ML20077K083

From kanterella
Jump to navigation Jump to search
Nonproprietary Technical Justification for Eliminating Pressurizer Surge Line Rupture as Structural Design Basis for Kewaunee Nuclear Plant
ML20077K083
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 06/30/1991
From: Palusamy S, Schmertz J
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML111661086 List:
References
WCAP-12874, NUDOCS 9108060350
Download: ML20077K083 (68)
v • d • e


Text

4 .

.....,.4 * . - .- , , e

, , * ., . f. , , ' . ' -

.< - y.. ,.

'f,,'

4, . %l ,, , ' ' ' , '

%g

.f.

..JCk. . . ~ - ' . -

=N ,, . y . [.' . ' . , . - ,

8' *

.t, ( - ,- .,'

,,A,: ,e ,. .e -

. . . ,yt*-., .-. 3 , ,s..,

+ i

s. .,.

,.4,+. et -

,~

i*

7 . ..,,. .

'* ' ,.8-i'.,,

, - . p. bi; ,7 4 . ,' 'r+

- \.gr+ . ' * ' '- 'S

.a '. ' 's * -. . ,

.. i ,<..  :

s

.g,' 4 .. . , .4.-

.g e.. .

. ...' -,.a,. - .

. . i

- . . ar. .. ,

.e.

. - )a

't y.. .

.i4 - - 3 4. ,

3 ,.

.s4.,, ',*...

7,; ,

- e, ,- t . .

'1 3

. 's 4

p

.- . . . ' .f

' . ,'/!<,- i4laa ' . :' y -

t "- - .

g- -

u,. .. . s

' .dyi ).  : bl(.3

e. 4 si , . ->' . ,,,,;. 41

,  ; , - t s, .'.t. -.:. o y.? .. ..s. . -.

. < 4. g ..;  ;

i, .4 s.. A,.. ,. . ,*(

. t,

r. r.

n . . , , ,; .. . . .,*5' . .. -' .

  • n + ' . . . . .: . i.

... -. .. . . * ;. 3',; -,..,, ..$.

g $..,

q , . g . - ..- ,,'(

, , 9'w

,ir g

y. ,' f -4.'.'...

,'. ,y d; i *. . . . *, .

t . ; 0'. ', i ' y- ,) -. ef . . . ,, 4 , . . ' , ' . t),93 , . s.f f.%g,g -

? t.,

, .l8

. a ,,. 1 i

. . , y o.,. - .,

.,s

, pp,

, + e.. ..

' . .4 * , .q '5 e P

.' * *. ,4 V .. *- . .s >. ,.a *

.' f- ,', , f . < _. ,-. 'yh..:o y Q, e , .., .. .

, e

. . ' . . . 3 f( . ,, ,,... c

.} "

., s

&. 3 lp; , f ' s% s'-( l;* ) ,8/ ** ..g.... *l g pA.. <. , p *{.. .* ~ -

g, 4 +s Q.9;J '- ' *.N i- g- *. . ,

o

.y' y ~qm.s i

....?* W. .

' Q' ..} s .l s. ,g ++s ., r_ ,

u,'"...a* '* . .:a

  • c.,. '..'g.' a'

..wQ Y.

  • y'&x&s t  ;

< . G . - w 'A

&l 90 .k i;y 9 4 4. q M .3,s.Y.y g;uwy .!o. v.W \.' .,.i y e 4 < *v .- .AdQ . v e.s' t.., t s : , - .

ss

.s ,

[ , ;' . [. : l

. { j ,'[.f j ,..p ., .hhhh N .5 h b h g. g'*g,. m y %.pft, %.s:-

w7 . . - s.' . ,

- kW D h.w+ [, t .z ' ek .k V ..

- .. ;j . cc

. i

,. r . . . . , ;g *; -

l'pp.y',,,'..'<.;'*'"[.;3,".L g. .[. .C j.

6f. }p# .[ . *, . f ' , . ..

4. ,,'..,'? *s,
  • 4 '. 4 ~1- ( (

. . ' . . , ,j

, 4' .

kf , . , ,

,e , ,

, e

'N ..W,*'.

[ Q 3, ' . ., ., '> . ,-

. . . ;{[ ',- -

c .

' l J. ) g[ 4 f't *

..l Y ,,

g.

%g ..1 ...s. +,

y .y 1

. t3..% s 9[

  • j *% , d .? ' v
  • l ~

l i(',

. h,. ;. ,}p wy(. .g.

. '-

  • v a . .  ;- . .

. ., .*/' .-

8 + n' :- . f;. / . ' b..

N *

. j. t

... h.

.y

.. <, i: 9..' .1 ',. * ,

, / j 9 '. . . . . .',, -.' '. ... ... .",.4.. #>4

  • s.,,.

j.

4

,g.*i ..,

w' A -@ , c .,

, . .s c.,r .d <f '

- ,n. g.,.%-

g,*,.*,. . *-

' , . ' ,. ;; .- '.., ;,c s,'.'.e -. .,',... , /

L,'

p%,

. e .

i

.6 w  ; &- .yK...

i 3 .

....,* *. . ,.  ; L.* '

4' . .*

  • .- e -

iq a -

!\'. . '

. . f.

.z. .

- - . c,,,- ..,. pr'- .. . eA l. . ,'

s t

I gii { . J'i M; l N o

.; . , (; ' , *- . .

Ji e

. ,g - -

,a . . , , s . .. . t .ic .<

o*** . ,. , -
  • l.

i T ) , ,, ' < , .'

. .; , . - . .t c

... t

. y hg

p. , " i
  • d.,..
  • h 1 ',

e' . ' p: 99, .

..,v.

. w. .. . ,. . . . . m+ ,a . .u. .i. . i p. . . : .

A . , , ,, ;.

. ..e.'..,.

%. <: M ;.,..'

s ,

,' s ,. . f' . , . ', 8

, 't

[ ".g. ,

, .; , t ,[ .*

. . .;, y h ' '. Y .

....v...

.Li .t.* 5' y

. "I

. , , . . . .._ . . , . -. .;., ., .x - ,,. <. s ' , . . y,: y. mq*.

. . 3. c ~ .

4s . *

'[ . ' J . . , ' .. . <.

.:.,?, .. .

(,.- .

. . . 3, .",

  • i . ,18 p . ,. .
, .?. *4.%. ..
- ,,1 , ,

. . , .,- .' ' 9' w.* .P

@ ' ' .' .,. .. .A .l .q . y,;,e . '!,-. ,. ,'. .'-

< . .; .. . +

,*e . . *. .- & -

6,

  • t 'e . *. ' . + ,

.1, , .. $g.. .g

.,e- t 9 .,, .-

..- - .-  : l 7. '.,p l .- .

, - q ,' p.v.

f ., ,

.,7

- - i .c . p ,, v.-

1,,,. . ; y .' . , ;,. ,. i, * ,, ,. , - : -

~

,7, : " . . ' . : .. ..

.),. , .

", , , [' ' ;Y . "..'f*,'*. . .4 -

! ..E i . ,

O. . '. -  ;- s t. ' '{..

, , ..' ,,, .. . [. i . .'.."

  • . .i . '.. .r, k, - ', S ' .

.. ,, 3

') . *

.; .' ' ** , ., . , ' , ' E *. , , , ..  ? f, q 4. " ' '* . *' : " '.e

'.,'-l' ' .* . ^

' - ' . ' * .'.,'I ,'" ,;*' j.

,.,'..c+*

'. .A

- . 7 6..

  • ?'.'. ',,+ .s  : . ,

,4 , . , ,.* - k. ,(  ; 1, .'.,.

ij. , . -

,1,.. *.

~-, .. , ,c y .

t . ,. . . ,

y r

} t e e-E .

't

+

.e g i.-

,, 4 - . . ,,, ,'. .j. .  : .a
.-. y . 3, e., **

c.,,

u. .

' , ..4 Q. '. ., '

g. . . .

3 *.

- ' '. s' j .

