ML20101H416

Suppl 2 to Pressurizer Surge Line Thermal Stratification for B&W 177-FA Nuclear Plants,Summary Rept Fatigue Stress Analysis of Surge Line Elbows

ML20101H416
Person / Time
Site:	Crystal River
Issue date:	05/31/1992
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20101H414	List: 3F0692-19, Forwards Suppl 2 to BAW-2127, Pressurizer Surge Line Thermal Stratification for B&W 177-FA Nuclear Plants,Summary Rept Fatigue Stress Analysis of Surge Line Elbows, to Address Util Actions to Comply W/Bulletin ML20101H416
References
BAW-2127-S02, BAW-2127-S2, IEB-88-011, IEB-88-11, NUDOCS 9206290339
Download: ML20101H416 (38)

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3 l I BAW-212 7 SUPPLEMENT 2 I

PRESSURIZER SURGE LINE l THERMAL STRATIFICATION FOR THE B&W 177-FA NUCLEAR PLANTS I

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SUMMARY

REPORT l

l FATIGUE STRESS ANALYSIS OF THE SURGE LINE ELBOWS lI

I B&lAl NUCLEAR I SERT / ICE COlVIPANY I 9206290339 920625 PDR G

ADOCK 05000202 PDR

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SUPPLEMENT 2 I

PRESSURIZER SURGE LINE l THERMAL STRATIFICATION FOR THE B&W 177-FA NUCLEAR PLANTS I

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SUMMARY

REPORT l FATIGUE STRESS ANALYSIS OF THE SURGE LINE ELBOWS I

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"- "B&lN NUCLEAR I C- SER1/ ICE COMPANY I 9206290339 PDR O

92062S ADOCK 05000302 PDR

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I BAW 2127 SUPPLEMENT 2 MAY 1992 I i I .

I PRESSURIZER SURGE LINE THERMAL STRATIFICATION l

I FOR THE B&W 177-FA NUCLEAR PLANTS 5

i SUM'iArd REPORT FATIGUE STRESS ANALYSIS OF THE SURGE LINE ELBOWS lI l

i Prepared for l Arkansas Power and Light Company l Duke Power Company Florida Power Corporation i General Public Utilities Nuclear Toledo Edison Company I B&W Nuclear Service Company

'I P.O. Box 10935 Lynchburg, Virginia 24506 0935 I

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EXECUTIVE

SUMMARY

tmm1 The purpose of this supplement is to report on the re-evaluation of the pressurizer surge line elbows fatigue usage in order to close out the Nuclear Regulatory Commission Safety Evaluation Report (SER) open item on BAW 2127 (Reference 13). _

Backaroumil On December 20, 1988, the Nuclear Regulatory Commission issued NRC Bulletin 88-i 11. The bulletin addressed technical concerns associated with thermal stratification in the pressurizer surge line and required utilities to establish and implement a program to ensure the structural integrity of the surge line.

The B&W Owners Group has developed a comprehensive program to address the requirements of the bulletin. This program and its results were summarized in BAW 2127, Final Submitt al for Nuclear Reculatory Commission Bulletin RE ll

" Pressurizer Surae line Thermal Stratification", for the B&W-designed lowered I loop plants, in September 1991. Supplement 1, Plant-Snecific Analysis in Resnonse to Nuclear Reaula_ tory Commission Bulletin 89-11 "Prgisurizer Surae line Thermal Stratification" Davis-Besse Nuclear Power Station UnitJ, presented the results for the B&W-designed raised loop plant.

The Nuclear Regulatory Commission Safety Evaluation Report (Reference 13) on BAW-2127 found the methodology uted to analyze and evaluate the stress and fatigue effects due to thermal stratification and thermal striping acceptable with one exception. The NRC staff and its consultant disagreed with the B&W I interpretation of the secondary stress index (C2 ) for elbows, thus, leaving the f atigue usage of the elbows as an open item to be resolved.

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Summary:

A re evaluation of the surge line elbows was performed using the Code table stress indices for elbows. The surgo.line was reevaluated to include shakedown, as described in NB 3228.4 of the ASME Code, and strain based f atigue of the surge line elbovis.

Detailed finite element analyses were performed on the limiting portions (the I

e elbows) of tha surge line piping, as well as on the entire surge line to is demonst rate shakedown. Shakedown was demonstrated to occur within 3 to 4 cycles and resulted in a maximum accumulated local strain of 1.07?; compared to the 5.0i; of ASME Code Cases N 47 28 and N-196-1.

I The fatigue usage for all surge line elbows was demonstrated to be less than 0.6 over the 40 year design life of the B&W 177 fuel assembly plants. Thus, at all surge line locations the 40 year cumulative fatigue usage factor remains less than the Code allowable fatigue usage factor of 1.0.

Conclusion:

Having previously demonstrated the f atigue usage of the nozzles and piping to be acceptable, excluding the elbows, and having demonstrated the acceptability of the elbows in this supplement, the B&WOG response to NRC Bulletin 88-11 is complete.

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I TABLE OF CONTENTS EXECUTIVE

SUMMARY

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1. INTRODUCTION .......................... 11 1-2 l

1.1 Back round . . . . . . . . . . . . . . . . . . . . . . . .

1.2 Conc usion . . . . . . . . . . . . . . . . . . . . . . . . 1-4

2. SHAKEDOWN ANALYSIS ....................... 2-1 21 Generalitics . . . . . . . . . . . . . . . . . . . . . . . 2-1 8 2.2 Shakedown Options .................... 2-1 2.3 Verification of Shakedown by Elastic Plastic Analysis .. 22 2.4 Verification of Shakedown by Use of the Bree Diagram I (lowered loop plants). . . . . . . . . . . . . . . . . . .

