ML19144A089

TLR-Criteria for Postulating Pipe Rupture Locations Background and History

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Issue date:	05/30/2019
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Criteria for Postulating Pipe Rupture Locations:

Background and History Technical Letter Report Michael Benson U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research May 2019

Executive Summary General Design Criterion (GDC) 4, Environmental and Dynamic Effects Design Bases, of Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, requires that nuclear plant components important to safety be designed to perform their intended functions under accident conditions. Branch Technical Position (BTP) 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, describes methods acceptable to the U.S. Nuclear Regulatory Commission (NRC) staff for demonstrating compliance with GDC 4. Among other principles, BTP 3-4 contains criteria related to the cyclic stress range and cumulative usage factor (CUF). If a given location in a pipe segment exceeds the criteria, the applicant must postulate a pipe break in that segment to comply with BTP 3-4. The applicant must then analyze the effects of the hypothetical rupture on surrounding components. However, stakeholders have questioned the technical basis behind the staffs positions in BTP 3-4. The purpose of this document is to record what is known about the stress range and CUF criteria in BTP 3-4 as part of a larger effort to reexamine these regulatory positions.

With this goal in mind, this document explores the historical development of BTP 3-4. The author found five versions of BTP 3-4 issued between September 1975 and December 2016.

This document describes the evolution of the stress range and CUF criteria over time. For example, the original BTP 3-4, issued in September 1975, indicated that only the CUF criterion was to be considered for higher stress ranges. In 1981, the staff revised the CUF criterion to be independent of stress range, perhaps to recognize that the calculation of CUF accounts for the effects of stress range on fatigue life. This historical background of BTP 3-4 may help the NRC staff in its efforts to explore, and perhaps improve upon, the technical basis behind the staffs positions.

In particular, stakeholders have criticized the CUF criterion as being overly conservative and technically unsupported. Stakeholders have argued that operating experience does not support the need for a CUF criterion. Therefore, they imply that new criteria based on other damage mechanisms may be needed and that the CUF criterion should be eliminated from BTP 3-4.

Stakeholders have also proposed probabilistic and risk-informed approaches as a way to investigate the efficacy of the CUF criterion. These approaches involve various assumptions to arrive at an estimated core damage frequency. For example, analysts often assume that leak probability is equal to rupture probability. The NRC staff should carefully consider the effect that these assumptions have on the conclusions of probabilistic studies.

A final complicating factor in the discussion of the CUF criterion is environmentally assisted fatigue (EAF). The fatigue design curves in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code do not account for the effect of the reactor coolant environment on fatigue life. License renewal commitments, new reactor design certifications and combined licenses, and NRC technical documents govern EAF. As part of the Economic Simplified Boiling Water Reactor (ESBWR) design certification, the applicant proposed to relax the CUF criterion in BTP 3-4 for cases in which the design accounts for EAF. Although the applicant offered no explicit basis for this approach, the implication is that the original CUF criterion was developed before EAF was understood. Therefore, accounting for EAF may make it especially onerous. Although the NRC staff accepted the industrys proposal, a documented basis for either the staffs original position or the relaxation of that position for the ESBWR is not i

clear. This action prompted the NRC staff to revise BTP 3-4 in 2016 to communicate its willingness to relax the CUF criterion in cases that account for EAF.

This document describes the concepts introduced above in greater detail and makes recommendations for future work in this area. The Extremely Low Probability of Rupture probabilistic fracture mechanics code could potentially be used to analyze rupture probabilities in a manner similar to that suggested by stakeholders that have criticized the CUF criterion.

Deterministic fatigue crack growth calculations may provide insight as well. Potential outcomes from this suggested future work include bolstering the technical basis of the current criteria or providing justification to update the existing criteria.

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Table of Contents Executive Summary ....................................................................................................................... i List of Tables................................................................................................................................ iv

1. Introduction ............................................................................................................................. 1

2. Branch Technical Position....................................................................................................... 2 2.1 Introduction to Branch Technical Position 3-4 ................................................................ 2 2.2 Summary ........................................................................................................................ 4

3. Historical Development of Branch Technical Position 3-4 ...................................................... 5 3.1 The Giambusso Letter .................................................................................................... 5 3.2 The OLeary Letter ......................................................................................................... 5 3.3 Revision History ............................................................................................................. 5 3.4 MEB 3-1, Revision 0....................................................................................................... 6 3.5 MEB 3-1, Revision 1....................................................................................................... 6 3.6 MEB 3-1, Revision 2....................................................................................................... 8 3.7 BTP 3-4, Revisions 2 and 3............................................................................................ 8 3.8 Summary of BTP 3-4 Historical Development ................................................................ 8

