LIC-13-0100, WCAP-17262-NP, Rev 1, Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun, Unit 1

WCAP-17262-NP, Rev 1, Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun, Unit 1

ML13220A074
Person / Time
Site:	Fort Calhoun
Issue date:	07/31/2013
From:	Kupper C Westinghouse
To:	Office of Nuclear Reactor Regulation
References
LIC-13-0100, LAR 13-06 WCAP-17262-NP, Rev 1
Download: ML13220A074 (64)

Text

LIC-13-0100 Enclosure Page 1 "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun Unit 1" WCAP-17262-NP, Revision 1 July 2013

Westinghouse Non-Proprietary Class 3 WCAP-17262-NP July 2013 Revision 1 Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun Unit 1 Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17262-NP Revision I Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun Unit 1 C. T. Kupper*

Piping Analysis and Fracture Mechanics July 2013 Reviewer: D. C. Bhowmick*

Piping Analysis and Fracture Mechanics A. Udyawar*

Piping Analysis and Fracture Mechanics Approved: S. A. Swamy*, Manager Piping Analysis and Fracture Mechanics

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2013 Westinghouse Electric Company LLC All Rights Reserved

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ii WESTINGHOUSE NON-PROPRIETARY CLASS 3 ii Acronym List ACRS Advisory Committee on Reactor Safeguards AIF Atomic Industrial Forum ANL Argonne National Laboratory AR Aspect Ratio ASME American Society of Mechanical Engineers CE Combustion Engineering CGR Crack Growth Rate CMTR Certified Material Test Report EPFM Elastic-Plastic Fracture Mechanics EPRI Electric Power Research Institute EPU Extended Power Uprate FCG Fatigue Crack Growth FCS Fort Calhoun Station IGSCC Intergranular Stress Corrosion Cracking LBB Leak-Before-Break LLNL Lawrence Livermore National Laboratory LOCA Loss of Coolant Accident MUR Measurement Uncertainty Recapture NRC Nuclear Regulatory Commission OD Outside Diameter OPPD Omaha Public Power District PCSG Pipe Crack Study Group PWR Pressurizer Water Reactor PWSCC Primary Water Stress Corrosion Cracking RCS Reactor Coolant System RMS Root Mean Square RT Room Temperature SAW Submerged Arc Weld SAM Seismic Anchor Motion SCC Stress Corrosion Cracking SMAW Shielded Metal Arc Weld SSE Safe Shutdown Earthquake WCAP-17262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii TABLE OF CONTENTS 1.0 Intro d u ctio n .................................................................................................................... 1-1 1 .1 P u rp o s e .............................................................................................................. 1-1 1.2 Background Information ..................................................................................... 1-1 1.3 Scope and Objectives ........................................................................................ 1-2 1.4 References ......................................................................................................... 1-3 2.0 Operation and Stability of the Reactor Coolant System ................................................. 2-1 2.1 Stress Corrosion Cracking .................................................................................. 2-1 2.2 W ater Hammer ................................................................................................... 2-2 2.3 Low Cycle and High Cycle Fatigue .................................................................... 2-3 2.4 W all Thinning, Creep, and Cleavage .................................................................. 2-3 2.5 References ......................................................................................................... 2-3 3.0 Pipe Geometry and Loading .......................................................................................... 3-1 3.1 Introduction to Methodology ............................................................................... 3-1 3.2 Calculation of Loads and Stresses ..................................................................... 3-2 3.3 Loads for Leak Rate Evaluation ......................................................................... 3-3 3.4 Load Combination for Crack Stability Analyses .................................................. 3-4 3.5 References ......................................................................................................... 3-4 4.0 Material Characterization ............................................................................................... 4-1 4.1 Primary Loop Pipe and Fittings Materials ........................................................... 4-1 4.2 Tensile Properties ............................................................................................... 4-1 4.3 Fracture Toughness Properties .......................................................................... 4-2 4.4 References ......................................................................................................... 4-5 5.0 Critical Location and Evaluation Criteria ........................................................................ 5-1 5.1 Critical Locations ................................................................................................ 5-1 5.2 Fracture Criteria ................................................................................................. 5-1 6.0 Leak Rate Predictions .................................................................................................... 6-1 6.1 Introduction ......................................................................................................... 6-1 6.2 General Considerations ...................................................................................... 6-1 6.3 Calculation Method ............................................................................................. 6-1 6.4 Leak Rate Calculations ...................................................................................... 6-2 6.5 References ......................................................................................................... 6-2 7.0 Fracture Mechanics Evaluation ...................................................................................... 7-1 7.1 Local Failure Mechanism ................................................................................... 7-1 7.2 Global Failure Mechanism .................................................................................. 7-2 7.3 Results of Crack Stability Evaluation .................................................................. 7-3 7.4 References ......................................................................................................... 7-4 8.0 Fatigue Crack Growth Analysis ...................................................................................... 8-1 WCAP-1 7262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv 8.1 References ......................................................................................................... 8-2 9.0 Assessment of Margins .................................................................................................. 9-1 9.1 References ......................................................................................................... 9-1 10.0 Conclusions .................................................................................................................. 10-1 Appendix A: Limit Moment ................................................................................................... A-1 WCAP-1 7262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V LIST OF TABLES Table 3-1 Dimensions, Normal Loads and Stresses for Fort Calhoun Unit 1 .......................... 3-5 Table 3-2 Faulted Loads and Stresses for Fort Calhoun Unit 1 .............................................. 3-6 Table 4-1 Measured Tensile Properties for Fort Calhoun Unit 1 Primary Loop Piping ............. 4-6 Table 4-2 Measured Tensile Properties for Fort Calhoun Unit 1 Primary Loop Elbows ........... 4-7 Table 4-3 Mechanical Properties for Fort Calhoun Unit 1 Materials at Operating Temperatures

........................................................................................................................... 4 -8 Table 4-4 Chemistry & Fracture Toughness Piping Properties of the Material Heats of Fort C a lh o u n Un it 1 .................................................................................................... 4 -9 Table 4-5 Chemistry & Fracture Toughness Elbow Properties of the Material Heats of Fort C a lh o u n Un it 1 .................................................................................................. 4 -10 Table 4-6 Fracture Toughness Properties Used to Evaluate Critical Locations .................... 4-11 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations ................ 6-3 Table 7-1 Stability Results for Fort Calhoun Based on Elastic-Plastic J-Integral Evaluations .7-5 Table 7-2 Stability Results for Fort Calhoun Based on Limit Load .......................................... 7-5 Table 8-1 Reactor Coolant System Operating Transients ....................................................... 8-3 Table 8-2 FCG at Alloy 82/182 Weld (Nozzle to Safe-end Weld) - Outlet Nozzle .................. 8-4 Table 8-3 FCG at Stainless Steel Weld (Safe-end to Pipe Weld) - Outlet Nozzle .................. 8-4 Table 8-4 FCG at Reactor Vessel Nozzle Location - Outlet Nozzle ....................................... 8-4 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes by Limit Load and Margins for Fort Calhoun

................................. ......................................................................................... 9-2 Table 9-2 Stability Results for Fort Calhoun Based on Elastic-Plastic J-Integral Evaluations.9-2 WCAP-17262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi WESTINGHOUSE NON-PROPRIETARY CLASS 3 VI LIST OF FIGURES Figure 3-1 Reactor Coolant System Pipe ........................................................................... 3-7 Figure 3-2 Schematic Diagram of Fort Calhoun Primary Loop Showing Weld Locations ...3-8 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ............. 6-4