2.5 Verification of Shakedown for Davis-Besse Unit 1 2-9 Surge Line . . . . . . . . . . . . . . . . . . . . . . . . 2-9

3. FATIGUE ANALYSIS .............,.......... 3-1 3.1 Generalities . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 Factors Contributing to Fatigue I 3.3 Shakedown Fatigue 3.4 Post-Shakedown Fatigue . . . . . . . . . . . . . . . . . .

3-1 32 3-3 3.5 Input Stress-Strain Curve for Post-Shakedown Fatigue . . . 3-3 I 3.6 Strain Correlations for Post-Shatedown Fatigue . . . . . .

3.7 Elbow Fatigue Analysis Results . . . . . . . . . . . ..

3-4 3-5

4.

SUMMARY

OF RESULTS ....................... 4-1

5. REFERENCES ........................... 5-1 l 6. DOCUMENT SIGNATURES . . . . . , . . . . . . . . . . . . . . . . . 6-1 ATTACHMENT A: Section III, Division 1. NB-3228.4 Code Inquiry and Response (pages A 1 through A-4)

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LIST OF TABLES lahl,t EiLit l-1 Overview of the Highest Total Cumulative Usage Factors in I 3-1 4-1 the Surge Line of the Seven B&W 177-FA Nuclear Plants Total Cumulative Usage factors for the Surge Line Elbows ...

Total Cumulative Usage f actors for the Surge Line Elbows ...

.... 1-6 36 4-2 I

LIST Or FIGURLS I ""*

2-1 PV# 4 X-Di spl acement . . . . . . . . . . .......... . 2-5 2-2 PV# 4 Y-Displacement . . . . . . . . . . . . . . . . . . . . . 26 I 23 3-1 3-2 PV# 4 Surge Surge Z-Displacement . . . .. . . . ... ........... 2-7 Line Mathematical Model . . . . . . . . . . . . . . . . . 3-7 Line Mathematical Model . . . . . . . . . . . . . . . . . 3-8 I

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1. INTRODUCTION This document is a supplement to B&WOG Topical Report BAW-2127, Final Submittal for Nuclear Reaulatory Commission Bulletin 88-11 "Pressuriggr Surae Line Thermal Stratification" (Reference 1). The analyses described in this supplement confirm I that all surge line elbows satisfy applicable ASME Code stress allowables for the operating B&W lowered and raised loop plants, cor.sidering the effects of thermal

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stratification and thermal striping.

This introduction briefly reiterates the background for the thermal stratification, striping and cycling issues, and provides a summary of the pressurizer surge line fatigue analysis results and conclusions contsined within I Reference I which address the technical issues described in NRC Bulletin 89-11 (Reference 2). The remaining sections of this supplement are as follows: .

9 Section 2 discusses the shakedown analysis performed, and provides a brief justification for the use 9f the ABAQUS finite element code, 4 Section 3 discusses the fatigue analysis performed, _

Section 4 provides summary results of the shakedown and fatigue evaluation I

4 of the surge line elbows, as well as the conclusion regarding the integrity of the entire pressurizer surge line, and 4 Section 5 lists all references.

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Ll__ Backaround The surge line in B&W 177 fuel assembly (177-FA) plants is approximately 50 feet I of piping which connects the pressurizer lower head and the reactor coulant hot leg piping. D"H ng plant operation, the reactor coolant system (RCS) is pressurized with a steam bubble in the pressurizer. Thus, the pressurizer contains saturated fluid while the remainder of the RCS is subcooled with temperatures cooler than the pressurizer fluid. During normal plant operation, this temperature dif ference is less than 50'f. However, during plant '.eatup and l cooldown conditions, the temperature difference can be much higher. The surge I line is the fluid flow path through which the pressurizer accommodates changes in RCS liquid volume. When the reactor coolant pumps are operating, there is normally a small outflow from the pressurizer due to continuous pressurizer spray flow through the spray bypass line. During plant heatup and cooldown conditions the surge line accommodates a significant number of fluid exchanges between the pressurizer and RCS, I Due to differences in density, the fluid temperature can vary in the horizontal surge line piping sections from top to bottom with the warmer fluid located above I the more dense (cooler) fluid. This phenomenon, known as thermal stratification, is most pronounced during outsurges from the pressurizer. During an insurge or outsurge under stratified conditions, thermal striping may occur at the fluid layer interf ace. Thermal striping is a rapid oscillation.of the thermal boundary interface caused by interfacial waves and turbulence eff ects. The original plant design fatigue analyses did not account for thermal stratification cycles in the I pressurizer surge line, which causes additional bending moments in the piping, nor did the analyses account for thermal striping which affects the fatigue usage at the inner surface of the pipe.

To assure pressurizer surge line integrity for the 40-year design life of pressurized water reactors (PWRs), the Nuclear Regulatory Commission issued NRC Bulletin Number 88-11, Pressurizer Surce _Line Thermal Stratification (December 20, 1988). This bulletin requires certain actions of licensees of all operating pressurized water reactors. The applicable actions are paraphrased below:

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la. At the first available cold shutdown af ter receipt of the bulletin, and which exceeds seven days, conduct a visual inspection of the pressurizer

<l surge line.

Ib. Within four months of receipt of the bulletin, licensees of plants in operation over ten years are requested to demonstrate that the pressurizer surge line meets the applicable design codes and other TSAR and regulatory I commitments for the licensed life of the giant, considering thermal stratification and thermal triping in the fatigue and stress evaluations;

, or provide the staff with a justification for continued operation while a detailed analysis of the surge line is performed that implements items Ic and Id below.

I Ic. If necessary, obtain plant specific surga line thermal and displacement data. Data can be obtained through collective efforts if sufficient similarities in geometry and operation can be demonstrated.

Id. Update the fatigue and stress analyses to ensure compliance with the applicable Code and Regulatory requirements within two years of receipt of the Bulletin or submit a justification for continued operation and a description of the proposed corrective actions for effecting long-term resolution.