4 - .-  ?. ,-

. l ,* , , , i -

.+

3 + , .

s

. . . y, e . . . ; >- a

. , p ~p g., , . . ,

,. . . . ..s

. . . . . . , -. . ~ ..

la *,* 6 - , , .

'- * . .* . ' , . . . ';' * ., 4 .,* '

,, .} , ,,

_ i ', ; - '. , ,,,

4 4 e . .

! J 'q , - ., .-  ?. . '.e'

.W . . *

./ . . se,

'. in ,. . ~,

  • 8

.h*..g- t . , * ,.

J .,.- t-4'.

sI

.W".+.. . ,. .

t..

  • 4

..s.

( - .

'..,* .b e4 *s- - - .' ** ; ' ' '

t , ;,, .' * .. ., $ ., 9 ... ,

+ 2 . . . . ; . , .. . - - . , ,

r. , ,,. . , ,

., .. . . . . sa, . ... .

,O.

. . - . . , .T

-,' ~ .- ,. .' . .,'*

's, 4 . ' ,

. , , 'w , . . ,*

4,

,, . .. ,. _ e.

. , , , . , A *c . . .

, . g C , ' : .*. 1 .,

,g, , , .. ... . ~ %. l .

p, . ,, ,. . . .

4 a- .

( '. . ' -

, - . , .

  • e .# t 4' '. .- , .., ,. ' . t .g, .
  • [ ' , Y l 'k# ,1,

.j = . . ., ,.. . ' . * -

  • e, s' .- *.. *

, ,i...,- '.' 4 - ,- '-

. .s

'. , . . ' :.['*'A..L'.+1,'8.'.,',1,'

iw,  ;.,. . . . , - . , - , . 1 f .

(.

, , q

. ,- {

, ' . ,.. ..' .l, . M, *)'..,i,..'.*.*,.'s 'ps'

'..,/ * * -  %

', ' . ' ' * .. [.* ,

s

.- '.<.- 4

.,,9...-- r. . . - .s s.

. - . n-. ... Y, *g Y;,.- , ;-

, ., - y ,. - . ,w ., .v.....  ;

g ; . / ,,,- *',,

-g / -

_p '1, , '" - ,. e .

j .. , .. .

7 . . . . ' .. :. '

, .k c n ,

.,y.,.......'.'.H'.,.J'. . ,. . . . . , . ,

M ,$ . .

,. . . \

..a.... , . y' c. . .

  • 4 ., ..,,.,..q..

. ..; . . .' . g. , . , , . . . . _ .. ..

, f .; . . , . . .e ,

. -. '.: .. . . * ,' - ' . ' . *4=

.. . . .e i ' 4

' .< .~

  • p.

. ,i s. : ,.- ,$c ' + + g/ r s'.

i,. ,' . . n. . . fr , . " -0 .- , 4* ' ,

,', ..'.. N, .f. ' > ,'t a, (< . ' : ' ' , I ' ,

. - . .. . , ' . ' . . , _ . . .,.,.a,( , -} . . ' '$* ', ,

s. ,, {7- ^ ' . ' , ;d.q . , . j .*, .'; . . .

,c .4 ' . . . , ..,',M.,,',,.,

, yJ ;4 3 . . j f ' . * ,. .} . . . ,

T.

..-' .- * ,.,

  • k \ ,,,, . . . , [*.*.

t p

W

[ *, , - . "; jy ' . , , [*y

, , ,3 . .

+, Y. 4 ,h . . .' .. ,,

,-%' . . . . . . ' , ' ,. . a, f, , ; , , . -

e y y

7. , s ;

-.. . 9,..,

~1  :

1: t  : a

n. -

> , ., . . ~.w .

7.,.,.3 . w .>.,..,.s. , c p' g , > , 3. . p . . q, c.y ,-,, c. , 3.y. _ n; p, . ~ . . .

.w; ,,.. m ; = . i ,i . ;. .. % . ;v ,,i,. i...y ."> , , c. %. , n .....s.

, . 9,. g - . ,yi,,, . , , . .e. ., .;. , ,p

.. 3 , . y . - hys_g c t +p%,r

, y . :._) H,9'dT',.

.y .9 -i . ,-.

,' 4 ..; .

b.*' ' 6 4 . .m' ,,h .p

  • t N.' '.g ~6.. y g[, ,* g ;9., !*[+. 4 4.. b
  • k- . ht .3,j,' E t

..+/...-# y if , * {t . .' - p;O Q  : .;x: . ; :._. _.pf..S ^

~ f"f' 4lk;, ?' ql'y[, tp s ;.7-f.'. p ' * .. ' y.?. ; V [4' I- ,' p  ;. ,; . , . .

w  ?

qd:(: n. d ' _ C ^ % ' ' f %y.. s.em+ 4 f,.;,yt :;;:,4.y.

.p . . y.. ..

..~-

n. n. .. ..,

'.. ; V. .lQld.

(;.. *

. :. %Q%. & fi i n. g. ,f.4.1,s%v,h ,

Q,, 'g.R Qj..'Q. .* ,' .4[.;' Y:.:g

.% %. .Q~.::gy ,'p.g ;% Y y.c y y,e,W ..

5f'Q) n 4*

.cy . ,1 9. .., w. c. r..g;- .. , 4- ..  %, .v. 4.1 4w(f. , # . q3 . 4,+e. , - . . v,~%r,.. .c y .3 y  ;+ 3.. . p >, ., g.wq. -  ; _ gfr y .m.4, ; ;a m>2 .. p.c 4; -

. . . 4 - e.Q, .

c..

3 4., yp . 4.., 9 ..

W. yr . - . ; 5 . ,t yL

4. qc q . .- .. .t 2 . . .t yp :

mi g .i', q

.w.  :. ,-  :

3 y k .: , . '

. ,, t Y.

we

~

maosom men -

PDR ADOCK 0300 v . w m n e w},n m w{ w.p.m;.7.((j '

y, p.p' p'.g.g g,g. Qgn] 4';

A .

- . -,. ems.#.w....< w w w w. w w w w ~

WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-12874 I

TECHNICAL JUSTIFICATION FOR ELIMINA. TING PRESSURIZER SURGE LINE RUPTURE AS THE STRUCTURAL DESIGN BASIS FOR KEWAUNEE NUCLEAR PLANT June 1991 D. C. Bhowmick S. A. Swamy Y. S. Lee D. E. Prager K. R. Hsu W . ) ,p" Verified: ~-

br/t bg Cb~

J./C. Schmert: )

Structural Mechanics Technology Approved: ) b /^' Y

' A . S. Polusamy. Manager Diagnostics and Monitoring Ternnology Work Performed Under Shop Order: KFBP-950

.- WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P,0. Box 2728

- Pittsburgh, Pennsylvania 15230-2728

@ 1991 Westinghouse Electric Corp.

3,- --- - , - - . - - - - - - - , - - - , - - - - - . - - - - . -- - - - - . - - . - - - - , - - - . - - - - - - - - - - - - - - - . - . - - . - - - - - - , - - - - - - - , - - - - . - . - - - - - . - - . - - - . - - - - - - - - - - - - - - - - - - - - - - , - - . , - - - . - - - - - . - - - - - - - . - - -

TABLE OF CONTENTS

, Section Title P_aje

1.0 INTRODUCTION

1-1 1.1 Background 1-1 l 1.2 Scope and Objective- 1-1 1.3 References 1-3 2.0 .0PERATION AND STABILITY OF THE PRESSURIZER SUPGE LINE AND THE REACTOR COOLANT SYSTEM 2-1 2.1 Stress Corrosion Cracking 2-1 2.2 Water Hammer 2-3 2.3 Low Cycle and High Cycle Fatigue 2-4 2.4 Summary Evaluation of Surge Line for Potential Degradation During Service 2-4 2.5 References 25 3.0 MATERIAL CHARACTER!ZATION 3-1 3.1 Pipe and Weld Materials 3-1 3.2 Material Properties 3-1 3.3 References 3-2 4.0 LOADS FOR FRACTURE MECHANICS ANALYSIS 4-1 4.1 Loads for Crack Stability Analysis 4 ?.