[ ]a,c,, Pressure Ratio as a Function of L/D ....................................... 6-5 Figure 6-2 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack ........................... 6-6 Figure 7-1 Fully Plastic Stress Distribution ....................................................................... 7-6 Figure 7-2 Critical Flaw Size Prediction - Location 1 ........................................................ 7-7 Figure 7-3 Critical Flaw Size Prediction - Location 6 ........................................................ 7-8 Figure 8-1 Reactor Vessel Outlet Nozzle with Stress Cut Locations ................................. 8-5 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels

.................. ........................................................................................................ 8 -6 Figure 8-3 Reference Fatigue Crack Growth Curves for Stainless Steels ......................... 8-7 Figure 8-4 Reference Fatigue Crack Growth Curves for Alloy 82/182 Welds .................... 8-8 Figure A-1 Pipe with a Through-Wall Crack in Bending .................................................... A-2 WCAP-1 7262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii EXECUTIVE

SUMMARY

The original structural design basis of the reactor coolant system for the Omaha Public Power District (OPPD) Fort Calhoun Station (FCS) Unit 1 required consideration of dynamic effects resulting from pipe break and that protective measure for such breaks be incorporated into the design. Subsequent to the original Fort Calhoun design, additional concern of asymmetric blowdown loads was raised as described in Unresolved Safety Issue A-2 (Asymmetric Blowdown Loads on the Reactor Coolant System) and Generic Letter 84-04. Fort Calhoun Unit 1 was part of the utilities which sponsored Westinghouse to resolve the A-2 issue. Generic analyses by Westinghouse to resolve the A-2 issue were approved by the NRC and documented in Generic Letter 84-04 (Reference 1-1).

Research by the NRC and industry coupled with operating experience determined that safety could be negatively impacted by placement of pipe whip restraints on certain systems. As a result, NRC and industry initiatives resulted in demonstrating that leak-before-break (LBB) criteria can be applied to reactor coolant system piping based on fracture mechanics technology and material toughness.

Subsequently, the NRC modified 10CFR50 General Design Criterion 4, and published in Federal Register (Vol. 52, No. 207) on October 27, 1987 its final rule, "Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures (Reference 1-2)." This change to the rule allows use of leak-before-break technology for excluding from the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in pressurized water reactors (PWRs).

This current report (WCAP-17262-NP Revision 1) demonstrates compliance with LBB technology for the Fort Calhoun Unit 1 reactor coolant system piping on a plant specific analysis. Inputs from the license renewal program are used in the LBB analysis. The report documents the plant specific geometry, operating parameters, loading, and material properties used in the fracture mechanics evaluation. Mechanical properties were determined at operating temperatures. Since the piping systems include cast stainless steel, fracture toughness considering thermal aging was determined for each heat of material for the fully aged condition.

Based on loading, pipe geometry and fracture toughness considerations, enveloping critical (governing) locations were determined at which leak-before-break crack stability evaluations were made. Through-wall flaw sizes were postulated which would cause a leak at a rate of ten (10) times the leakage detection system capability of the plant. Large margins for such flaw sizes were demonstrated against flaw instability. Finally, fatigue crack growth was shown not to be an issue for the primary loop piping.

The effects due to the license renewal program on the continued applicability of LBB for the reactor coolant loop piping at Fort Calhoun Unit 1 have been evaluated. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the reactor coolant primary loop piping need not be considered in the structural design basis for Fort Calhoun Unit 1 for the license renewal period.

WCAP-1 7262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Viii Revision 0 of this report is still applicable for the Extended Power Uprate (EPU) and Measurement Uncertainty Recapture (MUR) programs. Revision 1 is applicable for "current" operating conditions where "current" applies to normal operating conditions prior to EPU and MUR.

WCAP-1 7262-NP July 2013 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1.0 INTRODUCTION

1.1 PURPOSE This report applies to the Fort Calhoun Unit 1 Reactor Coolant System (RCS) primary loop piping. It is intended to demonstrate that for the specific parameters of Fort Calhoun, RCS primary loop pipe breaks need not be considered in the structural design basis. The LBB approach taken in this report has been accepted by the U.S. Nuclear Regulatory Commission (NRC) (Reference 1-2).

1.2 BACKGROUND

INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that RCS primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in the RCS primary loop was first presented to the NRC in 1978 in WCAP-9283 (Reference 1-3). That topical report employed a deterministic fracture mechanics evaluation and a probabilistic analysis to support the elimination of RCS primary loop pipe breaks. That approach was then used as a means of addressing Generic Issue A-2 and Asymmetric LOCA Loads.

Westinghouse performed additional testing and analysis to justify the elimination of RCS primary loop pipe breaks. This material was provided to the NRC along with Letter Report NS-EPR-2519 (Reference 1-4).

The NRC funded research through Lawrence Livermore National Laboratory (LLNL) to address this same issue using a probabilistic approach. As part of the LLNL research effort, Westinghouse performed extensive evaluations of specific plant loads, material properties, transients, and system geometries to demonstrate that the analysis and testing previously performed by Westinghouse and the research performed by LLNL applied to all Westinghouse plants (References 1-5 and 1-6). The results from the LLNL study were released at a March 28, 1983, ACRS Subcommittee meeting. These studies, which are applicable to all Westinghouse plants east of the Rocky Mountains, determined the mean probability of a direct LOCA (RCS primary loop pipe break) to be 4.4 x 1012 per reactor year and the mean probability of an indirect LOCA to be 107 per reactor year. Although FCS is not a Westinghouse plant, similar probabilities would be expected for a LOCA at FCS. Thus, the results previously obtained by Westinghouse (Reference 1-3) were confirmed by an independent NRC research study.

Based on the studies by Westinghouse, LLNL, the ACRS, and the AIF, the NRC completed a safety review of the Westinghouse reports submitted to address asymmetric blowdown loads that result from a number of discrete break locations on the PWR primary systems. The NRC Staff evaluation (Reference 1-1) concludes that an acceptable technical basis has been provided so that asymmetric blowdown loads need not be considered for those plants that can demonstrate the applicability of the modeling and conclusions contained in the Westinghouse response or can provide an equivalent fracture mechanics demonstration of the primary coolant loop integrity. In a more formal recognition of leak-before-break (LBB) methodology applicability for PWRs, the NRC appropriately modified 10 CFR 50, General Design Criterion 4, Introduction July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 "Requirements for Protection Against Dynamic Effects for Postulated Pipe Rupture" (Reference 1-2).

1.3 SCOPE AND OBJECTIVES The general purpose of this investigation is to demonstrate leak-before-break for the primary loop in Fort Calhoun Unit 1. The recommendations and criteria proposed in References 1-7 and 1-8 are used in this evaluation. These criteria and resulting steps of the evaluation procedure can be briefly summarized as follows:

1. Calculate the applied loads. Identify the locations at which the highest stress occurs.

2. Identify the materials and the associated material properties.

3. Postulate a surface flaw at a governing location. Determine fatigue crack growth. Show that a through-wall crack will not result.

4. Postulate a through-wall flaw at the critical (governing) locations. 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 a margin of 10 between the calculated leak rate and the leak detection capability.

5. Using faulted loads, demonstrate that there is a margin of at least 2 between the leakage flaw size and the critical flaw size.

6. Review the operating history to ascertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer, low and high cycle fatigue, wall thinning, creep, or cleavage.

7. For the materials actually used in the plant provide the properties including toughness and tensile test data. Evaluate long term effects such as thermal aging.

8. Demonstrate margin on applied load.

This report provides a fracture mechanics demonstration of primary loop integrity for Fort Calhoun consistent with the NRC position for exemption from consideration of dynamic effects.