I A portion of the B&W Owners Group program was presented to the Nuclea. Regulatory ,

Commission Staff en September 29, 1988 and April 7,1989. En interim evaluation, BAW 2085, dated May 1989, provided the staff with a justification for near term operation for all of the operating B&W 177 FA plants (Reference 3). T'ie NRC concluded that sufficient information had been provided to justify near term operation for B&W plants until the final report could be completed (References 4 and 15).

The final report for the lowered loop plants was completed and submitted to the NRC in December 1990 (Reference 1). Supplement I to this report, describing the plant-specific enalysis and providing a basis for a Justification for Continued 1-3

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Operation (JCO) iot the raised loop Davis-Besse Unit 1 plar,t, was submitted to the NRC in September 1991 (Reference 7). This report used the same methodology applied in Reference 1. Questions from the NRC, regarding these methods and the applied analysis, wer0 answered by Reference 11.

I The NRC SER, based on the lowered loop plant final report (Reference 1),

contained an open item regarding the analytical technique used to demonstrate I acceptable integrity of the surge line elbows. A similar open itu was included in the NRC acceptance of the JC0 for Davis Besse Unit 1 (Reference 12). As a result, the B&WOG performed a shakedown analysis of the surge line and 4 a rain based f atigue analysis of all surge line elbows. The preliminary results were presented to the NRC at a meeting with the B&W Owners Group Thermal Stratification Working Group on January 15, 1992. This second su;)plement to BAW-2127 summarizes the results from these analysos to demonstrate the integrity of the surge line elbows in order to resolve the NRC open item.

Only those B&W Owners Group program aspects directly contributing to the surge line shakedown and elbow fatigue are considered herein. All other prograt aspects and ger,eral methodologies, including the thermal loading conditions (thermal stratification peaks and valleys, thermal striping, fluid flow conditions), remain unchanged from the original PAW-2127 ano BAW-2127 Supplement 1 (References 1 and 7). The Reference 11 response (Responses to Nucian RtgylAtory Commission Ouestions), regarding earthqbake combinations, remains  ;

I unchanged as a result of this supplement.

1.2 Conclusion The limits on thermal stress ratchet (NB-3222.5), progressive distortion (NB- _

3227.3), local membrane stress (NB-3221.2), primary plus secondary stress intensity (NB-3222.3) and expansion stress intensity (NB-3222.3) need not be satisfied at a specific location if a shakedown analysis satisfying the requirements of NB-3228.4 is performed. hch an inalysis has shown that shakedown of tLa surge line is achieved within a few cycles and the maximum accumulated local strain is well below ASME Code Case allowables (References 9 and 10). The response to an ASME-Code Inquiry (Attachment A and Reference 14)

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has confirmed that the thermal expansion stress criterion of NB-3222.3 does not need to be satisfied since shakedown has been demonstrated in accordance with NB-3228.4(b).

All operating lowered loop plants have been shown to have acceptable cumulative fatigue usage for the full 40 year licensed plant life. Similarly, the elbow fatigue usage factors for Davis-Besse Unit 1, the only operating raised loop I plant, are within the allowable limit of 1.0 for the licensed 40 year plant lifetime based on the design transients defined in BAW 2127, Supplement 1 (Reference 7). In all cases, the structural analysis of the surge line has accounted for the thermal conditions (thermal stratification, thermal striping, and thermal cycling) which are expected to occur during the 40 year life of the plant.

Table 1 1 gives an overview of the highest Total Cumulative Usage Factors in the Surge Line of the seven B&W 177-fA nuclear plants. These factors are based on 240 heatup and cooldown cycles over the 40 year design life for all the plants except for the three Oconee Units which are based on 360 heatup/cooldown cycles over the l 40 year design life.

The Nuclear Regulatory Crmmission has approved operation ot' Davis-Besse Unit I through the 9th operating cycle based upon a Justification for Continued Operation submitted il Saptember, 1991 (Reference 12). An updated csmulative I fatigue usage factor ti r Javis-Besse Unit 1, considering the modifications to be completed prior to res tat t from the 9th refueling outage, will be provided at a later date.

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TABLE l-1. Overview of the Hiahest Total Cumulati.vf_tjase f actors in the Surge Line of the Seven B&W 177-FA Nuclear Plants.

< - --- ---- As Previou sly Reported -- -- -- --- - > - -- Revi sed -- >

Nuclear Pressurizer Surge Line to Surge Line Surge line #

Plants. Surge Nozzle Hot Leg Nozzle Non Elbow Elbows I _ _ _ _

Oconee 0.40 0.59 L_oc et i o_n s 0.48

_____J 0.49 Unit 1 ,

Oconee 0.41 0.59 0.48 0.50 '

. Unit 2 TMI 0.33 0.62 0.38 0.40 Unit 1 7 I C.R.

Unit 3 0.32 0.62 0.37 0.40 AND 0.32 0.62 0.38 0.40 Ur.it 1 .

Oconee 0.41 0.59 0.47 0.48 Unit 3 I Davis-Besse 0.93 0.76 0.62 0.59 Unit 1 (Note 1) (Note 1) (Note 1) (Note 21 j Rgtes relative to Davis-Besse Unit I surae linn Note 1: All of these usage factors assumed certain modifications including a the surge line support whip restraints and thermal insulation during -

< the 7th refueling outage. Additional analyses have been performed a to show that the cumulative usage factors up to the 9th refueling outage (without completion of modifications) are well within the allowable limit of 1.0.

Note 2: The usage factors for the elbows are for a 40 year plant life, and consider the committed modifications including the spring support

> I and pipe whip restraints during the 9th refueling autage to assure unrestricted motion of the surge line.

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General Davis-Be.sse Note:

The NRC Staff concluded (Reference 12) that continued operation of Davis-Cesse Unit 1 is acceptable through the oth refueling outage pending final I resolution of the pressurizer surge line stratification items which should be congleted by the end of the 8th refue ing outage, if possible.

