ML20137N126

Rev 0 to Leak-Before-Break Evaluation for Three Locations on Recirculation Sys at Fermi 2

ML20137N126
Person / Time
Site:	Fermi
Issue date:	03/25/1997
From:	Chen P, Cofie N, Miessi G STRUCTURAL INTEGRITY ASSOCIATES, INC.
To:
Shared Package
ML20137N122	List: ML20137N126 NRC-97-0043, Provides Info Re Recently Identified Issued Involving Effects of Postulated HELBs Inside Containment on EECWs & Utils Basis for Conclusion
References
SIR-97-029, SIR-97-029-R00, SIR-97-29, SIR-97-29-R, NUDOCS 9704080264
Download: ML20137N126 (54)

Text

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Report No.: SIR-97-029  :

Revision No.: 0 Project No.: DECO-01Q ProjectFileNo.: DECO-01Q-401 '

March 1997 -

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Leak-Before-Break Evaluation for Three Locations on the Recirculation System at Fermi 2 i

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Prepared for:

Detroit Edison Company Prepared by:

4 StructuralIntegrity Associates,Inc. ,

San Jose, CA Prep,ared by: D te. S 2F/97 l

@A. Messt Date: Shfl'll P. L.~ Chen j

[ Date: 3/Z.f/97

. G. f fic// " / f Reviewed and j

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Approved by: Date: 3/2r 97  !

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P. C. Riccardella i

9704080264 970404 PDR ADOCK 05000341 P PDR ;

f StructuralIntegrityAssociates,Inc.

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REVISION CONTROL SHEET Document Number- SIR-97-029

Title:

Leak-Before-Break Evaluation for Three Locations on the Recirculation System at Fermi 2 -

Client: Detroit Edison Comoany SI project Number: DECO-010 Section Pages Revision Date Comments i-v all 0 3/25/97 InitialIssue 1 1-1 to 1-4 7 3/25/97 InitialIssue 2 2-1 to 2-2 0 3/25/97 InitialIssue 3 3-1 to 3-2 0 3/25/97 InitialIssue 4 41 to 4-6 0 3/25/97 InitialIssue 5 5-1 to 5-20 0 3/25/97 InitialIssue 6 6-1 0 1/25/97 InitialIssue 7 7-1 to 7-2 0 3/25/97 InitialIssue Appendix A A-0 to A-10 0 3/25/97 InitialIssue O

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EXECUTIVE

SUMMARY

This report documents evaluations performed by Structural Integrity Associates to determine the ~

leak-before-break (LBB) capabilities at three locations on the recirculation system piping at Fermi

2. The three locations are adjacent to the reactor pressure vessel on two 12-inch discharge risers and one 28-inch suction line. At each of these locations, three closely spaced welds; nozzle-to-safe end (N-SE), safe end-to-pipe (SE-P) and pipe-to-elbow (P-E) were considered as potential break locations.

i The evaluations were performed taking guidance from NUREG-1061, Vol. 3 which provides NRC accepted methodology and criteria for performing LBB analysis. Additional guidance was taken from EPRI Report NP 4991 which provides specific application ofLBB to BWR piping. In summary, the LBB methodology involves the determination of through-wall critical flaw size (unstable throughwall flaw length) for the piping system component under normal plus SSE se:s nic loads and then computing the leakage under nonnal loads through a fraction of the critical flaw size (leakage flaw size). NUREG-1061 Vol. 3 requires a safety factor of two between the critical flaw size and the leakage flaw size. The computed leakage is then compared to the plant detection system to ensure that the leakage through the leakage flaw can be detected. Circumferential flaws are considered since they are more limiting than axial flaws.

Critical flaw sizes were calculated using elastic-plastic fracture mechanics and lower bound material propenies for the components. The calculated critical flaw sizes for the 12-inch risers ranged from 6.71 inches to 13.82 inches. The critini flaw sizes for the 28-inch suction line range from 24.53 inches to 27.97 inches. -

Leakage was calculated for throughwall flaws of various fractions of the critical flaw size. For the 12-inch risers the minimum leakage at one-quaner critical flaw size is about 0.2 gpm. The leakage increases to approximately 15 and 5 gpm at one half and three quarters the critical flaw size, respectively. The cor esponding leakages for the 28-inch suction line are about 1.5,10 and 35 gpm ,

respectively.

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Even though the IGSCC resistance of all the welds considered in this evaluation have been treated L'y NUREG-0313, Revision 2 accepted IGSCC improvement processes, a conservative IGSCC evaiuation was performed using the NUREG-0313, Revision 2 bounding crack growth law to determine times to grow from various sub-critical flaw sizes to the critical flaw sizes. The evaluation showed that it takes at least one cycle (over eighteen months) for a through-wall crack at half the critical flaw size to grow to three quarters the critical flaw size. This demonstrates that there is adequate time for the plant to take appropriate action before the critical flaw size is reached, once the leakage rates determined in this report are detected by the plant monitoring systems.

i The effect of degradation mechanisms which could impact the LBB evaluations were considered in the evaluation and it was concluded that they would have no effect on the results. These mechanisms are water hammer, corrosion (intergranular stress corrosion cracking (IGSCC) and erosion-corrosion) and fatigue. Experience from the BWR industry has shown that water hammer is not a concern for the recirculation system. All welds considered in this evaluation are resistant to IGSCC based on the fact that they have received mechanical stress improvement process (MSIP), induction heating stress improvement (IHSI) or were solution annealed. The material of the recirculation system is Type 304 stainless steel which is not susceptible to erosion-corrosion. Since severe thermal transients and other cyclic loads are not expected in the recirculation system, there is no active mechanism to lead to fatigue crack growth.

SIR-97-029, Rev. O il f StructuralIntegrityAssociates,Inc.

I Table of Contents -

Staunn P.agn

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1:

1.I' Backgro u nd . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............ 1 1.2 Leak-Before-Break Methodology . . . . . . . . . . . . . . ........... .. 1-1 2.0 CRITERIA FOR APPLICATION OF LEAK-BEFORE-BREAK APPROACH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.1 Criteria for Through-Wall Flaws . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2- Criteria for Part-Through-Wali Flaws . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.3 O th er Mechanisms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 3.0 CONSIDERATION OF WATER HAMMER, CORROSION AND FATIGUE . 3-l 3.1 Wa ter Ham m er . . . . . . . . . . . . . . . .. . . . . . . . . . . ... . ............ 3-1 3.2 Corrosion . .... ................... ....................... 3-1 3.3 Fa t i g u e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -2 4.0 PIPING MATERIALS AND STRESSES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 Ma t erial Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2 Pipin g S tress es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............. 4-2 4

5.0 LEAK-BEFORE-BREAK EVALUATION . . . . . ........ . .. ........ 5-1 5.1 Evaluation of Critical Flaw Sizes . ..... .............. ........ 5-1 5.2 Leak Rate Determination . . . . . . . . . . . . . . . . . . .. . . . . . . . . 5-3 5.3 Crack Growth Analysis for Circumferential Through-Wall Flaws . . . . 5-9 5.4 LBB Evaluation Results and Discussions . . . . . . . . . . . . . . . . . . . . . . . 5-10 6.0

SUMMARY

AND CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . 6-1

7.0 REFERENCES

. r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1

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, APPENDIX A Loads and Stresses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-0 SIR-97-029, Rev. O iii g gg3g3, g,

s List ofTables ,

l E b Table 4-1 Material Properties Used for Type 304 Stainless Steel in LBB Evaluation . . . 4-3 Table 4-2 Summanf of Stresses .......... .............. ............... 4-4 .

Table 51 Summary of Critical Flaw Size . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-11 g Table 5-2 LBB Evaluation Results for Detectable Leakage ,

(Pressure + Resultant Moment Stresses at Leak)'. . . . . . . . . . . . . . . . . . . . . 5-12

i Table 5-3 Summary of Through-Wall IGSCC Growth Evaluation . . . . . . . . . . . . . . . . 5-13 l

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a. j List ofFigures ,

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4-1. Ramberg-Osgood Tme Stress-True Strain Curve for Type 304 Stainless Steel . . . . . . . 4-5 4-2. Lower Bound J-Resistance Curve for Type 304 Stainles> 7 teel . . . . . . . .. . . . . . . 4-6 5-1. Flow of Subcooled Water Through a Crack . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-14 5-2. Leakage Versus Crack Size for Riser N2F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15 -:

i 5-3. Leakage Versus Crack Size for Riser N2D on Loop B . . . . . . . . . . . . . . . . . . . . . . . . 5-16 j 5-4. Leakage Versus Crack Size for Riser NIB on Loop B . . . . . . . . . . . . . . . . . . . . . . . 5-17 i

5-5. Crack Growth Analyses Results for Riser N2F on Loop A . . . . . . . . . . . . . . . . . . . . 5-18 1

3 5-6. Crack Growth Aaalyses Results for Riser N2D on Loop B . . . . . . . . . . . . . . . . . . . . 5-19 '

5-7. Crack Growth Analyses Results for Riser NIB on Loop B . . . . . . . . . . . . . . . . . . . . 5-20 i

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1.0 INTRODUCTION

i 1.1 - Background '

This report documents evaluations performed by Structural Integrity Associates (SI) to determine the leak-before-break capabilities at three locations on the recirculation sysu :t Fermi 2. These  ;

, evaluations are necessary because a pipe break at these locations could impact the safety function of the Emergency Equipment Cooling Water (EECW) system which is in the vic' m ity of these postulated breaklocations. -

1 The three locations considered in this evaluation are shown in Figure 1-1[11]. All three locations are in close proximity ofreactor pressure vessel nozzles. The first location is on 12-inch discharge riser N2F in the " A" loop. The second and third locations are on 12-inch discharge riser N2D and 28-inch section line NIB. on the "B" loop. At each of these locations, there are three closely spaced welds as shown in Figure 1-2; nozzle-to-safe end (N-SE), safe end-to-pipe (SE-P) and pipe-to-elbow (P-E).

