ML20155D996

Rev 0 to GE-NE-B13-01980-24, Fracture Mechanics Evaluation on Observed Indication at N3A Steam Outlet Nozzle to Shell Weld at Cooper Nuclear Station

ML20155D996
Person / Time
Site:	Cooper
Issue date:	10/31/1998
From:	Lai W, Mehta H, Stark R GENERAL ELECTRIC CO.
To:
Shared Package
ML20155D986	List: ML20155D982 ML20155D996
References
GE-NE-B13-01980, GE-NE-B13-01980-24, GE-NE-B13-1980, GE-NE-B13-1980-24, NUDOCS 9811040117
Download: ML20155D996 (18)

Text

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Attachment to NLS980182 16 Pages Total

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Fracture Mechanics Evaluation Main Steam Nozzle to Shell Weld N3A l

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l 9811040117 DR 991030 ADOCK 05000298 PDR

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7 GENuclearEnergy i

TECHNICAL SERVICES GE-NE-B13-01980-24, Rev. O GE Nuclear Energy \

DRF # B13-01980  !

175 Curtner Avenue, San Jose, CA 95125 ClassII i October 1998 l

l A FRACTURE MECHANICS EVALUATION ON OBSERVED INDICATION AT N3A STEAM OUTLET NOZZLE i TO SHELL WELD AT COOPER NUCLEAR STATION l t

October 1998 l

I Prepared for Nebraska Public Power District i

Prepared by GE Nuclear Energy l

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l GE Nucley Energy GE-NE-Bl3-01980-24. Rev. 0 l

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A FRACTURE MECHANICS EVALUATION ON OBSERVED INDICATION AT N3A STEAM OUTLET NOZZLE TO SHELL WELD AT COOPER NUCLEAR STATION l

October 1998 I

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Prepared by: M H.S. hiehta, Principal Engineer Structurall' techanjes & Materials Prepared by: v W.C. Lai, Engineer Structural Mechanics & Materials Verified by: b li R.R. Stark,tngineer Structural Mechanics & Materials W

Approved by: / b bM) .

T. A. Ct.ine, Manager Structural Mechanics & Materials

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GENxcle:r Exerxy GE NE-Bl3-01980-24 Rev. 0 '

IMPORTANTNOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this '

document are contained in the Agreement between Nebu t Public Power District (DISTRICT) and GE, effective September 1,1986, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than DISTRICT, or for any purpose other than that for which it is intended is not authorized: and with respect to any unauthorized

- use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document,' or that its use may not infringe privately owned rights.

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I GENuctrr Eccrgy GE-NE-Bl3-01980-24. Rev. 0

, Table of Contents

. Subiect Pane No.

1. PURPOSElOBJECTIVE 4

2. DESIGN INPUTS 4 l

3. ASSUMPTIONS 4

4. UNITS IN EQUATIONS

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5. CALCULATIONIANALYSIS METHODC _OGY 5 5.1. Analysis Methodology 5

5.2. Operating Conditions Considered 5

5.3. Stress Calculation Due to Internal Pressure 6

5.4. Stress Distribution Due to Thermal Gradient 6 5.5. Fracture Toughness 8

5.6. Fatigue Crack Growth Evaluation 8

5.7. Fracture Mechanics Evaluation Results 8

5.8. Local Membrane Stress Evaluation 9

6. CONCLUSIONS 9

7. REFERENCES 10 3

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- L Purpose / Objective The manual ultrasonic examination of the category B-D, N3 A nozzle to shell weld during Cooper Nuclear Station (CNS) fall 1998 outage (RF18) found a subsurface indication that appears unacceptable when evalered per the acceptance standards of ASME Section XI, IWB-3512-1 [ Reference 1). The o.. ils of the IWB-3500 evaluation are contained i examination report (Reference 2]. The N3 A nozzle is one of the fcur steam outlet nozzles in the reactor pressure vessel. The indication was characterized as a planar indication with a through-wall dimension of 0.88 inch, length of 12.25 inches with a surface separation of 2.56 l inches.

The Section XI procedures permit acceptance by analysis (Paragraph IWB-3600) of an 4

indication that is found to be unacceptable per the acceptance standards. This report 1

documents the results of a fracture mechanics evaluation of the observed indication ba the procedures ofIWB-3600.

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2. Design Inputs e

4 The design inputs and the associated references are indicated in the following:

(1) The indication geometry was obtained from the ultrasonic (UT) inspection report on I

the N3 A steam outlet noz7'- ~coared by GE [ Reference 2].  !

(2) The reference nil ductility temperature (RTm) for the weld, the nozzle forging and the shell course were obtained from References 3 and 4.

j (3) The pressure and temperature conditions for various operating conditions were obtained from References 5 and 6.

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3. Assumptions 1

l It wa9 assumed that the RTm of the weld between the nozzle and the vessel shell is less than l 18'F. t s

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4. Units in Equations English units (lbs, inches, psi, etc.) were used in the equations and the evaluations.