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2. SHAKEDOWN ANALYSIS I 2.1 Generalities The shakedown enalysis has been performed to satisfy the ASME Code (Refereate 6) via use of NB-3228.4, which replaces the requirements of NB-3f.,53.6(a) and/or t4B-3222.3 Thermal Expansion Stress Intensity. The response to an ASME Code Inquiry

< I (Reference 14) has confirmed that the thermal expansion stress criterion of NB-3222.3 does not need to be satisfied if shakedown can be demonstrated in 2 accordance with ND-3228.4(b). '

N8 3213.34 of the ASME Code, Reference 6, defines shakedown as follows:

I " Shakedown of a structure occurs if, after a few cycles of load application, ratcheting ceases. The subsequent structural response is elastic, or elastic plastic, and progressive incremental inebstic deformation is absent. Elastic shakedown is the case in which the subsequent response is elastic."

2.2 Shakqdpwn Options -

Shakedown of the surge line has been dem M strated using two methods. The most sophisticated method is the elastic-plastic analysis of the entire surge line piping system. The second method utilized the Bree-Diagram for the surge line

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g location which experiences the largest strain.

Since the Bree-Diagram only appears in the high temperature Code Case N-47-28 .

(Reference 9), both the Bree Diagram and the more sophisticated elastic-plastic analysis were used. In this elastic-plastic analysis, the surge line was analyzed as a system (surge line modeled from the pressurizer to the hot leg) [

with elastic-plastic properties for all elbows. This model was loaded with the I 2-1 -

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I most severe cyclic loads until shakedown was reached. The elastic-plastic shdedown analysis is described in Sub sections 2.3 through 2.6.

The Bree-Diagram method was also used for the most severe elbow ar.d cyclic loads l (see Sub-sections 2.4 and 2.5).

2.3 Verification of Shakedown by Elastic-Plattic Analysis 2 3._1 Inout Stress-Strain Curves Stress-strain curves (for austenitic stainless steel material) were developed for 150'f, 300F, and 450f and employed kinematic hardening for the -

" loading / unloading" behavior. These were the temperatures corresponding to the thermal conditions of the surge line for the most severe cyclic loads (Reference

1) to be applied in the elastic-plestic shakedown analysis. These stress-strain I curves have been developed using the general equations:

stress o - E

• c in the purely elastic domain, and '

stress o . K * (" in the elastic-plastic somain, where c is the total strain value, l and the exponent n is approximately equal to 0.3 for stainless steels.

These stress-strain curves are based on the following values f rom the ASME Code (Reference 6): the modulus of elasticity, E, the yield stress, Sy (at 0.2%

offset), and the ultimate tensile stress, Su.

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2.3.2 Input loads The most severe thermal stratification ranges for the shakedown analysis have I been defined in the previous fatigue evaluation of the surge line based on the operating design transient information (Pef rence 1). These transients were developed to conservatively reflect past varation. The most severe stress loading range, as identified in BAW-2127 (Reference 1) fatigue evaluation, was between the highly stressed Peak, PV4 of HUlAl heatup (top to bottom temperature difference of 393 F while the pressure is 578 psi ). and nominally stressed Valley, PV402 of C0181 cooldown (top to bottom temperature difference of 12 f I while the pressure is 379 psi.). This severe hypothetical stress load range was specified for 13 design cycles. These 13 cycles envelope the number of 2-2 I

occurrences for all lowered loop plants. Therefore, the shakedown analysis appl:ed this load cycle 13 times, sing the following cyclic loading map:

PV No.4 / PV No.402 / PV No.4 / PV No.402 / PV No.4 / PV No.402, etc...

I The concurrent thermal expansion, thermal stratification, internal pressure, and deadweight were simultaneously loaded on the surge line for each Peak or Valley condition. The model included the anchor motions resulting from the thermal expansion of the pressurizer and RCS hot leg.

2.3.3 Use of ABA0VS finite Element (pic The ABAQUS finite element code was utilized in constructing a model of the surge line to perform the shakedown analysis. ABAQUS (Reference 8), developed by Hi bt,i t t , Karlsson and Sorensen, Inc., is specifically designed for advanced structural analysis. These applications include nonlinear effects which dcminate the overall program design. ABAQuS development started t.arly in 1978. Electric Power Research Institute developmental funding has resulted in the implementation of capabilities that are generally useful in structural integrity evaluations of nuclear components.

Piping system design relies heavily on the use of elbows. This allows extra flexibility in the line to accommodate thermal loading. Therefore, the elbows themselves are the critical design components, it is essential to accurately -

predict their response in order to structurally qualify a pipeline. In the plastic regime, piping elbows achieve their flexibility through a sheli-type behavior, responding to bending loads with significant ovalization of the pipe cross-section. In contrast, the straight pipe cross-section does not deform to I any significant extent until Brazier buckling occurs.

The elbow elements of ABAQUS are intended for piping applications wherein nonlinear effe: . associated with ovalization and warping must be included. The l elbow elements use a combination of polynomial interpolation along the axis of the pipe and fourier interpolation around the circumference of the pipe. The elements contain nine integration points through the thickness and twenty-four 2-3 I

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I around the cross-section. The elbow elements of ABAQUS have been experimentally verified and lend themselves well to the shakedown analysis summarized in this I. document.

The ABAQUS shakedown model has been verified by comparison with the ANSYS mathematical model of the surge line (Reference 1), which was bench-marked to actual surge line displacements measured during the Oconee Unit 1, February,1989 heat-up. To demonstrate that the ABAQUS model yielded comparable displacements, this verification was performed using two purely elastic computer analyses: the I 1007. power thermal expansion load case and the most severe thermal stratification condition, Peak PV4 of HUl Al heat-up. For both load cases, the surge line displacements using ABAQUS agreed with the ones using ANSYS. Figures 2-1 through 2 3 show the results of the comparison between the displacements for the Peak PV4 l load case.

