ML20094B031

Structural Margin Evlauation for Hatch 2 Internal Core Spray Lines

ML20094B031
Person / Time
Site:	Hatch
Issue date:	10/31/1995
From:	Daniel R, Mehta H, Ranganath S GEORGIA POWER CO.
To:
Shared Package
ML20094B030	List: HL-5048, Provides Results of Internal Core Spray Piping Insps. GENE-523-A110-1095, Structural Margin Evaluation for Hatch 2 Internal Core Spray Lines, Encl ML20094B031
References
GENE-523-A110-1, GENE-523-A110-1095, NUDOCS 9510310264
Download: ML20094B031 (10)

Text

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, & GENuclearEnergy

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TECHNICAL SERVICES BUSINESS GENE-523-A110-1095 GE Nuclear Energy - DRF # 137-0010-8 175 Curtner Avenue, Sen Jose, CA 95125 October 1995 l i

- A STRUCTURAL MARGIN EVALUATION FOR HATCH 2

INTERNAL CORE SPRAY LINES 4

. October 1995

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Prepared by: 2_ .44

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Rache11a Daniel, Marhaaical Engineer

,. StructuralMechaniet Projects i

! Verified by: b_ v ' ^ -

H.S. Mehta, Principal Ergineer Structural Mechanics Pi qsts Approved by:

Dr. S. Ranganath, Engineering Fellow Nuclear Services Department GE Nuclear Energy San Jose, CA 9510310264 951031 PDR ADOCK 05000366

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GENuclear Energy GEbE-323-A110-1093 DRV137-0010 7

~. 1 TABLE OF CONTENTS Page il TABLE OF CONTENTS

1. INTRODUCTION AND

SUMMARY

' 1 1,1 STRUCTURAL ANALYSIS 1

1.2 CONCLUSION

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2. CORE SPRAY PIPE STRUCTURAL INTEGRITY 2 I

2.1 STRUCTURAL INTEGRITY 2 i

2.1.1 Summary 2 2.1.2 Allowable Flaw Size Determination 2 1 2.2

SUMMARY

AND CONCLUSIONS 5

3. REFERENCES 6 l.

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GENuclear Energy GENE-323.Al10-1093 DRF 137-0010-7 1.' INTRODUCTION AND

SUMMARY

This report presents the results of a structural margin evaluation of the Hatch 2 core spray line at the location of a reported flaw indication identified by IVVI. The indication is 1/2" long, is located on the elbow side of the fillet weld between the collar and the pipe in the region of the  !

downcomer coupling and is being treated as a crack. The fact that the indication is below the toe 1

of the fillet weld and in the heat affected zone (HAZ) of the weld suggests that the possible flaw, I ifit is a crack, may not have started from the crevice on the inside It is more likely that a crack would initiate from the weld HAZ on the outside of the pipe.

In order to address potential safety concerns related to this possible cracking at Hatch 2, GE Nuclear Energy has performed a preliminary evaluation to determine the safety significance of through-wall cracks having the same orientation and location as the observed indication.

The technical basis to support the continued structural integrity of the Hatch 2 core spray lines for a!! normal and injection conditions with a postulated crack is provided.

1.1 STRUCTURAL ANALYSIS The structural analysis, described in Section 2, examines whether the integrity of the core spray piping will be maintained in the presence of the observed indication.

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1.2 CONCLUSION

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A structural evaluation of the Hatch 2 core spray line has been performed to determine the impact on plant operation with a postulated circumferential crack near the downcomer coupling. Based on this analysis, it is concluded that Hatch 2 can safely operate with up to a 263 circumferential

crack and that no operational changes or restrictions are required at this time.

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GENuclear Energy GENE 323-Al10-1093 l DRF137-0010-7 1

i 2., CORE SPRAY PIPE STRUCTURAL INTEGRITY  !

l The structural integrity aspects of the core spray piping were reviewed to assess the impact a crack could have on the structural integrity of the piping. Structural analyses were performed to 4 determine the primary stresses in the piping. Although there is not enough information to g definitively classify the flaw in the core spray line, it is assumed to be a crack due to an

, Intergranular Stress Corrosion Cracking (IGSCC) mechanism.