4.2 Loads for Leak Rate Evaluation 4-2 4.3 Loading Condition 4-2 4.4 Summary of Loads and Geometry 4-4 4.5 Governing Locations 4-5

.pe

$143s/C11491 to j

TABLE OF CONTENTS (cont.)

Section Title Pace .,

t 5.0 FRACTURE MECHANICS EVALVATION 5-1 5.1 Global Failure Mechanism 5-1 5.2 Leak Rate Predictions 5-2 ,

5.3 Stability Evaluation 5-4 5.4 References 5-5 ,

6.0 , ASSESSMENT OF FATIGUE CRACK GROWTH 6-1 6.1 Introduction - 6-1 6.2 Initial Flaw Size 6-2 6.3 Results of FCG Analysis 6-2 6.4 References 6-3 7.0 ASSESSMENT OF MARGINS 7-1 -

8.0' CONCLUSIONS 8-1 APPENDIX A Limit Moment A-1

  • a l l

....,n,...

33

e LIST OF TABLES  :

,, Table Title Pace 3-1 Room Temperature Mechanical Properties of the Pressurizer Surge Line Materials and Welds 3-3 3-2 Room Temperature ASME Code Minimum Properties 3-4  ;

3-3 Representative Tensile Properties 3-5

. 3-4 Modulus of Elasticity (E) 3-6  ;

4-1 Tyoes of Loadings 4-6 4

4-2 Normal and_ Faulted Loading Cases for Leak-Before Break Evaluations 4-7 I

4-3 Associated Load Cases for Analyses 4-8 4-4 Summary of LBB Loads and Stresses by Case for Governing Locations 4-9 5-1 Leekage Flaw Size 5-6 5-2 Summary of Critical Flaw Size- 5-7 6-1 Fatigue Crack Growth Results for 10% of Wall Initial Flaw Size 6-4 7 Leakage Flaw Sizes, Critical Flaw Sizes and'Hargins 7-2 7-2 LBB Conservatisms 3 s m.w.si io $ji

. - - , .,,,,w., .-q-,r .,-m,4-,-.,,,-e .,-,...se-.- ,o-cw--r-,.,,,---.r,--.e- -v --r.-,w ww we--

l LIST OF FIGURES l 1

Figure Title Page

{

e 3-1 Kewaunee Surge Line Layout 37  !

4-1 Kewaunee Surge Line Showing the l Governing Locations 4-10 5-1 Fully Plastic Stress Distribution 5-8  !

5-2_ Analytical Predictions of Critical Flow Rates

. of_ Steam-Water Mir.tures 59 5 (------------------Ja c.e Pressure Ratio as a i Function of L/0 5 10 5-4 Idealized Pressure Drop Profile through a  !

Postulated Crack 5-11

+

5-5 Loads Acting on the Model at the Governing Location 5-12 e

5-6 Critical Flaw Size Prediction for Nede 1020 Case 0 5-13 5-7 Critical Flaw Size Prediction for Node 1020 Case E 5-14 ;

5-8 Critical Flaw Size Prediction for Node 1020 Case F 5-15 5-9 Critical Flaw Size Prediction for Node 1020 Case G 5-16 5-10 Critica1' Flaw Size Prediction for Node 1240. Case 0 5-17 i 11 Critical Flaw Size Prediction for Node-1240 Case E 5-18 vnvoven to 9

1

LIST OF FIGURES (cont.)  :

't Figure Title Page .,

t 5-12 Critical Fiaw Size Prediction for Node 1240 Case F 5-19

  • 5 13 Critical Flaw Size Prediction for Node 1240 Case G 5-20 i 6-1 Determination of the Effects of Thermal Stratification on f atigue Crack Growth 6-5  ;

6-2 Fatigue Crack Growth Methodology 6-6 6-3 Fatigue Crack Growth Rate Curve for Austenitic  :

Stainless Steel 6-7 6-4 Fatigue Crack Growth Rate Equation for Austenitic Stainless Steel 6-8 6-5 Fatigue Crack Growth Critical Locations 6-9 6-6 Fatigue Crack Growth Controlling Positions at each Location 6-10 A-1 Pipe with a Through-Wall Crack in Bending A-3 l

..l l

sissimiest to y

,. SECTION

1.0 INTRODUCTION

1.1 hCb2round The current structural design basis for the pressurizer surge line requires postulating non-mechanistic circumferential and longitudinal pipe breaks.

This results in additional plant hardware (e.g. pipe whip restraints) which would mitigate the dynamic consequences of the pipe breaks. It is, therefore, highly desirable to be realistic in the postulation.of pipe breaks for the surge line. Presented in this report are the descriptions of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type break will not occur within the pressurizer surge line. - The evaluations considering circumferential1y oriented flaws cover longitudinal cases. The pressurizar surge line is kr.own to be subjected to thermal stratification and the effects of -thermal stratification for Kewaunee surge line has been evaluated and documented in HCAP-12841, The results of the stratification evaluation as described in HCAP-12841 have been used in the leak-before-break evaluation presented in this report.

1.2 Scoce and Obhc1 hg The general purpose of this investigation is to demonstrate leak-before-break for the pressurizer surgo line. The scope of this work covers the entiro pressurizer surge line from the primary loop nozzle junction to the-pressurizernozzlejunction. A schematic drawing of the piping system is shown in Section 3.0. The recommendations and criteria prnposed in NUREG 1061 Volume 3-(1-1) are used in this evaluation. The criteria and the resulting steps of the t saluation procedure can be briefly summarized as follows:

  • 1) Calculate the applied. loads. Identi_fy the-location at which the highest stress occurs.
2) Identify the matcrials and the associated material properties.

5369s/062691:10- 1-1 1.__ _ _ - _ _ _ _ _ _ __ _ ~ . _ - . . _ _ . _ _ . . _ . _ _ _ . _ _ - _ . _ .

3) Postulate a' surface flaw at the govorning locotion. Determine fatigue crack growth. Show that a through wall crack will not result.
4) Postulate a through wall flaw at the governing location. The size of the flaw should be large enough so that the leakage is assured of s detection with margin using the installed leak detection equipment '

when the pipe is subjected to normal operating loads. A margin of 10 is demonstrated between the calculated leak rate and the leak detection capability.

5) Using maximum faulted loads, demonstrate that there is a margin of at least 2 between the leakage size flaw and the critical si e flaw.
6) Review the operating history to e.scertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer or low and high cycle fatigue.  ;

7)- For the materials actually in the plant provide the material properties and justify that the properties used in the evaluation are' representative of the plant specific material.

The flaw stability analyses is performed using the methodology described in SRP 3.6.3 (Reference 1-2).

The leak-rate-is calculated for the normal operating condition. The leak rate  ;

prediction model used in tnis evaluation is an (--- ------------------~~~---

.......................................ja,c.e The crack opening area requ*: red for calculating the leak rates is obtained by subjecting the postulated through wall flaw to normal operating loads (Reference 1-3).

, Surface roughness is accounted for in determining th6 leak rate through the postulated. flaw.

9 9 mwom., i.

1-2

. _ ,____.___ __ . _ _ . _ , . _ _ . . _ _ _ _ _ - ~ . . - _ _ - . _ _ _ _ .

r The computer codes used in this evaluation for leak rate and fracture

, mechanics calculations have been validated (bench marked). ,

, 1.3 References i

1-1 Report of the U.S. Nuclear Regulatory Commission Piping Review Committee  :

- Evaluation of Potential for Pipe Breaks, NUREG 1061, Volume 3. November 1984.

t 1-2 Standard Review Plan; public coiaents solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register /Vol. 52, No. 167/ friday, August ,

,28, 1987/ Notices, pp. 32626-326 "

1-3 NVREG/CR-3464. 1983 "The Application of fracture Proof Design Methods Using Tearing instability Theory to Nuclear Piping Postulated Circumferential Through Wall Cracks."