5. Calculation / Analysis Methodology 5.1. AnalysisMethodology The fracture mechanics methods used in the analysis are consistent with the procedures

- outlined in Section XI of the ASME Code [ Reference 1].l The primary stress requirements are based on the Code of Construction of the RPV [ Reference 7].

5.2. Operating Conditions Considered The operating conditions considered were: Hydrotest, Normrd (Level A), Upset (Level B),

Emergency (Level C), and Faulted (Level D). The thermal cycle drawing for the CNS

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[ Reference 6] does not explicitly identify the operating condition associated with each of the events covered. Therefore, the thermal cycle diagram of a similar but later built BWR plant was used as a guide. The stresses at t!v: locatin of the indication are primarily from:, internal pressure and thermal gradient. The steam outlet nozzle region experiences the pressure and temperature conaitions associated with ReFion A in the thermal cycles drawing.

The internal pressure for the hydrotest condition is 1100 psi [ Reference 5] and the thermal gradient _during this event is insigr.ificant. For the normal condition, the internal pressure is 1000 psi and the thermal gradient is 100'F, associated 3vith the heatup/ cool down event.

During the upset condit ion, the controlling event is ' Turbine Generator Trip, Feedwater On, Isolation Valves Stay open'. The internal pressure during this event is 1125 psi and the te aperature gradient is (565 538) or 27*F.

A thermal transient as shown in Figure I was conservatively used for the evaluation of emergency condition. This transient bounds the ' single relief or safety valve blow down' event in the CNS thermal cycle diagram. The internal pressure at the highest thermal stress level was conservatively taken as 400 psi.

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GENuclear EKergy GE.NE.Bl.101980-24. Rev. 0 Figure 2 shows the bounding thermal transient considered for the faulted condition. The event corresponds to ' pipe rupture and blow down'. The intemal pressure at the highest thermal stress level was taken as 22 psi.

5.3. Stress Calculation Due to InternalPressure The stress distributien due to internal pressure in the vicinity of the nozzle to shell weld is expected to be complex. Therefore, the stress distribution calculated by Gilman and Rashid

[ Reference 8] for a three dimensional analysis of a BWR feedwater nozzle under intemal pressure wa* reviewed. Figure 3 shows the distribution ofmaximum stress. The stress at the nozzle to weld location appears to vary from 25000 psi at the inside diameter to 12000 psi at the outer surface. The nominal circumferential stress in the vessel modeled in Refer was 17530 psi. The nominal size of the steam outlet nozzle is larger than that of the feedwater nozzle. However, the stress concentration effect introduced by the presence of a nozzle opening in a shell loaded under internal pressure is not a function of opening size.

1 This is evident from the fact that the peak stress to nominal stress ratio for internal pressure loading is the same (equal to 3.1) irrespective of the nozzle opening size. Therefore, for the internal pressure loading, the relationship between the stress distribution at the nozzle to shell weld junction and the nominal stress in the vessel shell is expected to be the same at the steam outlet nozzle as predicted by the three dimensional results shown in Figure 3.

Thus, the stresses at the nozzle to shell weld section due to internal pressure loading (i.e.,

membrane and bending stress magnitudes) were obtained by scaling up or down the Figure 3 stress magnitudes by the nominal pressure stress in the shell.

5.4. Stress Distribution Due to Thermal Gradient The bounding fluid temperature change rate during beat-up/ cool down is los tt /hr. The linear temperature gradient through the vessel wall was calculated using the f . lowing equation, based on one dimensional heat conduction equation:

AT = GC2/2p

-(1) where, G = Heat-up/ cool down rate (*F/hr)

C = Vessel thickness including clad (ft) 6 L

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!' 2 including clad, and p = 0.354 ft /hr, the AT value was calculated as 39.0*F. The bending

stress due to this temperature gradient was calculated using the following equation:

i ao = [EaAT/{2(1-v)}]

(2) where, E = Youngs modulus e = coefficient of thermal expansion v = Poisson's ratio = 0.3 During the bounding upset condition event, " turbine generator trip feedwater on, isolation valves stay open", the temperature change is (565-538) or 27'F. The AT value was also conservatively assumed to be the same.

For the emergency and faulted conditions, calculated finite element stress distributions from l previous analyses [ Reference 9) of the bounding transients were reviewed and characterized 3

in terms of membrane and bending components, as required by ASME Section XI, Appendix A procedures for the evaluation of subsurface indications (see Figure 4). The values I calculated are the following:

Emergency Condition o, = -3.0 ksi o, = ll.0 ksi i

, Pressure = 400 psi 4 Temp. = 259"F l Faulted Condition l

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o, = -8.0 ksi j o, = 26.0 ksi

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Pressure = 22 psi Temp. = 259 F j i

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,, GENucleir Enerty GE-NE-Bl3-01980 24. Rev. 0 5.5. Fracture Toughness The highest RTsor of the steam outlet nozzle forgings was 18'F. The highest RTyor ofthe upper shell plate was 14*F. The RTuorof the nozzle to shell weld itself could not be located.