I 2.1 STRUCTURAL INTEGRITY a

2.1.1 Summary ,

-1 All primary stresses expected during normal reactor operation were found to be small compared to the yield strength of the material. Therefore, it is reasonable to conclude that the normal operating loads by themselves do not result in stresses which are sufficient to cause IGSCC initiation. The addition of secondary stresses, thermal expansion, and weld residual stresses coupled with local cold work, could result in stresses exceeding the initiation threshold.

1 Once initiated, the stresses relax as the crack grows, and the compliance (or flexibility) of the pipe J

increases. Studies show that when the crack reaches 180 of the circumference, the compliance is reduced sufficiently to relieve almost all of the displacement controlled stresses. Therefore, crack growth is expected to be negligible or at virtual arrest once reaching 180 . In order to determine the integrity of the core spray line with a crack, a crack arrest evaluation was performed. Stresses due to pipe restraint are included in this evaluation.

! Based upon a review of these stresses, it is concluded that the structural integrity of the core spray piping including the crack will be maintained during core spray injection. The stresses 4

considered include those due to downcomer flow impingement loads, seismic loading, pressure, and weight.

2.1.2 Allowable Flaw Size Determination An evaluation was performed to determine the maximum allowable circumferential through-wall

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flaw size in the core spray pipe. This analysis will therefore provide an assessment of the safety margin in the pipe due to primary loads such as deadweight, pressure, flow impingement and seismic.

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i GENuclear Energy GENE-323-Al10-1093 DRF137-0010-7 The acceptable through-wall flaw size of the core spray line was determined utilizing the limit load formulation of References 1 and 2. To apply this methodology, the maximum primary membrane stresses in the longitudinal direction and primary bending stresses were determined for the pipe. A finite element model of the core spray pipe was developed to obtain the stresses due to deadweight, seismic and reactor vessel downcomer flow impingement on the pipe at the location of interest, and the resulting stresses were then combined with the stresses due to pressure and core spray flow loads in order to obtain the total stresses acting on the pipe.

Because the primary loads are small, a conservative analysis using the highest loads (SSE seismic and upset pressure) with upset allowables was performed. Stresses due to water hammer loads j were considered insignificant and neglected in this analysis since the cora pray inlet valve ramps  !

open over a period of twenty seconds upon system actuation. Additionally, the piping is full of water during actuation due to the presence of the vent hole on the top of the T-box. Previous analyses have shown that the water hammer loads in the core spray line were calculated to be less than 20 pounds of axial load on the pipe. Stresses due to thermal mismatch were also ruled insignificant based on previous GE core spray analyses.

By applying these resulting primary stresses, with a safety factor of 2.8, it was shown that the core spray pipe can tolerate a crack up to 263 through-wall at any location without incipient failure.

Analysis and Results A finite element model was constructed using the geometry of a core spray line very similar to

that of Hatch 2. Figure I shows the geometry of the model. The stress analysis was conducted

using ANSYS computer code (Reference 3). Loads due to the weight of the pipe (including 1 captured water in the pipe) were applied to the model along with vertical and horizontal s

ismic

loads and reactor vessel downcomer flow impingement loads. The largest resulting stresses were used from the finite element model results. These stresses were then combined with the stresses I

due to pressure and core spray flow loads. The resulting total stresses are shown in Table 1 on

! the next page as noted above, the loads due to thermal mismatch of the core spray line and reactor vessel need not be included as they are secondary in nature.