4. Cumulative Usage Factor Criterion ....................................................................................... 10 4.1 Known Technical Basis ................................................................................................ 10 4.2 American National Standards Institute/American Nuclear Society 58.2-1988 ............. 11 4.3 The Rodabaugh Approach ........................................................................................... 11 4.4 Electric Power Research Institute Technical Report on Postulating Break Locations.. 12 4.5 Gosselin and Simonens Approach .............................................................................. 13 4.6 Economic Simplified Boiling-Water Reactor Design Certification and Environmental Fatigue Effects ............................................................................................................. 14 4.7 Summary ...................................................................................................................... 15

5. Conclusions and Recommendations..................................................................................... 16 5.1 Summary ...................................................................................................................... 16 5.2 Future Work.................................................................................................................. 16

6. References ............................................................................................................................ 18 iii

List of Tables Table 1 Criteria under which Breaks Are Postulated Near Containment Penetrations ................ 2 Table 2 Criteria under which Breaks Are Postulated Away from Containment Penetrations ...... 3 Table 3 Revision History of BTP 3-4 ............................................................................................ 6 Table 4 Summary of BTP 3-4 Historical Development ................................................................ 9 iv

1. Introduction General Design Criterion (GDC) 4, Environmental and Dynamic Effects Design Bases, of Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities [1], states, in part, that components important to safety shall be designed to accommodate the effects ofpostulated accidents.

This means that safety-related equipment must be capable of performing their intended functions under accident conditions. Potential adverse conditions during a loss-of-coolant accident (LOCA) include missile generation, flooding, pipe whip, increased temperature and humidity, and discharging fluid jet impingement. To demonstrate compliance with GDC 4, applicants must postulate a number of pipe ruptures and assess the effects of the hypothetical scenario on surrounding equipment. A possible outcome of this analysis may be the installation of pipe whip restraints or jet impingement shields to protect nuclear plant components important to safety.

Section 3.6.3, Leak-Before-Break Evaluation Procedures [2], of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), describes the leak-before-break (LBB) methodology for pressurized-water reactors.

Here, applicants may demonstrate that the probability of rupture is extremely low, as mentioned in GDC 4, and thereby forego assessing the dynamic effects of a pipe rupture for that particular system. The U.S. Nuclear Regulatory Commission (NRC) staff accepts the LBB methodology as a basis for not installing (or, removing, if already installed) pipe whip restraints and jet impingement shields. Removing existing devices can provide a safety benefit by increasing access to equipment for nondestructive examination purposes.

SRP Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping [3], describes methods acceptable to the staff for complying with GDC 4 when LBB does not apply. Specifically, SRP Section 3.6.2 describes the NRC staffs positions on pipe rupture criteria and various dynamic stress analysis methods. Branch Technical Position (BTP) 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Revision 3, issued December 2016 [4], discusses pipe rupture criteria in detail.

NRC stakeholders have questioned the technical basis behind BTP 3-4 ([5]-[6]) The NRC staff recently updated BTP 3-4 to make it consistent with approved industry proposals for cases in which the cumulative usage factor (CUF) calculation accounts for environmentally assisted fatigue (EAF) ([7]-[9]). The purpose of this report is to discuss the pipe rupture criteria in BTP 3-4, including the historical development and documented basis for the criteria.

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2. Branch Technical Position 2.1 Introduction to Branch Technical Position 3-4 BTP 3-4, Section B.1, is titled High-Energy Fluid Systems Piping. High energy is defined as systems where temperature is greater than 93 degrees Celsius (C) (200 degrees Fahrenheit (F)) and gauge pressure is greater than 1.9 MPa (275 pounds per square inch, gauge) [10].

This document focuses solely on the pipe break criteria related to stress and fatigue. In addition, BTP 3-4, Sections B(1)(ii)2-B(1)(ii)7, cover piping system design and inspection criteria for excluding postulated high-energy line breaks in the high-energy piping systems near the containment penetration region. These other criteria within BTP 3-4, including the staffs positions on moderate-energy systems in BTP 3-4, Section B.2, Moderate-Energy Fluid System Piping, are outside the scope of this document. Table 1 and Table 2 summarize the break criteria discussed in this section, as expressed in BTP 3-4, Revision 3.

Table 1 Criteria under which Breaks Are Postulated Near Containment Penetrations Section of BTP 3-4 Criteria Notes Class 1 Piping B.1.(ii)(1)(a) If Sn > 2.4Sm, then Se > 2.4Sm or See Equations 2, 3, and 4.

Sn > 2.4Sm B.1.(ii)(1)(b) Cumulative Usage Factor (CUF) In 2016, the NRC staff added, 0.1 For new reactor design certification reviews, the staff has considered a CUF limit of 0.4 to be acceptable when the effects of environmental assited fatigue (EAF) are considered in the piping design.

B.1.(ii)(1)(c) S > 2.25Sm or S > 1.8Sy - BTP 3-4 provides exceptions to these criteria for the loads associated with pipe failure outside containment.

