ML031910661

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WCAP-16019-NP, Rev 0, Technical Justification for Eliminating 10 Accumulator Lines Rupture as the Structural Design Basis for Callaway Nuclear Power Plant.
ML031910661
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Site: Callaway Ameren icon.png
Issue date: 02/28/2003
From: Bhowmick D, Ching Ng, Petsche J, Swamy S
Westinghouse
To:
Office of Nuclear Reactor Regulation
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ML031910683 List:
References
FOIA/PA-2005-0108 WCAP-16019-NP, Rev 0
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Westinghouse Non-Proprietary Class 3 WCAP-16019-NP February 2003 Revision 0 Technical Justification for Eliminating 10" Accumulator Lines Rupture as the Structural Design Basis for Callaway Nuclear Power Plant O Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16019-NP Revision 0 Technical Justification for Eliminating 10" Accumulator Lines Rupture as the Structural Design Basis for Callaway Nuclear Power Plant D. C. Bhowmick C. K. Ng February 2003 Verified: 5i 7j~

(J*F. Petsche Structural Mechanics Technology Approved:

S.M A. m, ager Structural Mechanics Technology Westinghouse Electric Company LLC P.O. Box 355

- Pittsburgh, PA 15230-0355 0 2003 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES .............. v LIST OF FIGURES .............. vii 1 INTRODUCTION .. 1-1 1.1 Background .1-1 1.2 Scope and Objective .1-1 1.3 References .1-2 2 OPERATION AND STABILITY OF THE ACCUMULATOR LINES . .2-1 2.1 Stress Corrosion Cracking .2-1 2.2 Water Hammer .2-2 2.3 Low Cycle and High Cycle Fatigue .2-2 2.4 Other Possible Degradation During Service of The Accumulator Lines .2-3 3 MATERIAL CHARACTERIZATION .. 3-1 3.1 Pipe Materials And Weld Process .3-1 3.2 Material Properties .3-1 3.3 References .3-1 4 LOADS FOR FRACTURE MECHANICS ANALYSIS . .4-1 4.1 Nature Of The Loads .4-1 4.2 Loads for Crack Stability Analysis .4-2 4.3 Loads for Leak Rate Evaluation .4-2 4.4 Summary Of Loads And Geometry For The 1o" Accumulator Lines .4-2 4.5 Governing Locations For The Accumulator Lines .4-3 5 FRACTURE MECHANICS EVALUATION .. 5-1 February 2003

iv 5.1 Global Failure Mechanism ........................................ 5-1 5.2 Leak Rate Predictions............................................................................................5-2 5.2.1 General Considerations ........................................ 5-2 5.2.2 Calculation Method ........................................ 5-2 5.2.3 Leak Rate Calculations ........................................ 5-3 5.3 Stability Evaluation ......................................... 5-4 5.4 References ........................................... 5-4 6 ASSESSMENT OF FATIGUE CRACK GROWTH . .6-1 6.1 Introduction .6-1 6.2 Critical Location for Fatigue Crack Growth Analysis .6-1 6.3 Design Transients .6-1 6.4 Stress Analysis .6-1 6.5 OBE Loads .6-2 6.6 Total Stress For Fatigue Crack Growth .6-2 6.7 Fatigue Crack Growth Analysis .6-3 6.7.1 Analysis Procedure .6-3 6.8 Results...................................................................................................................6-4 6.9 References. 6-4 7 ASSESSMENT OF MARGINS .. 7-8 CONCLUSIONS .. 8-1 APPENDIX A - LIMIT MOMENT.............................................................................................. A-1 February 2003

V LIST OF TABLES Table 3-1: Room Temperature Material Properties for the 10" Accumulator Lines ................ 3-2 Table 3-2: Representative Tensile Properties for the 10" Accumulator Lines at Operating Temperatures ................................................................. 3-3 Table 3-3: Modulus of Elasticity (E) for the 10" Accumulator Lines ....................................... 3-3 Table 4-1 : Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10" Accumulator Line Loop I ........................................ 4-4 Table 4-2a: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 1 (Case A: 70 0F)........................................................ 4-5 Table 4-2b: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop I (Case B: 48 0F) ........................................................ 4-6 Table 4-3a: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop I (Case A: 70 0F)........................................................ 4-7 Table 4-3b : Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 1 (Case B: 48 0F) ............................................................ 4-8 Table 4-4: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10" Accumulator Line Loop 2 ........................................ 4-9 Table 4-5a : Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 2 (Case A: 70 0F)........................................................ 4-11 Table 4-5b: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 2 (Case B: 480 F) ........................................................ 4-12 Table 4-6a : Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 2 (Case A: 700F) ........................................................ 4-13 Table 4-6b : Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 2 (Case B: 480F) .......................................................... 4-14 Table 4-7: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10" Accumulator Line Loop 3 ....................................... 4-15 Table 4-8a: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 3 (Case A: 700 F)........................................................ 4-16 February 2003

vi Table 4-8b: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 3 (Case B: 480F) .................................................... 4-17 Table 4-9a: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 3 (Case A: 700 F)..................................................... 4-18 Table 4-9b: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 3 (Case B: 48 0F) .......................................................... 4-19 Table 4-10: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10" Accumulator Line Loop4 ..................................... 4-20 Table 4-11a :Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 4 (Case A: 700 F).................................................... 4-21 Table 4-11b :Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 4 (Case B: 480 F) .................................................... 4-22 Table 4-12a :Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 4 (Case A: 700F).................................................... 4-23 Table 4-12b :Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 4 (Case B: 480 F) .......................................................... 4-24 Table 5-1 : Leakage Flaw Sizes .............................................................. 5-5 Table 5-2: Summary of Critical Flaw Sizes .............................................................. 5-5 Table 6-1 : Design Transients Considered for Fatigue Crack Growth Evaluation ................... 6-5 Table 6-2: Accumulator Lines Fatigue Crack Growth Results............................................... 6-5 Table 7-1 : Leakage Flaw Sizes, Critical Flaw Sizes and Margins ......................................... 7-2 Table 7-2: LBB Conservatism .............................................................. 7-2 February 2003

vii LIST OF FIGURES Figure 3-1 Callaway Nuclear Power Plant 10" Accumulator Line Loop 1 Layout ................ 3-4 Figure 3-2 Callaway Nuclear Power Plant 10" Accumulator Line Loop 2 Layout ....... ......... 3-5 Figure 3-3 Callaway Nuclear Power Plant 10" Accumulator Line Loop 3 Layout ................ 3-6 Figure 3-4 Callaway Nuclear Power Plant 10" Accumulator Line Loop 4 Layout................ 3-7 Figure 4-1 Governing Weld Locations for 10" Accumulator Line Loop 2 .......................... 4-25 Figure 4-2 Governing Weld Location for 10" Accumulator Line Loop 3 ............................ 4-26 Figure 5-1 Fully Plastic Stress Distribution ................................................... 5-6 Figure 5-2 Analytical Predications of Critical Flow Rates of Steam-Water Mixtures ........... 5-7 Figure 5-3 ]a,ce Pressure Ratio as a Function of L/D .............................. 5-8 Figure 5-4 Idealized Pressure Drop Profile through a Postulated Crack ............................ 5-9 Figure 5-5 Loads acting on the Model at the Governing Locations .................................. 5-10 Figure 5-6 Critical Flaw Size Prediction for Node 3020 ................................................... 5-11 Figure 5-7 Critical Flaw Size Prediction for Node 3120 ................................................... 5-12 Figure 5-8 Critical Flaw Size Prediction for Node 3295 ................................................... 5-13 Figure 6-1 Schematic of Accumulator Line at RCL Cold Leg Nozzle Weld Location .......... 6-6 Figure 6-2 Reference Crack Growth Curves for Stainless Steel in Air Environment ........... 6-7 Figure A-1 Pipe with A Through-Wall Crack in Bending ................................................... A-2 February 2003

1-1 1 INTRODUCTION

1.1 BACKGROUND

The current structural design basis for the 10" Accumulator lines requires postulating non-mechanistic circumferential and longitudinal pipe breaks. This results in additional plant hardware (e.g. pipe whip restraints and jet shields) which would mitigate the dynamic consequences of the pipe breaks. It is therefore highly desirable to be realistic in the postulation of pipe breaks for the 10m Accumulator lines. Presented in this report are the descriptions of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type of break will not occur within the Accumulator lines. The evaluations consider that circumferentially oriented flaws cover longitudinal cases.

1.2 SCOPE AND OBJECTIVE The purpose of this investigation is to demonstrate Leak-Before-Break (LBB) of the 10.

Accumulator lines. The scope of this work covers the 10 Accumulator lines from the reactor coolant loop nozzle connection to the HV-8808 isolation valves. Schematic drawings of the piping systems are shown in Section 3. The recommendations and criteria proposed in SRP 3.6.3 (Reference 1-2) are used in this evaluation. The criteria and the resulting steps of the evaluation procedure can be briefly summarized as follows:

1. Calculate the applied loads. Identify the location(s) at which the highest faulted stress occurs.
2. Identify the materials and the material properties.
3. Postulate a surface flaw at the governing location. Determine fatigue crack growth.

Show that a through-wall crack will not result.