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'4 2.3.4 Shakedown Analysis Results Shakedown has been demonstrated for the most severe ranges of thermal stratification conditions (Peaks and Valleys) based on the following results evaluated at the location nf maximum plastic strain (within each elbow):

a.) Stress-strain Hysteresis and Stability CFeck b.) Maximum Accumulated Local Strain develop.ed through Shakedown c.) Sirain Range par complete cycle (used in Fatigue Analysis)

I t.) Stress-strain Hysteresis Results _

The stress-strain hysteresis was stable after 3 to 4 cycles of loading and unloading.

I B,) Maximum Accumulated local Strain Results I The maximum accumulated local strain was the maximum peak strain developed during the shakedown analysis. The resulting maximum accumulated local strain occurred at the peak stratification, PV4, and was 1.07% for the most critical elbow (Figure 3-1, Elbow B at the bottom of the riser).

Code cases N-47-28, approved July 27,1988, Appendix T (Reference 9), and N-196-1, approved January 21, 1982, Condition No. 2 (Reference 10) provide a maximum allowable accumulated local strain of 5.0%. These Code cases define the _

applicability of this maximum allowable accumulated local strain to any plastically analyzed location at which the calculated local strain is the result I of cyclic operations. The maximum accumulated local strain anywhere in the l shakedown model of the surge line was 1.0/% (1.07% u 5.0%). Therefore, the maximun accumulated local strain from th< shakedown analysis is acceptable, e C.) S.ty,tjn Ranoe Per_ omolete Cycl L esults R

Shakedown occurred within thra to four lotd cycles. After shakedown, the resulting strain range at the most critical elbow (Elbow B at the bottom of the riser) was 0.759%.

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?4 Verification of Shakedown by Use of the Bree Diagrain (lowered-loon plants) knakedown has been demonstrated for the most severe cyclic loads which lead to distortion. The most severe cyclic loads have been defined in the previous fatigue evaluation of the surge line (Reference 1). In the most severe heat-up I to be analyzed (associated with early plant life operation), the primary stress effects due to internal pressure become significant only at the end of this het-up. At that time, the r..ust severe thermal stratification peak stresses (which occur at the beginning of the heat-up) have been established and have shaken down. A Bree Diagram has been built for the surge line location undergoing the largest strain. On this Bree Diagram, the mmt severe thermal stratification cyclic loads (analyzed in the elastic plasti d iown analysis) have been shewn I to be the controlling conditions for shakem - c..:n compared to other conditions during the same heatup transient. Also, on the same Bree Diagt am, all of the stress points corresponding to the thermal stratification Peaks have been shown to be acceptable. This again demonstrates that shakedo- has occurred in the surge line.

2.5 Verification of Shakedown for Davis-Besse Unit 1 Surag Line A separate analysis to demonstrate shakedown of the Davis-Besse (raised loop plant) pressurizer urge line was not performed. The following is justification that the shakedown anal f sis results for the lowercJ loop plants provides bounding results for the Davis-Besse raised loop plant.

I The lowered and raised loop surge lines have the same pipe size (10 inch, schedule 140) and the same material, with similar length and flexibil:'y. Based on BAW-2127 and BAW-2127 Supplement No. 1 (References 1 and 7), the most severe thermal bending stress range (equation 12) for the Davis-Besse je line was less than that of the lowered loop plants. The ratio of elbow in-plane to out-of-plane moment was also lower for Davis-Besse. Therefore, the plastic strain I results in the Davis-Besse surge line elbows would be lower than the results from the analyzed lowered loop plant. This has been illustrated on a Bree diagram, where the corresponding thermal stratification peaks for the most severe Davis-Besse surge line heatup transient were lower than those analyzed in the elastic-plastic shakedown analysis of the lowered loop plant surge lines, Thus, the 2-9 lI

lowered loop plants provide a conservative demonstration of the shakedown for the Davis-Besse surge line, ,

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3. FATIGUE ANAL.YSIS 3.1 Generalities The total cumulative fatigue usage calculated in References 1 and 7 consisted of the sum of:

1) Main Fatigue Usage (Shakedown and Post-Silakedown Fatigue) and I 2) Additional Fatigue Usage (Items 3, 4, and 5 of Suo-section 3.2).

Only the main fatigue usage for cycles with Equation 10 Stress Range Intensity greater than the Code 3*Sm limit was recalculated for this analysis. All other fatigue usage values remained unchanged from Refecences 1 and 7. For Equation 10 Stress Ranges exceeding 3*Sm, fatigue was recalculated using the cyclic strain range as a function of the merr.ent and pressure terms along with a strain based penalty factor applied to the additional peak stresses of that cycle.

3.2 Factors Contributina to Fatique The total cumulative usage factor is equal to the sum of the following fatigue contributions:

1.) Shakedown Fatigue (see Sub-section 3.3).

2.) Post-Shakedown Fatigue (see Sub-section 3.4) of:

a.) Stratified Ranges Failing Equation 10, thus Recalculated b.) Stratified Ranges Passing Equation 10, thus No Recalculation c.) Non-Stratified Ranges Passing Equation 10, thus No Recalculation I 3.) Fatigue Usage for the non-stretified High Velocity Fluid Flow conditions.

4.) Thermal Striping Fatigue by itself.

5.) Fatigue Usage for the remaining Operating Basis Earthquake Ranges.

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. i I Only those contributions designated by items 1. and 2a. (listed above) were recalculated. All other fatigue contributions meet Code Equation 10 and I correspond to the original BAW-2127 and BAW-2127 Supplement 1 (Reference 1 and Reference 7).