- See Equation 1.

Class 2 Piping B.1.(ii)(1)(d) S + Sn > 0.8(1.8Sh + SA) - Sh and SA are allowable stresses defined in ASME Section III, paragraph NC-3600 [11].

- See Equations 1 and 2.

B.1.(ii)(1)(e) S > 2.25Sh or S>1.8Sy - See Equation 1.

- The same exceptions discussed in B.1.(ii)(1)(c) apply here.

• Sm is the allowable design stress intensity defined in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section II, Part D, Subpart 1, Tables 2A and 2B [12].

• • Sy is the yield stress.

+Section designations refer to BTP 3-4, Revision 3, only. Other versions may have differing conventions.

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Table 2 Criteria under which Breaks Are Postulated Away from Containment Penetrations Section of BTP 3-4 Criteria Notes Class 1 Piping B.1.(iii)(1)(a) At terminal ends None B.1.(iii)(1)(b) Sn > 2.4Sm and [Se > 2.4Sm or See Equations 2, 3, and 4.

Sn > 2.4Sm]

B.1.(iii)(1)(c) Cumulative Usage Factor (CUF) In 2016, the NRC staff added, 0.1 For new reactor design certification reviews, the staff has considered a CUF limit of 0.4 to be acceptable when the effects of EAF are considered in the piping design.

Class 2 and 3 Piping B.1.(iii)(2)(a) At terminal ends None B.1.(iii)(2)(b)(ii) SOL + SE > 0.8(1.8Sh+SA) - Sh and SA are allowable stresses defined in ASME Section III, paragraph NC-3652 and ND-3653 [11].

- See Equations 5 and 6.

• Sm is the allowable design stress intensity defined in ASME Code Section II, Part D, Subpart 1, Tables 2A and 2B

[12].

• • Sy is the yield stress.

+Section designations refer to BTP 3-4, Revision 3, only. Other versions may have differing conventions.

++Terminal ends are defined as the extremities of piping runs that connect to structures, components, or pipe anchors (see footnote 3 to BTP 3-4).

Table 1 references the quantities listed below from ASME Code Section III, paragraph NB-3652 and paragraph NB-3653 [11].

PDo D S = B1 + B2 o M i Equation 1 2t 2I where S is the stress intensity at the analyzed location, B1 and B2 are the primary stress indices, P is the design pressure, Do is the outer diameter of the pipe, t is the nominal wall thickness, I is the moment of inertia, and Mi is the resultant moment.

P0 Do D S n = C1 + C 2 o M i + C3 E ab x a Ta bTb Equation 2 2t 2I where Sn is stress intensity range for the load set under consideration; C1, C2, and C3 are secondary stress indices; Eab is the average elastic modulus; Mi is the resultant range of moment, excluding thermal expansion and thermal anchor movements; P0 is the range of service pressure; Ta (Tb) is the range of average temperature on side a (b) of a structural or material discontinuity; and a (b) is the coefficient of thermal expansion on side a (b) of a structural or material discontinuity.

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Do

• S e = C2 Mi Equation 3 2I where Se is the stress intensity range for loading conditions specified for Mi*, and Mi* is the same as Mi, with the exception that it only includes moments caused by thermal expansion and thermal anchor movements.

P0 Do D S n = C1 + C 2 o M i + C3 E ab x a Ta bTb Equation 4 2t 2I where Sn is the stress intensity range for primary plus secondary loading, and C3 is given as tabulated values in ASME Code Section III, paragraph NB-3681.

Table 2 references the quantities listed below from ASME Code Section III, NC-3653 [11].

Pmax Do M + MB S OL = B1 + B2 A Equation 5 2t n Z iM c SE = Equation 6 Z

where SOL is the stress intensity range caused by occasional loads, SE is the stress intensity range caused by thermal expansion loads, Pmax is the peak pressure, tn is the nominal wall thickness, MA is the resultant moment caused by weight and other sustained loads, MB is the moment caused by occasional loads, Mc is the moment caused by thermal expansion loads, and Z is the section modulus of the pipe.

2.2 Summary This section introduced the relevant stress and fatigue criteria in the current version of BTP 3-4.

For the readers reference, the discussion included the ASME Code Section III design equations referenced by BTP 3-4. Although BTP 3-4 is written in English without mathematical symbols, the criteria were presented here mathematically as inequalities. The criteria expressed here were written as the conditions under which breaks should be postulated. The actual language in BTP 3-4 expresses conditions under which breaks should be postulated and should not be postulated, depending on the section (e.g., contrast BTP 3-4, Section B.1.(ii)(1), with Section B.1.(iii)(1)). The remaining chapters of this document discuss the known historical, technical, and regulatory backgrounds behind these criteria.