4. Postulate a through-wall flaw at the governing location(s). The size of the flaw should be large enough so that the leakage is assured of detection with margin using the installed leak detection equipment when the pipe is subjected to normal operating loads.

Demonstrate that there is a margin of 10 between the calculated leak rate and the leak detection capability.

5. Using maximum faulted loads in the stability analysis, demonstrate that there is a margin of 2 between the leakage size flaw and the critical size flaw.
6. Review the operating history to ascertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer or low and high cycle fatigue.
7. For the materials types used in the Plant, provide representative material properties.

IntroductionFeray20 Introduction February 2003

1-2 The leak rate is calculated for the normal operating condition. The leak rate prediction model used in this evaluation is an [

]ace. The crack opening area required for calculating the leak rates is obtained by subjecting the postulated through-wall flaw to normal operating loads (Reference 1-3). Surface roughness is accounted for in determining the leak rate through the postulated flaw.

It should be noted that the terms flaw and "crack have the same meaning and are used interchangeably. "Governing location" and critical locations are also used interchangeably throughout the report.

1.3 REFERENCES

1-1 WCAP-7211, Revision 4, "Energy Systems Business Unit Policy and Procedures for Management, Classification, and Release of Information," January 2001.

1-2 Standard Review Plan; public comments solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal RegisterNol. 52, No. 167/Friday, August 28, 1987/Notices, pp.

32626-32633.

1-3 NUREG/CR-3464, 1983, "The Application of Fracture Proof Design Methods Using Tearing Instability Theory to Nuclear Piping Postulating Circumferential Through Wall Cracks."

IntroductionFeray20 Introduction February 2003

2-1 2 OPERATION AND STABILITY OF THE ACCUMULATOR LINES 2.1 STRESS CORROSION CRACKING The Westinghouse Reactor Coolant System (RCS) Class 1 lines have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking, IGSCC). This operating history totals over 1100 reactor-years, including 5 plants each having over 30 years of operation, 4 plants each with over 25 years of operation, 12 plants each with over 20 years of operation and 8 plants each with over 15 years of operation.

For stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. Since some residual stresses and some degree of material susceptibility exist in any stainless steel piping, the potential for stress corrosion is minimized by properly selecting a material immune to SCC as well as preventing the occurrence of a corrosive environment. The material specifications consider compatibility with the system's operating environment (both internal and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, and processing.

The elements of a water environment known to increase the susceptibility of austenitic stainless steel to stress corrosion are: oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides, sulfites, and thionatis). Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes and preoperational testing, water chemistry is controlled in accordance with written specifications. Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping.

During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. For example, during normal power operation, oxygen concentration in the RCS Class 1 lines is expected to be in the parts per billion (ppb) range by controlling charging flow chemistry and maintaining hydrogen in the reactor coolant at specified concentrations. Halogen concentrations are also stringently controlled by maintaining concentrations of chlorides and fluorides within the specified limits. This is assured by controlling charging flow chemistry. Thus during plant operation, the likelihood of stress corrosion cracking is minimized.

Wall thinning by erosion and erosion-corrosion effects will not occur in the Accumulator lines due to the low velocity and the material, austenitic stainless steel, is highly resistant to these degradation mechanisms. Therefore, wall thinning is not a significant concern in the portion of the system being addressed in this evaluation.

Operation and Stability of the Accumulator Lines ;perationandStabiityoftAccumulaorLinesFebruary 2003

2-2 As a result of the recent issue of Primary Water Stress Corrosion Cracking (PWSCC) occurring in V. C. Summer reactor vessel hot leg nozzle, the Alloy 82/182 weld is being currently investigated under the EPRI Materials Reliability Project (MRP) Program. It should be noted that the susceptible material under investigation is not found in the 100 Accumulator lines at the Callaway Nuclear Power Plant.

2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS and connecting accumulator lines since they are designed and operated to preclude the voiding condition in normally filled lines.

The RCS and connecting accumulator lines including piping and components are designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. Other valve and pump actuations are relatively slow transients with no significant effect on the system dynamic loads. To ensure dynamic system stability, reactor coolant parameters are stringently controlled. Temperature during normal operation is maintained within a narrow range by the control rod positions; pressure is also controlled within a narrow range for steady-state conditions by the pressurizer heaters and the pressurizer spray. The flow characteristics of the system remain constant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics are controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system and the connecting auxiliary lines. Preoperational testing and operating experience has verified the Westinghouse approach. The operating transients of the RCS primary piping and connected accumulator lines are such that no significant water hammer can occur.

2.3 LOW CYCLE AND HIGH CYCLE FATIGUE An assessment of the low cycle fatigue loading is discussed in Section 6 as part of this study in the form of a fatigue crack growth analysis.

Pump vibrations during operation would result in high cycle fatigue loads in the piping system.

During operation, an alarm signals the exceedance of the RC pump shaft vibration limits. Field measurements have been made on the reactor coolant loop piping in a number of plants during hot functional testing. Stresses in the elbow below the RC pump have been found to be very small, between 2 and 3 ksi at the highest. Field measurements on typical PWR plants indicate vibration amplitudes less than 1 ksi. When translated to the connecting Accumulator lines, these stresses would be even lower, well below the fatigue endurance limit for the Accumulator line material and would result in an applied stress intensity factor below the threshold for fatigue crack growth.

Operation and Stability of the Accumulator Unes February 2003

2-3 2.4 OTHER POSSIBLE DEGRADATION DURING SERVICE OF THE ACCUMULATOR LINES The 10" Accumulator lines and associated fittings for the Callaway Nuclear Power Plant are forged product forms, which are not susceptible to toughness degradation due to thermal aging.

The maximum normal operating temperature of the 10" Accumulator lines is about 558 0F. This is well below the temperature which would cause any creep damage in stainless steel piping.

Stability of and Stability Operation and the Accumulator of the Lines Accumulator Lines February 2003 Operation February 2003

3-1 3 MATERIAL CHARACTERIZATION 3.1 PIPE MATERIALS AND WELD PROCESS The material types of the 10i Accumulator lines for the Callaway Nuclear Power Plant are SA376-TP304, SA312-TP304, SA358 TP304 and SA403 TP304. They are wrought product of the types used for the piping of several PWR plants. The 10" Accumulator lines do not include any cast pipes or cast fittings. The welding processes used are Gas Tungsten Arc Weld (GTAW) and Shielded Metal Arc Weld (SMAW) combination or GTAW. Figures 3-1 to 3-4 show the schematic layouts of the 10" Accumulator lines Loops 1, 2, 3 and 4 and also identify the weld locations by node points.

In the following sections the tensile properties of the materials are presented for use in the Leak-Before-Break analyses.

3.2 MATERIAL PROPERTIES The room temperature mechanical properties of the Callaway Nuclear Power Plant 100 Accumulator lines material were obtained from the Certified Materials Test Reports (CMTRs) and are given in Table 3-1. The material properties at temperatures (700 F and 5580F) are required for the leak rate and stability analyses. The minimum and average tensile properties at the temperatures of interest stated above were calculated by using the ratio of the ASME Code Section II (Reference 3-1) properties and those tabulated in Table 3-1. Table 3-2 shows the representative minimum and average tensile properties at various operating temperatures.

The modulus of elasticity values were established at various temperatures from the ASME Code Section II (see Table 3-3). In the Leak-Before-Break evaluation, the representative minimum yield and minimum ultimate strengths at operating temperature were used for the flaw stability evaluations and the representative average yield strength properties were used for the leak rate predictions. These properties are summarized in Table 3-2.

3.3 REFERENCES

3-1 ASME Boiler and Pressure Vessel Code Section II, Part D - Material Properties, 2001 Edition, July 1, 2001, ASME Boiler and Pressure Vessel Committee, Subcommittee on Materials.

February 2003 Material Characterization Material Characterization Febnjary 2003

3-2 Table 3-1: Room Temperature Material Properties for the 10" Accumulator Lines Yield Ultimate Heat No. (SIN)* Material Strength Strength (psi) (psi) 5-789 (396) SA312 TP304 43900 85000 ERPB (439) SA403 WP304 37845 83960 ERLE (441) SA403 WP304 37195 80880 5-744 (396) SA312 TP304 44800 87700 ERPB (427) SA403 WP304 37845 83960 U4KY-H2 (371) SA403 WP304 39000 90500 25223 (315) SA358 TP304 37600 86900 24942 (315) SA358 TP304 39600 85600 61107 (162) SA358 TP304 48000 90000 45060 (99) SA403 WP304 38500 84500 F61056 (199) SA312 TP304 43200 89800 64034 (162) SA358 TP304 38400 85600 42276 (7) SA403 TP304 41600 91500 44838(243) SA403 WP304 37400 88200 U4KY-H2 (264) SA403 WP304 39000 90500 5-751 (396) SA312 TP304 46200 85600 ERPA (592) SA403 WP304 40600 85460 44429(83) SA403 WP304 35500 84500 43481(417) SA358 TP304 37400 88200 47741(278) SA403 WP304 41400 90800 43778(95) SA403 WP304 44800 85200 3083-6-2 (559) SA312 TP304 42600 82100 24955(488) SA358 TP304 39200 85700 45061 (209) SA358 TP304 37000 84000 45059 (160) SA403 WP304 37000 84000 43481(243) SA403 TP304 37400 88200 41960(7) SA403 WP304 33200 80600 ERDK (121) SA403 WP304 36800 84380