All these welds are considered as potential break locations in this evaluation.

1.2 Leak-Before-Break Methodology The concept ofleak-before-break (LBB) implies that any crack or defect wijch develops in a component will grow to a through-wall configuration, and thus be detected by plant monitoring systems before reaching a size that would significantly reduce margins to component rupture in the plant. NUREG-1061, Vol. 3 (1) provides the methodology and criteria for application of LBB for the elimination of protective stmetures such as pipe whip restraints, jet impingement shields etc.,

which provide protection against pipe break. In the context used in this evaluation, LBB is part of an overall assessment to determine suitability for continued plant operation with respect to the safety function of the EECW system, assuming postulated through-wall flaws, using the criteria outlined in NUREG-106'1, Vol. 3. Additional guidance on the application ofLBB specifically to BWR piping is provided in Ref. 2. In summary, the approach provided in NUREG-1061, Vol. 3 approach involves the determinaion of brouah-wall critical flaw size for the piping system component under normal SIR-97-029, Rev. O l-I h StructuralIntegrityAssociates,Inc.

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[

i plus SSE seismic loads and then computing leakage under only normal loads through a flaw with .a length equal to one half of the critical Baw size (the leakage flaw size). The computed leakage is then

compared to the plant detection systems to ensure that the leakage throuah this flaw can be detected. .

l. The acceptance criteria for LBB evaluation per NUREG-1061, Vol. 3 [1] are provided in Section 2 l of this report. The applicable criteria will be used for the LBB evaluation for the three break i

1 locations on the recirculation system at Fermi 2.

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i Yd Break Lccati:n on N2D (B Loop) 270* . 0*

, c RISER #5 12 Break Location X on N1B C (B Loop) RISER #4 I

90 CU - Break Location on N2F

} CD A^

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I PI RlSER #1 12*

x s RISER #2 -

12*

y*" _ m.

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n-h C 28 C3 j 28*

Loop 'S' Only C3 S 1 V

ca

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kj C3

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-r &

Recirc Loop Piping Une A (25 Quadrant)

Identical for Une 'B' '

Except as Noted Figure 1-1.

Postulated Break Locations for the LBB Evaluations in the Recirculation System Loops A and B at Fermi 2 SIR-97-029, Rev. 0 1-3 f Structors!IntegrityAssociates,Inc.

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RPV  !

Reactor Recirc. i Nozzle ,

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s SE-P N-SE p.c Weld Wld Weld

+, i - / V s Y

SA-182-F304 SA-508 CL2 e71sino A-358 CL 1 l TYPE 304 SS .

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Figure 1-2. Details ofWeld Configurations at the Locations for LBB Evaluations I

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2.0

' CRITERIA FOR APPLICATION OF LEAK-BEFORE-BREAK APPROACH ~

NUREG-1061, Volume 3 [1]i ' dentifies several criteria to be considered in determining applicability of the leak-before-break approach to piping systems. -

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Section 5.2 ofReference 1 provides an extensive discussion of the criteria for performing leak-before- -!

break analyses 'Ihe details of that dimmion will not be repeated here, but a summary of the various ,

requirements as applied to the three locations on the recirculation system at Fermi 2, is provided. ,

2.1 Criteria for Through-Wall Flaws l'

' Acceptance criteria for critical stresses and critical flaws are:

t 1.

The flaw size that is required to produce an " acceptable leakage rate" shall be smaller than  !

1 the critical flaw length associated with the maximum stress, including safe shutdown i

earthquake (SSE) seismic loadings, by a factor'of 2.

2.

The stress required to make the " acceptable leakage rate" flaw critical shall be greater than the maximum stress (with SSE) by a factor of at least [i. '

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3.

The net section collapse criterion (NSCC) evaluation approach may be used to compute the l

• critical flaw size provided a safety factor of 3 is placed on the combination of normal operation plus safe shutdown earthquake stresses.

It has been Aund in previous LBB evaluations conducted by Structural Integrity Associates (SI), that

, the second 1d third criteria stated above are generally not bounding. The method described in the first criteria provides a smallerleakage rate than the second criteria. Therefore, only the first criteria will be considered in this report. Furthermore, elastic-plastic fracture mechanics (EPFM) approach is generally conservative relative to the NSCC approach when applied to stainless steel piping.

Therefore, only EPFM principles will be applied in this evaluation.

SIR-97-029, Rev. 0 2-1

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2.2 Criteria for Part-Through-Wall Flaws NUREG-1061 requires demonstration that a long pan-through-wall flaw which is detectable by ultrasonic means will not grow due to fatigue to a depth which would produce instability over the life of the plant. Previous studies performed by Structural Integrity (SI) have shown that in all cases, fatigue crack growth is not very significant and does not affect the outcome of the LBB studies  !

especially for the recirculation system of a BWR in which there is little cyclic loading. Hence, in this .

evaluation, fatigue crack growth evaluation will not be performed.

2.3 Other Mechanisms

, NUREG-1061 limits applicability of the leak-before-break approach to those locations where -

degradation or failure by mechanisms such as water hunmer, erosion / corrosion, and intergranular stress corrosion cracking (IGSCC) is not a significant possibility. These mr-hanisms were considered for the locations under evaluation and are discussed in Section 3 of this :epon.

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3.0 CONSIDERATION OF WATER HAMMER, CORROSION AND FATIGUE NUREG-1061, Volume 3 [1] states that LBB should not be applied to high energy lines susceptible to failure from the effect of water hammer, corrosion or fatigue. These potential failure mechanisms are thus d'.:ussed below with regard to the recirculation system at Fermi 2. It is concluded that the above failure mechanisms do not invalidate the use of LBB for the affected locations on the recirculation system at Fermi 2.

3.1 Water Hammer i

Several studies performed on water hammer experience .in the industry have indicated that the l recirculation system of a BWR is not susceptible to water hammer events. A very comprehensive discussion of the industry experience is provided in NUREG-1061, Vol. 4 [3] and NUREG-0927 [4]. '

Section 4.2 ofReference 2 provides specific evaluation ofwater hammer potential in the recirculation system of a BWR and concludes that water hammer is not a significant event for this system.

3.2 Corrosion i

Two corrosion damage mechanisms which can lead to rapid piping failure are intergranular stress  ;

corrosion cracking (IGSCC) in austenitic stainless steel pipes and flow assisted corrosion (erosion-corrosion or cavitation, primarily)in carbon steel pipes. Although the Type 304 stainless steel piping material for the recirculition system at Fermi 2 is susceptible to IGSCC, the following processes have l

been applied to the weldments shown in Figure 1-2 to make them resistant to IGSCC.

, The nozzle-to-safe end welds received mechanical stress improvement process (MSIP)

The safe end to pipe welds received induction heating stress improvement (IHSI) treatment The pipe to elbow welds were connected in the shop and solution annealed after welding SIR-97-029, Rev. 0 3-1 h StructuralIntegrityAssociates,Inc.

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All these processes are considered acceptable means by NUREG-0313, Revision 2 [5] for mitig these welds against IGSCC. ,

Erosion-corrosion is a mechanism limited to carbon steel piping. It is not a problem for stainless steel pipes and components. Cavitation damage occurs only at valves used for flow control. As such neither of these mechanisms is anticipated to be an active degradation mechanism for the recirculation!

system at Fenni 2. '

3.3 Fatigue i

As mentioned in Section 2.2 of this report, fatigue analyses will not be performed in this report s previous studies performed by SI have shown that fatigue crack growth is generally insignifican' Since severe thermal transients and other cyclic loads are not expected in the recirculation sys there is no active mechanism to lead to fatigue crack growth.