1-4 WCAP 12841 Structural Evaluation of the Kewaunee Pressurizer Surge Line, Considering the Effects of Thermal Stratification I-

O .

on,w." '

1-3

_ _ _ ___ _~ _ _ _ _ _ . . . _ . _ _ . _ _ . _ _ . _ . _ _ _ ~ . . . _ . _ _ _ . _ _ _ _ _ _ . _ _ _ , _ . _ _ _

SECTION 2.0 t ,.

OPERATION AND STABILITY OF THE PRES $URIZER SURGE LINE

, AND THE REACTOR COOLANT SYSTEM 2.1 Stress Corrosion Cracking The Westinghouse reactor coolant system primary loop and connecting Class 1

. lines have an operating history that demonstrates the inherent uporating

' stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking). This operating history totals over 400 reactor years,

, l including five plants each having over 15 years of operation and 15 other plants each with over 10 years of operation.

In 1978. the United States Nuclear Regulatory Commission (USNRC) formed the second Pipe Crack Study Group. (The first Pipe Crack Study Group established in 1975 addressed cracking in boiling water reactors only.) One of the objectives of the second Pipe Crack Study Group (PCSG) was to include a review of the potential for stress corrosion cracking in Pressurized Water Reactors (PWP's). The results of_the study performed by the PCSG were presented in NUREG-0531 (Reference 2-1) entitled " Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants." In that report the PCSG stated:

"The PCSG has determined that the potential for stress-corrosion cracking in PWR primary system piping is extremely low because the ingredients that produce IGSCC are not all presant. The use of hydrazine additives and-a

-hydrogen overpressure limit the oxygen in the coolant to very low levels.

Other impurities that might cause. stress-corrosion cracking. such as halides or caustic, are also rigidly controlled.- Only for brief periods during reactor rhutdown when the coolant-is exposed to the air and during

~the subsequent startup are conditions even marginally capable of producing stress-corrosion cracking in tha crimary systems of PWRs.

s m . o .o ie 2-t

Operating experience in PHRs supports this determination. To date, no stress-corrosion cracking has been reported in the primary piping or safe ends of any PHR."  %

During 1979, several instances of cracking in PHR feedwater piping led to the establishment of the third PCSG. The investigations of the PCSG reported in NUREG-0691 (Referenec 2-2) further confirmed that no occurrences of IGSCC have been reported for PHR primary coolant systems.

As stated above, for tne Westinghouse plants there is no history of cracking failure in the reactor coolant system loop or pressurizer surge line piping.

The discussion below further qualifies the PCSG's findings, for stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. Since some residual stresses and some degree of material susceptibility exist in any stainless steel piping, the .

potential for stress corrosion is minimized by properly selecting a material immune to SCC as well as preventing the occurrence of a corrosive -

environment. The material specifications consider compatibility with the system's operating environment (both internal and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, _

eelding, fabrication, and processing.

The elements of a water environment known to increase the susceptibility of austenitic stainless steel to stress corrosion are: oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g.,

sulfides, sulfides, and thionates). Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes and preoperational testing, water chemistry is controlled in accordance with written specifications. Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping. -

5369s/06269):10 2-2

, - - - ~ ~ - . - . - - . _ - - - - - - . -

l i

a During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits. Contaminant concentrations are kept I j below the thresholds known to be condacive to stress corrosion cracking with the major water chemistry control standards being included in the plant i operating procedures as a condition for plant operation. For example, during >

normal power operation, oxygen concentration in the RCS and connecting Class 1 ,

lines is expected to be in the ppb range by controlling charging flow chem-istry and maintaining hydrogen in the reactor coolant at specified concenti e tiens. Halogen concentrations are also stringently controlled by maintaininj l

concentrations of chlorides and fluorides within the specified limits. Thi) is assured by controlling charging flow chemistry. Thus during plant oper -

tion, the likelihood of stress corrosion cracking is minimized, 2.2 Water Hamme; ,

Overall. there is a low potential for water hammer in the RCS and connecting  !

$9rge lines since they are designed and operated to preclude the voiding rendition in normally filled lines. The RCS and connecting surge line including piping and components, are designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative .

relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. Other valve and pump actuations are relatively slow transients with no significant effect on the system dynamic loads. To ensure dynamic system stability, reactor coolant parameters are stringently controlled. Temperature during normal operation is maintained within a narrow range by control rod position; pressure is controlled by pressurizer heaters and pressurizer spray also within a narrow range for steady state conditions. The flow characteristics of the system remain conttant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics are controlled in the design process. Additionally. Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system and connecting surge lines. Preoperational testing ri.9 operating experience have verified the Westinghouse approach. The operating transients Oth 00ft110 g ,,3

_ - ~ _ . _ . , _ _ _ . _ - - . _ - . _ _ , . . _ . . . - . - - , , , _ . - _ . _ . _ - . . - _ _ , _ _ , - - - , , _ . .

t l

of the RCS primary piping and connected surge lines are such that no significant water hammer can occur. ., ;

2.3 Low Cycle and Hioh Cycle Fatioue ,

Low cycle fatigue considerations are accounted for in the design of the piping system through the fatigue 9 sage factor evaluation to show compliance with the rules of Section !!! of the ASME Code. A further evaluation of the low cycle fatigue loading is discussed in Section 6.0 as part of this study in the form i of a fatigue crack growth analysis.

Pump, vibrations during operation would result in high cycle fatigue loads in the piping system. During operation. an alarm signals the exceeaance of the __

RC pump shaft vibration-limits. Field measurements have been made on the reactor cor knt loop piping of a number of plants during hot functional '

testing. Stresses in the elbow below the RC pump have been found to be very small, between 2 and 3 ksi at the highest. Recent field measurements on typical PWR plants indicate vibration amplitudes less than 1 ksi. When '

translated to the connecting surge line, these stresses would be even lower.

well below the fatigue endurance limit for the surge line material and Wuld result in an applied stress intensity factor below the threshold for fatigue <

crack growth.

2.4 Summary Evaluation of Surge Line for Potential Degradation Durino Service t

(

There has never been any service cracking or wall thinning identified in the pressurizer surge lines of Westinghouse PWR design. Sources'of such degradation are mitigated by the design construction, inspection, and operatien of the pressurizer surge piping.

There is no mechanism for water hammer in the pressuri:er/ surge system. The pressuri:er safety and relief piping system which is connected to the top of the pressurizer could-have-loading from water hammer events, However, these loads are effectively mitigated by the pressurizer and have a negligible ,,

j effect on the surge line.

L ...>.,...o.

2-4 l

L . . _ . _ _ . . _ . _ . _ _ _ _ . . . . _ . . .. _ _ .- . _ . _ _ _ __._. _ _ _ . . _ - - _ __. _ ,.... - _ ._ . - -,

Wall thinning by erosion and erosion-corrosion effects will not occur in the surge line due to the low velocity, typically less than 1.0 ft/sec and the material, austenitic stainless steel, which is highly resistant to these f degradation mechanisms. Per NUREG 0691, a study of pipe cracking in PWR piping, only two incidents of wall thinning in stainless steel pipe were l reported and these were not in the surge line. Although it is not clear from  !

the report, the cause of the wall thinning was related to the high water f velocity and is therefore clearly not a mechanism which would affect the surge  !

line.

It is well known that the pressurizer surge lines are subjected to thermal stratification and the effects of stratification are particularly significant during certain modes of heatup and cooldown operation. The effects of  !

stratification have been evaluated for the Kewaunee surge line and the loads, accounting for the stratification effects, have been derived in WCAP-12841.  ;

These loads are used in the leak-before-break evaluation described in this report. 1 The Kewaunee Nuclear Plant surge line piping and associated fittings are i

forged product forms (see Section 3) which are not susceptible to toughness degradation due to thermal aging.-

Finally, the maximum operating temperature of the pressurizer surge piping, which is about 650'F, is wall below the temperature which would cause any

  • creep damage in stainless steel piping.