3.3 Shakedown Fatiaue For the most severe thermal stratification ranges (PV No.4 - PV No.402) analyzed in the ABAQUS shakedown analysis, the total elastic plastic strain range occurring between the maximum Peak (PV4) and the minimum Valley (PV402) was retained. The purely elastic strain range for the thermal stratification cycle PV4-PV402 was also calculated in an elastic structural analysis of the surge line I mathematical model, and the corresponding Kpp penalty factor (elastic-plastic strain divided by the purely elastic strain) was calculated. Using the total elastic-plastic strain range (for PV4-PV402) and the corresponding Kpp penalty factor, the alternating stress, S,i,, was calculated using the equation below.

S,,, =

• (E7c7

• strain range +

• K, = additional peak stress )

8verage where Kpp was the plastic penalty Octor, and the additional peak stress was the combination of the peak stresses due to fluid flow (through-wall .hermal gradients), thermal striping, and the non-linearity of the temperature profile (combination as performed in Reference 1).

The shakedown fatigue from the resulting alternating stress was performed to retain consistency between the shakedown analysis and the fatigue analysis, even I though the ASME Code only requires the post-shakedown strain ranges to be considered [NB-3228.4(c)]. The shakedown fatigue was also included in response to NRC comments during the January 15, 1992 meeting with the B&W Owners Group Thermal Stratification Working Group.

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I 3.4 Post-Shakedgwn Fatiaue l

l In the Post-Shakedown Fatigue, the calculation of the alternating stress S,g, results from the same formula given in Sub-section 3.3:

I

• S, , , = *(Em ~ strain range +

• Q = additional peak stress )

everage I The values for the strain range and the plastic penalty factor, Kpp (elastic-plastic strain divided by the purely elastic strain), were calculated through a detailed finite element ABAQUS elbow model. The additional peak stress was the I combination of the peak stresses due to fluid flow (through-wall thermal gradients), thermal striping, and the non-liaearity of the temperature profile (combination as performed in Reference 1).

For the thermal stratification ranges which meet Equation 10 Stress Range Intensity of the ASME-Code (within the 3*Sm allowable), the alternating stress values and the usage factors are calculated using the simplified equations of the ASME-Code.

3.5 innut Stress-Strain Curve for Post-Shakedown Fatiaue.

For each thermal stratification cycle, the temperature of interest is the mean valm of the temperature during the cycle [see NB-3222 51.1d NB-3228.4(c)]. This I mean value has been found to be 300 F or less, depending on the location of the elbow analyzed. Therefore, the applicable stress-strain curve will be taken at 300 F in the finite element elastic-plastic analysis of the surge line elbow.

l The input stress-strain data for the finite element elastic-plastic analysis of the elbow (Post-shakedown) are defined as follows: linear purely clastic regime up to 1.5

• Sm (modulus of elasticity E from the ASME Code), and linear stress-strain relationship in the elastic-plastic regime intersecting the nonstrain-hardened stress-strain curves at the 1.0 % total strain location. This second l

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I slope is conservatively low in that no strain-hardening is included for the stress at 1.0 % total strain.

Note that in the fatigue analysis, the moment ranges applied af ter shakedown on each one of the surge line elbows were calculated " purely elastically" in the structural analysis of the surge line. Reduction of the moments due to plasticity in the elbows is not taken into account, 3.6 Strain Correlations for post-Shakedown Fatiaya The ABAQUS surge line elbow finite-element model has been loaded by combinations of bending moments and internal pressure values (the internal pressure values I considered are 0 psi., 600 psi., 1500 psi. and 2250 psi.).

From the different load cases, the highest strain value anywhere in the elbow was empl oyed. An analysis has been performed to compare (for the same internal pressure values) the highest strain range. due to a combined "in-plane opening moment / in-plane closing moment" load case with the highest strain range due to "out-of-plane moment" in the elbow mid-section. From this analysis it has been I found that the "out-of-piane moment" load case leads to a lower value for the lg highest strain range in the elbow. Therefore, using the "in-plane opening moment 5 / in-plane closing moment" load case, correlation tables were built for the calculation of the highest strain range anywhere in the elbow as a function of the elastically calculated moment range and of the internal pressure in the elbow. Similarly, correlation tables were built for the plastic penalty factor Kpp to be applied on the additional peak stresses.

l For each thermal stratification cycle, the strain range and the plastic penalty I factor, Kpp, are calculated through a conservative linear interpolation between i

values in the correlation tables mentioned above. Knowing this strain range and the corresponding Kpp value, the alternating stress is calculated using the formula given earlier in Sub-section 3.4.

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4 I }.7 Elbow Fatiaue Analysis Results For the surge line elbows of the lowered loop plants, the highest 40-year total cumulative usage factor occurred in elbow B of Figure 3-1 (vertical elbow at the bottom of the surge line riser). It was equal to 0.50 for Oconee Unit 2, 0.49 I for Oconee Unit 1, 0.48 for Oconee Unit 3, and 0.40 for TMI Unit 1, Crystal River Unit 3 and Arkansas Nuclear One Unit 1.

For the surge line elbo...; of Davis-Besse Unit 1, the shakedown fatigue was based on the maximum strain range calculated in the surge line elbows of the lowered loop plants. The strain range of the most highly stressed elbow in the lowered loop plarits was conservatively applied to each elbow of the Davis-Besse Unit 1 I plant. The remaining fatigue (post-shakedown) for Davis-Besse utilized Davis-Besse loads and the elastic-plastic finite element elbow results. This resulted in a maximum total cumulative usage facto of 0.59. The maximum total cumalative usage factor for Davis-Gesse Unit 1 occur red in the vertical elbow at the bottom of the pressurizer, elbow A.

Table 3-1 on the following page gives the resulting Total Cumulative Usage Factors for all elbows (see Figure 3-1 and Figure 3-2) and highlights the maximum values for each plant.