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3. Historical Development of Branch Technical Position 3-4 3.1 The Giambusso Letter BTP 3-3, Protection against Postulated Piping Failures in Fluid Systems Outside Containment, Revision 3, issued March 2007 [13], contains the publicly available portions of a letter from A. Giambusso to NRC applicants on the subject of postulated piping failures outside containment (see Appendix A of this report). This letter was dated December 1972, before the Atomic Energy Commission split into the NRC and the Energy Research and Development Administration according to the Energy Reorganization Act of 1974 [14]. The criteria listed in Giambussos letter, including stress range and CUF criteria, are very similar to those found in BTP 3-4. Therefore, this document is a precursor to BTP 3-4.

3.2 The OLeary Letter OLeary letter, dated July 12, 1973, is a followup letter to Giambussos letter (see Appendix C of BTP 3-3 [13]). This letter clarified the Atomic Energy Commissions expectations for new plants with regard to compliance with GDC 4. Much of the guidance provided in this letter is related to optimizing the physical plant layout to ensure that safety-related components can perform their intended functions under accident conditions. Appendix A to OLearys letter contains criteria for postulating break locations. For this reason, this letter is also a precursor to BTP 3-4.

3.3 Revision History The first historical version of BTP 3-4 appears in Section 3.6.2, Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, of NUREG-75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Revision 0, issued September 1975 ([15]). At that time, the cognizant NRC organization, which was the Mechanical Engineering Branch (MEB), designated BTPs. For this reason, MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, describes the staffs positions that would later become BTP 3-4.

Table 3 lists several versions of BTP 3-4 ([15]-[18],[4]).

The NRC revised MEB 3-1 twice between 1975 and 1987. The author did not locate any evidence of intermediate versions between MEB 3-1, Revision 2, issued June 1987, and BTP 3-4, Revision 2, issued March 2007. BTP 3-4, Revision 3, issued December 2016, is the latest revision at the time of publication of this document. This section summarizes certain features of the various revisions of MEB 3-1 and BTP 3-4. This discussion will primarily explore how the stress- and fatigue-related criteria have evolved over time. For this reason, this chapter does not provide a comprehensive discussion of all changes to BTP 3-4. The comparisons drawn in this section are only for historical context and demonstrate the evolution of BTP 3-4.

Section 3.8 summarizes the historical development of BTP 3-4.

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Table 3 Revision History of BTP 3-4 3.4 MEB 3-1, Revision 0 MEB 3-1 was originally written for break criteria outside containment only, although the current version of BTP 3-4 applies to inside and outside containment. It references SRP Section 3.6.1, Plant Design for Protection against Postulated Piping Failures in Fluid Systems Outside Containment, which was the responsibility of the Auxiliary and Power Conversion Systems Branch (APCSB). APCSB 3-1 [15] (now BTP 3-3 [13]) discusses various plant design features that, when followed, demonstrate compliance with GDC 4. APCSB 3-1, Section B.2.c.(1),

specifically references MEB 3-1 for stress limits. A reference to BTP 3-4 stress limits no longer exists in the current version of BTP 3-3. In the current version of BTP 3-3, Section B.3.(a) states that BTP 3-4 only applies during normal conditions, not seismic events.

Although they are similar, the stress and fatigue criteria in MEB 3-1 for high-energy Class 1 piping near containment penetrations, differ from those found in the current version of BTP 3-4.

For example, MEB 3-1, Revision 0, Section B.1.b.(1)(b), states that the CUF criterion of 0.1 applies only if the 2.4Sm limit is violated. MEB 3-1, Revision 0, Section B.1.b.(1)(c), where the CUF criterion first appears, reinforces this notion by stating that the applicant should not consider the CUF criterion unless MEB 3-1, Revision 0, Section B.1.b.(1)(b), directs it to do so.

3.5 MEB 3-1, Revision 1 MEB 3-1, Revision 1, issued July 1981, modifies and expands upon Section A: Background.

In this section, the NRC staff first described pipe ruptures as being rare event(s) which may only occur under unanticipated conditions. The NRC states that piping failures generally occur at high stress and fatigue locations. These statements may provide some context to the staffs rationale behind the regulatory positions described in the document.

In MEB 3-1, Revision 1, the staff modified the content of MEB 3-1, Revision 0, Section B.1.b(1).

A portion of MEB 3-1, Revision 0, Section B.1.b(1), reads as follows:

(a) The maximum stress range should not exceed 2.4Sm (b) The maximum stress range between any two load sets (including the zero load set) should be calculated by Eq. (10) in Paragraph NB-3653, ASME Code,Section III, for normal and upset plant conditions and an operating basis earthquake (OBE) event transient.

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If the calculated maximum stress range of Eq. (10) exceeds the limit of B.1.b(1)(a) but is not greater than 3 Sm, the limit of B.1.b(1)(c) should be met.