  • S/N: Serial Number Febmary 2003 Characterization Material Characterization February 2003

3-3 Table 3-2: Representative Tensile Properties for the 10" Accumulator Unes at Operating Temperatures Minimum Average Minimum Temperature Yield Yield Ultimate Material (OF) (psi) (P si)

SA376/SA358/SA312 TP304 or SA403 WP304 558 20827 24936 68134 SA376/SA358/SA312 TP304 or SA403 WP304 70 33200 39749 80600 Table 3-3 : Modulus of Elasticity (E) for the 10" Accumulator Lines Material Temperature (OF) E (106 psi)

SA376/SA358/SA312 TP304 or SA403 WP304 l 558 l 25.510 SA376/SA358/SA312 TP304 or SA403 WP304 70 28.300 Characterization February 2003 Material Characterization Material February 2003

3-4 HV-8808A

,i Y -3290 3245 3240 3215 3200 3175 3050 3065~~~~~ 3 31 300 4 308031 314 Lo 3035 ILyou

< ~~~~~~~3120 LOD EG LOOP I Figure 3-1 Callaway Nuclear Power Plant IO" Accumulator Line Loop I Layout Chraterzt FebrAL 2003i Matna Material Characterization February 2003

3-5 3085 3100 3150 3180 3185 3045 3120 3170 3040 3205 3035 3220 3235 COLD LEG LOOP 2 3240 3265 3270 4 3275

- 3280 HV-8808B Figure 3-2 Callaway Nuclear Power Plant 10" Accumulator Line Loop 2 Layout Fe.u r Chr.erz to

_2003 Maera Characterization Material February 2003

3-6 3045 3050 3040 /

3065 CODLEG LOOP 3 3 03 3077 302 3090 3120 31, 25-_3190 3170 L3215 3240 3270 ACCMLJ 3235 0-

, t 3275 3330 3335_

3295 3300 3310 HV-8808C Figure 3-3 Callaway Nuclear Power Plant 10" Accumulator Une Loop 3 Layout Ma. a .n r ce z to Fe ru. 2003._.

Material Characterization February 2003

3-7 HV-8808D 3410 3300 3310 3400 3290 3280 3340 3260 3060 3250 3250 9 3803 3190

<3170 3110 4 3050 3140 3130 3 0 COLD LEG LOOP 4 Figure 3-4 Callaway Nuclear Power Plant 10" Accumulator Line Loop 4 Layout materiai Characterization February 2003

4-1 4 LOADS FOR FRACTURE MECHANICS ANALYSIS 4.1 NATURE OF THE LOADS Figures 3-1 to 3-4 show the schematic layouts of the 10. Accumulator lines Loops 1, 2, 3 and 4 and also identify the weld locations by node points. The stresses due to axial loads and moments were calculated by the following equation:

F M a=-+ (4-1)

A Y

where, G = Stress F = Axial Load M = Moment A = Metal Cross-Sectional Area z = Section Modulus The moment for the desired loading combinations were calculated by the following equation:

M = VM2+MI+Ml (4-2)

where, M = Moment For Required Loading Mx = Torsional Moment My = Y Component of Bending Moment Mz = Z Component of Bending Moment The axial load and moments for crack stability analysis and leak rate predictions are computed by the methods to be explained in Sections 4.2 and 4.3.

February 2003 Loads for Loads Mechanics Analysis Fracture Mechanics for Fracture Analysis February 2003

4-2 4.2 LOADS FOR CRACK STABILITY ANALYSIS In accordance with Standard Review Plan 3.6.3 the absolute sum of loading components can be applied which results in higher magnitude of combined loads. If crack stability is demonstrated using these loads, the LBB margin on loads can be reduced from 42 to 1.0. The faulted loads for the crack stability analysis were calculated by the absolute sum method as follows:

F = IFowi + IFTHI + IFpI + IFSSEI (4-3)

MX= IMxDwI + IMxTHI + IMxssEI (4-4)

My= IMyDwI + IMYTHI + IMyssEl (4-5)

Mz= IMzDwI + IMZTHI + IMzssEI (4-6) where DW = Deadweight TH = Normal Thermal Expansion Load P = Load Due To Internal Pressure SSE = Safe Shutdown Earthquake Loading Including Seismic Anchor Motion 4.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for the leak rate predictions were calculated by the algebraic sum method as follows:

F = FoW + FTH + FP (4-7)

MX = MX DW + MX TH (4-8)

MY = MY DW + MY TH (4-9)

MZ= MZ DW + MZTH (4-10)

The parameters and subscripts are the same as those explained in Sections 4.1 and 4.2.

4.4

SUMMARY

OF LOADS AND GEOMETRY FOR THE 10" ACCUMULATOR LINES The load combinations were evaluated at the various weld locations. Normal loads were determined using the algebraic sum method whereas the faulted loads were combined using the absolute sum method. The normal operating loadings for the 10' Accumulator lines are Loads for Fracture Mechanics Analysis February 2003

4-3 Pressure (P), Deadweight (DW) and Normal Operating Thermal Expansion (TH) loads. The faulted loadings consist of Normal Operating loads plus Safe Shutdown Earthquake (SSE) loads including the Seismic Anchor Motion. The effects of two normal operating temperature scenarios (Case A: 700 F and Case B: 480F) for the portion of the piping between the isolation valves HV-8808 and the accumulator tanks are considered in the LBB evaluation.

Tables 4-1, 4-4, 4-7 and 4-10 show the piping geometry and normal operating condition for the 10" Accumulator lines Loops 1, 2, 3 and 4 respectively at the weld locations. The minimum pipe wall thickness at the weld counterbore is used in the analysis. The normal loads and stresses for the 10" Accumulator lines Loops 1, 2, 3 and 4 at the weld locations are tabulated in Tables 4-2, 4-5, 4-8 and 4-11 respectively, while Tables 4-3, 4-6, 4-9 and 4-12 are for the faulted loads and stresses.

4.5 GOVERNING LOCATIONS FOR THE ACCUMULATOR LINES The welds at the 10' Accumulator lines for the Callaway Nuclear Power Plant are fabricated using the GTAW and SMAW combination or GTAW. The governing locations were established on the basis of the pipe schedules, material type, operating temperature, operating pressure, and the highest faulted stresses at the welds. All the Accumulator lines for the Callaway Nuclear Power Plant were investigated and the governing locations were identified and found to be located in Loops 2 and 3. These governing locations enveloped the 10" Accumulator lines loops 1, 2, 3 and 4 for the Callaway Nuclear Power Plant in the Leak-Before-Break analyses.

Figures 4-1 and 4-2 show the schematic layout of Accumulator lines Loops 2 and 3 for the Callaway Nuclear Power Plant and also identify the governing weld locations.

The governing weld locations enveloping 10"Accumulator lines Loops 1, 2, 3 and 4 are found to be located in Loops 2 and 3 and are shown below:

Node 3020 (Accumulator Loop 2)

Node 3120 (Accumulator Loop 2)

Node 3295 (Accumulator Loop 3)

Analysis Mechanics Analysis Fracture Mechanics Febwary 2003 Loads for Fracture Loads for February 2003

4-4 Table 4-1: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10" Accumulator Line Loop 1 Minimum Normal Operating Weld Outer Wall Location Material Type Diameter Thickness Pressure Temperature Node (in) (in) (Psig) ( 0F) 3020 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3035 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3040 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3045 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3050 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3065 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3080 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3085 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3120 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3125 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3140 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3160 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3170 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3175 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3200 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3215 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3240 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3245 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3290 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 Loads for Fracture Mechanics Analysis L oa ds for FractureMech anics An alysis Feb ruary 20 03

4-5 Table 4-2a: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 1 (Case A:

700F)

Weld Axial Moment Total Location Axial Force Moment Stress Stress Stress Node (Ibs) (ibs) (psi) (psi) (psi) 3020 135569 428875 4890 6794 11684 3035 139602 452190 5035 7163 12199 3040 139682 476672 5038 7551 12589 3045 144010 457585 5194 7249 12443 3050 144010 420456 5194. 6660 11855 3065 144010 365628 5194 5792 10986 3080 144010 372476 5194 5900 11095 3085 146085 318952 5269 5052 10322 3120 144769 404706 5222 6411 11633 3125 136646 512165 4929 8113 13042 3140 136646 241046 4929 3818 8747 3160 143777 437760 5186 6934 12121 3170 143777 183789 5186 2911 8097 3175 145378 158004 5244 2503 7747 3200 145411 116594 5245 1847 7092 3215 45494 53795 1641 852 2493 3240 45527 63335 1642 1003 2645 3245 46302 73682 1670 1167 2837 3290 42932 92867 1549 1471 3020 February 2003 Loads for Mechanics Analysis Fracture Mechanics for Fracture Analysis February 2003

4-6 Table 4-2b: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 1 (Case B:

48-F)