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!I Elbow Fatique Analysis:

TABLE 3_1. TOTAL CUMUL ATIVE USAGE FACTORS FOR THE SURGE LINE ELBOWS Elbows ---> Elbow A Elbow B Elbe v C Elbow D Elbow E I Oconee Unit 1 (4 elbows) 0.13 0.49 0.34 0.38 does not apply.

Oconee Unit 2 0.13 0.50 0.35 0.39 does not I (4 elbows)

TMl Unit 1 0.10 0.40 0.29 0.33 apply.

does not (4 elbows) apply.

Cr. River Unit 3 0.10 0.40 0.29 0.33 does not _

(4 elbows) . apply.

I ANO Unit 1 (4 elbows) 0.10 r-0.40 0.29 0.33 does not apply.

I Oconee Unit 3 (4 elbows) 0.13 0.48 0.33 0.38 does not apply.

Davis-Besse Unit 1 0.59 0.57 0.44 0.15 0.30 (5 elbows)

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Figure 3-1 Surge Line Mo theno tical Mociel ELBOV A 32 I 5 4

HDT LEG 6

elements 1 through 5 7

Jt B = -550' jy y= - 1.5 0 '

8 'Z= 24S656' I Z,/ * -

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,_ ete-ents

/' D througn 13 10 f

11 p' ELBOV B 12 90'  ! (x = + 53 0')

14 13 ' Dr c.n No :le 5 /'for Ns:S - 3, 4 5 ano 9 I \

16 42 g f elements 23 in-ougn 26 .\

g g, 18 NSSS PLANT x x 39

\ r

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3 OCONEE 1 (X= 41.0 ') 20 Droin Nozzle I 4 5

7 DCONEE 2 TMI 1 CR 3 for NSSS- 7 and 8. y\ l 22 #

8 ANO 1 "n p",

9 OCONCE 3 ELBOV C -

p Pressurizer 26'5 elements 29 through 32 27 12 / /

/ 23 I .

ELBOV D 31 Q 39 29 I 3-7 I

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num sum aus uma man uma num sms sum aus uma uma amm ama amm mas - -

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ums Figure 3-2 Surge Line Mathematical Model NSSS PLANT EWOW D 14 DB1 34 33 35 32 36 37 PRESSURIZER 4i

,2 ao N-Q x43 44 w ,3

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iD EWOWD 'N?

N HOT LEG '

29 13 12 11 10  !

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87 28 '

18 EWOW A EWOWE

' 19 20 4 27 21 \

26 25 2, 23 22

\x y Drain Line '~~

EWOW C A Z .

X /

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4.

SUMMARY

OF RESULTS Surae Line Shakedown Analysis:

Surge line shakedown occurred within the first three to four thermal stratification cycles. The maximum peak strain anywhere in the elbow, and at any time during the shakedown, occurred in elbow B for the lowered loop plants (vertical elbow at the bottom of the surge line riser). It was equal to 1.07%,

I to be compared with the maximum allowable peak strain of 5.0%. Therefore, the shakedown analysis has satisfied the Primary Plus Secondary and the Thermal Expansion Stress Intensity requirements of the ASME Code (References 6 and 14).

Fatiaue Usaae:

The fatigue usage for all surge line elbows was demonstrated to be less than 0.6 over the 40 year design life of the B&W 177 fuel assembly plants. Thus, the cumulative fatigue usage factor remains less than the Code allowable fatigue usage factor of 1.0 at all surge line locations. The results from the elbow fatigue analysis are presented on the following page, Table 4-1.

~

Conclusion:

Having previously demonstrated the fatigue usage of the nozzles and piping to be I acceptable, excluding the elbows, and having demonstrated the acceptab'ility of the elbows in this supplement, the B&WOG response to NRC Bulletin 88-11 has been completed. The highest Total Cumulative Usage Factors for all surge line locations are given in Table 1-1 of Section 1.

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TABLE 4-1. _ TOTAL CUMU.LATIVE USAGE FACTORS FOR TH LSURGt LINL [LB0WS El bows ---> Elbow A Elbow B Elbow C Elbow D Elbow E I Oconee Unit 1 0.13 0.49 0.34 0.38 does not (4 elbows) apply.

I Oconee Unit 2 (4 elbows) 0.13 0.50 0.35 0.39 does not apply.

TMI Unit 1 0.10 0.40 0.29 0.33 does not (4 elbows) apply.

Cr. River Unit 3 0.10 0.29 0.33 does not I (4 elbows)

ANO Unit 1 0.10 0.40 0.40 0.29 0.33 apply.

does not

~

(4 elbows) ,

apply.

_ Oconee Unit 3 0.13 0.48 0.33 0.38 does not J elbows) _ __.

apply.

Davis-Besse Unit 1 0.59 0.57 l 0.44 0.15 0.30 (5 elbows) l The elbows are shown on Figure 3-1 and Figure 3-2.

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5. References I 1. BAW-2127, Final Submittal for Nuclear Reaulatory Commission Bulletin 88-11

" Pressurizer Surae Line Thermal Stratification", December 1990.

I " Pressurizer Surge Line Thermal Stratification",

2, NRC Bulletin 88-11 December 20, 1988.

3. BAW-2085, Submittal in Resoonse to NRC Bulletin 88,-11. " Pressurizer Surae Line Thermal Stratification", May 1989.

4. NRC Letter dated May 18, 1990, J.T. Larkins to M. A. Haghi, " Evaluation of Babcock and Wilcox Owners Group Bounding Analysis Regarding NRC Bulletin 88-11".

5. "ANSYS" Computer Code, Versions 4.lc and 4.3. Engineering Analysis System, User's Manual Volumes I and II, Swanson Analysis Systems, Inc.