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

Introduction July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3

1.4 REFERENCES

1-1 USNRC Generic Letter 84-04,

Subject:

"Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1, 1984.

1-2 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

1-3 WCAP-9283, "Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," March, 1978.

1-4 Letter Report NS-EPR-2519, Westinghouse (E. P. Rahe) to NRC (D. G. Eisenhut),

Westinghouse Proprietary Class 2, November 10, 1981.

1-5 Letter from Westinghouse (E. P. Rahe) to NRC (W. V.Johnston) dated April 25, 1983.

1-6 Letter from Westinghouse (E. P. Rahe) to NRC (W. V.Johnston) dated July 25, 1983.

1-7 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.

1-8 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

Introduction July 2013 WCAP-1 7262-NP Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2.0 OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM 2.1 STRESS CORROSION CRACKING The Westinghouse and Combustion Engineering (CE) designed reactor coolant system (RCS) primary loops 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 1400 reactor-years, including 16 plants each having over 30 years of operation, 10 other plants each with over 25 years of operation, 11 plants each over 20 years of operation and 12 plants each over 15 years of operation.

In 1978, the United States Nuclear Regulatory Commission (USNRC) formed the second Pipe Crack Study Group. (The first Pipe Crack Study Group (PCSG) established in 1975 addressed cracking in boiling water reactors only.) One of the objectives of the second PCSG was to include a review of the potential for stress corrosion cracking in Pressurized Water Reactors (PWR's). The results of the study performed by the PCSG were presented in NUREG-0531 (Reference 2-1) entitled "Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants." In that report the PCSG stated:

"The PCSG has determined that the potential for stress-corrosion cracking in PWR primary system piping is extremely low because the ingredients that produce IGSCC are not all present. The use of hydrazine additives and a hydrogen overpressure limit the oxygen in the coolant to very low levels. Other impurities that might cause stress-corrosion cracking, such as halides or caustic, are also rigidly controlled. Only for brief periods during reactor shutdown when the coolant is exposed to the air and during the subsequent startup are conditions even marginally capable of producing stress-corrosion cracking in the primary systems of PWRs. Operating experience in PWRs supports this determination. To date, no stress corrosion cracking has been reported in the primary piping or safe ends of any PWR."

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 thionates). Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to Operation and Stability of the Reactor Coolant System July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 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 is expected to be in the 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. Thus during plant operation, the likelihood of stress corrosion cracking is minimized.

During 1979, several instances of cracking in PWR feedwater piping led to the establishment of the third PCSG. The investigations of the PCSG reported in NUREG-0691 (Reference 2-2) further confirmed that no occurrences of IGSCC have been reported for PWR primary coolant systems.

Primary Water Stress Corrosion Cracking (PWSCC) occurred in V. C. Summer reactor vessel hot leg nozzle, Alloy 82/182 welds. It should be noted that this susceptible material is found at the Fort Calhoun Unit 1 Reactor Vessel Inlet and Outlet nozzle locations. Mitigation will be implemented as required to minimize PWSCC at the Reactor Vessel nozzle locations.

2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. The reactor coolant system, including piping and primary components, is 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 control rod position; pressure is controlled by pressurizer heaters and pressurizer spray also within a narrow range for steady-state conditions. 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 and CE designs have instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping are such that no significant water hammer can occur.

Operation and Stability of the Reactor Coolant System July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 2.3 LOW CYCLE AND HIGH CYCLE FATIGUE An evaluation of the low cycle fatigue loadings was carried out as part of this study in the form of a fatigue crack growth analysis, as discussed in Section 8.0.

High cycle fatigue loads in the system would result primarily from pump vibrations. These are minimized by restrictions placed on shaft vibrations during hot functional testing and operation.

During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for fatigue crack growth. Fort Calhoun RCS configurations are similar and the results are concluded to be similar.

2.4 WALL THINNING, CREEP, AND CLEAVAGE Wall thinning by erosion and erosion-corrosion effects should not occur in the primary loop piping due to the low velocity, typically less than 1.0 ft/sec and the stainless steel material, which is highly resistant to these degradation mechanisms. The cause of wall thinning is related to the high water velocity and is therefore clearly not a mechanism that would affect the primary loop piping.

The maximum operating temperature of the primary loop piping, which is less than 6000 F, is well below the temperature that would cause significant mechanical creep damage in stainless steel piping. Cleavage type failures are not a concern for the operating temperatures and the stainless steel material used in the primary loop piping.

2.5 REFERENCES

2-1 Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants, NUREG-0531, U.S. Nuclear Regulatory Commission, February 1979.

2-2 Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.

Operation and Stability of the Reactor Coolant System July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3.0 PIPE GEOMETRY AND LOADING

3.1 INTRODUCTION

TO METHODOLOGY The general approach is discussed first. As an example a segment of the primary coolant loop pipe is shown in Figure 3-1. The as-built outside diameter and minimum thickness of the pipe are, as shown in the figure. The normal stresses at the weld location are from the load combination procedure discussed in Section 3.3 whereas the faulted loads are as described in Section 3.4. The components for normal loads are internal pressure, dead weight, pressure expansion, and normal thermal expansion. An additional component, Safe Shutdown Earthquake (SSE), is considered for faulted loads. As seen from Table 3-2, the highest stressed location in the entire loop is Location 1 (Figure 3-2) at the reactor vessel outlet nozzle to pipe weld. The highest stressed location in the cold/cross-over leg piping is Location 6 at the cross-over leg elbow weld. These are the critical locations at which, as enveloping locations, leak-before-break is to be established. Essentially a circumferential flaw is postulated at each location which is subjected to both the normal loads and faulted loads to assess leakage and stability, respectively. The loads (developed below) at the critical locations are also given in Figure 3-1. Loads from Replacement Steam Generator (RSG) that are applicable for current power operating conditions are used in the LBB analysis.

Since the piping for Fort Calhoun Unit 1 is made of cast stainless steel, thermal aging effects must be considered (Section 4.0). Thermal aging results in lower fracture toughness; thus, locations must be examined taking into consideration both fracture toughness and stress. Once loads (this section) and fracture toughness (Section 4.0) are obtained, the critical locations are identified in Section 5.0. At these locations, leak rate evaluations (Section 6.0) and fracture mechanics evaluations (Section 7.0) are performed per the guidance of References 3-1 and 3-2.

Fatigue crack growth (Section 8.0) and stability margins are also evaluated (Section 9.0).

The critical locations are determined based on stresses and material properties for Fort Calhoun Unit 1. All the weld locations for evaluation are those shown in Figure 3-2. It should be noted that the Fort Calhoun Primary loop consists of 2 loops with 24 weld locations in each loop. Due to the similarity in geometry and materials, only one loop with 14 weld locations will be analyzed for the hot, cold, and crossover legs with the enveloping loads being used at each weld location from both loops. Therefore, all 24 weld locations in each loop are accounted for with the 14 weld locations analyzed.

Pipe Geometry and Loading July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 3.2 CALCULATION OF LOADS AND STRESSES The stresses due to axial loads and bending moments are calculated by the following equation:

F M (3-1)

A Z

where, a = stress F = axial load M = moment A = pipe cross-sectional area Z = section modulus The moments for the desired loading combinations are calculated by the following equation:

M= Mx+M2+M (3-2)

where, M) X component of moment, Torsion my Y component of bending moment Mz Z component of bending moment The axial load and moments for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.3 and 3.4.