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, .d.0 FIFING MATERIALS AND STRESSES -

b 4.1 ' MaterialProperties  :

1 f - The piping matenals at the locations considered in the LBB evaluation are shown in Tgure 1-2. The pipe and safe and materials are Type 304 stainless steel. The safe end is welded to the SA 508 Class - {

]

- 2 nozzle using Inconel 182 weld metal. Per the welding speci6 cation which was used for fabrication, )

several processes were used for the welding inchmEng submerged arc weldag (SAW), shielded snetal  !

I are welding (SMAW), gas tungsten are  :"=g (OTAW) and gas metal are welding (GMAW).  ;

' Weidments resuhing from the non-flux processes (OTAW and GMAW) have been shown to exhibit j very high teW= and therefore have superior Saw tolerance. These materials typically fhil by net i j section plastic collapse (limit load) (6]. However, the flux weidenants (SAW and SMAW) have  ;

relatively low ductility and toughness. As such, they exhibit unstable crack eve-ian before the  !

flawed pipe section reaches the limit load associated with the non-flux weld flow stresses (6). In order to perform a bounding evaluation, the flux weldment properties will be used in this report, i

The material properties used in the elastic plastic fracture mechanics analyses are shown in Table 4-1, and represent lower bound flux weldment toughness properties published in the literature (6,7]. The i

elastic modulus (E), Code allowable stress intensity (Sm), and lower bound yield strength (o,=o,) and

j. uhimate strength (a) were taken from Section HI of the ASME Boiler dc Pressure Vessel Code [8]

for the temperatures ofinterest. De flow stress is computcd as an average of o, and o , although

. this does not influence the stability analysis results. Ramberg-Osgood tme stress-strain parrasters,

~

a and n, and the J resistance material properties, are taken from Reference 7. Plots of the r: tress-l strain curve and the J-resistance curve are shown in Figures 4-1 and 4 2, respectively.

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I e w -e +,w --- - - - - - v -

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'l 4.2- ' Piping Stresses i

i The piping stresses which are normally ccasidered in a LBB evaluation are due to normal operating  !

condition (NOC) which includes pressure, deadweight and thermal expansion while the reactor is at 'f full power combined with Safe Shutdown Earthquake (SSE). The piping loads for the various weld {

locations were obtained from Reference 9 and are presented in Appendix A. Geometric data for calculating the stresses was obtained from Reference 10. Stresses for LBB evaluation were '  !

conservatively calculated considering the torsional moment. For calculation of critical flaw size, the  !

stress combination due to NOC and SSE loads is used. For leakage calculations, the stress' combination due to NOC is used.- These stress combinations are shown in Tables 4-2. i The axial stress due to internal pressure is calculated from the expression: I P? R l

%et *

(Rl-Rl) ,

where p is the internal pressure, R, is the outside radius, and R,is the inside radius. A pressure of l 1047 psig was used for the suction side while 1254 psig was used for the discharge piping [11].  !

The dead weight and thermal stresses were calculated from the relationship:

5

• a=,+y F

• A Z -

where: -

~ ~

F,is axial force M is the resultant moment = /Ml + Ml + Ml f t

M is the torsional moment  !

M., M., are the bending moments in b and c directions, respectively.

Ais pipe cross-sectional area  !

Z is the section modulus.

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' A summar/ of the stresses calculated from these loads for the various welds is presented in Table 4-2, which also shows the ~ stress combinations for calculation of critical flaw size as well as the stress combinations for determination ofleakage. j Table 4-1 ^  :

Material Properties Used lor Type 304 Stainless Steel in LBB Evaluation >

i E (ksi) 25,570 i

, o, (ksi)

(=o,) 18.85 i o,(ksi) .

63.0 0% (ksi) 41.18-a 11.56

'n 2.88 9 Ju(in-kip /m2) 0.30 .

J. (in-kipfm 2) 5.0  :

C 2.673 N 0.3162.

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3 ' Table 4-2 6

g Summary ofStresses

.0 y STRESS (km)

S Location Wekt -

.o -

PRESSURE DEAD WT. TilERMAl, ' S.S E. P+DW+Til P+DW+Til+SSE - -

Imp A N2F N_-SE 2.883 0.339 3.104 2.587 6.326 8.913 SE-P 4.694 0.512 5.055 4.393 10.261 14.654 P-E ,

4.694 0.173 2.909 3.895- 7.776' I1.67I Imp B N2D N-SE 2.883 0.359 2.440 3.042 5.682 8.724 SE-P 4.694 0.531- 4.012 4.790 9.236 14.026 P-E 4 694 0.136 2.065 2.578 6.894 9.472 Imp B NIB N-SE 3.548 0.102 1.354 2.356 5.004 7.360

.i SE-P 3.994 0.117 1.578 2.430 5689 8.119 p.E 3 994 0 106 1.757 1.366 5 856 ' 7.223 L

j E,

E -

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R c-E -

Ja N

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

True Stress - True Strain Curve Typ. 304 staini. . st i 90 80 -

~

70 -

60 -

T d -

i . 50 -

1 3

40 -

e 30 -

20 -

10 -

j 1

o ,

0 0.2 0.4 True Strain (in/in)

Figure 4-1. Ramberg-Osgood True Stress-Tme Strain Curve for Type 304 Stainless Steel SIR-97-029, Rev. 0 45

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Lower Bound J-Resistance Curve Type 304 Stainless Steet 2.5 -

2.6 -

2.4 -

2.2 -

2 -

g 1.5 -

tr

(~

1.6 -

i 1.4 -

g 1.2 -

n 1

0.8 -

0.6 -

0.4 -

l 0.2 -

0 O 0.2 0.4 0.6 0.8 1 Crock Extension (in.)

Figure 4-2. Lower Bound J-Resistance Curve for Type .'04 Stainless Steel SIR-97-029, Rev. 0 4-6 f StructuralIntegrityAssociates,Inc.

5.0 LEAK-BEFORE-BREAK EVALUATION The LBB approach involves the determination of critical flaw sizes, critical stresses and leakage through postulated flaws. The critical flaw length for a through-wall flaw is that length for which, under a given set of applied stresses, the flaw would become marginally unstable. Similarly, the critical stress is that stress at which a given flaw size becomes marginally unstable. NUREG-1061, Volume 3 (1) defines required margins of safety on both flaw length and applied stress. However, as explained in Section 2, safety margins based on flaw length have been found in previous evaluations to be the more conservative of the two and therefore, only the criterion based on flaw length will be used in this evaluation. Furthermore, previous evaluations have demonstrated that circumferential flaws are more restrictive than postulated axial flaws. , For this reason, the evaluation presented herein will concentrate on circumferential flaws.

5.1 Evaluation of Critical Flaw Sizes Critical flaw sizes may be determined using the net section collapse criterion (NSCC) approach or l the J-Integral / rearing Modulus (J/T) methodology. NSCC is particularly suited for materials with extreme amounts of ductility and toughness such as wrought stainless steel materials, since it assumes that the cross section of the pipe becomes fully plastified at the onset of failure. For materials with lower toughness, J/r methodology is used. In this evaluation, the critical flaw sizes will be determined based on the J/r approach.  !

Methods for calculations ofJ and T for the stability assessment of through-wall circumferential flaws in cylindrical geometries such as pipes has been developed by EPRIin References 12,13, and 14.

This methodology was used for the determination ofcritical flaw sizes using the Structural Integ+y Associates computer program, pc-CRACK (15].

The expression for the J-integral for a through-wall circumferential crack under tension loading (12]

whichis applied in this analysis is:

SIR-97-029, Rev. 0 5-I f StructuralIntegrityAssociates,Inc.

(..,

2

' a' '

R'

. J = f, ' a, , R ' : P y a a, e, c .

'a P h, - .

, t, E .,b, , b , n, -t,. P,.

4 where 1 8 c y a,F -"E f, a,,

R = ' b, t '

s, t, . 4kg2:2 ,

a, =

effective crack length including small scale yielding correction

. R- = nominal pipe radius t = pipe wall thickness F' = elasticityfactor P =

applied load = o. 2n Rt ; _where o is the remote tension stress in the

,i uncracked section.  !

a =

Ramberg-Osgood material coefEcient a' E =

elastic modulus

-l s

o, = yield stress . l e, =  !

yieldstrain c = b-R 2a =

cracklength 2b =

2n R h, =

plasticity factor P,-. =

lhdt load corresponding to a perfectly plastic material n =

Ramberg-Osgood strain hardening exponent.

' Similar equations [12,13,14) are used to compute critical flaw sizes for circumferential through-wa cracks under bending stresses. Crack extensions during stable ductile tearing in the EPFM analyses are conservatively subtracted from the critical flaw length computations.

SIR-97-029, Rev. 0 5-2 i

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{ StructuralIntegrityAssociates,Inc.