Weld Axa oc oet Axial Moment Total Location Axial Force Moment Stress Stress Stress Node (Ibs) (in(Ibs) (psi) ps (psi) 3020 135590 426215 4891 6752 11642 3035 139600 449585 5035 7122 12157 3040 139680 474033 5038 7509 12547 3045 143988 455106 5194 7209 12403 3050 143988 418145 5194. 6624 11817 3065 143988 363949 5194 5765 10959 3080 143988 370896 5194 5875 11069 3085 146087 318129 5269 5039 10309 3120 144771 403926 5222 6399 11620 3125 136678 510733 4930 8090 13020 3140 136678 239255 4930 3790 8720 3160 143899 438829 5190 6951 12142 3170 143899 179683 5190 2846 8037 3175 145539 152737 5250 2419 7669 3200 145573 112550 5251 1783 7034 3215 45656 50952 1647 807 2454 3240 45689 67256 1648 1065 2713 3245 46408 76921 1674 1218 2892 3290 44167 80885 1593 1281 2874 Loads for Fracture Mechanics Analysis February 2003

4-7 Table 4-3a: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 1 (Case A:

700F)

Weld Axial Moment Total Location Axial Force Moment Stress Stress Stress Node (Ibs) (in-lbs) (p si) (psi) (psi) 3020 154100 561781 5558 8899 14458 3035 150151 550347 5416 8718 14134 3040 150095 564475 5414 8942 14356 3045 148512 535092 5357 8476 13833 3050 148344 495706 5351 7852 13203 3065 146994 426654 5302 6759 12061 3080 146945 443200 5300 7021 12321 3085 151104 399270 5450 6325 11775 3120 153162 500729 5525 7932 13457 3125 151994 629038 5482 9964 15447 3140 152216 359467 5490 5694 11185 3160 146449 531704 5282 8423 13705 3170 146399 292397 5281 4632 9912 3175 147294 266790 5313 4226 9539 3200 147236 203688 5311 3227 8537 3215 47514 134359 1714 2128 3842 3240 47791 154985 1724 2455 4179 3245 48276 156900 1741 2485 4227 3290 48263 186629 1741 2956 4697 February 2003 Loads for Fracture Mechanics Analysis Fracture Mechanics Analysis February 2003

4-8 Table 4-3b: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 1 (Case B:

0 48 F)

Weld Axial Force Moment Axial Moment Total Location (Is I-b) Stress Stress Stress Node (Ibs) (in-Ibs) (psi) (psi) (psi) 3020 154079 559025 5558 8855 14413 3035 150153 547863 5416 8679 14095 3040 150097 561761 5414 8899 14313 3045 148534 532769 5358 8439 13797 3050 148366 493590 5352 7819 13170 3065 147016 425056 5303 6733 12036 3080 146967 441695 5301 6997 12298 3085 151106 398526 5450 6313 11763 3120 153164 499890 5525 7919 13443 3125 151962 627795 5481 9945 15426 3140 152184 357891 5489 5669 11159 3160 146327 532758 5278 8439 13717 3170 146277 288274 5276 4566 9843 3175 147455 261460 5319 4142 9460 3200 147398 199560 5317 3161 8478 3215 47676 131553 1720 2084 3804 3240 47953 158138 1730 2505 4235 3245 48382 160144 1745 2537 4282 3290 47166 175010 1701 2772 4474 Analysis Mechanics Analysis Fracture Mechanics Loads for Fracture February 2003 February 2003

4-9 Table 4-4: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10' Accumulator Line Loop 2 Minimum Normal Operating Weld Outer Wall Location Material Type Diameter Thickness Pressure Temperature Node (in) (in) (psig) (OF) 3020 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 558

____ SA403 WP304 3035 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 558 3040___ SA403 WP304 t0.750 086 2558 3040 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 558 SA403 WP304 1708255 3045 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70 3045 SA376SA403 WP304 10.750 0.896 2285 55 3050 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 558 SA403 WP304TP304 or 1.08627 3065 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70 SA403 WP304 3075 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70 3180___ SA403 WP304 1.0 .9287 3085 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70 SA403 WP304 .

3100 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70 SA403 WP304TP304 o 170800 3120 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70

______ SA403 WP304_____ ___

310 SA376/SA358/SA312 TP304 or 1.5 .9 257 3150 ~SA403 WP304 1.5 .9 257 3170 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70 SA403 WP304 ______________

310 SA376/SA358/SA312 TP304 or 1.5 .9 257 3180 ~SA403 WP304 1.5 .9 257 3185 SA376/SA358/SA312 TP304 or 10.750 0.896 2285 70

_______ SA403 W P304__ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _

325 SA376/SA358/SA312 TP304 or 1.5 .9 257 3205 ~SA403 WP304 1.5 .9 257 3220 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70 SA403 WP304 _ _ _ _ _ _ _ _ _ _ _ _ _ _

3235 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70

_______ SA403 W P304 _ _ _ _ _ _ _ _ _

3240 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70

_______ SA403 W P304 _ _ __ _ _ _ I__ _ I__ __ I__

Mechanics Analysis forFracture Loads February 2003~~~~~~~~~~~~~~~~~~~~~~

Loads for Fracture Mechanics Analysis February 2003

4-10 Table 4-4: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10,, Accumulator Line Loop 2 Minimum Normal Operating Weld Outer Wall Location Material Type Diameter Thickness Pressure Temperature Node (in) (in) (psig) (OF) 3265 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70 SA403 WP304 10.750 0.896 700_70 3270 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70 SA403 WP304 1070 086 707 3275 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70 SA403 WP304 _____

3280 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70

_______ SA403 WP304 ____

3300 SA376/SA358/SA312 TP304 or 10.750 0.896 700 70

_______ SA403 W P304__ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _

February 2003 Loads for Loads Mechanics Analysis Fracture Mechanics for Fracture Analysis February 2003

4-11 Table 4-5a: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 2 (Case A: 700F)

Weld Axial Force Moment Axial Moment Total Stress Location Stress Stress Node (Ibs) (in-lbs) (psi) (psi) (psi) 3020 135921 457121 4903 7241 12144 3035 140738 489594 5076 7756 12832 3040 140817 517471 5079 8197 13276 3045 143286 503550 5168 7977 13145 3050 143286 463535 5168 7343 12511 3065 143286 393868 5168 6239 11408 3075 143286 397994 5168 6305 11473 3085 144950 327660 5228 5190 10419 3100 143954 447673 5192 7092 12284 3120 136012 552719 4906 8756 13662 3150 136012 142016 4906 2250 7156 3170 143735 304783 5185 4828 10013 3180 143735 131024 5185 2076 7260 3185 144670 132956 5218 2106 7324 3205 144764 80918 5222 1282 6503 3220 44847 38689 1618 613 2231 3235 44961 12451 1622 197 1819 3240 45454 18369 1640 291 1931 3265 45454 76657 1640 1214 2854 3270 43271 56155 1561 890 2450 3275 42315 77640 1526 1230 2756 3280 45454 122442 1640 1940 3579 3300 45454 88435 1640 1401 3040 M c a is dFractureeMechanics fo:rcu n l ssF b ay2 0 LoadsLofor Analysis February 2003

4-12 Table 4-5b: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 2 (Case B: 480F)

Weld Axial Force Moment Axial Moment Total Stress Location (ls i-b) Stress Stress Node (Ibs) (in-Ibs) (psi) (psi) (psi) 3020 135905 455720 4902 7219 12121 3035 140728 488417 5076 7737 12813 3040 140807 516466 5079 8181 13260 3045 143275 503066 5168 7969 13137 3050 143275 463193 5168 7337 12505 3065 143275 394618 5168 6251 11419 3075 143275 399012 5168 6321 11489 3085 144960 329472 5229 5219 10448 3100 143964 449375 5193 7118 12311 3120 136000 554473 4906 8783 13689 3150 136000 144348 4906 2287 7192 3170 143767 308671 5186 4890 10075 3180 143767 131206 5186 2078 7264 3185 144751 130836 5221 2073 7294 3205 144847 77069 5225 1221 6446 3220 44930 35367 1621 613 2231 3235 45047 13152 1625 208 1833 3240 45541 19247 1643 305 1948 3265 45541 78778 1643 1248 2891 3270 43304 55575 1562 880 2442 3275 42348 77463 1528 1227 2755 3280 45541 123668 1643 1959 3602 3300 45541 88544 1643 1403 1 3045 Mechanics Analysis February 2003 Loads for Fracture Mechanics for Fracture Analysis February 2003

4-13 Table 4-6a: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 2 (Case A: 700F)

Weld Axial Force Moment Axial Moment Total Stress Location (ls i-b) Stress Stress(pi Node (in-!bs (Ibs) ) (psi) 3020 155514 682504 5609 10811 16421 3035 149777 674575 5403 10686 16088 3040 149714 679570 5400 10765 16165 3045 148133 636724 5343 10086 15429 3050 148036 596112 5340 9443 14783 3065 147651 530372 5326 8402 13727 3075 147659 542645 5326 8596 13922 3085 149787 478419 5403 7579 12981 3100 150929 604084 5444 9569 15013 3120 155118 779295 5595 12345 17940 3150 154825 316493 5585 5014 10598 3170 148357 421113 5351 6671 12022 3180 148266 343333 5348 5439 10787 3185 149597 371173 5396 5880 11276 3205 149382 366060 5388 5799 11187 3220 48383 349067 1745 5530 7275 3235 48355 298075 1744 4722 6466 3240 47576 293722 1716 4653 6369 3265 49189 412327 1774 6532 8306 3270 47878 373345 1727 5914 7641 3275 48796 264504 1760 4190 5950 3280 50133 340956 1808 5401 7209 3300 50257 277652 1813 4398 6211 Mechanics Analysis February 2003 Loads for Fracture Mechanics Analysis February 2003