If the calculated maximum stress range of Eq. (10) exceeds 3 Sm, the stress ranges calculated by both Eq. (12) and Eq. (13) in Paragraph NB-3653 should meet the limit of B.1.b(1)(a) and the limit of B.1.b(1)(c).

(c) The cumulative usage factor should be less than 0.1 if consideration of fatigue limits is required according to B.1.b(1)(b).

The NRC staff combined Sections (a) and (b) above in MEB 3-1, Revision 1. The CUF criteria of MEB 3-1, Section B.1.b(1)(c), which were conditional in MEB 3-1, Revision 0, became unconditionally applicable in MEB 3-1, Revision 1. Hence, the NRC needed to modify the language in Sections (a) and (b). The criteria with regard to ASME Code Section III, Equations (12) and (13) (Equations 3 and 4 of this report, respectively), were made more restrictive by lowering the threshold where they apply. With these changes, this portion of MEB 3-1, Revision 1, Section B.1.b, reads as follows:

(a) The maximum stress range between any two load sets (including the zero load set) should not exceed 2.4Sm and should be calculated by Eq. (10) in Paragraph NB-3653, ASME Code,Section III, for those loads and conditions thereof for which level A and level B stress limits have been specified in the systems Design Specification, including an operating basis earthquake (OBE) event transient. The Sm is design stress intensity as defined in Article NB-3600 of the ASME Code Section III.

If the calculated maximum stress range of Eq. (10) exceeds 2.4Sm, the stress ranges calculated by both Eq. (12) and Eq. (13) is Paragraph NB-3653 should meet the limit of 2.4 Sm.

(b) The cumulative usage factor should be less than 0.1.

MEB 3-1, Revision 1, is the first document in which a separate set of criteria for areas away from containment penetrations appears. These early criteria are similar to those found in the current version of BTP 3-4, such as at terminal ends, for locations where S > 2.4Sm, and for locations where CUF is greater than 0.1. MEB 3-1, Revision 1, Section B.1.c.(1)(d), states that the highest stressed locations should be considered break locations if the Sm and CUF criteria are always satisfied for a given piping run:

(d) If two intermediate locations cannot be determined by (b) and (c) above, two highest stress locations based on Eq. (1) should be selected. If the piping run has only one change or no change of direction, only one intermediate location should be postulated.

MEB 3-1, Revision 0, does not provide criteria for these so-called arbitrary break locations.

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3.6 MEB 3-1, Revision 2 MEB 3-1, Revision 2, was released to the public under the cover of Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, dated June 19, 1987 [17].

This modification removed the MEB 3-1, Revision 1, Section B.1.c.(1)(d), criteria discussed in Section 3.5. However, a few other changes from MEB 3-1, Revision 1, to MEB 3-1, Revision 2, are noteworthy.

The NRC staff changed the maximum stress criteria for Class 1 piping in MEB 3-1, Revision 1, Section B.1.b.(1)(c), from S > 2.25Sm to S > min(2.25Sm,1.8Sy). The staff made the same change to MEB 3-1, Revision 1, Section B.1.b.(1)(e). The staff altered the stress range criteria in MEB 3-1, Revision 1, Section B.1.b.(1)(d), from S+Sn > 0.8(1.2Sh + SA) to S+Sn > 0.8(1.8Sh + SA).

3.7 BTP 3-4, Revisions 2 and 3 The author found no evidence of a Revision 0 or Revision 1 of BTP 3-4. As Table 3 suggests, roughly 20 years separates MEB 3-1, Revision 2, and BTP 3-4, Revision 2 [18]. BTP 3-4, Revision 2, contains an expanded background section that discusses the LBB, new reactor design certifications, and replacement earthquake loading. In addition to certain formatting changes, the staff changed the S > min(2.25Sm,1.8Sy) criterion to S > 2.25Sm and S > 1.8Sy.

Revision 3 of BTP 3-4 [4] is the final version currently in use as of the writing of this report. In this version, the staff expanded upon the CUF criteria in BTP 3-4, Revision 2, Sections B.1.(ii)(1)(b) and B.1.(iii)(1)(c), as follows:

(b) The cumulative usage factor (CUF) should be less than 0.1. For new reactor design certification reviews, the staff has considered a CUF limit of 0.4 to be acceptable when the effects of environmental assisted fatigue (EAF) are considered in the piping design.

As discussed in Section 4.6, this statement was added as a result of the precedent set by the NRC staffs review of the ESBWR design certification application ([8]-[9]). The staff also implemented additional formatting improvements in this version.

3.8 Summary of BTP 3-4 Historical Development This section has shown that the staffs positions on pipe break criteria in BTP 3-4 have evolved over time. In many cases, the staffs rationale behind the updates to BTP 3-4 is not currently known. Even so, the history behind BTP 3-4 may prove useful when considering further improvements to the document. Table 4 summarizes the major changes to BTP 3-4 for the stress- and CUF-based pipe break criteria.