2.5 References 2-1 investigation and Evaluation of Stress-Corrosion Cracking in Piping of-Light Water Reactor Plants. NUREG-0531, U.S. Nuclear Regulatory Commission, February 1979.

2-2 Investigation and Evaluation of Cracking Incidents in Piping in ,

Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission. September 1980, wwwse ,o g.s

.=a--.- a.- - . _ .. _ - - -._ - ... - . - ~ . ~

SECTION 3.0 MATERIAL CHARACTERIZATION 3.1 Pipe and Weld Materials The pipe material of the pressuri'er surge line for the Kewaunee Nuclear Plant are t.376/TP316 and A403/WP316. h M are a wrought prcduct form of the type ustd for tre primary leap piping of several PWR plants. The surge line is connected to the primary loop nozzle at one end and the other end of the surge line is connected to the pressurizer noz:le. The surge line system doet not include any cast pipe or cast fitting. The welding processes used are shielded n.etal arc (SMAW) and submerged arc (SAW). Weld locations are

, identitled in Figure 3-1.  ;

In tha following section the tensile properties of the materials are presented for use in the leak-before-break analyses, a

c

- 3.2 Material Properties g 1

The room temperature mechanical properties of the Kewaunee Nuclear Plant surge a line matrials were ottained from the Certified Materials Test Reports and are given in Table 3-1. The room temperature ASME Code minimum preperties are given in Table 3-2. It is seen that the measured properties well exceed those of the 9. n. :he representative minimum and v erage tensile properties were establ' ae i se Table 3-3). The material properties at temoeratures (135'F, 205*F, 455*F, and 653*F) are required for the leak rate and stability analyses discussed later. The minimum and average tensile preperties were calculated by using the ratie of the ASME Code Section 111 properties at the temperatures of interest stated above. Table 3-2 shows the tensile properties at various temperatures. The mcdulus of elasticity values were established at various temperatures from the ASME Code Section Ill (Table 3-4). In the leak-before-break evaluation, the representative minimum properties at temperature are used for the flaw stability evaluations and the representative ou, cm.i io 31

average properties are used for the leak rate predictions. The min mum i

ultimate stresses are used for stability analyses. These properties are .,

summarized in Table 3-3.

3,3 References 3-1 ASME Boiler and Pressure Vessel Code Section III, Division 1, Appendices July 1, 1989.

> J 4

I G 9 mwom" ' 3-2 l

1 TABLE 3-1

, Room Temperature Mechanical Properties of the Pressurizer Surge Line Materials and Helds YIELD ULTIMATE ID_ HEAT NO./ SERIAL NL BATERIAL STRENGTH STRENGTH [LQNL R/2 (psi) (psi) (%) (%)

1 J2338/6302 A376/TP316 41,800 84,900 44.4 52.5 43,700 90,400 48.1 67.9 2 .J2338/6303 A376/TP316 41,600 86,900 46.6 68,9 44,300 86,400 49.1 69,7 3 J2338/6303 A376/TP316 41,600 86,900 46.6 68.9 44,300 86,400 49.1 69.7 4 J2009/5794 A376/TP316 41,900 86,400 51.8 65.9 57,300 81,400 57.4 71'.4 5 J2471/1048- A403/HP316 41,600 84,400 51.2 70.2 Shop Held (SH) - Fabricated by SAH (worst case selected for conservatism)

Field Held-(FH) - Fabricated by GTAH/SMAH combination 5369s/062691:10 3-3

- . ~ . . . . _ . __ . . . - - . _ , _ _ . _ _ _ . - - _ _ . .._. _- .. _.. - . . . _ . . . _ _ . . _ _ . _ _ . _ . _ . . _ . . . . . _ . _

TABLE 3-2

Room Temperature ASME Code Minimum Properties Material Yield Stress Ultimate Stress (psi)

(psi) 1 4

A376/TP316 30,000 75,000 A403/WP316 30,000 75,000 4

b 4

o e

w i

on.om io 3-4 l I

TABLE-3-3 Representative Tensile Properties for Kewaunee

Minimum Temperature Minimum Average Ultimate Material ('F)- Yield (psi) Yield (psi) (psi)

A376/TP316 -135 39,560 42.370 81,400 205. 35,610 38,140 81,300 455 -28,520 30,550 77,920

'653- 25,630 27,490 77,920 A403/WP316 135 39,560 39,560 84,400

.205. 35,610 35,610 84,310 455 28,520 28,520 80,790 653 25,630 25,530 80,790 i**

1143,/031491 to 3-5

TABLE-3-4 Modulus of Elasticity (E)

Temperature E (ksi)

(*F) 135 27,950 205 27,600 455 26.115 653. 25,035 e

't l -

l.

i- ..

,5103s/031491 to 3-6

FW: FIELD WELD e SW: SHOP WELD I

FW RCL () SW s

PZR t SW FW v .

Figure 3-1 Kewaunee Surge Line Layout son..cn w ,a 3,y

SECTION 4.0-LOADS FOR FRACTURE MECHANICS ANALYSIS

. Figures 3-1=shows schematic layout of the surge line for Kewaunee and identify the weld locations.

-The stresses due to axial loeds and bending moments were calculated by the following equation:

s=k+ (4-1) where, o =- stress F = axial lead M = bending moment A = metal cross-sectional area -

~Z = section modulus 1The bending moments for the desired loading combinations were calculated by

-the_following equation:

MB

  • IN Y "Z ) (4-2)-

where-.

M B

= bending moment for required loading My = Y component of bending moment N

Z

= Z c mp nent f bending moment

-The axial load rnd bending moments for crack stability analysis and leak rate predictions are computed by the methods to be explained in Sections 4.1 and 4.;2 which follow.

. siswniano 43  ;

4!1L Loads for Crack Stability Analysis

'The faulted loads for the crack stability analysis were calculated by the absolute sum method as follows: ,

=

F_ IF DW I + IF7gl + lF pl + IFSSE I (4-3) r My =

IMYDWl + (MY TH I + IN Y SSE I (4~4) j + IM ZDW I + IN M =

7 IM Z TH Z SSE l (4-5)

DW = Deadweight TH = Applicable thermal load (normal or stratified)

P = Load due to internal pressure SSE = SSE loading including seismic anche ration 4.2 Loads for Leak Rate Evaluation The normal operating loads for' leak rate predictions were calculated by the algebraic sum method as follows: *

=

F FDW + FTH + pF (4-6)

N

  • Y INY)DW * (NY)TH (4-7)

M =

z (MZ )DW + (NZ )TH (4-8)

The parameters and subscripts are the same as_those explained in Section 4.1.

4.3 Loading Conditions B cause thermal stratification can cause large stresses at heatup and cooldown temperatures in the range of 455'F, a review of stresses was used to identify

.the worst situations for LBB applications. The loading states so identified are given in Table 4-1.

4 e

(

- 518WC3149110 4.g

Seven loading cases were identified for LBB evaluation as given in Table 4-2.

Cases A, B, C are cases for leak rate calculations with the remaining cases being the corresponding faulted situations f cr stability evaluations.

The cases postulated for leak-before-break are sumarized in Tab > 4-3. The cases of primary interest are the postulation of a detectublo leak at normal power conditions (----------------------------------------- ~--'--------------

.................................... .... _ ..~... ._..........,........

..........__...........................__................... 3a,c,c The combination [-----------------------------------------------------------

_____________................... ..............___....)a,c.e 4 a

mwm.' " 4-3

The more realistic cases (.---------_-..--- _----_-__-..-._.-.-_-------- .

_____._____.__________ _________._____ _ ______________ a,c.e

[.______________________________________________________________________

_-------------_------------]a.c.e The logic for this AT (_-_____]a,c.e is based on the following:

Actual practice, based on experience of other plants with this type of situation, indicates that the plant operators complete the cooldown as quickly as possible once a leak in the primary system is detected. Technical Specifications require cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> but actual practice is that the plant depressurizes the system as soon as possible once a primary system leak is detected. Therefore, the hot leg is generally on the warmer side of the limits ()200'F) when the pressurizer bubble is quenched. Once the bubble is quenchen, the pressurizer is cooled down fairly quickly reducing the AT in the system.