4-14 Table 4-6b: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 2 (Case B: 480F)

Weld Axial Force Moment Axial Moment Total Stress Location Stress Stress Node (Ibs) (in-lbs) (psi) (psi) (psi) 3020 155530 681213 5610 10791 16401 3035 149787 673505 5403 10669 16072 3040 149724 678635 5401 10750 16151 3045 148144 636226 5344 10078 15422 3050 148047 595730 5340 9437 14777 3065 147662 531270 5326 8416 13742 3075 147670 543738 5327 8613 13940 3085 149797 480561 5403 7612 13016 3100 150939 605883 5444 9598 15042 3120 155130 781229 5596 12375 17971 3150 154837 318330 5585 5043 10628 3170 148325 424703 5350 6728 12078 3180 148234 343293 5347 5438 10785 3185 149678 369295 5399 5850 11249 3205 149465 362300 5391 5739 11130 3220 48466 344053 1748 5450 7198 3235 48441 302899 1747 4798 6545 3240 47663 296151 1719 4691 6411 3265 49276 414028 1777 6559 8336 3270 47911 374634 1728 5935 7663 3275 48829 264082 1761 4183 5945 3280 50220 341662 1811 5412 7224 3300 50344 277370 1816 4394 6210 Analysis for Fr Mechancs Loads cture ebruary 200 Loads for Fracture Mechanics Analysis February 2003

4-15 Table 4-7: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10" Accumulator Line Loop 3 Weld Outer Minimum Normal Operating Location Material Type Diameter Wall Pressure Temperature Node (in) Thickness (psig) (OF)

(in) 3020 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3035 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3040 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3045 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3050 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3065 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3077 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3090 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3120 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3125 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3150 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3170 SA3581SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3190 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3215 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3235 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3240 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3270 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3275 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3295 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3300 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3310 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3315 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3330 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3335 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3345 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 eray20 Lod o rcueMcaisAayi Loads for Fracture Mechanics Analysis February 2003

4-16 Table 4-8a: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 3 (Case A:

700F)

Weld Axial Force Moment Axial Moment Total Stress Location Qs i-b) Stress Stress(pi Node (Ibs) (in-(bs) (psi) (psi) (psi) 3020 138044 500776 4979 7933 12912 3035 144561 478496 5214 7580 12794 3040 144640 473619 5217 7503 12720 3045 137711 402925 4967 6383 11350 3050 137711 232099 4967 3677 8644 3065 137711 392808 4967 6222 11190 3077 137711 528543 4967 8373 13340 3090 140117 572772 5054 9073 14127 3120 139123 555303 5018 8796 13815 3125 148315 516518 5350 8182 13532 3150 148315 490085 5350 7763 13113 3170 148291 329945 5349 5227 10575 3190 148300 233326 5349 3696 9045 3215 48383 128699 1745 2039 3784 3235 51886 338696 1872 5365 7237 3240 53502 356284 1930 5644 7574 3270 53502 338044 1930 5355 7285 3275 52427 323487 1891 5124 7015 3295 51647 357000 1863 5655 7518 3300 53502 386620 1930 6124 8054 3310 53502 308964 1930 4894 6824 3315 50535 234591 1823 3716 5539 3330 50535 210144 1823 3329 5152 3335 44259 103312 1596 1637 3233 3345 44259 74449 1596 1179 2776 February 2003 Loads for Mechanics Analysis Fracture Mechanics for Fracture Analysis February 2003

4-17 Table 4-8b: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 3 (Case B:

48 0F)

Weld Axial Force Moment Axial Moment Total Stress Location Stress Stress Node (Ibs) (in-(bs)(psi) 3020 138043 500624 4979 7930 12910 3035 144553 478405 5214 7578 12792 3040 144632 473552 5217 7501 12718 3045 137719 402883 4968 6382 11350 3050 137719 231490 4968 3667 8635 3065 137719 392016 4968 6210 11177 3077 137719 527639 4968 8358 13326 3090 140125 571933 5054 9060 14114 3120 139131 555047 5018 8792 13811 3125 148323 516501 5350 8182 13532 3150 148323 490120 5350 7764 13114 3170 148296 330085 5349 5229 10578 3190 148305 233780 5349 3703 9053 3215 48388 129634 1745 2054 3799 3235 51899 337874 1872 5352 7224 3240 53505 355411 1930 5630 7560 3270 53505 338396 1930 5360 7290 3275 52409 323401 1890 5123 7013 3295 51629 356319 1862 5644 7507 3300 53505 386318 1930 6120 8050 3310 53505 308777 1930 4891 6821 3315 50499 236197 1822 3742 5563 3330 50499 216364 1822 3427 5249 3335 44491 107001 1605 1695 3300 3345 44491 72831 1605 1154 2759 Mechanics Analysis February 2003 for Fracture Loads for Fracture Mechanics Analysis February 2003

4-18 Table 4-9a: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 3 (Case A:

70 0F)

Weld Axial Force Moment Axial Moment Stress Total Stress Location Stress Node (Ibs) (in-lbs) (psi) (psi) (Ps!)

3020 152011 689048 5483 10915 16398 3035 148507 603444 5357 9559 14916 3040 148404 576812 5353 9137 14490 3045 152648 471370 5506 7467 12973 3050 151928 430425 5480 6818 12298 3065 150978 568765 5446 9010 14456 3077 150934 619565 5444 9814 15259 3090 149412 620550 5389 9830 15219 3120 150556 605417 5431 9590 15021 3125 154154 556474 5560 8815 14375 3150 154176 532586 5561 8437 13998 3170 154225 400158 5563 6339 11902 3190 154249 296064 5564 4690 10254 3215 54544 202083 1967 3201 5169 3235 53279 397271 1922 6293 8215 3240 54293 413091 1958 6544 8502 3270 54481 396246 1965 6277 8242 3275 55068 394666 1986 6252 8238 3295 54094 497330 1951 7878 9829 3300 54953 487298 1982 7719 9701 3310 55003 418735 1984 6633 8617 3315 52470 367349 1893 5819 7712 3330 52523 306748 1895 4859 6754 3335 47376 228691 1709 3623 5332 3345 47275 243639 1705 3859 5565 Analysis Mechanics Analysis February 2003 Loads for Fracture Mechanics for Fracture February 2003

4-19 Table 4-9b: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 3 (Case B:

48 0F)

Weld Axial Force Moment Axial Moment Stress Total Stress Location Stress Node (Ibs) (in-Ibs) (psi) (psi) (psi) 3020 152012 688883 5483 10912 16396 3035 148499 603296 5356 9557 14913 3040 148396 576718 5353 9136 14488 3045 152640 471324 5506 7466 12972 3050 151920 429820 5480 6809 12289 3065 150970 567968 5446 8997 14443 3077 150926 618650 5444 9800 15244 3090 149404 619684 5389 9816 15205 3120 150548 605128 5430 9586 15016 3125 154162 556463 5561 8815 14375 3150 154184 532625 5561 8437 13999 3170 154230 400243 5563 6340 11903 3190 154254 296319 5564 4694 10258 3215 54549 202101 1968 3201 5169 3235 53292 396447 1922 6280 8202 3240 54296 412357 1958 6532 8491 3270 54484 396504 1965 6281 8246 3275 55050 394566 1986 6250 8236 3295 54076 496316 1951 7862 9813 3300 54956 485647 1982 7693 9675 3310 55006 416684 1984 6601 8585 3315 52434 368446 1891 5836 7728 3330 52487 311920 1893 4941 6834 3335 47608 224299 1717 3553 5270 3345 47507 236261 1714 3743 5456 February 2003 Loads for Mechanics Analysis Fracture Mechanics for Fracture Analysis February 2003

4-20 Table 4-10: Summary of Callaway Nuclear Power Plant Piping Geometry and Normal Operating Condition for 10 Accumulator Line Loop 4 Weld Outer Minimum Normal Operating Location Material Type Diameter Wall Pressure Temperature Node (in) Thickness (psig) (OF) 3030 SA3581SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3050 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3060 SA3581SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3070 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3080 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 558 3110 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3130 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3140 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3170 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3190 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3200 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 2285 70 3230 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3250 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3260 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3280 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3290 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3300 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3310 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3330 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3340 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3360 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3400 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3410 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 3450 SA358/SA312 TP304 or SA403 WP304 10.750 0.896 700 70 Mechanics Analysis February 2003 Loads for Fracture Mechanics for Fracture Analysis February 2003

4-21 Table 4-1la: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 4 (Case A:

700F)