I 6. "ASME Boiler and Pressure Vessel Code", Section Ill,1986 Edition with no i

Addenda, i

1

7. BAW-2127, Supolement 1, Plant-Soecific Analysis in Resoinse to Nucleer Reaulatory Commission Bulletin 88-11 " Pressurizer Surge Line Thermal Stratification" Davis-Besse Nuclear Power Station Unit 1, September 1991, I B&W Owners Group letter number OG-970.

"ABA0VS" Computer Code, Version 4.G.5, Hibbitt, Karlsson & Sorensen Inc.,

I 8.

Pawtucket, RI, User's Manual Volume I,1989.

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9. ASME Code Case N 47-28, " Class 1 Components in ElevateJ Temperature Service", approved July 27, 1988, Appendix T, Paragraph T-1310, " Limits I for inelastic Strains".

10. ASME Code Case N-196-1, " Exemption from the Shakedown Requirements When Plastic Analysis is Performed", approved January 21, 1982, Condition Number 2.

Responses to Nuclear Reaulatory Commission Questions on B&W Owners Group I 11.

Report BAW-2127, Final Submittal for Nuclear Reaulatory Commission _

Bulletin 88-11 " Pressurizer Surae Line Thermal Stratification", Octot,ar 1991, B&W Owners Group letter number OG-961.

12. NRC Letter dated November 6, 1991, Jon B. Hopkins, S'., to Donald C.

Shel ton, Docket No. 50-346,

Subject:

" Pressurizer Surge Line Thermal Stratification, NRC Bulletin 88-11 (TAC NO. 72128)".

13. NRC Letter dated July 24, 1991, Joseph W. Shea to James A. Taylor, I

Subject:

"NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, Safety Evaluation Report".

14. Response to a Code Inquiry relative tc Section III, Division 1, NB-3228.4, File N192-6, March 26, 1992. This Code Inquiry and Response is provided 2 herein as Attachment A.

I 15. NRC Letter dated August 7,1990, M. D. Lynch, Sr. to D. C. Shelton, Toledo Edison Company, ' "NRC Bulletin 88-11, Pressurizer Surge t.ina Thermal Stratification - Evaluation of Babcock and Wilcox Owners Group Bounding Analysis" (TAC No. M72128).

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6. DOCUMENT SIGNATURES I This document prepared by:

A

~

l'A lo//0 9R K./FI Bratcher, i.ngineer 11 '

/ 7 Material ructural Analysis

). -

f o ! 42. -

R. J. Gurd EngineeV1V / '

Material & tructurar Analysis This document reviewed for technical content and accuracy by:

g # L S& 44a J d //e z-W. D. Maxham, Supervisor' Material & Structural Analysis Verification of independent review:

hh_

K. E. Moore, Manager

$ /6 -[/1 Material & Structural Analysis Manager This document approved for release: /

I ,-

> AuuA C40/92 A. W. Robinion, Program Manager B&W Owners Group Thermal Stratification I Working Grouo I

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

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l ATTACHMEt<T A Section 111, Division 1, j NB-3228.4 Code Inquiry and Response I

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--,m---' - - . , - - w- g y-w-- r

oove r LMocas, e==i rttrovwe tscistasi:No stucts 130 SECONo AvtNut W ALTMM MA%ACHUSETTS C22'd 9195

" (617) thB3350 T*x (7101324 7'40 December 30, 1991 7569-2 I

Secretary ,

i ASME Boiler and Pressure Vessel Committee 345 East 47th Street I New York, NY 10017 Subiect: Technical Inquiry - ASME BPVC Section III Gentlemen:

The writer respectfully requests that the attached Technical Inquiry be considered by Section III, Very truly yours, 5

I DFL/tmo Attachment I

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Secretary ASME Boiler and Pressure Vessel Committee 7569-1 Attachment SCOPE Additinnal guidance is requested regarding paragraph NB-3228.4, Shakedown Analysis (19 Edition with Addendum).

BACKGROUND I The structural integrity of a pressurizer surge line undergoing _

thermal loading (including expansion bending moments and forces) as a result of flow stratification has been demonstrated by performing a Shakedown Analysis in accordance with NB 3228.4 conservatively using kinematic hardening. Shakedown occurred in a few cycles and a cumulative usage factor of < l.0 over the design life was c al cul ated . The deformations prior to shakedown are well within specified limits.

l Subparagraph (b) of NB-3228.4 recognizes that the following limits have been satisfied by the Shakedown Analysis:

NB-3221.2 - Local Membrane Stress Intensity NB-3222.2 - Primary Plus Secondary Stress Intensity NB-3222.5 - Thermal Stress Ratchet NB-3227.3 - Progrenive Distortion of Nonintegral Connections However, satisfaction of NB-3222.3 Expansion Stress Intensity is not -

specifically exempted even though in satisfying NB-3222.2 for piping, loadings categorized as expansion must be included.

LNAIJIRY I In demons crating Shakedown in accordance with NB-3228.4(b) are the expansion stress criterion of NB-3222.3 satisfied?

RESPONSES Yes, as long as the range of strain calculated on a plastic isit includes the effect of all cyclic loads which lead to distortion.

A-3

E.-- -

31 t'. 3 The American Society of 0::e::. : 0::..:::::

• Me:han::al b;inear; in

45 Em c:s 5:ree:

New en NY '0017 Donald F. Landers President Telee/ne Engineering Services 130 Second Ave.

I Waltharn, MA 02254

Subject:

Section lit, Division 1. NS 3228.4 File #: N192-6 Your letter dated December 30,1991 Reference 4

Dear Mr. Landers:

Our understanding of the questions in your inquiry, and our reply, is as follows:

Question: In demonstrating shakedown in accorcance with NS-3228.4(b), does the e@ans.on stress enterion of NS-3222.3 need to be satisfied?

Reply: No.

I Very truly you.%

I

/

i enristian sanna

'g' Assistant Secretary, Boiler & Pressure Vessel Committee _

g (212) 605-4705 I

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