The piping stresses consist of tension, bending and torsion stresses. The tension stress is due tc, internal pressure while the bending and torsion stresses are caused by deadweight, thermal and seismic loadings. As described above, only the pure tension or pure moment models were determined to be applicable. Thus, fracture mechanics solutions were developed for two cases: 1) the total axial stress was considered to be tension and 2) the total axial stress was considered to be bending. Then, the critical flaw sizes obtained with the tension model (a,) and with the bending model(a )3 were combined to determine the actual critical flaw size (a,) due to combined tension and bending stress. Linear interpolation was used as described by the following equation:

n I ' I o, 06 a" = a' + a*

o, + o,, o, + o,,

This approach has been shown to provide a good representation of the critical flaw size for combined state of stress. Ciearly, it provides the exact answer at the extremes of pure tension and pure bending.

l For calculation of critical flaw sizes, the worst loading condition associated with using the resultant bending moment (including torsion effects) have been included. This is conservative and will result in the smallest possible critical flaw size. The critical flaw sizes are shown in Table 5-1.

5.2 Leak Rate Determination

~

The determination ofleak rate is performed using the Structural Integrity Associates proprietary program, pc-LEAK (16]. The methodology employed in pc-LEAK involves the determination of crack opening area (COA), with consideration oflocal plasticity at the crack tip. Then, the flow rate is determined based on classical thermal-hydraulic expressions for single and two-phase flow.

Crack opening area under the influence of steady-state operating stress (combined tension and bending)is computed from References 17 and 18 as:

SIR-97-029, Rev. 0 5-3 StructuralIntegrity Associates, Inc.

A, = 1 (nR 2) I, (0) 1+3 3 +cos0' '

E o, r 4 ,

- where:

A, = 2 crack opening area (in ) including plastic zone correction, assuming plane stress o, =

steady-state tension stress (psi) 03 =

steady-state axial bending uress (psi)

E. =

elastic modulus (psi) .

R =

nominal pipe radius (in.), and 0 =

the angle describing half the through-wall crack length (radians).

The term I,(0) is computed for varying R/t (pipe radius / thickness) in accordance with the equations ofReference 17.

The plastic zone correction for the effect of yielding near the crack tip is incorporated by the following equation [17]:

0'=0+

2nRa,'

SIR-97-029, Rev. 0 5-4 h StructuralIntegrityAssociates,Inc

i

j. a where:

~

. 0, . =

effective half-length of angle through-wall crack, assuming plane stress Kg = stress intensity factor due to combined tension and bending o, =

material reference stress (flow stress = % (o, + o,) was used in this study, _

where o = ultimate tensile strength

=

o, yield strength (0.2%)

, The flow rate through the' crack is based on classical thermal-hydraulic methodology. The development of the approach is detailed in the following section. The methodology includes.

considerations ofboth liquid and vapor flow ofwater, including the consideration of two phase flow

- within the crack.

The crack is considered to have a total length cf 2a either around the circumference or axially along

~

the pipe wall. The crack has an average opening width w, and the flow path length through the wall 4

is taken as L.

The hydraulic diameter [19] of the flow path is:

4A D

a=P where:

D.u

hydraulio. diameter A

cross-sectional area

~

P = perimeter For a narrow crack oflength 2a,

SIR-97-029, Rev. 0 ' - 5-5 f Structors!IntegrityAssociates,Inc.

i D 4xA A i g = (2) (2a) a

• If w is the average crack opening width, then i A = 2aw and .

f Dg= 2w *

. j The frictional loss in the constant area channel will be assumed to be that between parallel plates with a surface roughness. The parameter ofinterest to characterize the flow resistance per unit of arcais:

)

Kg + K ,,= DG + E + K ,,

Da where:

Kg =

effective total pressure loss coefficient l

K, =

individual discontinuity total pressure loss coefficient  !

=

/ friction factor i

'L =

flow path length, (pipe wall thickness) j

.Dy = hydraulic diameter l

}

K., =

exit loss coefficient = 1.0 i

The pressure loss coefficients for the entrance and flow direction changes must be computed separately from the friction loss parameters. For example, Rererence 20 recommends a discontinuity  :

loss coefficient of 0.5 for a sharp entrance channel. Reference 21 recommends a value of 2.7 to

. 1 I

l SIR-97-029, Rev. 0. 5-6 -

l f StructurallatogrityAssociates,Inc.

properly account for the vena contracta (reduction in cross section) when dealing with near saturated water entering a narrow crack.

The friction factor for turbulent flow (Reynolds number > 4000) is determined from Reference 22.

1 = - 2 logg ( 3.7Dg +

2.52 )

[f pe{f where:

=

f friction factor c = surface roughness Dy = hydraulic diameter Re = Reynolds number.

For laminar flow between parallel plates, Reference 20 recommends, 96 f= _Re which occurs below about R, = 2000. In the transition range between 2000 < R, < 4000, a best enimate friction factor is determined by interpolating between the friction factor at R, = 2000 and that at R, = 4000.

For the turbulent friction factor equation, and for the transition range, an iterative approach must be taken to solve for the friction factor.

Reference 16 recommends a value of 5 pm (0.000197 inches) for the surface roughness of fatigue cracks. For more tonuous paths, and extremely small crack opening displacements, additional losses might be input with increased values for K. However, this effect will be quite small for crack opening widths which will produce detectable leakage in a power piping system.

SIR-97-029, Rev. 0 5-7 f StructuralIntegrityAssociates,Inc.

For the pipe region filled with subcooled water, the flow can be determined by standard incompressible flow methodology. For saturated steam flow, the mass flow rate versus inlet total pressure may be determined directly from the charts _of fUD from Reference 23. Similarly, Reference- 23 provides charts for the blowdown of water and steam-water mixtures. These are incorporated as tables in pe-LEAK In evaluating the flow of subcooled water, which flashes as the static pressure reaches saturation, a two-step approach is used. For the subcooled portion of the flow, the incompressible-flow equation is used: '

Pg ,,u, - P,,, = (K,,,3, + 1.0 + b).1/2 p V2 H

where:

Pn ,,3,

=

pressure inside pipe P,,, =

saturation pressure associated with water temperature in pipe p =

liquid density V = velocity K,,,3, =

inlet plus discontinuity loss coefficient 1.0 =

total to static pressure loss coefficient at the downstream end of the flow From this equation, the length (fLi /D) of channel to bring fluid from its subcooled condition to a flashing saturated mixture may be determined as a function of mass flux. This is illustrated in Figure 5-1.

In length L ,2 a two-phase homogeneous mixture flows and this length may be determined for saturated water from the Reference 22 charts. For small values of fL/D, the saturation flashing .

point may occurjust at the exit of the crack, such that the flow can be approximately determined solely based upon flow ofliquid water. When the inlet pressure is near saturation pressure, the flow SIR-97-029, Rev. 0 5-8 h StructuralIntegrityAssociates,Inc.

i e

may be approximately determined from the Reference 22 charts. In between, a combined flow situation exists. '

t The leakage was calculated for a BWR operating temperature of 550*F and normal operating pressure of1047 psig for the suction side and 1254 psig for the discharge. Since torsional moments I

were included in the critical flaw size calculations, the leakage was calculated considering nonnal operating stresses (without SSE loads) including the torsional moments. The leakage results are  !

presented in Figures 5-2 through 5-4 as a function of crack size for the three locations. Notice that at each location, the leakage is determined separately for the three welds. Table 5-2 shows the predicted leakage at various fractions of the critical flaw size for each case. I

. t e

5.3 Crack Growth Analysis for Circumferential Through-Wall Flaws l

In order to help make decisions relative to the requirements of the leakage detection system at Fermi 2, crack growth analyses were performed to determine the time it takes for an initial through-wall flaw of half the critical flaw size to propagate around the circumference. The analyses were performed using the pc-CRACK computer software [15]. The fracture mechanics model. selected i

from pc-CRACK library for use in the analyses was a circumferential through-wall crack in a cylindrical shell under tension and bending.

i Even though all welds at the affected locations have been treated against IGSCC using NUREG-W13 accepted processes, it was conservatively assumed that crack growth is due to IGSCC. The crack growth law used was obtained from NUREG-0313, Revision 2 [5] and is given by:

I b = 3.59 x 10'8 K"f inches per hour dt where Ki = stress intensity factor (ksi /in).

Because the initial flaw is a through-wall flaw, butt weld residual stresses are expected to be relieved, and, therefore, only the sustained piping stresses are considered in the evaluation.

SIR-97-029, Rev. 0 5-9 h StructuralIntegrityAssociates,Inc.

e Results of the analyses are presented in Figures 5-7 through 5-9 for the three locations. The results are further summarized in Table 5-3. It can be seen from this table that it takes on the order of months for a crack to grow from one fraction of the critical flaw size to another. In fact, it can be seen that it takes at Irast one cycle (over eighteen months) for a through-wall crack at half the critical flaw size to grow to three quaners the critical flaw size.