Pipe Geometry and Loading July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 3.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:

F = FDw + FTH + Fp + Fpexp (3-3)

Mx= (Mx)Dw + (MX)TH + (Mx)Pexp (3-4)

MY= (My)Dw + (MY)TH + (My)Pexp (3-5)

Mz= (Mz)ow + (MZ)TH + (Mz)Pexp (3-6)

The subscripts of the above equations represent the following loading cases:

DW = deadweight TH = normal thermal expansion P = load due to internal pressure Pexp = pressure expansion This method of combining loads is often referred to as the algebraic sum method (References 3-1 and 3-2).

The loads based on this method of combination are provided in Table 3-1 at all the weld locations identified in Figure 3-2. The as-built dimensions are also given in Table 3-1.

Pipe Geometry and Loading July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 3.4 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the margin in terms of applied loads needs to be demonstrated by crack stability analysis. Margin on loads of 1.4 (4/2) can be demonstrated if normal plus Safe Shutdown Earthquake (SSE) are applied. The 1.4 (4I2) margin should be reduced to 1.0 if the deadweight, thermal expansion, internal pressure, pressure expansion, SSEINERTIA and seismic anchor motion (SAM) loads are combined based on individual absolute values as shown below.

The absolute sum of loading components is used for the FCS LBB analysis which results in higher magnitude of combined loads and thus satisfies a margin on loads of 1.0. The absolute summation of loads is shown in the following equations:

F =IFDW I+I FTH I+ I Fp I+ I FSSEINERTA I+ IFSSEAM I + IFpexp I (3-7)

Mx = I (Mx)Dw I+ I (MX)TH I + I (MX)SSEINERTIAI + I (MX)SSEAMI + I (Mx)Pexp I (3-8)

MY= I (My)Dw I+ I (MY)TH I + I(My)ssEINERTIAI + I (MY)SSEAMI + I (My)Pexp I (3-9)

Mz= I(Mz)Dw I+ I (MZ)TH I + I (MZ)SSEINERTIAI + I (Mz)ssEMI + I (Mz)pexp I (3-10) where subscript SSEINERTIA refers to safe shutdown earthquake inertia, SSEAM is safe shutdown earthquake anchor motion.

The loads so determined are used in the fracture mechanics evaluations (Section 7.0) to demonstrate the LBB margins at the locations established to be the governing locations. These loads at all the weld locations (see Figure 3-2) are given in Table 3-2.

3.5 REFERENCES

3-1 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.

3-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

Pipe Geometry and Loading July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 3-1 Dimensions, Normal Loads and Stresses for Fort Calhoun Unit I Minimum Axial Bending Outside Diameter Thickness Loadb Moment Total Stress Locationa (in) (in) (kips) (in-kips) (ksi) 1 38.50 2.688 1822 21287 14.43 2 38.50 2.688 1771 806 6.18 3 38.50 2.688 1771 9396 9.57 4 39.00 2.938 1682 18368 11.63 5 29.25 2.063 1019 2965 8.43 6 29.00 1.938 1019 2732 8.80 7 29.00 1.938 1053 2828 9.10 8 29.00 1.938 1042 1763 8.01 9 29.00 1.938 1025 2178 8.31 10 29.25 2.063 995 3530 8.80 11 29.00 1.938 1044 1345 7.62 12 29.00 1.938 1023 681 6.86 13 29.00 1.938 1039 629 6.91 14 29.00 1.938 1028 1550 7.72 Notes:

a. See Figure 3-2

b. Included Pressure Pipe Geometry and Loading July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-6 Table 3-2 Faulted Loads and Stresses for Fort Calhoun Unit I Locationa'b Axial Loadc (kips) Bending Moment (in- Total Stress (ksi) kips) 1 1964 23806 15.90 2 1962 2777 7.59 3 1961 11586 11.06 4 2039 21235 13.73 5 1078 4934 10.52 6 1075 4368 10.71 7 1074 4200 10.54 8 1058 2528 8.84 9 1058 3130 9.42 10 1077 4604 10.23 11 1067 2471 8.84 12 1069 1670 8.09 13 1061 1860 8.22 14 1062 3884 10.16 Notes:

a. See Figure 3-2

b. See Table 3-1 for dimensions C. Included Pressure Pipe Geometry and Loading July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 F>

MP K OD I Location 1 OD' = 38.50 in ta = 2.688 Normal Loadsa Faulted Loadsb Forcec: 1822 kips Forcec: 1964 kips Moment: 21287 in-kips Moment: 23806 in-kips ODa = 29.00 in Location 6 ta = 1.938 Normal Loadsa Faulted Loadsb Forcec: 1019 kips Forcec: 1075 kips Moment: 2732 in-kips Moment: 4368 in-kips a See Table 3-1 b See Table 3-2 c Includes the force due to a pressure of 2100 psia Figure 3-1 Reactor Coolant System Pipe Pipe Geometry and Loading July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 Pump Steam Generator

• Alloy 82/182 Weld Cross-Over Leg HOT LEG Temperature: 593 0 F* Pressure: 2100 psia CROSS-OVER LEG Temperature: 545 0 F* Pressure: 2100 psia COLD LEG Temperature: 545 0 F* Pressure: 2100 psia

• Temperatures applicable to Fort Calhoun current operating conditions (pre-EPU+MUR)

Figure 3-2 Schematic Diagram of Fort Calhoun Primary Loop Showing Weld Locations Pipe Geometry and Loading July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS The primary loop pipe materials are A-451 CPF8M, and the elbow fittings are A-351-65 CF8M.

4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) for Fort Calhoun primary loop piping and elbow fittings were used to establish the tensile properties for the leak-before-break analyses. The tensile properties for the pipe material are provided in Table 4-1; while the tensile properties for the elbow fittings material are provided in Table 4-2.

For both the A351-65 CF8M and A451 CPF8M materials, the representative properties at Fort Calhoun current operating temperatures were established from the tensile properties at room temperature by utilizing Section IIof the ASME Boiler and Pressure Vessel Code (Reference 4-1). Code tensile properties at the operating temperatures of 593 0F for the hot leg and 545 0F for the cold/crossover legs were obtained by interpolation. Ratios of the ASME Code tensile properties at operating temperature to the corresponding properties at room temperature were then applied to the room temperature tensile properties obtained from CMTRs (Tables 4-1 and 4-2) to obtain the Fort Calhoun specific properties at operating temperatures.

The average and lower bound yield strengths and ultimate strengths for the pipe and elbow fitting materials are tabulated in Table 4-3. The ASME Code modulus of elasticity values are also given, and Poisson's ratio was taken as 0.3.

Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 4.3 FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast stainless steels that are of interest are in terms of Jjc (J at Crack Initiation) and have been found to be very high at 6000 F. Cast stainless steel is susceptible to thermal aging during service. Thermal aging of cast stainless steel results in a decrease in the ductility, impact strength, and fracture toughness, of the material. Depending on the material composition, the Charpy impact energy of a cast stainless steel component could decrease to a small fraction of its original value after exposure to reactor temperatures during service.

The method described below was used to calculate the end of life toughness properties for the cast material of the Fort Calhoun Unit 1 primary coolant loop piping and elbows.

In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials. The ANL research program measured mechanical properties of cast stainless steel materials after they had been heated in controlled ovens for long periods of time. ANL compiled a data base, both from data within ANL and from international sources, of about 85 compositions of cast stainless steel exposed to a temperature range of 290-400 0C (550-750 0F) for up to 58,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (6.5 years).

From this database, ANL developed correlations for estimating the extent of thermal aging of cast stainless steel (References 4-2 and 4-3).