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Table 4 Summary of BTP 3-4 Historical Development Section in BTP 3-4, Rev. 3 MEB 3-1, Rev. 0 MEB 3-1, Rev. 1 MEB 3-1, Rev. 2 BTP 3-4, Rev. 2 BTP 3-4, Rev. 3 Expanded Background section Section A: Background Relatively short introduction No change No change No change Added a discussion of pipe failures occurring at high stress and fatigue locations if 2.4Sm < Sn 3Sm, then CUF 0.1 if Sn > 2.4Sm, then Se > 2.4Sm Added statement and Sn >2.4Sm concerning CUF when EAF B.1(ii)(1)(a)-(b) if Sn > 3Sm, then Se > 2.4Sm No change No change is considered in design and Sn >2.4Sm CUF 0.1 certification applications and CUF 0.1 B.1(ii)(1)(c) S > 2.25Sm No change S > min(2.25Sm, 1.8Sy) S > 2.25Sm and S > 1.8Sy No change B.1(ii)(1)(d) S+Sn > 0.8(1.2Sh + SA) No change S+Sn > 0.8(1.8Sh+SA) No change No change B.1(ii)(1)(e) S > 1.8Sh No change S > min(2.25Sh, 1.8Sy) S > 2.25Sh and S > 1.8Sy No change Essentially, no At terminal ends for Class 2 and 3 changes were made.

Extended to Class 1 piping Clarified the discussion of piping and non-Class piping B.1(iii)(1)(a) non-Class piping to apply enclosed in protective structures or Recommendations for No change B.1(iii)(2)(a) Removed caveat on to seismically analyzed located adjacent to protective non-Class piping now protective structures non-ASME Class piping.

structures refer to Class 2 and 3 recommendations 9

4. Cumulative Usage Factor Criterion BTP 3-4, Sections B.1.(ii)(1)(b) and B.1.(iii)(1)(c), provide the CUF criterion. Various stakeholders have questioned the technical basis of this regulatory position ([5]-[6]).

Complicating factors include accounting for EAF in CUF calculations and determining how this criterion will be applied in license renewal and subsequent license renewal space. An overly conservative CUF criterion may be of particular concern for boiling-water reactor licensees seeking subsequent license renewal. Pressurized-water reactors have many piping systems approved for LBB, as briefly explained in Chapter 1. Approval for LBB means that the licensee does not have to analyze for dynamic effects of a pipe break, such as pipe whip. However, line breaks still must be analyzed as a design-basis accident, as explained in Volume 52 of the Federal Register, pages 41288-41294 (52 FR 41288-41294; October 27, 1987) [10], even with LBB approval. The safety evaluation report on license renewal application for the Edwin I.

Hatch Nuclear Plant [19] illustrates the NRC staffs position on break postulations (i.e., postulation of pipe breaks under the CUF criterion is a time-limited aging analysis because of the required CUF calculation). However, even if a boiling-water reactor cannot meet the CUF break criterion under subsequent license renewal, applicants still have the option to manage the effects of aging under 10 CFR 54.21(c)(1)(iii) and Aging Management Program X.M, Fatigue Monitoring. This section of the report documents the known background behind the CUF criterion and the technical concerns raised by stakeholders.

4.1 Known Technical Basis The review of the history of BTP 3-4 in Chapter 3 shows that the CUF criterion has evolved over time. In its original form, the CUF criterion was only to be applied for higher stress ranges (i.e., when Sn > 2.4Sm). The CUF criterion became unconditionally applicable in the first revision to MEB 3-1 in July 1981. This could have been done to recognize the fact that the calculation of the CUF takes into account the variation of fatigue life with stress range. The lack of supporting documentation from that time makes it difficult to ascertain the staffs true reasoning. In any event, all subsequent revisions of BTP 3-4 retained the CUF criterion as written in 1981.

On January 5, 2012, the NRC held a public meeting on nuclear power plant piping fatigue issues [20]. During this meeting, the Electric Power Research Institute (EPRI) described a potential technical approach for arriving at a CUF criterion (see Section 4.4). As a followup to that meeting, the NRC staff provided the following perspective on the basis behind the BTP 3-4 CUF criterion:

• The criteria in BTP 3-4 provide conservative margin relative to the ASME Code design approach.

• The added margin may account for unanticipated events and uncertainties in design inputs.

• In 1986, the NRC staff considered whether the CUF criterion should be increased to 0.4 (see Section 4.3) and determined that the CUF criterion should not be increased [21].

• The ASME Code design fatigue curves do not account for EAF and, therefore, may not be conservative.