5.4 LBB Evaluation Results and Discussions As expected, the leakage calculated for the 28-inch suction line is significantly greater than the a

leakages calculated for the 12-discharge risers at various fractions of the critical flaw size. In fact the leakage for the 28-inch suction line is greater than 1.5 gpm from through-wall flaw sizes as low as one-quaner of the critical flaw size. The most limiting location on the discharge risers are the safe end to-pipe welds where the leakage could be as low as 0.2 gpm at one-quaner the critical flaw size.

However, at half the critical flaw size, the leakage on the discharge lines are all greater than 1.5 gpm.

The 'relatively low leakage at the safe end-to-pipe weld on the discharge risers can be attributed to small critical flaw sizes resulting from the high stresses calculated at these locations. Nevenheless, the leakages presented in Table 5-2 can be compared to the leak detection system of the plant to determine at which fraction of the critical tlaw size can be detected. Having determined the detectable leakage, the IGSCC growth evaluation presented in Table 5-3 can be used to determine how long it will take for a through-wall flaw at the detectable leakage size to grow to the critical flaw size. The plant in this way can determine if there is adequate time to take timely action once a leak is detected.

i SIR-97-029, Rev. 0 5-10 f Structura!IntegrityAssociates,Inc.

l Table 5-1 i Summary of Critical Flaw Size  !

Total Stress Tension Stress Cntical flaw size (inch) .

Location Weld (ksi) (ksi) Tension Bending Combined .  :

Imp A N2F N SE 8.913 2.883 11.170 14.755 13.595

- SE-P 14.654 4.694 4.950 7.541 6.711 .

P-E 11.671 4.694 7.333 10.290 9.101 l Loop B N2D N-SE 8.724 2.883 11.381 15.019 13.817 i SE-P 14.026 4.694 4.644 8.060 6.917 P-E 9.472 4.694 9.423 12.682 11.067 j

> Loop B N1B N-SE 7.360 - 3.548 24.083 31.589 27.971 SE-P 8.119 3.994 20.926 28.023 24.532 PE 7.223 3.994 23.543 30.889 26.827 ,

t

.k 6

l f

i

)

! i

- l i

i 4

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l

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Table 5-2 1

-l LBB Evaluation Results for Detectable Leakane  !

(Pressure + Resultant Moment Stresses at Leak) i

.g Leakage at Fraction of Critical Flaw Size j (gpm). j Location Weld Critical Flaw One-quaner ' One-half Three - l Size (in) quarters j Loop AN2F N-SE 13.60 0.58 4.20 . 13.31 i-

- SE-P 6.71 0.25 1.69 5.22 1 P-E 9.10 0.36 2.57 8.12 l Loop B N2D N-SE 13.82 0.51 3.78 12.05 .

SE-P 6.92 0.23 1.57 4.87 i P-E 11.07 0.52 3.74 12.30-t Loop B NIB N-SE '27.97 1.74 11.92 40.19 SE-P 24.53 1.59 10.95 35.22- i P-E- 26.83 2.13 14.85 46.99 l I

i l

I i

l 1

i l

I

)

i

'i SIR-97-029, Rev. 0 5 '

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' i f

Table 5-3 .

I Summary of Through-Wall IGSCC Growth Evaluation Location Weld Critical Flaw Time from % Time from %

Size (in) to % Critical to Critical '

Flaw Size Flaw Size (months) . (months) 1 Loop AN2F M SE 13.60 41 19 dE-P 6.71 2,1 11 .

6 P.E 9.10 32 15 Loop B N2D N-SE 13.82 50 23 -

SE-P 6.92 25 14 .

P-E 11.07 34 15 Loop B N1B N-SE 27.47 56 25  !

i

, - SE-P 24.53 46 21 P-E 26.83 39 18  ;

i J

'I 4

SIR-97-029, Rev. 0 5-13 f StructuralIntegrityAssociates,Inc.

l t  ;

\

4 6

i

?

I P.  ;

1 fL, -

s

- ' fL, .

K + 1.0 +  :

P.-P.

p- D. . .

l D

t fjow of  ; critical -  ;

T~ .

' subcooled water ' two-phase flow flashing  :'

point -

96182:0 i

i Figure 5-1. Flow of Subcooled Water Through a Crack I

l I

t i

}

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+

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

4-LEAKAGE EVALUATION  !

NOZZLE N2F WELDS

~

60 3

(DD  : Critical Flaw Length)  !

50 Safe-End to Pipe l

h 40 -

l l

7 g Pipe to Elbow g 30 --  ;

i

.g, 4  !

20 --  !

P 10 --

N Nozzle to Ssfe-End i

I i

0 t 0 2 4 6 8 10 12 14  :

Flaw Size (in.)  !

Figure 5-2. Leakage Versus Crack Size for Riser N2F on Loop A

(

F f

SIR-97 029, Rev. 0 5-15 I h StructuralIntegrityAssociates,Inc.

h

LEAKAGE EVALUATION NOZZLE N2D WELDS 60 (DD  : Critical Flaw Length) t 50 c Safe-End to Pipe 40

~

Pipe to Elbow Tc.

2 g 30 -

5 .

l 20 --

l I

1 I

10 -

i

. Nozzle to Safe-End

\

l 0 l 0 2 4 6 8 10 12 14 l Flaw Size (in.)

l Figure 5-3. Leakage Versus Crack Size for Riser N2D on Loop B l

9 SIR-97 029, Rev. 0 5-16 StructuralIntegrity Associates, Inc. l

6 LEAKAGE EVALUATION NOZZLE NIB WELDS 140' (0D  : Critical Flaw Length) 120 PE Pipe to Elbow N 4

100 .

s SE-P

- Safe End to Pipe 80 1

5 e

3 60 -

40 20 Nozzle to Safe-End 0

0 5 10 15 20 25 30 Flaw Size (in.)

Figure 5-4. Leakage Versus Crack Size for Riser NIB on Loop B L

SIR-97-029, Rev. 0 5-17 h StructuralIntegrityAssociates,Inc.

)

Crack Growth Evaluation  ;

RPV NOZZLE N2F WELDS  ;

18.00  !

i t

Safe-End to Pipe i

^

~

14.00 -- l l

~

3 12.00 - i 2 10.00 -- Nozzle to Safe-End m.

8.00 --

I 6.00 --

Pipe to Elbow i

4.00 )

l i

2.00

(@: Critical Flaw) l 0.00 0E+00 lE+04 2E+04 3E+04 4E404 SE+04 6E+04 j Time (hours)

Figure 5-5. Crack Growth Analyses Results for Riser N2F on Loop A l

SIR-97-029, Rev. 0 5-18 i h StructuralIntegrityAssociates,Inc.

3

_ . _. _ _ . ~ _ _ _ . _ _ _

1

- 2 Crack Growth Evaluation RPV NOZZLE N2D WELDS '

20.00 ,

Pipe to Elbow I8.00 -

l 16.00 i

14.00 -

l 12.00 --

d

=,

2 Nozzle to Safe-End g 10.00 --

s 8'.00 -- ,

l 6.00 --

L 4.00 --

2.00 -

Safe-End to Pipe 0.00 OE+00 lE+04 2E+04 3E+04 4E+04 SE+04 6E+04 7E+04 Time (hours)

Figure 5-6. Crack Growth Analyses Results for Riser N2D on Loop B t

1 SIR-97-029, Rev. 0 '

5-19 f StructuralIntegrityAssociates,Inc.

.g Crack Growth Evaluation l RPV N0ZZLE NIB WELDS  !

30.00 Pipe to Elbow -

28.00 -

P-E '

26.00 -

N-SE 24.00 _

i  !

22.00 --

7

• ~

u

.5 20.00 -- i j Nozzle to Safe-End l d

18.00 -

16.00 --

14.00 -

Safe-End to Pipe

(@: Critical Flaw]l 10.00 OE+00 1E+04 2E+04 3E+04 4E+04 5E+04 6E+04 7E+04 Time (hours)

Figure 5-7. Crack Growth Analyses Results for Riser NIB on Loop B SIR-97-029, Rev. 0 5-20 f StructuralIntegrityAssociates,Inc.

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l 1

6.0

SUMMARY

AND CONCLUSIONS 1

Leak-before-break (LBB) evaluations are performed for recirculation piping welds at Ferm-2 in accordance with the requirements ofNUREG-1061, Volume 3. In the evaluations, circumferential flaws were considered, since they are more liniting. Cdtical flaw sizes and leakage rates through i fractions of the critical flaw sizes were calculated for three welds at each of three locations adjacent l

to the reactor pressure vessel on discharge risers N2F and N2D as well as suction line N!B (total i of nine welds). Times to grow from various sub-entical flaw sizes to critical flaw sizes were also ,

detennined. .

\

Based on these evaluations, the following conclusions are provided.