Weld Axial Force Moment Axial Moment Total Stress Location (ls i-b) Stress Stress(pi Node (Ibs) (inlbs) (psi) (psi) 3030 137555 233713 4962 3702 8664 3050 141865 190923 5117 3024 8141 3060 141945 190758 5120 3022 8142 3070 135913 243581 4902 3859 8761 3080 135913 352384 4902 5582 10484 3110 135913 398865 4902 6318 11221 3130 135913 411503 4902 6519 11421 3140 142490 313707 5140 4969 10109 3170 142490 72191 5140 1144 6283 3190 145574 215910 5251 3420 8671 3200 145574 187986 5251 2978 8229 3230 45657 92736 1647 1469 3116 3250 45657 110803 1647 1755 3402 3260 49055 194923 1769 3088 4857 3280 49055 131755 1769 2087 3857 3290 48693 148654 1756 2355 4111 3300 48693 298476 1756 4728 6484 3310 48713 300320 1757 4757 6514 3330 47822 194359 1725 3079 4804 3340 48349 215916 1744 3420 5164 3360 48693 202913 1756 3214 4971 3400 40926 344303 1476 5454 6930 3410 44353 288289 1600 4567 6167 3450 44353 101359 1600 1606 3205 d cu eMechanics Fr fo Fracture Ayi M c a is Analysis e r a y2 0 LoadsLo for February 2003

4-22 Table 4-lib: Summary of Callaway Nuclear Power Plant Normal Loads and Stresses for 10" Accumulator Line Loop 4 (Case B:

480F)

Weld Axial Force Moment Axial Moment Total Stress Location Qb) Q-b) Stress Stress Node (Ibs(psi) (psi) (psi) 3030 137548 233598 4961 3700 8662 3050 141865 190693 5117 3021 8138 3060 141945 190486 5120 3017 8137 3070 135906 243457 4902 3857 8759 3080 135906 352948 4902 5591 10493 3110 135906 399657 4902 6331 11233 3130 135906 412356 4902 6532 11434 3140 142483 314568 5139 4983 10122 3170 142483 72727 5139 1152 6291 3190 145567 216329 5251 3427 8677 3200 145567 188330 5251 2983 8234 3230 45650 92831 1647 1471 3117 3250 45650 110949 1647 1758 3404 3260 49062 195252 1770 3093 4863 3280 49062 132392 1770 2097 3867 3290 48694 149263 1756 2364 4121 3300 48694 299183 1756 4739 6496 3310 48719 301146 1757 4770 6528 3330 47828 195320 1725 3094 4819 3340 48360 216715 1744 3433 5177 3360 48694 203709 1756 3227 4983 3400 40819 342010 1472 5418 6890 3410 44535 281806 1606 4464 6070 3450 44535 121082 1606 1918 3524 Loads for Fr Mechancs cture Analysis ebruary 200 Loads for Fracture Mechanics Analysis February 2003

4-23 Table 4-12a: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 4 (Case A:

700F)

Weld Axial Force Moment Axial Moment Stress Total Stress Location Stress Node (Ibs) (in-Ibs) (psi) (psi) (psi) 3030 153191 399047 5526 6321 11847 3050 147270 284358 5312 4504 9817 3060 147174 268941 5309 4260 9569 3070 156367 316425 5640 5012 10653 3080 155760 465984 5618 7382 13000 3110 154977 533440 5590 8450 14040 3130 154965 542948 5590 8601 14190 3140 149176 418049 5381 6622 12003 3170 148902 217947 5371 3452 8823 3190 147963 355128 5337 5626 10963 3200 147927 304034 5336 4816 10152 3230 47407 192438 1710 3048 4758 3250 47302 205145 1706 3250 4956 3260 49584 281010 1789 4451 6240 3280 49758 202161 1795 3202 4997 3290 49801 224149 1796 3551 5347 3300 49811 417677 1797 6616 8413 3310 50814 418794 1833 6634 8467 3330 51657 307971 1863 4879 6742 3340 51858 327125 1871 5182 7052 3360 50442 286211 1819 4534 6353 3400 49285 481617 1778 7629 9407 3410 46590 407948 1681 6462 8143 3450 46420 205039 1674 3248 4922 Loads for Fracture Mechanics Analysis February 2003

4-24 Table 4-12b: Summary of Callaway Nuclear Power Plant Faulted Loads and Stresses for 10" Accumulator Line Loop 4 (Case B:

480F)

Weld Axial Force Moment Axial Moment Stress Total Stress Node (Ibs) (in-lbs) (psi) (psi) (psi) 3030 153198 398961 5526 6320 11846 3050 147270 284168 5312 4501 9814 3060 147174 268670 5309 4256 9565 3070 156374 316305 5640 5011 10651 3080 155767 466544 5619 7390 13009 3110 154984 534237 5590 8463 14053 3130 154972 543804 5590 8614 14204 3140 149183 418898 5381 6636 12017 3170 148909 218347 5371 3459 8830 3190 147956 355546 5337 5632 10969 3200 147920 304292 5336 4820 10156 3230 47400 192704 1710 3053 4762 3250 47295 205214 1706 3251 4957 3260 49591 281348 1789 4457 6246 3280 49765 202668 1795 3210 5005 3290 49802 224814 1796 3561 5358 3300 49812 418435 1797 6628 8425 3310 50820 419659 1833 6648 8481 3330 51663 308854 1864 4892 6756 3340 51869 327742 1871 5192 7063 3360 50443 286771 1820 4543 6362 3400 49392 479391 1782 7594 9376 3410 46772 401713 1687 6363 8051 3450 46602 220517 1681 3493 5174 Febwary 2003 Loads for Mechanics Analysis Fracture Mechanics for Fracture Analysis February 2003

4-25 3085 3150 3180 3185 3045 3120 3170 3040 Critical Weld Location 3205 3035 Critical Weld Location 3220 z;;~ a) .53235 COLD LEG LOOP 2 3240 3265 3270 4 3275 3280 HV-8808B Figure 4-1 Governing Weld Locations for 10" Accumulator Line Loop 2 Loads for Fracture Mechanics Analysis February 2003

4-26 3050

'COLD LEG LOOP 3 3125 3170 3240 3270 3235 3275 3330 3335 3295 3315 Critical Weld Location 3300 3310 HV-8808C Figure 4-2 Governing Weld Location for 10" Accumulator Line Loop 3 Analysis Febwary 2003 Fracture Mechanics Loads for Fracture Mechanics Analysis February 2003

5-1 5 FRACTURE MECHANICS EVALUATION 5.1 GLOBAL FAILURE MECHANISM Determination of the conditions which lead to failure in stainless steel should be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. One method for predicting the failure of ductile material is the [ ]atce method based on traditional plastic limit load concepts, but accounting for [ ace and taking into account the presence of a flaw. The flawed component is predicted to fail when the remaining net section reaches a stress level at which a plastic hinge is formed. The stress level at which this occurs is termed as the flow stress. [

]a.C~e This methodology has been shown to be applicable to ductile piping through a large number of experiments and is used here to predict the critical flaw size in the Accumulator lines. The failure criterion has been obtained by requiring equilibrium of the section containing the flaw (Figure 5-1) when loads are applied. The detailed development is provided in Appendix A for a through-wall circumferential flaw in a pipe section with internal pressure, axial force, and imposed bending moments. The limit moment for such a pipe is given by:

[ ]a,c,e (5-1) where:

]a~ce (5-2)

Evaluation Mechanics Evaluation February 2003 Fracture Mechanics Fracture February 2003

5-2 The analytical model described above accurately accounts for the internal pressure as well as imposed axial force as they affect the limit moment. Good agreement was found between the analytical predictions and the experimental results (Reference 5-1). The Flaw stability evaluations, using this analytical model, are presented in Section 5.3.

5.2 LEAK RATE PREDICTIONS Fracture mechanics analysis shows that postulated through-wall cracks in the 10. Accumulator lines would remain stable and would not cause a gross failure of this component. However, if such a through-wall crack did exist, it would be desirable to detect the leakage such that the Plant could be brought to a safe shutdown condition. The purpose of this section is to discuss the method, which will be used to predict the flow through such a postulated crack and present the leak rate calculation results for through-wall circumferential cracks.

5.2.1 General Considerations The flow of hot pressurized water through an opening to a lower backpressure (causing choking) is taken into account. For long channels where the ratio of the channel length, L, to hydraulic diameter, DH, (L/DH) is greater than [ ]a,c,e, both [ ]ace must be considered. In this situation the flow can be described as being single-phase through the channel until the local pressure equals the saturation pressure of the fluid. At this point, the flow begins to flash and choking occurs. Pressure losses due to momentum changes will dominate for [ ]a,c,e. However, for large UDH values, the friction pressure drop will become important and must be considered along with the momentum losses due to flashing.

5.2.2 Calculation Method In using the l jace The flow rate through a crack was calculated in the following manner. Figure 5-2 from Reference 5-2 was used to estimate the critical pressure, Pc, for the primary loop enthalpy condition and an assumed flow. Once Pc was found for a given mass flow, the l

]ace was found from Figure 5-3 taken from Reference 5-2. For all cases considered, since [ laCe Therefore, this method will yield the two-phase pressure drop due to momentum effects (AP 2 ,) as illustrated in Figure 5-4. Now using the assumed flow rate, G.the frictional pressure drop can be calculated using Aps =[ ]a,c,e (5-3)

Fracture Mechanics Evaluation February 2003

5-3

'where the friction factor f was determined using the [ ]a C.e The crack relative roughness, e, was obtained from fatigue crack data on stainless steel samples. The relative roughness value used in these calculations was [ ]ac,e RMS.