ANL developed the fracture toughness estimation procedures by correlating data in the database conservatively. After developing the correlations, ANL validated the estimation procedures by comparing the estimated fracture toughness with the measured value for several cast stainless steel plant components removed from actual plant service. The ANL procedures produced conservative estimates that were about 30 to 50 percent less than actual measured values. The procedure developed by ANL in Reference 4-3 was used to calculate the end of life fracture toughness values for this analysis. ANL research program was sponsored and the procedure was accepted (Reference 4-4) by the NRC.

The chemical compositions of the Fort Calhoun primary loop pipe and elbow fitting material are available from CMTRs and are provided in Table 4-4 and Table 4-5 of this report. It should be noted that the Fort Calhoun elbows are CF8M steel whereas the piping segments are CPF8M steel. The following equations are applicable for both CF8M and CPF8M type materials.

The following equations are taken from Reference 4-3:

Creq = Cr + 1.21 (Mo) + 0.48(Si) - 4.99 (4-1)

Nieq = (Ni) + 0.11(Mn) - 0.0086(Mn) 2 + 18.4(N) + 24.5(C) + 2.77 (4-2) where Creq = (Chromium equivalent); Nieq = (nickel equivalent);

6c (ferrite content) in percent volume is given by:

Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 6c=100.3(Creq / Nieq )2-170.72(Creq / Nieq )+74.22 (4-3)

The saturation room temperature (RT) impact energies of the cast stainless steel materials were determined from the chemical compositions.

For CF8M steel with < 10% Ni, the saturation value of RT impact energy CVsat (J/cm 2) is the lower value determined from Io9goCVsat = 1.10 + 2.12 exp (-0.0414) (4-4) where the material parameter 4 is expressed as 4 = 8, (Ni + Si +Mn) 2(C + 0.4N)/5.0 (4-5) and from logioCvsat = 7.28 - 0.0116c - 0.185Cr - 0.369Mo - 0.451Si- 0.007Ni - 4.71(C + 0.4N) (4-6)

For CF8M steel with > 10% Ni, the saturation value of RT impact energy CVsat (J/cm 2) is the lower value determined from IogioCvsat = 1.10 + 2.64 exp (-0.064¢) (4-7) where the material parameter 4 is expressed as S= 5 (Ni + Si +Mn) 2 (C + 0.4N)/5.0 (4-8) and from IogioCvsat = 7.28 - 0.011 0 - 0.185Cr - 0.369Mo - 0.451Si- 0.007Ni - 4.71(C + 0.4N) (4-9)

The saturation J-R curve at RT, for static-cast CF8M steel is given by Jd = 16(Cvsat) 0 67(Aa)n (4-10) and for centrifugally cast CF8M steel, by Jd = 20(Cvsat)0 67(Aa)n (4-11) where the exponent n for CF8M steel is expressed as n = 0.23 + 0.08 loglo (Cvsat) (4-12) where Jd is the "deformation J" in kJim 2 and Aa is the crack extension in mm.

The saturation J-R curve at 2900C (554 0 F), for static-cast CF8M steel is given by Jd = 49 (Cvsat) 0 4 1(Aa)n (4-13) and for centrifugally cast CF8M steel, by Jd = 57 (Cvsat) 0 4 1(Aa)n (4-14) where the exponent n for CF8M steel is expressed as n = 0.23 + 0.06 log1 o (Cvsat) (4-15) where Jd is the "deformation J" in kJ/m 2 and Aa is the crack extension in mm.

Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4 Note: Conservatively static cast equations are used for J calculations.

[

]a,c,e The results from the ANL Research Program indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of Submerged Arc Welds (SAWs). The applied value of the J-integral for a flaw in the weld regions will be lower than that in the base metal because the yield strength for the weld materials is much higher at the temperature 1 .

Therefore, weld regions are less limiting than the cast base material.

In the fracture mechanics analyses that follow, the fracture toughness properties given in Table 4-6 will be used as the criteria against which the applied fracture toughness values will be compared.

1 All the applied J values were conservatively determined by using base metal strength properties.

Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5

4.4 REFERENCES

4-1 ASME Boiler and Pressure Vessel Code Section II, Part D. "Materials," 2001 Edition, July 1,2001.

4-2 0. K. Chopra and W. J. Shack, "Assessment of Thermal Embrittlement of Cast Stainless Steels," NUREG/CR-6177, U. S. Nuclear Regulatory Commission, Washington, DC, May 1994.

4-3 0. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems," NUREG/CR-4513, Revision 1, U. S. Nuclear Regulatory Commission, Washington, DC, August 1994.

4-4 "Flaw Evaluation of Thermally Aged Cast Stainless Steel in Light-Water Reactor Applications," Lee, S.; Kuo, P. T.; Wichman, K.; Chopra, 0.; Published in International Journal of Pressure Vessel and Piping, June 1997.

Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 Table 4-1 Measured Tensile Properties for Fort Calhoun Unit I Primary Loop Piping Yield Strength (psi) Ultimate Strength (psi)

Heat No. Room Temp. Room Temp.

J-26078901 46000 83600 J-21278901 43110 83200 K-3559012 43100 83700 J-289567890 53100 84700 J-289567890 53100 84700 J-283789012 42500 82100 J-355123456 47800 85500 J-288234567 46600 89220 J-285012345 45600 85700 J-288234567 46600 89220 J-26201234 44460 83920 J-285678901 47600 85700 J-287567890B 46120 85210 J-286234567 53900 85100 J-288901234 47900 83000 J-285678901 47600 85700 J-284456789 51000 84700 J-286901234 49400 78700 J-287567890 46120 85210 J-284456789 51000 84700 Note: The Fort Calhoun Unit 1 Pipe Material is A451-CPF8M Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-7 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-7 Table 4-2 Measured Tensile Properties for Fort Calhoun Unit I Primary Loop Elbows Yield Strength (psi) Ultimate Strength (psi)

Heat No. Room Temp. Room Temp.

27129-1 43500 85500 24968-1 31500 86750 32785-1 44300 83800 26895-2 48000 84000 26895-1 41400 82400 33801-2 39800 78300 32515-1 43100 84000 33756-1 42900 85000 33418-1 42000 85250 33867-2 46600 88100 33975-1 45000 85500 33712-1 46000 88300 33676-1 47750 87900 27418-1 39600 78600 26307-2 49100 87500 32515-2 40200 78100 33041-1 45800 86100 36972-3 45000 84000 Note: The Fort Calhoun Unit 1 Elbow Material is A351-CF8M Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 Table 4-3 Mechanical Properties for Fort Calhoun Unit I Materials at Operating Temperatures Lower Bound Average Ultimate Yield Modulus of Strength Temperature Strength Elasticity Yield Stress (OF) (psi) (psi) (psi) (psi)

Hot Leg 593 28716 25.335 x 106 19821 74976 Cold/Crossover 545 29519 25.575 x 106 20375 74976 Leg I I I I Poisson's ratio: 0.3 Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3H4-9 4-9 Table 4-4 Chemistry & Fracture Toughness Piping Properties of the Material Heats of Fort Calhoun Unit I a,c,e 7

Material Characterization July 2013 WCAP-1 7262-NP Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-10 4et10otCahuUi Maera CLASSe Tale45heitr &FWcueSToughOUssElo NO-RPrIpeTrYie Table 4-5 Chemistry & Fracture Toughness Elbow Properties of the Material Heats of Fort Calhoun Unit I a,c,e Material Characterization July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-11 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-11 Table 4-6 Fracture Toughness Properties Used to Evaluate Critical Locations Location i 2)

(in-lb/in TraJc (non-Tmat dimensional) Jax (in-lb/in Heat Number a,c,e t I t Material Characterization July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA 5.1 CRITICAL LOCATIONS The leak-before-break (LBB) evaluation margins are to be demonstrated for the critical locations (governing locations). Such locations are established based on the loads (Section 3.0) and the material properties established in Section 4.0. These locations are defined below for the Fort Calhoun primary loop piping. The faulted loads from Table 3-2 and the weld locations from Figure 3-2 are used for this evaluation.