Predicted leakage rates in the 28-inch suction line are relatively large for all flaw sizes. At one-quaner the cdtical flaw size, the leakage in this pipe is at least 1.5 1

gpm. At one-half the entical flaw size, the leakage is at least 10 gpm. The leakage l 1

rate is at least 35 gpm at three quarter the critical flaw size in this pipe.

l l

The safe end-to-pipe weld is the most limiting location on the discharge risers. At one quaner the critical flaw size, the calculated leakage is about 0.2 gpm and increases to 1.5 gpm at one half the critical flaw size. The leakage at this location is approximately 5 gpm at three quarter the critical flaw size. The other weld '

locations on the two discharge lines have leakage rates of at least 0.3 gpm at one quaner the critical flaw size,2.5 gpm at one half the critical flaw size and 12 gpm at three quaner the critical flaw size.

IGSCC evaluation performed using the NUREG-0313, Revision 2 bounding crack growth law showed that it takes at least one cycle (over eighteen months) for a through-wall crack at half the cdtical flaw size to grow to three quaners the critical flaw size. This demonstrates that there is adequate time for the plant to take j appropriate action before the critical flaw size is reached once the leakage rates detennined in this report are detected by the plant monitoring systems.

SIR-97 029. Rev,0 6-1 f StructuralIntegrityAssociates,Inc.

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5 k

7.0 - REFERENCES

l. NUREG-1061, Volume 3, " Report of the U. S. Nuclear Regulatory Cortmussion Piping Review Comrnittee Evaluation of Potential for Pipe Breaks", prepared by the Piping _

Review Committee, NRC, April 1985.

2. EPRI Repon NP-4991," Application of the Leak-Before-Break Approach to BWR Piping" D M 7 1966.

3. NUREG 1061, Volume 4, "Repon of the U.S. Nuclear R$*ory Commission on Piping -

Review Committee - Evaluation of other Dynamic Loads and Load Combinations",

, preyW by the Piping Review Committee, NRC, April 1985. ,

4. NUREG-0927, Revision 1, " Evaluation of Water Hammer Occunences in Nuclear Power

. Plants, Technical finding Relevant to USI A-1", March 1984. ,

5. Hazelton, W. S. and Koo, W.H., " Technical Repon on Material Selection and Processing  :

, Gidelines for BWR Coolant Pressure Boundary Piping", NUREG-0313, Rev. 2, USNRC,  !

January 1988.

6. ASME Section XI Task Group for Piping Flaw Evaluation, " Evaluation of Flaws in Austenitic SteelPiping", Journal ofPresare Venei Technology, Volume 108, August 1986, i

7. " Evaluation and Discussion of EPRI's High Energy Pipe Rupture Experimems", EPRI Repon No. NP-5531, by Structural Integnty Associates and S. Levy, Inc., SI Repon No.  ;

, SIR-86-034, September,1987. l l

8. ASME Boiler and Pressure Vessel Code,Section III Appendices and Section II, Part A, l 1989 Edition. i

9. Detroit Edison Calculation and Drawings (SI File Nos. DECO-01Q 201 and DECO-01Q. 1 203).

DC-2674 Vol. V DCD-RPS Loop 'A' DC 2675 Vol. III DCD-RPS Loop 'B' M 5356 5 Rev. 0 ISIDrawing for Loop 'A'

, M-5358-5 Rev. 0 - ISI Drawing for Loop 'B' l M 5359-5 Rev. B - ISI Drawing for Loop 'A'

10. Combustion Engineering Drawings (SI File No. DECO-01Q-202). l 232 908 Rev. 3 (DECO FileNo. RI-90) 232 897 Rev. 3 (DECO File No. RI 132) 1 1

SIR-97-029, Rev. 0 7-1 f StructuralInterityAssociates,Inc.

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

l 1 .t 4

6

'i 11; General Electric Drawings /Speci6 cation (SI File No. DECO-01Q-202 and DECO-01Q-204),

i 117C4384, Rev. 2 & 117C4384AB, Rev. 3 (DECO File # R1-294) 1 l 761E214 Page 1 Rev.11, Page 2 Rev.10 & Page 3 Rev. 7 (DECO File # Rl-174) i 21 A9318 Rev. 3 & 21 A9318AB Rev.1 (DECO File RI-198) 12.- Kumar, V., et al., "An Engineering Approach for Elastic-Plastic Fracture Analysis," EPRI NP-1931, July 1981.  !

' I

13. - Kumar, V., et al., " Advances in Elastic-Plastic Fracture Analysis", EPRI NP-3607, August  !

1984.-

r

14. Kumar, V., et al.. " Elastic Plastic Fracture Analysis of Through-wall and Surface Flaws in .

, Cylinders," EPRI NP-5596, January 1988.

• I

15.

i Stmetural Integrity Associates, Inc., "pc-CRACK Fracture Mechanics Software", Version 2.1, User's Manual, May 1992.

t .

16. "pc-LEAK Calculation of Leakage Rates From Through-Wall Cracks", Version 1.0, StructuralIntegrity Associates, August 1990.

- j

17. Paris, P.,C., and Tada, H. , "The Application of Fracture Proof Design Methods Using j Tearing Instability Theory to Nuclear Piping Postulating Circumferential Through-Wall '

Cracks", NUREG/CR-3464, September 1983.

18. Klecker, R., Brust, F., and Wilkowski, G., "NRC, Leak-Before-Break (LBB.NRC) Analysis Method for Circumferentially Through-Wall Cracked Pipes Under Axial Plus Bending Loads", NUREG/CR-4572, BMI-2134, May 1986.

19. Rohsenow and Choi, " Heat, Mass, and Momentum Transfer", Prentice-Hall, New Jersey, l

1961.

20. . Blevins, R. D., " Applied Fluids Dynamics Handbook", Van Nostrand Reinhold Co., New York,1984

21. " Calculation of Leak Rates Through Cracks in Pipes and Tubes", EPRI Report NP-3395, Electric Power Research Institute, December 1983.

22. Marks Standard Handbook for Mechanical Engineers, Eighth Edition, McGraw Hill, New York,1978.

23. Lahey, R. J., and Moody, F. J. " Thermal Hydraulics of Boiling Water Reactors", American Nuclear Society,1977.

i SIR-97-029, Rev. 0 7-2 StructuralIntegrity Associates, Inc. 1

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

9 r _ 'f, ,

e,1 l

,  ;. t i

.j

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(- - )

f. . . L!

APPENDIX A ~t t- Loads and Stresses I

~

l i

p  !

i t

i I

1 i

1 2

k

.i 5

9 l

. i 8

i 1

1 J

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f DECO-01Q, FERMI NUCLEAR PIANT Unit 2 RECIRCULATION SYSTEM RV NOZZLE N2F(INLET)

Nonle to Safe-End Weld- LOOP A NODE NO.378 OD(IN): '14375 T(IN): 1.1875 ID (IN):. 12 ,

P (KSI): 1.254 A (IN^2): 49.198 Z(IN^3): 150.007 Moment Arm 4.0 FORCES AND MOMENTS TOTAL AXIAL FA FB FC Faxial 'MA MB MC MOMENT STRESS LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN-KIPS) (IN-KIPS) (IN-KIPS) (TN-KIPS) (KSI)

P l - - - - - - - 2.883 THERMAll l 1.1 8,5 3.9 1.10 1993 -144.9 391.2 462 3.104 THERMAL 2 l 1.5 3 1 1.50 48.7 -4.5. 181.7 188 1.285 THERMAL 3 l 1.3 5.8 3.2 130 157.6 115.7 289.7 350 2356 DW 0 13 -0.2 0.00 8.4 -10.4 -49.0 51 0.339 i SSEl l 3.2 3.1 3.8 3.20 185.2 2173 114.1 307 2.115 SSE12 0.6 1.4 0.5 0.60 27.1 31.6 54.6 69 0.470 SSEI3 1.8 3.2 2.6 1.80 132.9 138.1 103.0 218 1.487

]

SSED1 -0.1 0.2 0.1 0.10 5.5 5.6 12.2 15 0.099 SSED 3 0 -0.2 -0.1 0.00 5.1 -3.5 7.0 9 0.062 SSED 6 0 -0.7 0.2 0.00 13.6 -8.0 23.9 29 0.191 Max. SSED 3&6 0 -0.7 -0.2 0.00 13.6 -8 23.9 29 0.191 SSE X l 2.117 0.470 l SSE Y SSE Z l.499 SSE_XY 2.587 l SSE_YZ 1.969 l THERMAL 3.104 i SElSMIC 2.587 a

SIR-97-029, Rev. O A-1 h StructuralIntegrityAssociates,Inc. l

L

- DECO 01Q, FERMI NUCLEAR PLANT Unit 2 RECIRCULATION SYSTEM RV NOZZLE N2F(INLET)

Safe-End TO Pipe Weld. LOOP A NODE NO.378 OD (!N): 12.875 T(IN): 0.71875 ID (IN): 11.4375 P(KSI): 1.254 A (IN^2): 27.449 2 (IN^3): 79.038 Moment Arm 4.0 6

FORCES AND MOMENTS TOTAL AXIAL FA FB FC Faxial MA MB MC MOMENT STRESS LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN-KIPS) (IN-KIPS) (IN-KIPS) (IN-KIPS) (KST)

P - - - - - - - 4.694 THERMAll 1.1 8.5 -3.9 1.10 199.3 -l13.7 323.2 3% 5.055 THERMAL 2 1.5 3 l 1.50 48.7 3.5 157.7, 165 2.143 THERMAL 3 1.3 5.8 -3.2 1.30 157.6 90.1 243.3 304 3.888 DW 0 1.3 -0.2 0.00 8.4 8.8 38.6 40 0.512 SSE11 3.2 3.1 3.8 3.20 185.2 186.9 89.3 278 3.632 SSE12 0.6 1.4 0.5 0.60 27.1 27.6 43.4 58 0.757 SSE13 1.8 3.2 2.6 1.80 132.9 117.3 77.4 193 2.513 SSEDI 0.1 0.2 0.1 -0.10 5.5 4.8 10.6 13 0.166 SSED 3 0 0.2 -0.1 0.00 5.1 2.7 5.4 8 0.100 SSED 6 0 -0.7 -0.2 0.00 13.6 -6.4 18.3 24 0.300 .

