOUS N0ZZLE REGIO!
g.-
AT THE END OF 200 SEISMIC FATIGUE CYCLES Applied Failure Criterion Squared Dist.
Std.b Squared Dist.
Probability Mean Std. Dev.a Std.
Mean Dev.
Std.
of Wall l'
Location ksi/Tii ksi/in Mean Dev.
ksi E ksi G Mean Dev.
_Z_
Reactor Vessel e
\\
Outlet-Safe End 155.2 38.8 25592 12230 300 20.4 90416 12254 3.74 9.2 x 10-5 l
e Inlet-Safe End 41.9 10.5 1866 894 300 20.4 90416 12254 7.21 2.8 x 10-13
[
y
~
V i-Steam Generator d
[
Outlet 21.0 5.25 469 224 300 20.4 90416 12254 7.34 1.1 x 10-13 d
Inlet 27.7 6.93 815 390 300 20.4 90416 12254
- 7. 3.1 1.3 x 10-13
?
Primary Punp h
C Outlet 34.6 8.65 1272 608 300 20.4 90416 12254 7.27 1.8 x 10'I3 d
Inlet 31.8 7.95 1074 51 3 300 20.4 90416 12254 7.28 1.7,x 10-I3 i
i i
a - 25% of mean.
~
b - 6.8% of mean.
c - Outside surface.
h
{
d - Inside surface 5
i' D7:CUJSI0th PROBABILITY OF THROUGH WALL PENETRATION DURING seismic EVE E IS QUITE LOW.
5
=-..:.--.
.. = = _.... - _. -.
.=
= -
32 PROBABILITY OF lDSS OF C00LAtU ACCIDETE IN THE PhESEfCE OF Im0 UGH lhLL F CRITICAL BENDIfG II) MENT _
RANDCM INDEPENDENT VARIABLES: PhESSURE,P Ftm SmESS, o j
g RANDOMDEPENDENTVARIABLE: BEtolts fben, M l
INDEPENDENTVARIABLE: mLF-CRACK ANGLE, a-M = 4- (w-a)2 p 2 2 -J
.,2 42 g
Rt p 2
1 Ro (2 COS B - SIN a) 2 (n-a)2 p 2 t,
g s
APPLIED BENDItG ID1ENT i
RANDOM VARIABLE: BENDItG N0f1ENT, ME I D 0DOLOGY 1
IMM)RTANCE SAMPLIfG APPLIED TO PRESSURE (P),
l i
FLOW STRESS (o ) AND APPLIED M0fea Ci )
g g
i EVALUATE: PR (M - M >0) g y.
f e
?
o L
f.
t
?',
f d
1' o
l hobability of Pipe Rupture fbm Earthquake Assuming a Circumferentia11.'
Oriented Through Wall Flaw Flaw Size as a S'. of Circumference Probability of Failure
~
{
8 Reactor Vessel Steam Generator. f Primary Pump i j
7 Ler.gths t Outlet Inlet 7
i St e End Outlet Inlet Outlet Inlet (1.)
Safe End 2
5 4.95-5.4 2.2 x 10-3 7.6 x 10-I 10 9.9-10.8 1.1 x 10-2 1.6 x 10-Ib 15 14.8-16.2 4.4 x 10-2 1.2 x 10-9' 20 19.8-21.6 1.7 x 10-I 5.5 x 10-I 1.3 x 10-7' j;
30
.32.4 6.9 x 10-I 2.2 x 10-2.6 x 10'I 1.5 x 10 1.5 x 10-9 I,
-f u
l39.6-43.2 9.8 x 10-I 4.6 x 10-II l1.1 x 10-8l 4 x 10-7 l4.8x10-b 3 x 10-6 i
40 i
I l
t 8
i-
" Mean radius applied.
- Less than 1 y 10-I2 l'
CGiCLUSIGh PRTABILIT( OF T} ROUGH WALL FLAW 15 IN. LONG IN RV Otm_ET IbZZLE LEADING TO FAILURE DJRING SE!SMIC EVENT IS FOUR CHANCES IN A H'i':DRF3.
j
T r
~,
Summary of Worst Case Probabilistic Structural Reliability Evaluation W
{
LOCATION:
Reactor Vessel Outlet Nozzle Safe End i
Item No.
Condi ti or.
Des cription Probability l
1 DDE Probability of Occurrence During Plant Life (Calculated) 5x10-3
\\
6x10' 9 Probability)of Crack Initiating During DDE 2
No Flaw Present (Calculated a
~
1 lx10-3b 3
Circumferential Probability of quarter thickness circumferential Flaw Present Flaw being in worst location (Estimated) 4 Quarter Thickness Probability of Flaw Leading to Leakage (cal'culated) lx10~4 Circumferential Flaw Growing During DDE
[
L f
5 Leakage, Through Wall Probability of Through Wall Flaw Leading to 4x10-2 Circumferential Flaw Six Rupture (calculated)
Times Thickness in Length (15 in.)
b
- a. For or at 200 cycles i
b.