4.4 Summary of Loads and Geometry The load combinations were evaluated at the various weld locations. Normal loads were determined using the algebraic sum method whereas faulted leads eere combined using-the absolute sum method.

1 5369s/062691:10 4-4

4.5 Governing Locations 2

All the welds at Kewaunee surgeline are fabricated using the SMAW and SAW procedure. The following governing locations were established for the welds.

Figure 4-1 shows the governing locations.

SHAW Weld Node 1020.

SAW Weld Node 1240.

The loads and stresses at these governing locations for all the loading combinations are shown in Tables 4-4.

~

4-5

TABLE 4-1 Types of Loadings Pressure (P) <

Dead Weight (DW)

Normal Operating Thermal Expansion (TH)

Safe Shutdown Earthquake and Seismi: Anchor Motion (SSE)a

. a,c.e a

5SE is used to refer to the absolute sum of these loadings.

saww.. io 4-6

TABLE'4-2 Normal and Faulted Loading Cases for Leak-Before-Break Evaluations CASE A: This is the normal operating case at 653'F consisting of the algebraic sum of the loading components due to P, DW and TH, a,c.e CASE B:

I CASE C:

CASE D: This is the faulted oporating case at 653*F consisting of the a. solute sum (every component load is taken as positive) of P, DW, TH and SSE.

CASE E: 8'C

CASE F:

CASE G:

1 e t 4603,'102290 $ 0 4y

TABLE 4-3 Associated Load Cases for Analyses A/D This is here-to-fore standard leak-before-break evaluation.

a,c.e

-A/F B/E B/F a

B/G l

a C/G a

These are judged to be low probability events.

t.

l

.acwiemo io 4.g

TABLE 4-4 Summary of LBB Loads and Stresses by Case for Governing Locations Node Case FX (ibs) SX (psi) MB (in-lb) SB IP8i) ST (psi) 1020 A 134150 4943 835533 13468 18411 1020 - ------ ---- ------- ----- ----- "'C

1020 - ----- --- ------- ----- -----

~ ~

1020 D 151864 5596 1150890 18551 24146 l 1020 - ------ ---- ------- ----- -----

a,c.e 1020 - ----- ---- ------ ----- -----

1020 .

1240 A 140976 5194 564140 9093 14287

~

-1240 - ------ ---- ------ ---- -----

a,c.e 1240 "

" ~

1240 D 149846 5521 1169853 18856 24377

~

- 1240 - ------ ---- ------- ----- --...

a,c e 1240 - ----- ---- ------ ----- -----

1240 ----- ---- ------- ----- -----

e '

e e m2ccaini to 4.g

o PIPE 10" SCHEDULE 140 o MINIMUM WALL THICKNESS 0.875" HIGHEST STRESSED WELD LOCATION (SMAW)

?

1020 s

RCL

, PZR 1240 HIGHEST STRESSED U WELD LOCATION (SAW) .

Figure 4-1 Kewaunee Surge Line Showing Governing Locations swa.m miio 4-10 i

SECTION 5.0 FRACTURE MECHANICS EVALUATION 5.1 Global failure Mechanism Determination of the conditions which lead to failure in stainless steel should be done with plastic fracture methodology becauss of the large amount of deformation accompanying fracture. One method for predicting the failure of ductile material is-the {-------------------Ja,c.e method, based on traditional plastic limit load concepts, but accounting for (----------------Ja,c.e and taking into account the presence _of a flaw, -The flawed component is predicted to fail when the remaining

. net section reaches a stress level at which a plastic hinge is formed. The stress level at which this occurs is termed as the flow stress. (------------------

]a c.e This methodology has been shown to be applicable to_ ductile piping through a large number of experiments and is used here'to predict the critical flaw size in the pressurizer surge line. The failure criterion has been obtained by requiring equilibrium of the section

~

_containing the flaw (Figure 5-1) when loads are applied. The detailed development is provided in Appendix A for a through-wall circumferential flaw in a pipe section with internal pressure, axial force, and imposed bending moments. The limit moment for such a pipe is given by:

a,c.e

[.. ,

. . . .] (5-1) where:

[. ..................................................................

A

..............]a,C,e

+*

e*

5-1

f-ja,c,e (5-2)

. The analytical model described above accurately accounts for the internal pressure as well as imposed axial force as they affect the limit moment. Good agreement was found between the analytical predictions and the experimental results (reference 5-1). Flaw stability evaluations, using this analytical model, are presented in section 5.3.

~

5.2 Leak Rate Predictions Fracture mechanics analysis shows in general that postulated through-wall cracks in the surge line would remain stable and do not cause a gross failure ,

of this conaonent. However, if such a through wall crack did exist, it would be desirable to detect the leakage such that the plant could be-brought to a safe shutdowr condition. The purpose of this section is to discuss the method which will be used to predict the flow through such a postulated crack and present the ieak rate calculation results for through-wall circumferential cracks.

5.2.1 General Considerations The flow of hot pressurized water through an opening to a lower back pressure (causing choking) is taken into account. For long channels where the ratio of the channel length, L, to hydraulic diameter, DH ',c$l/U H ) is greater than

[ _)a,c.e, both (------------------------------Ja .e must be considered.

In.this situ: tion the flow can be described as being single phase through the channel until the local pressure equ:!: the :sturation pressure of the fluid. ,,

l

! 5 " 2 ' * "' ' '

5

.At this point, the flow begins to flash and choking occurs. Pressure losses due t'o momentum changes.will dominate for (------------Ja.c,e However, for

-large L/Dg values, the friction pressure drop will become important and must 3 be considered along with the Nmentum losses due to flashing.

5.2.2 Calculational Method in using the (---------------------------------------------------------------

................................................)a,c,e ,

The flow rate through a crack was calculated in the following manner. Figure 5-2 from reference 5-2 was used to estimate the critical pressure, Pc, for the primary loop enthalpy condition and an assumed flow. Once Pc was found for a given mass flow, the (------------------------------------------------]a c.e j was found from figure 5-3 taken from reference .5-2. For all cases considered, since(--------------------------Ja,c,e Therefore, this method will-yield the two phase pressure drop due to momentum effects as illustrated-in figure 5-4. Now using-the assumed flow rate, G, the frictional pressure drop can be calculated using.

~

APf = [- . ..

...]8'C (5-3) where the friction factor f is determined using the (--------------Ja,c,e The crack relative roughness, c, was obtained from fatigue crack data on stainless steel samples. The relative roughness value used in these calculations was (----------------Ja,c,e RMS.

The frictional pressure drop using Equation 5-3-is then calculated for the assumed flow and added to the (-------------------------------------------

t'

............Ja.c,e to obtain the total pressure drop from the system under t

considoration to the atmosphere. Thus, m wo w .i '

5-3

Absolute Pressure - 14.7 = [4-----------------------------------------Ja,c.e (5-4) s for a given assumed flow G. If_the right-hand side of equation 5-4 does not agree with the pressure difference between the piping under consideration and the atmosphere, then the procedure is repeated until equation 5-4 is satisfied to withir. an acceptable tolerance and this results in the flow value through the crack.

5.2.3 Leak Rate Calculations Leak rate calculations were performed as a function of postulated through-wall crack length for the critical locations previously identified. The crack opening area was estimated using the method of reference 5-3 and the leak rates were calculated using the calculational methods described above. The leak rates were calculated using the normal operating loads at the governing node identified in section 4.0. The crack lengths yielding a leak rate of 10

-gpm (10-times the leak detection capability of.1.0 gpm) for critical location at the Kewaunee Nuclear Plant pressurizer surge line are shown in Table 5-1. '

The Kewaunea plant has an RCS pressure boundary leak detection system which is consistent with the guidelines of Regulatory Guide 1.C for detecting leakage '

of 1 gpm in one hour.