Max. SSED 3&6 0 -0.7 -0.2 0.00 13.6 -6.4 18.3 24 0.300 ,

SSE X 3.636 SSE Y 0.757 SSE Z 2.531 SSE_XY 4.393 l SSE_YZ 3.258 THERMAL. 5.055 SEISMIC 4.393 SIR-97-029, Rev. O A-2 Structural Integrity Associates, Inc.

. .. - . .. . . . .. . ~ . -- -- .-.

, j I

DECO-01Q, FERMI NUCLEAR PLANT Unit 2 '

RECIRCULATION SYSTEM - RV NOZZLE N2F (INLET) i Pipe to Elbow Weld. LOOP A NODE NO.376 I

.i OD(!N): 12.875 T(IN): 0.7I875 - .  !

ID (IN): 11.4375 PiKSI): '

1.254 7

. A (INa2): 27.449 79.038 f

Z(INa3):

I i FORCES AND MOMENTS TOTAL AXIAL l FA FB FC Faxial MA MB MC MOMENT STRESS i

. LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN KIPS) s!N KIPS) (IN-KIPS) (IN-KIPS) (1sSI)

P 4.694 THERMALI 1.1 8.5 3.9 1.10 1993 14.1 107.2 227' 2.909 l THERMAL 2 1.5 3 1 1.50 48.7 -27.7, 81 98, 1301 l 7EERMAL3 IJ 5.8 3.2 130 157.6 9.8 -97.6 186 2396 I DW 0 -0.9 0.2 0.00 8.4 3: 10.4 14 0.173 l SSE11 3.2 3.1 3.8 3.20 185.2 91 81.2 222 2.922

]

SSE12 0.6 1.4 0.5 0.60 27.1 17.3 10.7 34 0.451 SSE13 1.8 3.2 2.5 1.80 132.9 58.7 41.1 151 1.976 SSED1 0.1 -0.2 -0.1 -0.10 -5.5 -1.6 143.7 144 1.823 SSED 3 0 0.2 0.1 0.00 5.1 0.6 26.7 27 0344 SSED 6 0 0.7 0.2 0.00 13.6 O.5 76.6 78 0.984 Max. SSED 3&6 0 0.7 0.2 -0.00 13.6 0.6 76.6 78 0.984 I

SSE X 3.444 SSE Y I 0.451 SSE Z 2.208~

SSE_XY 3.895 SSE_YZ 2.858 DEAMA1. 2.909 SEISMIC JJ95 l

1 l

51R-97-029, Rev. O A.3 gg,,,q,,,y 9,g,y,ggy,,,ggg,7,,, jag,

DECO-01Q, FERMI NUCLEAR PLANT Unit 2 RECIRCULATION SYSTEM - RV NOZZLE N2D (INLET) -

Nozzle to Safe-End Weld- LOOP B -

NODE NO.318 OD (!N): 14.375 T(!N): 1.1875 ID (!N): 12 P (KSI): 1.254 A (INa2): 49.198 -

Z (IN^3): 150.007 Moment Arm 4.0 l

FORCES AND MOMENTS TOTAL AXIAL FA FB FC Faxial !dA MB MC MOMENT STRESS ,

LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN-KIPS) (IN KIPS) (IN-KIPS) (IN-KIPS) (KSI)

P - - -

l - - - - 2.883 THERMALI -0.4 6.1 -0.1 -0.40' 7 6.5 -364.6 365 2.440 THERMAL 2 0.4 4.7 2.4 0.40 -113.9 -243.6 292 1.954 4 113.5 ]

THERMAL 3 0 4 0.1 0.00 -3 10.2 -235.7 236 1.573  ;

DW 0.1 1.5 0.1 0.10 p -5.5 0.8 53 3 54 0.359 SSEI1 13 2.7 2.3 130 4 96.1 1393 171.8 241 1.634 ,

SSE12 0.6 43 0.6 0.60 24.9 25.5 178.2 182 1.224 SSE13 2.3 5.4 1.1 230l 5.1 62.8 2513 265 1.812 SSEDI O 03 -0.2 0.00  ; 7.8 -73 12.7 17 0.111 SSED 3 0 -0.1 0.1 0.00 $,).9 2.4 8.0 9 0.059 SSED 6 0.1 03 0.1 0.10 [.j;t 4.1 20.7 22 0.152 Max. SSED 3&6 0.1 03 -0.1 I 0.10l g/., -4.1 20.7. 22 - 0.152 SSE X 1.638 SSE Y l.224 SSE Z l.819 SSE_XY 2.861 SSE_YZ 3.042  ;

DIERMA1. 2.440 i SEISMIC 3.042 I I

l l

I S1R-97-029, Rev. O g.4 gg,yggy,,, y,g,,,;fy p,gggy,g,,, jag _

1 DECO-01Q, FERMI NUCLEAR PLANT Unit 2 RECIRCULATION SYSTEM - RV NOZZLE N2D (INLET)

Safe-End to Pipe Weld- LOOP B NODE NO.318 OD(!N): 12.875 T(IN): 0.7I875 ID (IN): 11.4373 P (KSI): 1.254 A (INa2): 27.449 Z (IN^3): 79.038 Moment Arm 40

\ .

FORCES AND MOMENTS TOTAL AXIAL FA FB FC Faxial MA MB MC MOMENT STRESS LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN KIPS) (IN KIPS) (IN-KIPS) (IN KIPS) (KSI)

P - - - - - - - 4.694 THERMALI -0.4 6.1 -0.1 0.40l 7 5.7 315.8 316 4.012 THERMAL 2 0.4 4.7 -2.4 0.4 01 113.5 94.7 206.0 254 3.222 THERMAL 3 0 4 d.' O.00 3 9.4 -203.7 204- 2.580 DW 0.1 1.5 ' O.1 0.10 5.5 0.0 -413 42 0.531 SSEIi 1.3 2.' 23 130 96.1 120.9 150.2 215 2.773 SSEI2 0.6 4J 0.6 0.60 24.9 20.7 143.8 147 1.887 ,

SSEI3 2.3 5.4 1.1 230 55.1 54.0 208.1 222 2.892 l l

SSED1 0 03 0.2 0.00 7.8 5.7 103 14 0.179 _

SSED 3 0 -0.1 -0.1 0.00 2.9 -1.6 , 7.2 8 0.100 l SSED 6 0.1 -03 0.1 0.10 7.8 -33 183 20 0.259 4 Max. SSED 3&6 0.1 -03 -0.1 0.10 7.8 -33 1851 20 0.259 SSE X 2.779 )

SSE Y 1.887 SSEE 2.903 SSE) Y 4.666 SSEJT. 4.790 THERMAL 4.012 SEISMIC 4.790 O .

1 I

i 1

SIR-97-029, Rev. O A.5 { StructuralIntegrityAssociates,Inc.

i 1

DECO-01Q, FERMI NUCLEAR PLANT Unit 2 RECIRCULATION SYSTEM - RY NOZZLE N2D (INLET) {

Pipe to Elbow Weld- LOOP B -

l NODE NO.316  !

OD(IN): 12.875 f T(IN): 0.71875 ID(IN): 11.4375

} P(KSI): 1.254 -

A (IN^2): 27.449 j Z (IN^3): 79.038

~

l

- FORCES AND MOMENTS TOTAL AX1AL {

FA FB . FC Faxial MA MB MC MOMENT STRESS 1 LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN-KIPS) (IN K!PS) (IN-KIPS) (IN-KIPS) (KSI) .l.

P - - - - - - - ,

4.694 j THERMAll 0.4 -6.11 0.1 -0.40 7 -9.4 161.6 162 2.065 .!