Ccnser.ative estimate based on two units (8 loops).
i I
~
w
-s
..w.
=
.=....-:-wa
_1-_.
..f-...;a.
6 A
-s g
e
/
,e
,4.
litt3 it,,u,C Ettot elf ttis et l
[ esatta tat $ ]
g Quattiv allvesett 3
gg,.se Maltasat
' PtCC98 tut a f a$ut C00C. 50t. IIC.
Ottttf est SUS-instavsCE l I retStetsCg h i
asstnetits ano PiPies $fsitu
$Unitiv a1SveseCf QUsttTT 455Unanct g,g*.le wtL9196 aat y
ggg.se of testsee: etGutateet C0elitWCTI0er
'It9UlttuteTS leltatsCE alast C006. 50t. (TC.
le5Pitit0s. ETC.
r i P.O.
- 8.3s10lte. CALCUtarts I_
- E f e s0*3ga s, _ _ 3
' { '
/
s0 9tf ECT le PsPt OLFECT se PIPE
/\\
p._
3 lC.P., P.O. ClaCE latilaf ssGI Quatite f aitsetS3 8
l cs.:,-s (200 CvCtts _ _.]'*""
a ours_
_ _ _ _ L_ _ _ _,
_ _ _ _.L _ _ _,
I P.O.* 8 10*3. t $
- - --- - - _t ewatto I IP.O. (taCI Islit afip4 0 t
- t0-10
,,I
}
".P 4;
6 6 _C.* Sel0*
_ -.s ett Ctatt GaeWie suaLL Dutiss e
LEGis3 SElsalic Evtst O.0tFtt3! 14 CEPf e 10tettFile 8* " *ans "Il 18 DIPf"MSI' IIII P.O.
mall Ptstita110s 8 lC.8..
a
- a la * *
. :. w.e. s u tra_ _ __;
LOO: Lle! 0F Miteit
- s. u. D&LL Pte,,f f eaflee i, n0 f UEtutt,C09,510ta leg RFC g
isisis n 0 L-___,,,,._.-__#
P.O. :Peesatitliv 0F OCCutAinCf P. C. :Psega8tLIIT Of CeaLLieGE lEAa 9tTICit0 C.P.:Cor3tissent Peceasittiv 0F att Patta tutett (Pit REG.6v10E I.4%. (TC)
OCCL4 B 34 useissunt Celt stal f*t0Wie
'4 pall Flas Silt
- 25.51a.
l C.P.
P.O.
PIPE BVPfuat 3 I 8_e s IO.2 C AL Ctta f E D I
,,____I___,
1 P.O.. PIPE 4.Piget I *10 50 Cattplat t e 1
.__^~T~~~~'
p0ca 1 U
(LeCLD360s: LOCL 6s1718:59 tv f asT*.
QUsat il tit? 9tf* El4Pt L-Logic Diagram cnd Lines of Defense Ar'not. ted with Psobabilir. tic Rostritr
O' 4
y.
UKillSIQE.
- 1. HIGH QUALITY, HIGH INTEGRITY PRIfMRY PIPING t.YSTEMS EXIST IN Ti!E PLAllTS.
- 2. ASStNIfG A ]/4 t FLAtt Jfl WORST LOCATIOfb FATIGUE CRACK GROWTH IS LIMITED TO LESS Dwi 0.1 IrEH DURING SEISMIC EVENT.
- 3. CRITICAL CIRCLNFERENTIAL TFROUGH VALL FLAW SIZES EXCEED 2 FEET.
- 4. PROBABILIT( OF PIPE RUPTURE DUE TO SEISMIC EVEf6 IS LESS TIMN 10-10 gog 50 YEAR PLAIR LIFE.
GBEMI. C010_USIQh THE SIMULTANEOUS OCCtRRENCE OF THE EARTrRUAKE AND TE I.0CA CAN APPROPRIATELY BE DELETED FROM THE SET OF EVENTS AfD EVENT COMBitMTIONS EVAllRTED FOR DIABLO CAtlYON UNITS 1 AND 2 Wincur diDUE RISK
{
TO TE HEALTH NID SAFETY OF THE PUBLIC.
6 e
l 0
t l
i
v -..
.-.=~:.
...i.....c--..--.
,i N.
-1 i
i.c.
).
-)
o.
.)
I, I
MEETING SUW1ARY Docket File' NRC PDR
~
NRR Reading LWR-!1 File E. Case P,.
Boyd R. DeYoung D. Vassallo J. Stolz K. Kniel
- 0. Parr S. Varga L. Crocker D. Crutchfield F. Williams R. Mattson H. Denton D. Muller Project Manager:
Attorney, ELD E. Hylton IE (3)
ACRS (16)
L. Oreher NRC
Participants:
S. Rubenstein N. Newmark, Univ. of Illinois J. Murphy I. Wall D. Jeng P. T. Kuo l
k Fora % -sr/
e4F