The frictional pressure drop using Equation 5-3 was then calculated for the assumed flow and added to the [ lace to obtain the total pressure drop from the system under consideration to the atmosphere. Thus, Absolute Pressure - 14.7 = [ ]ace (5-4) for a given assumed flow G. If the right-hand side of Equation 5-4 does not agree with the pressure difference between the piping under consideration and the atmosphere, then the procedure is repeated until Equation 5-4 is satisfied to within an acceptable tolerance and this results in the flow value through the crack.

For the locations at the lower temperatures, the leak rate is calculated by using the simple orifice type flow formula given by [

(5-5) jaxce 5.2.3 Leak Rate Calculations Leak rate calculations were performed as a function of postulated through-wall crack length for the governing locations previously identified. The crack opening area was estimated using the method of Reference 5-4 and the leak rates were calculated using the calculation methods described above. The leak rates were calculated using the normal operating loads at the governing locations identified in Section 4. Average yield strength properties from Table 3-2 were used in the leak rate calculations. The crack lengths yielding a leak rate of 10 gpm (10 times the leak detection capability of 1.0 gpm) for the governing weld locations in the 10" Accumulator lines at the Callaway Nuclear Power Plant are shown in Table 5-1.

Mechanics Evaluation Fracture February 2003~~~~~~~~~~~~~~

Fracture Mechanics Evaluation February 2003

5-4 The Callaway Nuclear Power Plant has an RCS pressure boundary leak detection system which is consistent with the guidelines of Regulatory Guide 1.45 for detecting leakage of 1 gpm in one hour.

5.3 STABILITY EVALUATION A typical segment of the pipe under maximum loads of axial force F and moment M is schematically illustrated in Figure 5-5. In order to calculate the critical flaw size, plots of the limit moment versus crack length are generated as shown in Figures 5-6 to 5-8. The critical flaw size corresponds to the intersection of this curve and the maximum load line. The critical flaw size is calculated using the lower bound base metal tensile properties tabulated in Table 3-2.

The welds at the governing locations are GTAW/SMAW combination or GTAW. Therefore, in order to envelop all the weld process types, the Z' factor correction for SMAW was conservatively applied (Reference 5-5) as follows:

Z = 1.15 [1 + 0.013 (OD - 4)] (for SMAW) (5-6) where OD is the outer diameter in inches. Substituting OD = 10.75 inches, the Z factor was calculated for the 10. Accumulator lines to be 1.25 for SMAW. The "Z" correction factor for GTAW is 1.0. The applied loads were conservatively increased by the Z factors for SMAW and the plots of limit load versus crack length were generated as shown in Figures 5-6 to 5-8 for the critical locations. Table 5-2 shows the summary of critical flaw sizes.

5.4 REFERENCES

5-1 Kanninen, M. F. et al., Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks EPRI NP-192, September 1976.

5-2 [

]a,c,e 5-3 Crane, D.P., Handbook of Hydraulic Resistance Coefficient."

5-4 Tada, H., The Effects of Shell Corrections on Stress Intensity Factors and the Crack Opening Area of Circumferential and a Longitudinal Through-Crack in a Pipe,"

Section 11-1, NUREG/CR-3464, September 1983.

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

FractureMechaics Evaluatio February 200 Fracture Mechanics Evaluation February 2003

5-5 Table 5-1: Leakage Flaw Sizes Temperature Crack Length (in.)

Node Point (F) (for 10 gpm leakage) 3020 558 3.87 3120 70 3.70 3295 70 6.20 Table 5-2: Summary of Critical Flaw Sizes Temperature Critical Node Point (O Flaw Size (in) 3020 558 11.39 3120 70 12.54 3295 70 16.36 Fracture Mechanics Evaluation February 2003

5-6 (f

Figure 5-1 Fully Plastic Stress Distribution February 2003 Fracture Mechanics Evaluation Mechanics Evaluation February 2003

5-7 a,c,e I-U 0

LU aI STAGNATION ENTHALPY (102 Btu/ib)

Figure 5-2 Analytical Predications of Critical Flow Rates of Steam-Water Mixtures Fracture Mechanics Evaluation February 2003

5-8

a. ce a.

Uj a.

2j 4C

'U F.

'U LENGTH/DIAMETER RATIO WLiD)

Figure 5-3 [ Jace Pressure Ratio as a Function of L/D Fracture Mechanics Evaluation February 2003

5-9 a,c,e

_ S_

a,c,e

[

Figure 5-4 Idealized Pressure Drop Profile through a Postulated Crack

5-10

- r7 a0 01

~~2

\ief Figure 5-5 Loads acting on the Model at the Governing Locations Evlato Feru. 200 Fratr Mehnc Fracture Mechanics Evaluation February 2003

5-11 a,c,e OD = 10.75 in ao = 20.827 ksi F = 155.514 kips t = 0.896 in a,= 68.134 ksi M = 682.504 in-kips SA376/SA358/SA312 TP304 or SA403 WP304 with SMAW weld Figure 5-6 Critical Flaw Size Prediction for Node 3020 February 2003 Fracture Mechanics Evaluation Fracture Mechanics Evaluation February 2003

5-12 a,c,e OD = 10.75 in ay = 33.200 ksi F = 155.130 kips t = 0.896 in o; = 80.600 ksi M = 781.229 in-kips SA376/SA358/SA312 TP304 or SA403 WP304 with SMAW weld Figure 5-7 Critical Flaw Size Prediction for Node 3120 vlaio era20 Fracture Mehnc Frctr Mechanics Evaluation February 2003

5-13 a,c,e OD = 10.75 in cy = 33.200 ksi F = 54.094 kips t = 0.896 in cr = 80.600 ksi M = 497.330 in-kips SA376/SA358/SA312 TP304 or SA403 WP304 with SMAW weld Figure 5-8 Critical Flaw Size Prediction for Node 3295 Evaluation Mechanics Evaluation Febmary 2003 Fracture Mechanics Fracture February 2003

6-1 6 ASSESSMENT OF FATIGUE CRACK GROWTH

6.1 INTRODUCTION

The fatigue crack growth of the Callaway Nuclear Power Plant Accumulator lines was determined by comparison with a generic fatigue crack growth analysis of a similar piping system. The details of the generic fatigue crack growth analysis are presented below. By comparing all parameters critical to the fatigue crack growth analysis between Callaway and the generic analysis, it was concluded that the generic analysis would adequately cover the fatigue crack growth of the Callaway Nuclear Power Plant Accumulator lines.

Due to similarities in Westinghouse PWR designs, it was possible to perform a representative fatigue crack growth calculation which would be applicable to the Callaway Nuclear Power Plant. A comparison was made of the number of cycles, material, geometry, and types of discontinuities.

6.2 CRITICAL LOCATION FOR FATIGUE CRACK GROWTH ANALYSIS The weld location at the RCL cold leg nozzle to accumulator pipe was determined to be the most critical location for the fatigue crack growth evaluation. The nozzle configuration and weld location are shown in Figure 6-1. The geometry of the accumulator pipe was identical between the Callaway Nuclear Power Plant and the generic model (10 Schedule 140). Both analyses used austenitic stainless steel at the critical location.

6.3 DESIGN TRANSIENTS The transient conditions selected for this evaluation are based on conservative estimates of the magnitude and the frequency of the temperature fluctuations documented in various operating plant reports. These are representative of the conditions which are considered to occur during plant operation. The normal operating and upset thermal transients, in accordance with the design specification and the applicable system design criteria document, were considered for this evaluation. Out of these, 15 transients were used in the fatigue crack growth analysis and are listed in Table 6-1. There are some differences between the generic transients used in the fatigue crack growth evaluation and the Callaway Nuclear Power Plant transients but these differences will have insignificant impact on the fatigue crack growth results.

6.4 STRESS ANALYSIS A thermal transient stress analysis was performed for a typical plant similar to the Callaway Nuclear Power Plant to obtain the through-wall stress profiles for use in the fatigue crack growth analysis. The generic Accumulator line design transients described in Section 6.3 were used.

Growth Crack Growth Fatigue Crack of Fatigue February 2003 Assessment of February 2003

6-2 A simplified analysis method was used to develop conservative maximum and minimum linear through wall stress distributions due to minor thermal transients. In this method, a 1-D computer program was used to perform the thermal analysis to determine the through wall temperature gradients as a function of time. The inside surface stress was calculated by using an equation, which is similar to the transient portion of ASME Section III NB 3600, Equation (11). The effect of discontinuity was included in the analysis by performing a separate 1-D thermal analysis for the pipe and nozzle. The maximum and minimum inside surface stresses were then obtained by searching the inside surface stress values calculated for each time step of the transient solution. The outside surface stresses corresponding to the maximum and minimum inside surface stresses were then calculated by a similar method. The maximum and minimum linear through wall stress distribution for each thermal transient was obtained by joining the corresponding inside and outside surface stresses by a straight line. These two stress profiles are called the maximum and minimum through wall stress distributions respectively, for convenience.