THERMAL 2 0.4 4.7 2.4- 0.40 113.5 33.1 853 146' l.859 I DIERMAL3 0 -4 -0.1 0.00 3 7.6 100.9 101 1.231 l DW 0.1 1.1 -0.1 0.10 5.5 1.2 3.8 10 0.136 l SSE!1

1.3 2.7 2.2 130 96.1 70.7 91 150 1.946  !

SSE12 0.6 43 0.6 0.60 24.9 93 39.9 48 0.628  !

.SSE13 2.3 5.4 1.1 230 55.1 42.2 111.1 131 1.741 l

, SSED1 0 -03 0.2- 0.00 7.8 1.7 4 9 0.113 l 2

SSED 3 0 0.1 0.1 0.00 2.9 03 -4.2 5 0.065 SSED 6 0.1 031 0.1 0.10 7.8 0.5 -10.8 13 0.172 '!

Max. SSED 3&6 0.!: 03; 0.1 0.10 7.8 -0.5 -10.8 13 0.172 j SSE X 1.949 j SSE Y 0.628 l

! SSEZ l.750 SSE_XY 2.578 SSE_YZ 2.378 THEIO4AL 2.065

, SEISh0C 2.578

- e 1

!~

l SIK-91-029, Rev. 0 .A-6 Structural lategrity Associates, Inc.

DECO.01Q, FERMI NUCLEAR PLANT Unit 2  ;

I RECIRCULATION SYSTEM - RV NOZZLE NIB (OUTLET)

Nozzle to Safe-End Weld. LOOP B '

NODE NO.001  !

OD (IN): 29375 T(IN): 1.78125 ID(IN): ' 25.8125 P (KSI): 1.047 i

, A (IN^2): 154.414 .

Z (IN^3): 1004.79 Moment Arm i 4.5 l FORCES AND MOMENTS TOTAL AXIAL FA FB FC Faxial MA MB MC MOMENT STRESS LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN KIPS) (IN KIPS) (IN-KIPS) (IN-KIPS) (KSI)

P - - - - - - - 3.548 THERMALI 5.3 14.7 3.5 5.30 1239.8 342.5 323.8 1326 1354 TtIERMAL2 -6 8.2 0.3 -6.00 584 111.4 561.4 818 0.853 THERMAL 3 3.6 10.4 2.6 -3.60 892.4 -261.7 235.3 959 0.978 DW 0.1 0.4 0.2 -0.10 21.9 24.4 96.2 102 0.102 SSE11 3.1 11- 3.8 3.10 265.1 226.9 667.0 753 0.769 SSEI2 2.5 16.6 2.ll 2.50 103.7 167.5 1056.4 1075 1.086 l

SSE!3 3.2 19 3.9 3.20 182.1 246.9 1210.0 1248 1.263 ]'

SSEDI 0.2- 0 0 -0.20 2.1 -1.5 29.2 29 0.030 SSED 3 0 -0.1 -0.2; 0.00 33.8 12.9 7.1 37 0.037 i SSED 6 0.1 -0.8 0.8 0.10 114.5 46.6 -48.2 133 0.133 Max. SSED 3&6 0.1 0.8 -0.8 0.10 114.5 46.625 -48.225l 133 0.133 SSE X 0.770 SSE Y 1.086 SSE Z l.270 SSE_XY 1.856 SSE_YZ 2.356 THERMAL 1.354 SE15MIC 2J56 l

l 1

SIR-97-029, Rev. 0 A-7 StructuralIntegrity Associates, Inc.

s DECO-01Q, FERMI NUCLEAR Pl. ANT Unit 2 RECIRCULATION SYSTEM . RV NOZZLE NIB (OUTLET)

Safe-End to Pipe Weld- LOOP B NODE NO. 001 <

OD (IN): 28.4375 T (IN): 1.5625 ID (IN): 25 3 125 P (KSI): 1.047 A (IN^2): 131.922 Z(IN^3): 840.484 Moment Arm 4.5 g

FORCES AND MOMENTS TOTAL AXIAL FA FB FC Faxial MA MB MC MOMENT STRESS LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN-KIPS) (IN KIPS) (IN-KIPS) (IN-KIPS) (KSI)

P - - - - - - - 3.994 THERMALI 5.3 14.7 3.5 530 1239.8 310.7 190.6 1292 1.578 THERMAL 2 -6 8.2 0.3 ' -6.00 L 584 108.6 -487.0 768 0.959 THERMAL 3 3.6 10.4 2.6 -3.60 892.4 -238.1 141.1 934 1.139  ;

DW 0.1 0.4 -0.2 0.10 21.9 22.6 92.6

• 98 0.117 l SSEI1 3.1 11 3.8 3.10 265.1 192.5 567.4 655 0.803 906.0 924 1.118 SSE12 2.5 16.6 2.1 2.50 103.7 148.5 f SSEI3 3.2 19 3.9 3.20 182.1 211.5 1037.8 1075 1303 SSED1 0.2 0 0 -0.20 2.1 -1.5 29.2 29 0.036 SSED 3 0 -0.1 -0.2 0.00 33.8 11.1 -6.1 36 0.043 l SSED 6 0.1 0.8 -0.8 0.10 114.5 39.4 -41.0 128 0.153 l Max. SSED 3&6 0.1 -0.8 -0.8 0.10, 1 I4.5 39375 -40.975 128 0.153 )

SSE X 0.804 SSE Y 1.118 SSE Z l.312 l 1.922 SSE_XY 2.430 SSE_YZ THERMAL t.578 i SElsMIC 2.430 SIR-97 029, Rev. O A.g h StructuralIntegrityAssociates,Inc

DECO-01Q, FERMI NUCLEAR PLANT Unit 2  :

RECIRCULATION SYSTEM - RV NOZZLE N18 (OUTET) l Pipe to Elbow Weld LOOP B . l

' NODE NO. 003 i

OD(IN): 28.4375.  !

T (!N): 1.5625

{

25.3125  !

ID (IN):

P (KSI): 1.047 131.922 l A (IN^21:

Z (IN^3): 840.484  !

\.

FORCES AND MOMENTS TOTAL AXIAL j i FA FB FC - Faxial MA MB MC MOMENT STRESS LOAD (KIPS) (KIPS) (KIPS) (KIPS) (IN-KIPS) (IN-KIPS) (IN-KIPS) (IN-KIPS) (KSI)

P - - - - - -

l - 3.994  ;

THERMAL 1 -5.3 14.7, 3.5 5.30 -1239.8 220 704.4l 1443 1.757 THERMAL 2 -6 8.2 -0.3 -6.00 584 100.8 773.7 975 1.205 ,

THERMAL 3 3.6 -10.4 2.6' 3.60 -892.4 170.2' 505.8 1040. 1.264

DW 0.1 -1.1 0.2 -0.10 -21.9 -17 -83.6 88 0.106 f SSEII 2.9 10.9 3.8 2.90 265.I' 228.9 293.9 457 0.566 SSE12 2.5 16.5 2 2.50 103.7 113 480.7 505 0.619 SSEI3 3.2 19 3.7 3.20 182.1 147.9 548.2 596 0.734

l. SSED1 -0.2 0 0 -0.20 2.1 0.5 27.7 28 0.035 SSED 3 0 0.1 0.2 'O.00 33.8 -5.2 2.2 34 0.041 SSED 6 0.1 0.8 0.8 0.10 114.5 -

19.5 18.8 118 0.141 l l Man. SSED 3&6 0. I ' O.8 0.8' O.10 114.5 19.5 18.8 118 0.141 4

SSE X 0.567 SSE Y 0.619 SSE Z 0.747 SSE XY 1.166 SSE._YZ- 1.366  ;

i- THERMAL 1.757 )

SEISMIC 1.366 l 1

. 1 i

+

SIR-97-029, Rev. 0 Structors! Integrity Associates, Inc.

. A-9

t

~

DECO-01Q, FERMI NUCLEAR PLANT Unit 2  !

~

< RECIRCULATION SYSTEM - RV N0ZZLES N2F, N2D and NIB SUSTAINED STRESSES t

STRESS (ksi)

~

Location Weld PRESSURE DEAD WT. 'IEERMAL S.S.E. P+DW+TH P+DW+TH+SSE Loop A N2P N-SE 2.883 0339 3.104 2.587 6.326 8.913 SE-P 4.694 0.512 5.055 4393 10.261 14.654 i

' P.E 4.694 0.173 2.909 3.895 7.776 11.671 l Loop B N2D N SE 2.883 0359 2.440 3.042 5.682 8.724 SE-P 4.694 0.531 4.012 4.790 9.236 14.026 i 9.472 I P-E 4.694 0.136 2.065 2.578 6.894 Loop B N1B N SE 3.548 0.102 1.354 2356 5.004 7.360 i SE-P 3.994 0.117 1.578 2.430 5.689 8.119  !

PE 3.994 0.106 1.757 1366 5.856 7.223 f

I G

9 SIR-97-029, Rey, o A.to f StructuralIntegrityAssociates,Inc.

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