• When the effects of EAF are accounted for in CUF calculations, the staff has demonstrated a willingness to accept a CUF of 0.4 as a criterion.

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• The staff was not aware of compelling reasons to change the existing criterion.

• BTP 3-4 criteria are not requirements; instead, they are methods acceptable to the staff to demonstrate a safety case. Therefore, other methods may be acceptable as long as they are adequately justified.

4.2 American National Standards Institute/American Nuclear Society 58.2-1988 The American National Standards Institute (ANSI) and the American Nuclear Society (ANS) published ANSI/ANS-58.2-1988, Design Basis for Protection of Light Water Nuclear Power Plants against the Effects of Postulated Pipe Rupture, on January 1, 1988 [22]. This standard largely mirrors the guidance in BTP 3-4. However, the CUF criterion is 0.4 in this standard. The forward to ANSI/ANS-58.2-1988 states, [The NRCs] 0.1 fatigue usage factor is considered to be conservative. ANSI/ANS-58.2-1988 acknowledges that the ASME Code fatigue design curves do not account for environmental effects and imply that the NRC retained the CUF criterion of 0.1 to bound this concern. However, no documented technical basis is available for the CUF criterion of 0.4 in ANSI/ANS-58.2-1988.

4.3 The Rodabaugh Approach In 1986, E.C. Rodabaugh proposed a technical basis for the proposed change to the CUF criterion [21]. Rodabaugh stated that a reduced CUF, Ur, is related to a reduced allowable stress range, Sr, and the unreduced ASME Code-allowable stress range, Sm, through Equation 7.

5 S

U r = r Equation 7 Sm He then stated that the stress range criterion is Sr = 0.8Sm in MEB 3-1. Substituting into Equation 7 yields Equation 8:

U r = (0.8 ) = 0.328 5

Equation 8

Rodabaughs argument is that a CUF criterion of 0.33 would correspond exactly to the stress range criterion of 0.8. The NRC staff rejected the idea of increasing the CUF criterion on this basis at the time. However, the staff has stated that the CUF criterion could be increased once the ASME Code,Section III, design curves incorporate the effects of reactor coolant environment on fatigue lives. To date, the ASME Code has not incorporated EAF. License renewal commitments, new reactor design certifications and combined licenses, and NRC technical documents that cover the subject currently govern EAF issues ([23]-[24]).

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4.4 Electric Power Research Institute Technical Report on Postulating Break Locations In October 2011, EPRI published Technical Report (TR) 1022873, Improved Basis and Requirements for Break Location Postulation [5]. TR 1022873 states that the NRCs guidelines on break locations are based mainly on the idea that thermal fatigue is the damage mechanism that is most likely to cause a pipe rupture. The work suggests that other damage mechanisms may be more likely to cause a rupture than thermal fatigue. TR 1022873 further evaluates the CUF criteria in particular and proposes an alternative approach. This section provides a brief overview of EPRIs findings in TR 1022873. This document does not comment on the validity of EPRIs claims; instead, it only reports them as part of the historical background of this topic.

4.4.1 Review of Operating Experience EPRI reviewed the literature and existing operating experience databases to examine which damage mechanisms are most likely to lead to pipe rupture. EPRI considered rupture events, leakage events, and instances whereby nondestructive examination methods found cracks.

EPRI concluded that vibration fatigue, flow accelerated corrosion, and stress-corrosion cracking events are more likely to occur than thermal fatigue.

4.4.2 Probabilistic Approach EPRIs approach involved the use of the probabilistic fracture mechanics (PFM) software pc-PRAISE. EPRI evaluated pipe leakage probability and assumed that rupture probability was equal to leakage probability. In addition, it drew upon the work discussed in NUREG/CR-6674, Fatigue Analysis of Components for 60-Year Plant Life, issued June 2000 [25], for the following aspects of the study:

• estimating core damage frequency (CDF) from leakage probability

• accounting for EAF effects

• identifying fatigue-sensitive locations

• obtaining stresses and the number of cycles for various locations After describing the results of the PFM analysis, the report concludes that there is no direct correlation between CDF and the CUF with environmental effects.

4.4.3 Electric Power Research Institutes Proposed Alternative Approach In this report, EPRI proposes the following four-phase approach to postulating break locations, which serves as an alternative to BTP 3-4:

• Phase 1. Eliminate locations where the consequence of failure is low.

• Phase 2. Identify relevant damage mechanisms and eliminate locations where the damage rate is low.

• Phase 3. Eliminate locations where damage mechanisms propagate rapidly if managing such mechanisms can be demonstrated.

• Phase 4. Perform a probabilistic evaluation to assess risk.

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EPRI provides more details on each of these phases in its report [5]. For example, the report mentions damage mechanisms other than fatigue as potentially relevant for determining break locations.