However, the other welds for wH h information is available have RTuor values of-50'F.

Based on the previous data on file with GE on the measured RTsor values of various RPV

j. welds, it is reasonable to assume that the weld RTuor will be less than 10*F. Therefore, the limiting RTsor was determined to be 18'F based on the nozzle forging material. This value was used in calculating the reference fracture toughness or Km value for each of the operating conditions.

5.6. Fatigue Crack Growth Evaluation The fatigue crack growth was calculated using the following crack growth rate relationship for subsurface flaws given in Reference 1 (Figure A-4300-1):

da/dN = 2.67x10 "(AK)n2' (3) where, da/dN = Crack growth rate in in/ cycle AK = Stress intensity factor range in ksiVin.

The major contributors to fatigue crack growth are the start-up/ shut down cycles. The number of such cycles specified for the design life are 120. Therefors this value was conservatively used in updating the through-wall depth of the subject indication.

5.7. Fracture Mechanics Evaluation Results The applied stress magnitudes described in the preceding Subsections were used to calculate the applied stress intensity factor value at the subject indication for various operating conditions. The K, values were calculated using the following equation from Reference 1:

K, = n,M VnV(a/Q) + c,M,Vn4(a/Q) where, c., e, = membrane and bending stresses, psi.

a = minor half-diameter, in., of subsurface flaw 8

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l Q = flaw shape parameter M, = correction factor for membrane stress M, = correction factor for bending stress l

Table I shows the calculated values of the applied iK values for various operating conditions.

The last column shows the allowablei K values which were obtained by dividing the Ka value by the safety factor. A safety factor of 410 was used for hydrotest, normal and upset j

conditions and a safety factor of 42 was used for emergency and faulted conditions.

A comparison of the applied and allowable K ivalues shows that the applied K values i are less than the allowable Ki values for all operating conditions. Based on this, it is concluded that the subject indication meets the criteria ofIWB-3612 of Reference I and on this basis continued operation is acceptable.

S.8. LocalMembraneStress Evaluation The procedures in IWB-3610 require a primary stress evaluation in addition to the fracture '

mechanics requirements ofIWB-3612. The maximum primary membrane stress cannot exceed 1.5 S,. Assuming that the clad does not bear any part of the load, the maximum through-wall flaw depth must therefore be limited to 1/3 the low alloy steel (LAS) wall thickness. For the UT reported LAS wall thickness of 6 inches, the maximum allowable through-wall dimension for a subsurface flaw is 2.0 inches. Since the subject indication dimension of 0.883 inch, including projected fatigue crack growth, is less than this value, the primary membrane stress requirements are satisfied.

6. Conclusions The manual ultrasonic examination of the category B-D, N3A nozzle to shell weld during CNS fall 1998 outage (RFl8) found a subsurface indication that appears unacceptable when evaluated per the acceptance standards of ASME Section XI, IWB-3512-1. Fracture mechanics and primary stress evaluations per the requirements ofIWB-3610 were conducted the results of which are documented in this report. The evaluation results show that subject indication meets criteria ofIWB-3612,Section XI, and the prima < stress requirements of Section III of the ASME Code. Therefore, continued operation "as is"is acceptable.

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7. References

[1] ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition, No Addenda.

[2] GE Nuclear Energy UT Examination Report No. R-196 for Cooper Nuclear Station RE18 (October 1998).

[3] GE Design Record File No. B13-01389,"RPV Surveillance Test".

[4] " Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis," GE Report No. GE-NE-523-159-1292, February 1993.

[5] (a) Cooper Nuclear Station USAR Volume II, Section IV-2.6.1, " Design Loadings".

(b) Paragraph IWB 5000 of Reference 1.

[6]- GE Drawing No. 729E762, " Reactor Thermal Cycles".

[7] " Analytical Report for Consumers Reactor Vessel Cooper Station," Combustion Engineering Report No. CENC 1150, April 1971.

[8] J.D. Gilman and Y.R. Rashid, "Three Dimensional Analysis of Reactor Pressure Vessel Nozzles," Transactions of the 1" Internationt! %nference for Structural )

Mechanics in reactor Technology (SMiRT), Vol. 4, Part G, September 1971. l i

[9]

r GE Design Record File No. A00-05611," Equivalent Margin Analysis".

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Table 1 Comparison of Allowable and Applied K values Operating Condition K, Applied K, Allowable j (KsiVin.) (KsiVin.)

Hydro Test 32.9 59.7 Normal Operation (Level A) 29.5 63.2 Upset Condition (Level B) '34.0 63.2 l

Emergency Condition (Level C) 12.3 141.4 l Faulted Condition (Level D) 3.7 141.4 1

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N' ATTACHMENT 3 LIST OF NRC COMMITMENTS l l

-Correspondence No:NLS990192 e

The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this docunent or any associated i regulatory commitments. l COMMITTED DATE COMMITMENT OR OUTAGE ,

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l l PROCEDURE NUMBER 0.42 l REVISION NUMBER 6 l PAGE 9 OF 13 l