Critical Locations The primary loop is made of cast stainless steel for both pipes and elbows. The highest stressed locations for the primary loop are at Location 1 (in the hot leg) at the reactor vessel outlet nozzle to pipe weld and Location 6 at the crossover leg elbow weld (for the cross-over and cold leg). Locations 1 and 6 (Figure 3-2) are the critical locations for all the weld locations in the primary loop piping. Enveloping materials from all the heats (see Section 4) are used for the LBB fracture mechanics leak rate (see Section 6) and stability (see Section 7) analyses.

5.2 FRACTURE CRITERIA As will be discussed later, fracture mechanics analyses are made based on loads and postulated flaw sizes related to leakage. The stability criteria against which the calculated J and tearing modulus are compared are:

(1) IfJapp < J1, then the crack will not initiate and the crack is stable; (2) If Japp > Jjc; and Tapp < Tmat and Japp < Jmax, then the crack is stable.

Where:

Japp = Applied J J = J at Crack Initiation Tapp = Applied Tearing Modulus Tmat = Material Tearing Modulus Jmax = Maximum J value of the material For critical locations, the limit load method discussed in Section 7.0 was also used.

Critical Location and Evaluation Criteria July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6.0 LEAK RATE PREDICTIONS

6.1 INTRODUCTION

The purpose of this section is to discuss the method which is used to predict the flow through postulated through-wall cracks and present the leak rate calculation results for through-wall circumferential cracks.

6.2 GENERAL CONSIDERATIONS The flow of hot pressurized water through an opening to a lower back pressure causes flashing which can result in choking. For long channels where the ratio of the channel length, L, to hydraulic diameter, DH, (L/DH) is greater than a,c,e 6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by

]a,c,e The flow rate through a crack was calculated in the following manner. Figure 6-1 from Reference 6-2 was used to estimate the critical pressure, P,, for the primary loop enthalpy condition and an assumed flow. Once Pr was found for a given mass flow, the [

]a~c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [ ]ac.e Therefore, this method will yield the two-phase pressure drop due to momentum effects as illustrated in Figure 6-3, where P, is the operating pressure. Now using the assumed flow rate, G, the frictional pressure drop can be calculated using APf = [ ]a,c,e (6-1) where the friction factor f is determined using the [ ]ac'e The crack relative roughness, c, was obtained from fatigue crack data on stainless steel samples. The relative roughness value used in these calculations was [ ]a,c,e The frictional pressure drop using equation 6-1 is then calculated for the assumed flow rate and added to the [ ]ace to obtain the total pressure drop from the primary system to the atmosphere.

Leak Rate Predictions July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 That is, for the primary loop:

Absolute Pressure - 14.7 = [ ]a,c,e (6-2) for a given assumed flow rate G. If the right-hand side of equation 6-2 does not agree with the pressure difference between the primary loop and the atmosphere, then the procedure is repeated until equation 6-2 is satisfied to within an acceptable tolerance which in turn leads to flow rate value for a given crack size. It should be noted that for FCS, the defined atmospheric pressure is 14.2 psi rather than 14.7 psi. This difference in pressure of 0.5 psi will have negligible impact on leak-rate calculations.

6.4 LEAK RATE CALCULATIONS Leak rate calculations were made as a function of crack length at the governing locations previously identified in Section 5.1. The normal operating loads of Table 3-1 were applied, in these calculations. The crack opening areas were estimated using the method of Reference 6-3, and the leak rates were calculated using the two-phase flow formulation described above. The average material properties of Section 4.0 (see Table 4-3) were used for these calculations.

The flaw sizes to yield a leak rate of 10 gpm were calculated at the governing locations and are given in Table 6-1 for Fort Calhoun Unit 1. The flaw sizes so determined are called leakage flaw sizes.

The Fort Calhoun Unit 1 RCS pressure boundary leak detection system meets the intent of Regulatory Guide 1.45, and the plant leak detection capability is 1 gpm. Thus, to satisfy the margin of 10 on the leak rate, the flaw sizes (leakage flaw sizes) are determined which yield a leak rate of 10 gpm.

6.5 REFERENCES

6-1 a,c,e 6-2 M. M, EI-Wakil, "Nuclear Heat Transport, International Textbook Company," New York, N.Y, 1971.

6-3 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 I1-1, NUREG/CR-3464, September 1983.

6-4 D. Rudland, R. Wolterman, G. Wilkowski, R. Tregoning, "Impact of PWSCC and Current Leak Detection on Leak-Before-Break," proceedings of Conference on Vessel Head Penetration, Inspection, Cracking, Repairs, Sponsored by USNRC, Marriot Washingtonian Center, Gaithersburg, MD, September 29 to October 2, 2003.

Leak Rate Predictions July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations Location Leakage Flaw Size (in) 1 5.28 6 6.64 Reactor Vessel inlet and outlet nozzle locations have Alloy 82/182 welds. These locations are also analyzed for LBB with Alloy 82/182 material properties. Location 1 has higher faulted stress than location 14; therefore, Location 1 is selected for the LBB evaluations for the Alloy 82/182 welds.

II a,ce Leak Rate Predictions July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 a, c, e Ic STAGNATION ENTHALPY (102 tw/lb Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 a.-l w

Ir.

15 LEN42THADIAMETER RATIO MIDI Figure 6-2 [ ]ac~e Pressure Ratio as a Function of LID Leak Rate Predictions July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 a,c,e Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7.0 FRACTURE MECHANICS EVALUATION 7.1 LOCAL FAILURE MECHANISM The local mechanism of failure is primarily dominated by the crack tip behavior in terms of crack-tip blunting, initiation, extension and final crack instability. The local stability will be assumed if the crack does not initiate at all. It has been accepted that the initiation toughness measured in terms of J1, from a J-integral resistance curve is a material parameter defining the crack initiation. If,for a given load, the calculated J-integral value is shown to be less than the Jic of the material, then the crack will not initiate. If the initiation criterion is not met, one can calculate the tearing modulus (see equation A-14a of Reference 7-1) as defined by the following relation:

dJ E Tapp da x where:

Tapp = applied tearing modulus E = modulus of elasticity af 0.5 (oy + (yu) = flow stress a crack length

('y, (Yu - yield and ultimate strength of the material, respectively Stability is said to exist when ductile tearing does not occur if Tapp is less than Tmat, the experimentally determined tearing modulus. Since a constant Tmat is assumed a further restriction is placed in Japp. Japp must be less than Jmax where Jmax is the maximum value of J for which the experimental Tmat is greater than or equal to the Tapp used.

As discussed in Section 5.2 the local crack stability criteria is a two-step process:

(1) If Japp < J1c,then the crack will not initiate and the crack is stable; (2) If Japp > Jj; and Tapp < Tmat and Japp < Jmax, then the crack is stable.

Fracture Mechanics Evaluation July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 7.2 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 plastic instability method, based on traditional plastic limit load concepts, but accounting for strain hardening and taking into account the presence of a flaw. The flawed pipe 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. The flow stress is generally taken as the average of the yield and ultimate tensile strength of the material at the temperature of interest. This methodology has been shown to be applicable to ductile piping through a large number of experiments and will be used here to predict the critical flaw size in the primary coolant piping. The failure criterion has been obtained by requiring equilibrium of the section containing the flaw (Figure 7-1) when loads are applied. The detailed development is provided in Appendix A for a through-wall circumferential flaw in a pipe with internal pressure, axial force, and imposed bending moments. The limit moment for such a pipe is given by:

[ Ia,c,e where:

a,c,e Gf 0.5 (oy + ca,) = flow stress, psi I

]ace The analytical model described above accurately accounts for the piping 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 7-2). For application of the limit load methodology, the material, including consideration of the configuration, must have a sufficient ductility and ductile tearing resistance to sustain the limit load.