The above methodology was used for minor thermal transients. For severe thermal transients, 11-D axisymmetric finite element model of the accumulator piping was used to determine the nonlinear stress distributions. The effects of discontinuity at the critical location was included by increasing the magnitude of 11-D nonlinear through wall stress by 20 percent at the inside one third thickness of the pipe wall.

The stresses due to the generic pressure and the generic moment loading were superimposed on the through wall cyclical stresses to obtain the total maximum and minimum stress profile for each transient.

6.5 OBE LOADS The stresses due to OBE loads were neglected in the fatigue crack growth analysis since these loads are not expected to contribute significantly to crack growth due to the small number of cycles.

6.6 TOTAL STRESS FOR FATIGUE CRACK GROWTH The total through wall stress at a section was obtained by superimposing the generic pressure stress and the generic moment stresses on the thermal transient stresses. Thus, the total stress for fatigue crack growth at any point is given by the following equation:

Total Stress Stress due to Stress due to For Fatigue Internal + Moment (DW + Thermal Crack Pressue + Thermal Transient Stress Growth ressure Expansion)

February 2003 Assessment of Crack Growth Fatigue Crack of Fatigue Growth February 2003

6-3 6.7 FATIGUE CRACK GROWTH ANALYSIS The fatigue crack growth analysis was performed to determine the effect of the design thermal transients. The analysis was performed for the critical cross section identified in Figure 6-1. A range of crack depths was postulated, and each was subjected to the thermal transients in Table 6-1, which included pressure and moment loads.

6.7.1 Analysis Procedure The fatigue crack growth analyses presented herein were conducted in the same manner as suggested by Section Xi, Appendix A of the ASME Boiler and Pressure Vessel Code (Reference 6-1). The analysis procedure involves assuming an initial flaw exists at some point and predicting the growth of that flaw due to an imposed series of fluctuating stresses. The growth of a crack per loading cycle is dependent on the range of applied stress intensity factor AK,, by the following:

da = COAKn (6-1) where SCO and the exponent and are material properties, and AK, is defined as (AK, = K. -

Kmin). For inert environments these material properties are constants, but for some water environments they are dependent on the level of mean stress present during the cycle. This can be accounted for by adjusting the value of SCOW by a function of the ratio of minimum to maximum stress for any given transient. Fatigue crack growth properties of stainless steel in a pressurized water environment have been used in the analysis.

The input required for a fatigue crack growth analysis is basically the information necessary to calculate the parameter AK,, which depends on crack size and structure geometry and the range of applied stresses in the area where the crack exists. Once AK, is calculated, the growth due to that particular cycle can be calculated by Equation (6-1). This increment of growth is then added to the original crack size, the AK, adjusted, and the analysis proceeds to the next transient. The procedure is continued in this manner until all the transients have been analyzed.

The reference crack growth law used for the stainless steel accumulator pipe system was taken from that developed by the Metal Properties Council - Pressure Vessel Research Committee Task Force In Crack Propagation Technology. The reference curve has the equation:

[ (6-2)

Growth Crack Growth Fatigue Crack of Fatigue February 2003 Assessment of Assessment February 2003

6-4 Iace This equation appears in Appendix C of ASME Section Xl for air environments and its basis is provided in Reference 6-2, and shown in Figure 6-2. For water environments, an environmental factor of [ afcce was used, based on the crack growth tests in PWR environments reported in Reference 6-3.

6.8 RESULTS Fatigue crack growth analyses were carried out at the critical cross section. Analysis was completed for a range of postulated flaw sizes oriented circumferentially, and the results are presented in Table 6-2. The postulated flaws are assumed to have an aspect ratio of six to one.

Even for the largest postulated flaw of 0.30 inch, which is about 33 percent of the wall thickness, the result projects that flaw growth through the wall will not occur during the 40 year design life of the plant. Therefore fatigue crack growth should not be a concern for the Callaway Nuclear Power Plant Accumulator lines.

6.9 REFERENCES

6-1 ASME Boiler and Pressure Vessel Code Section Xl, 2001 Edition, "Rules for Inservice Inspection of Nuclear Power Plant Components" 6-2 James, L. A., and Jones, D. R, "Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air," in Predictive Capabilities in Environmentally Assisted Cracking."

ASME publication PVP-99, Dec. 1985..

6-3 Bamford, W. H., 'Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment," Trans ASME, Journal of Pressure Vessel Technology, Feb. 1979.

Engineering Development Labs Report HEDL-TME-76-43, May 1976.

Assessment of Fatigue Crack Growth Assessment of FatigueCrackGrowth February 2003

6-5 Table 6-1 : Design Transients Considered for Fatigue Crack Growth Evaluation Trans. No. Description No. of Occurrences 1 Unit Loading 13200 2 Unit Unloading 13200 3 Step Load Increase 2,000 4 Step Load Decrease 2,000 5 Feedwater Cycling 2,000 6 Reactor Trip with Cooldown No Safety Injection 160 7 Inadvertent RCS Depressurization 20 8 Control Rod Drop 80 9 Turbine Roll Test 20 10 Accumulator Actuation, Accident Operation 21 11 Accumulator Actuation, Inadvertent During Cooldown 4 12 High Head Safety Injection 110 13 Steady-State and Random Fluctuations 3.2 x 106 14 RHR Operations During Plant Cooldown 200 15 RHR Operations During Refueling 80 Table 6-2: Accumulator Lines Fatigue Crack Growth Results Initial Crack Depth (in) After a,c,e Growth February 2003 Assessment of Fatigue Crack of Fatigue Crack Growth February 2003

6-6 Critical Section for RCL Cold Leg Fatigue Crack Growth Nozzle I I -Weld kvbterial Pipe l

5.375* R 4.4795" R R is the Pipe Radius Figure 6-1 Schematic of Accumulator Line at RCL Cold Leg Nozzle Weld Location Assessment of Fatigue Crack Growth February 2003

6-7 3.0 X 10 4 a

I

,I ia io-r(L IL jLY . A..IIIL 2.0 X100 101 102 A (ksi a -n Figure 6-2 Reference Crack Growth Curves for Stainless Steel in Air Environment Crack Growth Fatigue Crack of Fatigue Febwary 2003 Assessment of Assessment Growth February 2003

7-1 7 ASSESSMENT OF MARGINS In the preceding sections, the leak rates calculations, fracture mechanics analysis and fatigue crack growth assessment were performed. The results of the leak rates of Section 5.2 and the corresponding stability results of Section 5.3 are used in performing the assessment of margins. Margins are shown in Table 7-1.

In summary, at all the critical locations relative to:

1. Flaw Size - Using faulted loads obtained by the absolute sum method, a margin of 2 or more exists between the critical flaw and the flaw having a leak rate of 10 gpm (the leakage flaw).
2. Leak Rate - A margin of 10 exists between the calculated leak rate from the leakage flaw and the leak detection capability of 1 gpm.
3. Loads - At the critical locations the leakage flaw was shown to be stable using the faulted loads obtained by the absolute sum method (i.e., a flaw twice the leakage flaw size is shown to be stable; hence the leakage flaw size is stable). Therefore a margin on loads of 1.0 (see Section 4.2 for explanation) using the absolute summation of faulted load combinations is satisfied.

All the LBB recommended margins are satisfied.

In this evaluation, the Leak-Before-Break methodology is applied conservatively. The conservatism used in the evaluation is summarized in Table 7-2.

Margins of Margins February 2003 Assessment Assessment of February 2003

7-2 Table 7-1 : Leakage Flaw Sizes, Critical Flaw Sizes and Margins Node Critical Flaw Leakage Flaw Margin

_____ ____ Size (in) Size (in) Mr i 3020 11.39 3.87 2.94 3120 12.54 3.70 3.39 3295 16.36 6.20 2.64 Table 7-2: LBB Conservatism Factor of 10 on Leak Rate Factor of approximately 2 on Flaw Size for All Locations Algebraic Sum of Loads for Leakage Absolute Sum of Loads for Stability Average Material Properties for Leakage Minimum Material Properties for Stability February 2003 Assessment of Margins Assessment of Margins February 2003

8-1 8 CONCLUSIONS This report justifies the elimination of 10" Accumulator line breaks as the structural design basis for the Callaway Nuclear Power Plant as follows:

a. Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation.
b. Water hammer should not occur in the RCS piping (primary loop and the attached class 1 auxiliary lines) because of system design, testing, and operational considerations.
c. The effects of low and high cycle fatigue on the integrity of the Accumulator lines were evaluated and shown acceptable.
d. Ample margin exists between the leak rate of small stable flaws and the capability of the Callaway Nuclear Power Plant reactor coolant system pressure boundary leakage detection system.
e. Ample margin exists between the small stable flaw sizes of item (d) and the critical flaw size.

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

Based on the above, it is concluded that 10" Accumulator line breaks should not be considered in the structural design basis of the Callaway Nuclear Power Plant.

ConclusionsFeury20 Conclusions February 2003

A-1 APPENDIX A - LIMIT MOMENT I

j a,c,e A~pni - Lii Moen Februar 2003 AppenGlX A - Limit moment February 2003

A-2 FIgure A-i Pipe with A Through-Wall Crack In BendI Figure A-1 Pipe with A Through-Wall Crack in Bending Moment Limit Moment February 2003 Appendix AA - Limit February 2003

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