5.3 Stability Evaluatien

'A' typical segment of the pipe under maximum loads-of axial force F and bending-moment M-is schematically illustrated as shown in figure 5-5. -In order to ,. ; ,

calculate the critical flaw size. plots of the limit moment versus crack length are generated _as shown in figures 5-6 to 5-13. The critical flaw size corresponds to the intersection'of this curve and the maximum load line. The critical flaw size is calculated using the lower bound base metal tensile ,

properties-established in section 3.0.

9 4 mwmoi ta 5-4 l . -

The welds at-the location of-interest (i.e. the governing location) are SAW

,. and SMAW. Therefore, "Z" factor correction for SMAW and SAW welds were applied (references 5-5 and 5-6) as follows:

& = 1.15 (1 + 0.013 (0.0. - 4)] (for SMAW) (5-5)

Z = 1.30 (1 + 0.010 (0.D. - 4)] (for SAW) (5-6) where OD is the outer diameter in inches. Substituting OD = 10.75 inches, the 2 factor was calculated to be 1.25 for SMAW and 1.39 for SAW. The applied loads were increased by the Z factors and the plots of limit load versus crack leng,th were generated as shown in figure 5-6 to 5-13. Table 5-2 shows the summary of critical flaw sizes for Kewaunee.

5.4 References 5-1 Kanninen,_M. F. et al., " Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks' i.PRI 'D-192, September 1976.

5-2 (---------------------------------- -----------------------------------

............____ ..............ja,c.e 5-3 Tada, H., "The Effects of Shell Corrections on Stress Intensity Factors and the Crack Opening Area of Circumferential and a longitudinal Through-Crack in a Pipe," Section 11-1, NUREG/CR-3464, September 1983.

5 NRC letter from M. A. Miller to Georgia Power Company, J. P. O'Reilly, dated September 9, 1987.

5-5 ASME Code Section XI, Winter- 1985 Addendum, Article IWB-3640.

5-6 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register /Vol. 52, No. 167/ Friday, August 28, 1987/ Notices, pp. 32626-32633.

vn,mauc io 5-5

TABLE 5-1 Leakage flaw Size Node Point 1.oad Case Temperature Crack length (in.)

(*F) (for 10 gpm leakage) 1020 a,c.e e

1240 5

e

( *.

l l'

l-5092s <C21291 10 5-6

y

-i TABLE 5-2 e.

Summary of Critical Flaw Size Critical Node Point Load Case Temoerature Flaw Size (in)

(*F) 1020 a,c.e

.. 1240 muses m O

A y 0 0 %

$092s/021291 10 5-7

s

._ a,c,e 6

4 Figure 5-1 Fully Plastic Stress Distribution uu m"' '

5-8

e

- a,c.e 1

i n

=

E l

a 8

a

  • W M

~

x Figure 5-2 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures SCS2 s /C 20891 10 g,n J 3

s 5

4 a,c,e I

e z

W s

3 .

w s

4. .

a M

a u

l l

l l

1 Figure 5-3 [ Critical or Choked]a,c.e Pressure Ratio as a Function of L/D wu, um* 'o 5-10

i 4

a,c,e a.:,e

.  ! /

4

'w- s Figure 5-4 Idealized Pressure Drop Profile Through a Postulated Crack w.v:m. : 5 ,11

[qg ,

Cr ~ ( ~

DL -

).f, ci

\Lf_k

/ y N

~ .- m ,

(CQ * -

Q I

\

! l ,

I I

i l

I l

1 I .

I l

.. d .__I I

I I l

<l l I I I I I

I I I A

G\ j/M-s s Figure 5-5. Loads Acting on the Mocel at the Governing Location mmmm ic 5- M

I d

e

a. C. e

~ -

4 Ib P!PE OD=10.75 T= .800 SICY=25.6 SIGU=77.9 Fa=152. H=.115E+04 Figure 5-6. Critical Flaw Si:e Prediction for Node 1020 Case D im.wni se 5-13

l s

9

a. C, e i

PIPE OD=19.75 T= .880 SICV=25.6 SICll=77.9 Fa=152. H= .12.t E+ 04 Figure 5-7. Critical Flaw Size Prediction for Node 1020 Case E m2. caos , in 5-14

l l

l 4

a,c,e PIPE OD=10.75 T= .000 SIGY=35.6 SIGU:01.3 Fa=34.6 H:915.

figure 5-8 Critical flaw Size Prediction for Node 1020 Cr.se F scia. u .. ,e b '. 5

4 C. e PIPE OD=10.75 T= .800 SIGY=39.6 S!GU=01.4 Fa=36.4 H=.155E+04 Figure 5-9 Critical Flaw Size Prediction for Nude 1020 Case G u>.wm.. 1 5-16

a. C. e M

.. l PIPE OD=10.75 T= .800 SIGY=25.6 SICU:77.9 Fa=150. M=.117E+04 Figure 5-10 Critical Flaw Size Prediction for Node 1240 Case D i m .s m .iio 5-17

  • . l l

l 8, C, e

?

r OI PIPE OD=19.75 T= .889 SIGY=25.6 SIGU=77.9 Fa=159. M=.191E+94 Figure 5-11 Critical flaw Size Prediction for Node 1240 Case E nia.4:'*

5-18

l e

_ d, C, e n

I Eb PIPE OD=10.75 T= .800 SICY=28.5 S I Cll = 7 7 . 9 Fa=32.0 M:042.

Figure 5-12 Critical Flaw Sizo Prediction for Node 1240 Case F mr umiio 5-19

'c i

s.

a, c. e i

PIPE OD=19.75 T= .889 SICY=28.5 SIGU=77.9 Fa=37.3 M=.172E+94 l

l Figure 5-13 Critical Flaw Size Prediction for Node 1240 Case G snwesus, ss 5.gn Y ..._ .-- ..- - -...-.-- - -- . - - - - - . - .. -- - - . - -- -

WEsTINGHouss P .oPRIETARY class 2

}

SECTION 6.0 y ASSESSMENT OF FATIGUE CRACK GROWTH 6.1 Introduction To determine the sensitivity of the pressurizer surge line to the presence of small cracks when subjected to the transients discussed in WCAP-12841, fatigue crack growth analyses were performed. This section summarizes the analyses and results.

Figure 6-1 presents a general flow diagram of the overall process. The methodology consists of seven basic steps as shown in figure 6-2. Steps 1 through 4 are discussed in WCAP-12841. Steps 5 through 7 are specific to fatigue crack growth and are discussed in this section.

There is presently no fatigue crack growth rate curve in the ASME Code for austenitic stainless steels in a water environment. However, a great deal of work has been done (References 6-1 and 6-?) which supports the development of

, such a curve. An extensive study was performed by the Materials Property Council Working Group on Reference Fatigue Crack Growth concerr.ing the crack growth behavior of these steels in air environments, published in referente 6-1. A reference curve for stainless steels in air environments, based on

~

this work, is in the 1989 Edition of Section XI of the ASME Code. This curve is shown in figure 6-3.

A compilation of data for austenitic stainless steels in a PWR water environment was made by Bamferd (reference 6-2), and it was found that the effect of the environment on the crack growth rate was very small. For this reason it was estimated that the environmental factor should be set at 1.0 in the crack growth rate equation from reference 6-1. Based on these works (references 6-1 and 6-2) the fatigue crack growth law used in the analyses is

,, as shown in figure 6-4.

mww.im 6-1

6.2 initial _ Flaw Size Various initial surface flaws were assumed to exist. The flaws were assumed to be semi elliptical with a six-to-one aspect ratio. The largest initial ,,

flaw assumed to exist was one with a depth equal to 10% of the nominal wall thicknet,, the maximum flaw size that could be found acceptable by Section XI of t.se ASME Code.

6.3 Results of FCG Analy111 Fatigue crack growth analyses were performed at locatiens 1 and 2 where detailed fracture; mechanics analyses as described in Sections 5 were completed. It should be noted that location 1 is near the reactor coolant loop nozzle and location 2 is near the pressurizer nozzle.