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I GENuclear Energy GENE-323-A110-1093 DRF137-0010-7

, Stress Type Stress Magnitude (psi)

Membrane 1000 Bending 455 Table 1: Calculated Primary Stresses The stresses shown in Table I were then utilized to determine the acceptable through-wall flaw  ;

l size based on the methods of References 1 and 2. The acceptable flaw size was determined by requiring a suitable design margin on the critical flaw size. The critical flaw size was determined by using limit load-concepts. In the limit load theory, it is assumed that a pipe with a l

circumferential crack is at the point ofi , ient failure when the net section at the crack develops a plastic hinge. Plastic flow is assumed to occur at a critical stress level, or,, called the flow stress l of the material. For ASME Code analysis (Section XI, Appendix C), or may be taken as equivalent to 3S. (S.=14.4 ksi for 304L at 550 F, Ref. 4). A santy factor of 2.8 is used on the flow stress, for an adjusted value of 15.43 ksi. This results in considerable simplificaden of the analysis.

Consider a circumferential crack oflength,1 = 2Ra, and cor , tant depth, d, located as shown in Figure 2. In order to determine the point at which collaps cccurs, it is necessary to apply the equations of equilibrium assuming that the cracked section behaves like a hinge. For this condition, the assumed stress state at the cracked section is as shown in Figure 2 where the maximum stress is the flow stress of the material, op. Equilibrium oflongitudinal forces and moments about the axis gives the following equations:

(For neutral axis located such that a + p < x) p = [(x - ad/t) - (P./or)n]/2 Pb= (2af /x)(2 sin p - d/t sin a) where, t = pipe thickness, inches a = crack half-angle as shown in Figure 2 p = angle that defines the location of the neutral axis P. = Membrane axial stress 4

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GENulear Euray GENE-323,Al10-1093 DRP137-0010-7

, P6 = Bending stress Using the stresses of Table 1 and a d/t ratio of 1.0 (through-wall flaw), the allowable through-wall crack for which failure by collapse might occur is 263*.

Since the applied stresses are displacement controlled, the compliance will increase enough by the time a 180 through-wall crack develops that crack arrest is expected.

2.2

SUMMARY

A.ND CONCLUSIONS The potential sources of stress in the Hatch 2 core spray piping resulting from various plant conditions were reviewd The postulated cracking mechanism is assumed to be IGSCC.

Due to the predominant secondary stresses, the crack in the core spray line is expected to arrest prior to reaching 180 An assessment was made to determine the critical flaw size of the core spray pipe by treating stresses associated with the design loading as primary stresses and performing a limit load evaluation. The results of this evaluation confirm that in the core spray i lines a through wall crack of up to 263 (approx. I1.5 inches) around the circumference would I not cause pipe failure. This length is the maximum allowable crack lengths at the end of the next fuel cycle. If the observed length of the indication is less than this during future inspections, continued operation is justified. Using the crack growth rate of SE-5 in/hr, the predicted time to reach the maximum allowable flaw size of 11.5 inches is 13 years.

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l GENuclear Energy GENE-333-Al10-1093 DRF137-0010-7

3. REFERENCES

1. Ranganath, S. and Mehta, H. S., " Engineering Methods for the Assessment of Ductile Fracture Margin in Nuclear Power Plant Piping," Elastic-Plastic Fracture: Second Symposium, e Volume II - Fracture Resistance Curves and Engineedng Applications, ASTM STP 803, C.F.

Shih and J. P. Gudas, Eds., American Society for Testing and Materials,1933, pp. II-309 - II-330.

2. ASME Boiler and Ptessure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, American Society of Mechanical Engineers,1989 Edition, l

Paragraph IWB 3640. J l

3. DeSalvo, G. J., Ph.D. and Swanson, J. A., Ph.D., ANSYS Engineering Analysis System User's Manual. Revision 4.4. Swanson Analysis Systems, Inc., Houston, PA, May 1,1989.

4. Sun Shipbuilding drawing 50753-1 sheets I and 5 (RPF 3743-1(1)-7 and 3743-1(5)-7).

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Hatch Core Spray Line Analysis Figure 1 Finite Element Model of Core Spray System 7

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