Fracture Mechanics Evaluation July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 7.3 RESULTS OF CRACK STABILITY EVALUATION J-Integral Analysis:

Stability analyses were performed at the governing locations established in Section 5.1. The elastic-plastic fracture mechanics (EPFM) J-integral analyses for through-wall circumferential cracks in a cylinder were performed using the procedure in the EPRI fracture mechanics handbook (Reference 7-3).

The lower-bound material properties were used. The fracture toughness properties established in Section 4.3 and the normal plus seismic loads given in Table 3-2 were used for the EPFM calculations. The postulated flaw size was 2 times (for flaw size margin of 2) the leakage flaw size established in Section 6.0 (see Table 6-1). Evaluations were performed at the critical locations identified in Section 5.1. The results of the elastic-plastic fracture mechanics J-integral evaluations are given in Table 7-1.

Limit Load Analysis:

A stability analysis based on limit load was also performed for these locations as described in Section 7.2. The weld process type, at the critical locations 1 and 6, is used as Shielded Metal Arc Welding (SMAW). The "Z" correction factor for SMAW (References 7-4 and 7-5) is as follows:

Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW where OD is the outer diameter of the pipe in inches.

The Z-factors were calculated for the critical locations, using the dimensions given in Table 3-1.

The applied faulted loads (Table 3-2) were increased by the Z-factors and plots of limit load versus crack length were generated as shown in Figures 7-2 and 7-3. Lower bound material properties were used from Table 4-3. Table 7-2 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented on the same table.

The Alloy 82/182 weld has high toughness and it does not degrade due to the thermal aging and therefore the LIMIT load method with a weld-process 'Z' factor of 1.0 should be used to calculate the critical flaw sizes. However, a 'Z' factor was conservatively applied as shown below to account for the elastic-plastic consideration of the Alloy 82/182 weld material based on Reference 7-6. The critical flaw size and the leakage flaw size for the Alloy 82/182 welds are shown at the bottom of Table 7-2.

Z = 0.0000022 x (OD)3 - 0.0002 x (OD)2 + 0.0064 x OD + 1.1355 where OD = Outside Diameter.

Fracture Mechanics Evaluation July 2013 WCAP-1 7262-NP Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4

7.4 REFERENCES

7-1 NUREG-1061 Volume 3, "Report of the U. S. Nuclear Regulatory Commission Piping Review Committee," November 1984.

7-2 Kanninen, M. F., et. al., "Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks," EPRI NP-192, September 1976.

7-3 Kumar, V., German, M. D. and Shih, C. P., "An Engineering Approach for Elastic-Plastic Fracture Analysis," EPRI Report NP-1931, Project 1237-1, Electric Power Research Institute, July 1981.

7-4 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Registerfol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

7-5 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

7-6 ASME Pressure Vessel and Piping Division Conference Paper PVP2008-61840, "Technical Basis for Revision to Section XI Appendix C for Alloy 600/82/182/132 Flaw Evaluation in Both PWR and BWR Environments," July 28-31, Chicago IL,USA.

Fracture Mechanics Evaluation July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 Table 7-1 Stability Results for Fort Calhoun Based on Elastic-Plastic J-Integral Evaluations Fracture Criteria Calculated Values Flaw Jic Jmax Japp Location Size* (in) (in-lb/in 2 ) Tmat (in-lb/in 2 ) (in-lb/in 2 ) Tapp a,c,e K

Table 7-2 Stability Results for Fort Calhoun Based on Limit Load Critical Location Critical Flaw Size (in) Leakage Flaw Size (in) 1 35.27 5.28 6 34.16 6.64 I

]a,c,e Fracture Mechanics Evaluation July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3

• v ON-POPRITAR WESTNGHOSE CLAS 37-6 of Figure 7-1 Fully Plastic Stress Distribution Fracture Mechanics Evaluation July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-7 a,c,e OD = 38.50 in. Uy-min = 19.82 ksi F = 1964 kips t = 2.688 in. Ou-min = 74.98 ksi M = 23806 in-kips SA451-CPF8M with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-2 Critical Flaw Size Prediction - Location I Fracture Mechanics Evaluation July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8 a,c,e OD = 29.00 in. Gy-min = 20.38 ksi F = 1075 kips t = 1.938 in. ou-min = 74.98 ksi M = 4368 in-kips SA-351 CF8M with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-3 Critical Flaw Size Prediction - Location 6 Fracture Mechanics Evaluation July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8.0 FATIGUE CRACK GROWTH ANALYSIS To determine the sensitivity of the primary coolant system to the presence of small cracks, a fatigue crack growth analysis was carried out at the reactor vessel outlet nozzle safe-end region for Fort Calhoun (see Location 1 in Figure 3-2). This region was selected because crack growth calculated here will be typical of that in the entire primary loop; crack growths calculated at other locations can be expected to show less than 10% variation. Fatigue crack growth is the only credible crack growth mechanism.

A finite element stress analysis was carried out for the outlet nozzle safe end region with representative nozzle dimensions to determine the stresses resulting from thermal transients and mechanical loading. These stresses were then combined with conservative welding residual stress distributions at the stainless steel and Alloy 82/182 welds to determine the fatigue crack growth for postulated flaws in the various materials at the nozzle safe end region.

Table 8-1 summarizes the transients and design cycles applicable for Fort Calhoun Unit 1 (Reference 8-2). Extended Power Uprate (EPU) and Measurement Uncertainty Recapture (MUR) transients were not considered in Revision 1 of this report.

Circumferentially oriented surface flaws were postulated at three different locations at the outlet nozzle, these locations are: safe end to pipe stainless steel weld, safe end to nozzle Alloy 82/182 weld, and the ferritic steel nozzle (Figure 8-1). The total stress at each region was used to generate crack tip stress intensity factors (K1) for circumferential flaw Aspect Ratios -AR (flaw length/flaw depth) of 2, 6, and 10. Fatigue crack growth analyses were performed for the postulated flaws with initial flaw depths of 10% through-wall thickness.

Fatigue crack growth rate laws were used from the ASME Code Section Xl (Reference 8-3) for the ferritic steel and stainless steel as shown in Figures 8-2 and 8-3 respectively. The fatigue crack growth law for the safe end to nozzle region (Alloy 82/182) was derived from Reference 8-1 and illustrated in Figure 8-4. The laws were all structured for applicability to pressurized water reactor environments.

The calculated fatigue crack growth results for the semi-elliptic surface flaws of circumferential orientation and various depths are summarized in Tables 8-2 through 8-4, and the results show that the crack growth is very small. Therefore, it is concluded that surface flaws will not become through-wall flaws for the end of service life of the plant and also applicable for the period of extended operation (Section 4.3.1 of Reference 8-4 has indicated that OPPD does not expect the number of design cycles for the transients that are counted to be exceeded during the period of extended operation).

Fatigue Crack Growth Analysis July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-2

8.1 REFERENCES

8-1 NUREG/CR-6721, ANL-01/07, "Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds," April 2001.

8-2 Fort Calhoun Station Updated Safety Analysis Report (USAR-4.2), Revision 12.

8-3 ASME Boiler and Pressure Vessel Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition.

8-4 NUREG-1782, "Safety Evaluation Report Related to the License Renewal of the Fort Calhoun Station, Unit 1," Docket No. 50-285.