Results of the fatigue crack growth analysis are presented in table 6-1 for an initial flaw of 10% minimum wall thickness.

Conservatisms existing in the fatigue crack growth analysis are listed below.

1. Plant operational transient data has shown that the conventional design transients contain significant conservatisms.

[_ _________________________..________ ....._____ ..___________

____________ _.___________________]a,c.e

4. FCG neglects fatigue life prior to initiation.

5183 /070191:10 6-2

WESTINGHOUSE PRoPRitTARY CLASS 2 6.4 References ,

~

6-1. James, L. A. and Jones, D. P., " Fatigue Crack Growth Correlations for

  • Austenitic Stainless Steel in Air," in Predictive Capabilities in Environmentally Assisted Cracking, ASME publication PVP-99, December 1985.

6-2. Bamford, W. H., " Fatigue Crack Growth of Stainlect Steel Reactor Coolant Piping in a Pressurized Water Reactor Environment," ASME Trans. Journal of Pressure vessel Technology, Feb.1979.

4 4

sen ,$3a o ie 6-3

! TABLE 6-1 FATIGUE CRACK GROWTH RESULTS FOR 10% OF WALL INITI AL FLAW SI2E Initial Initial final (40 yr) Final Flaw Location Position Si:e (in) (7. Wall) Size (in) (% Wall)

~ ~

a,c.e

=

e -

I 518 3 t 'C 31491 10 g.4

. _ _ _ _ - _ _ _ _ - _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _

i l

,, OETERMINAflON OF THE EFFICT5 0F TH(RM AL STR ATIFICATICQ , g M

Figure 6 1 Determination of the Effects of Thermal Stratification on Fatigue I.rs:t Growth so.mua. i i.

6-5

l i

i

.. .................... a,c.e 9

Figure 6-2 Fatigue Crack Growth Methodology s m. '" '" ' ' 6-6

_ _ = _ . .

I A

seair*

i/ lV/Y

. ~ . . , . i y

/

))

,e If! > .,

1 > 1. r l

I +

ll l ll l l/ - {i /

1 / / J i / i

!/ //// i

! !/ //!  !

/

.i. . .

//// . . . , ,

,t*

, /'"

- 1 s ll lll l

} >I Ii // i j i / !l i/ i 2 // / // / i

~

\ / / //ll '

'/ / if w ....

'd i, ,

, , , , "I f I f I / "

i^

. t ,  : si '

/ / f I I/

l l / /} l j fjjj f

) ) ,l 4

,,->< / / fl l ,,,

t... ,,,

4 4 ite M Figure 6-3 Fatigue Crack Growth Rate Curve for Austenitic Stainless Steel sootsionna to 9y

3 lj=CFSEAK.30 where

  • Crack Growth Rate in inches / cycle hk C = 2.42 x 10 20 F = Frequency factor (F = 1.0 for temperature below 800=F)

S = R ratio correction ($ = 1.0 for R = 0; S

  • 1 + 1.5R for 0 < R < .8; and S = 43.35 + 57.97R for R > 0.8) .

E = Environmental Facter (E = 1.0 for PWR) ,

4X = Range of stress intensity f actor, in psi .'in R

= The ratio of the minimum K; (K! min) to the maximum K; (K;,, )

Figure 6-4. Fatigue Crack Growth Equation for Austenitic Stainless Steel uowem $ io 6-8

1' LOCATION 1 s

s RCL PZR LOCAT10t4 2 Figure 6-5. Fatigue Crack Growth Critical Locations m.,em.i io 6-9

a -

9 i

't

~

. a.c.e A

Figure 6-6. Fatigue Crack Growth Controlling Positions at Each Location

>= we =. i .

6-10

SECTION 7.0 ASSESSMENT OF MARGINS s'

In the preceding sections, the leak rate calculations, fracture mechanics analysis and fatigue crack growth assessment were performed. Margins at the critical location are 6ummarized below:

in Secton 5.3 using the IWB-3640 approach (i.e. "Z" factor approach), the

" critical" flaw sizes at the governing locations aie calculated. In Section 5.2 the crack lengths yielding a leak rate of 10 gpm (10 times the leak detection capability of 1.0 gpm) for the critical locations are calculated.

The leakage size flaws, the instability flaws, and margins are given in Table 7+1. The margins are the ratio of instability flaw to leakage flaw.

The margins for analysis combinatioli cases A/D, (- ----------------Ja.c.e well exceed the factor of 2. The margin for the extremely low probability event defined by (--------Ja.c.e 43 (,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,

j ....................................ja.C,0 As stated in Section 4.3, the

.. probability of simultaneous occurrence of SSE and maximum stratification due to shutdown because of leakage is estimated to be very low.

In this evaluation, the leak-before-break methodology is applied conservatively. The conservatisms used in the evaluation are summarized in Table 7-3.

1 5141s/031891 10 7,

l

._ . . . . . . . 9' ..

TABLE 7-1 Leakage flaw Sizes. Critical flaw Si:ss and Margins a ed Critical Flaw Leakage Flaw Si:e (in) Size (in) Margin Nr

-.es-- i.stle 1020 A/D 10.08 3.10 3.3 a,c,e 1240 A/D 9.19 3.75 2.5 I ... ..... .... ...

a,c.e .

~

8 These are judged to be low probability events un. nini io 7-2

TABLE 7-3 I

LBB Conservatisms 4

o Factor of 10 on Leak Rate o Factor of 2 on Leakage Flaw for all cases (except for B/G which has 1.9 and it is a low probabilty case) o Algabraic Sum of Loads for Leakage o Absolute Sum of Loads for Stability o Average Material Properties for Leakage o Minimum Material Properties for Stability e

5143 n r031491 10

l SECTION

8.0 CONCLUSION

S l'

. This report justifies the elimination of pressuri:er surge line pipe breaks as the structural design basis for Kewaunee Nuclear Plant as follows:

a. Stress corrosion cracking is precludeC by use of fracture resistant materials in the piping system anc controls on reactor coolant chemistry, temperature, pressure, and flow during normal cperation,
b. Water. hammer should not occur in the RCS piping (primary loop and the attached class 1 auxiliary lines) because of system design, testing, and operational considerations,
c. The effects of low and high cycle fatigue on the integrity of the surge line were evaluated and shown acceptable. The effects of thermal stratification were evaluated and shown acceptable, i
d. Ample margin exists between the leak rate of small stable flaws and the criterion of Reg. Guide 1.45.
e. Ample margin exists between the small stable flaw sizes of item d and the critical flaw si:e.

The postulated reference flaw will be stable because of the ample margins in d, e and will leak at a detectable rate which will assure a safe plant shutdown.

Based on the above, it is concluded that pressuri:er surge line breaks should not be considered in the structural design basis of Kewaunee Nuclear Plant.

l l

i m, e..' "

8-1

A. *--__4 -.A., _. - - -, . , - . - - e. m.2.- -. ._2a E A--4 , - - . u .-s 1 --

l APPENDIX A

+

LIMIT MOMENT 9

  • 4 tills /031491 10 A-1

APPENDlX A LIMli MOMENT 3

[

Ja.c.e

, e.

I

{

l l

veww ,, ,o g.2

m. -,.,- .---.--- - - - - - - - - - - - - - . - - - - , - . - . - , - - - - - - - - - - - - - - - - - - - - , . - - - - - - - . - - - - - - , - . - - - , - - - - - - - , . - - . - . - . - - - . - - - - - - - - , - - - - - - - - - - - - - - - - - - - - - - - - - , - - - - - - - - - - - - - - - - - - -

C *.

9 I I r

C 0)

CC

~

c

. e sad V

9 6

%)

P*

3 5

0 6

4*.

w 4

A w

e en

+

3 k

b e

em4 4

EL L.

b

. - l 1 f

j t

h A-3

. - - - . . . -

Retrieved from "https://kanterella.com/w/index.php?title=ML20077K083&oldid=2764226"