Fatigue Crack Growth Analysis July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3 Table 8-1 Reactor Coolant System Operating Transients Number Transient Identification Design Cycles 1 Plant Heatup at 100°F/hr 500 2 Plant Cooldown at 100°F/hr 500 3 Plant Loading at 10% of Full Power/min 15000 4 Plant Unloading at 10% of Full Power/min 15000 5 Step Load Increase of 10% of Full Power 2000 6 Step Load Decrease of 10% of Full Power 2000 7 Normal Plant Variation* 106 8 Reactor Trip 400 9 Loss of Reactor Coolant Flow 40 10 Abnormal Loss of Load 40 11 Loss of Secondary Pressure 5 12 Hydrostatic Test, 3125 psia 10 13 Plant Leak Test, 2250 psia 200

• The RCS average temperature for purpose of design is assumed to increase and decrease a maximum of 6°F in 1 minute. The corresponding RCS pressure variation is less than 100 psi. This transient has an insignificant effect on fatigue crack growth and was not included in this evaluation.

Fatigue Crack Growth Analysis July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4 Table 8-2 FCG at Alloy 82/182 Weld (Nozzle to Safe-end Weld) - Outlet Nozzle Thickness = 3.031 in., Inside Radius = 16.219 in.

Initial Flaw Size = 10% Through Original Wall Thickness Initial Flaw Final Flaw Depth at Final Flaw Ratio (alt)

Flaw Configuration Depth (in.) End of Service Life at End of Service (in.) Life a,c,e m

L Table 8-3 FCG at Stainless Steel Weld (Safe-end to Pipe Weld) - Outlet Nozzle Thickness = 2.688 in., Inside Radius = 16.563 in.

Initial Flaw Size = 10% Through Original Wall Thickness Initial Flaw Final Flaw Depth at Final Flaw Ratio (alt)

[

Flaw Configuration Depth (in.) End of Service Life (in.)

at End of Service Life 1I a,c,e Table 8-4 FCG at Reactor Vessel Nozzle Location - Outlet Nozzle Thickness = 3.031 in., Inside Radius = 16.219 in.

Initial Flaw Size = 10% Through Original Wall Thickness Initial Flaw Final Flaw Depth at Final Flaw Ratio (a/t)

Flaw Configuration Depth (in.) End of Service Life at End of Service Life (in.)

,ce

__ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _1

[

m I Fatigue Crack Growth Analysis July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Figure 8-1 Reactor Vessel Outlet Nozzle with Stress Cut Locations Fatigue Crack Growth Analysis July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-6 1000 700 500 200 S 100 0

U 0

U 70 C

0 b

50 0

b 0

20 0

U 0

c-i 10 7

5 *Linear interpolation is recommended to account for R ratio dependence of water environment curves, for 0.25 < R < 0.65 for steep slope:

5 9 5 da = (1.02 X 10"6) 01 AK "

dN 1

2 5 7 10 20 50 70 100 Stress Intensity Factor Range (AK 1 ksi 41'.)

Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels Fatigue Crack Growth Analysis July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 3 X 10-4 10o-4 10,5 T

C b*

10.6 10-7 6 10 20 50 100 AK (ksi ')

Note: A Factor of 2.0 is applied to the Air Environment Curve to Represent crack growth rate in PWR Environment Figure 8-3 Reference Fatigue Crack Growth Curves for Stainless Steels Fatigue Crack Growth Analysis July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-8 10-6 EE 1 -1l I/ , . . J .... ...... 10-111/ " . ,. ... j . .... J 10.11 10-10 10-9 10-8 10-7 10-6 10-11 10.10 10-7 10 .g 10-8 10-5 CGRair (MIs) CGRuir (r/s)

Alloy 600

- C-4.M34e-14+1.6216e-i6 T

-1,4896e-1BT 2 +t4.3546e-21 T3 10-13 L),

o ANL 0 James Ix Amnzallag et al.

L Naganto et at 1It U- ~.&..&.I..4.. .I.4..~..

0 100 200 300 400 500 600 Temperature (°C)

Figure 8-4 Reference Fatigue Crack Growth Curves for Alloy 821182 Welds Fatigue Crack Growth Analysis July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9.0 ASSESSMENT OF MARGINS The results of the leak rates of Section 6.4 and the corresponding stability evaluations of Section 7.3 are used in performing the assessment of margins. Margins are shown in Tables 9-1 and 9-2. All of the LBB recommended margins are satisfied.

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 plant leak detection capability of 1 gpm according to FCS Technical Specification 2.1.4 (Reference 9-1).

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). A margin of 1 on loads using the absolute summation of faulted load combinations is satisfied.

9.1 REFERENCES

9-1 Omaha Public Power District, Fort Calhoun Station Unit 1, Renewed Facility Operating License No. DPR-40, Appendix A, Technical Specification 2.1.4, "Reactor Coolant System Leakage Limits."

Assessment of Margins July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes by Limit Load and Margins for Fort Calhoun Critical Critical Flaw Size* Leakage Flaw Size Margin 1 35.27 5.28 6.7 6 34.16 6.64 5.1

• Linmit Load Method I

a,c,e Table 9-2 Stability Results for Fort Calhoun Based on Elastic-Plastic J-Integral Evaluations Fracture Criteria Calculated Values Critical Flaw Jic Jmax Japp Location Size* (in) (in-lblin 2 ) Tmat (in-lb/in 2 ) (in-lb/in 2) Tapp a,c,e

.-- II I-L i-i Assessment of Margins July 2013 WCAP-17262-NP Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1

10.0 CONCLUSION

S This report justifies the elimination of RCS primary loop pipe breaks from the structural design basis for the Fort Calhoun Unit 1 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.

Mitigation measures for PWSCC sensitive Alloy 82/182 welds will be implemented as required on the Fort Calhoun Reactor Vessel nozzles weld locations.

b. Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.

c. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.

d. Ample margin exists between the leak rate of small stable flaws and the capability of the Fort Calhoun reactor coolant system pressure boundary Leakage Detection System.

e. Ample margin exists between the small stable flaw sizes of item (d) and larger stable flaws.

f. Ample margin exists in the material properties used to demonstrate end-of-service life (fully aged) stability of the critical flaws.

For the critical locations, flaws are identified that will be stable because of the ample margins described in d, e, and f above. The LBB assessment was performed based on deterministic fracture mechanics methods that have been previously approved by the NRC.

Based on the above, the leak-before-break conditions and margins are satisfied for the Fort Calhoun Unit 1 primary loop piping. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the reactor coolant primary loop piping need not be considered in the structural design basis for Fort Calhoun Unit 1 for the license renewal period with current power operating conditions.

Conclusions July 2013 WCAP-1 7262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A: LIMIT MOMENT I

a,c,e Appendix A: Limit Moment July 2013 WCAP-17262-NP Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment July 2013 WCAP-1 7262-NP Revision 1

Revise the structural design basis for reactor coolant system (RCS) piping described in USAR Section 4.3.6.

NRC and industry initiatives using fracture mechanics technology and material toughness demonstrate that leak-before-break (LBB) criteria can be applied to RCS piping.

Eliminate RCS piping rupture from consideration in the FCS structural design basis LBB methodology performed by Westinghouse using the latest LBB criteria.

The analysis was performed using the existing RCS leak detection capability and piping stress analysis loads. Thermal aging considerations were performed using approved Westinghouse methodology.

The analysis is applicable for the period of extended operation (PEO), which begins at midnight on August 9, 2013.

No Technical Specification (TS) changes are proposed in this license amendment request (LAR).