4.4.4 Discussion Commenting on the validity of EPRIs proposal ([5]) is beyond the scope of this report. The focus of this chapter is to present what is known about the CUF criterion in BTP 3-4, including questions that have been raised about its technical basis. The EPRI document represents a risk-informed approach to the topic. In future consideration of risk-informed approaches, EPRIs study reveals three technical topics that may warrant further thought. First, EPRI assumed that leakage probability is equal to rupture probability. It states that this assumption is conservative; however, it may only be conservative for an LBB condition. Second, EPRI does not explicitly address defense-in-depth in its report. Typically, risk-informed approaches that the NRC approves are accompanied by an assessment of the five principles of risk-informed decision-making, including defense-in-depth. Finally, any risk-informed approach requires acceptance criteria. Additional consideration of the appropriate acceptance criteria may be warranted.

4.5 Gosselin and Simonens Approach 4.5.1 Overview S.R. Gosselin and F.A. Simonen [6] also critiqued the technical basis behind the CUF break criterion. Like EPRI, they used operating experience data to suggest that thermal fatigue was a relatively small contributor to pipe degradation, as compared to other damage mechanisms.

They asserted that fatigue usage factors were meant to serve as a conservative design methodology instead of an accurate predictor of failure probability. For quantitative arguments, they performed a damage tolerance assessment (Section 4.5.2) and a CDF calculation (Section 4.5.3).

4.5.2 Damage Tolerance Assessment Gosselin and Simonen [6] performed a deterministic analysis to investigate the margins to leakage upon reaching a given CUF. The results of that work suggested that, even at a CUF of 1.0, the design margin protects against leakage for a 60-year operating period. This result may bolster their assertion about the conservative nature of the CUF approach to fatigue design.

4.5.3 Core Damage Frequency Calculation Gosselin and Simonens calculation [6] of the CDF used the following relationships for determining probability of a LOCA, PLOCA, and LOCA frequency, LOCA/WELD-YR:

PLOCA = (PSBLOCAlTWC + PLBLOCAlTWC )x PTWC Equation 9 LOCA / WELD YR = (PSBLOCAlTWC + PLBLOCAlTWC )x TWC / WELD YR Equation 10 where PXlY means probability of X given Y, SBLOCA is small-break loss-of-coolant accident, LBLOCA is large-break loss-of-coolant accident, and TWC is through-wall crack. Using the conditional break probabilities in NUREG-1829, Estimating Loss-of-Coolant Accident Frequencies through the Elicitation Process, issued April 2008 [26], and the break size definitions in NUREG/CR-6674 [25], they estimated PSBLOCAlTWC = 6.4205x10-2 and 13

PLBLOCAlTWC = 8.7590x10-4. To use Equation 9, Gosselin and Simonen [6] had to estimate PTWC, but they did not provide details on how they determined PTWC. From that value, they calculated the probability that the initiated cracks will grow through the wall. Gosselin and Simonen [6] did this for several values of Fen, which is the ratio of the fatigue life in air and the fatigue life in reactor water.

Finally, having obtained estimates of all parameters in Equation 9, Gosselin and Simonen [6]

determined LOCA probabilities and frequencies as a function of Fen. They found LOCA probabilities in the range of 1x10-5 or lower and were able to estimate the CDF with the following equations:

CDFSBLOCA = CCDPSBLOCA x SBLOCA x N WELDS Equation 11 CDFLBLOCA = CCDPLBLOCA x LBLOCA x N WELDS Equation 12 where, CCDP is conditional core damage probability, and NWELDS is the number of welds in the analyzed piping system. Based on NUREG/CR-6674, two assumptions were made:

(1) CCDPSBLOCA = 1x10-6 and (2) CCDPLBLOCA = 5x10-4. Finally, with this methodology, Gosselin and Simonen [6] determined that CDFs were 1x10-8 or less when the CUF was less than 1.0.

4.5.4 Discussion As in Section 4.4, this report does not address the technical adequacy of Gosselin and Simonens [6] approach. However, it does represent a critique on the current BTP 3-4 CUF criterion and another risk-informed approach to the topic. Equation 9 does not consider the probability of a LOCA without a through-wall crack. This is similar to EPRIs assumption that leakage probability is equal to rupture probability (see Section 4.4), which may be strictly true only for an LBB condition. This is a potential weakness in the approach because the BTP 3-4 criteria do not apply to systems approved for LBB.

4.6 Economic Simplified Boiling-Water Reactor Design Certification and Environmental Fatigue Effects Chapter 4 of this report has been dedicated to a discussion of the CUF criterion in BTP 3-4.

The intent is to document various questions that NRC stakeholders have raised on the technical basis behind this position. The CUF limit referenced in the ESBWR Design Control Document (DCD) [8] is another topic of interest related to BTP 3-4.

The ESBWR obtained designed certification under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, as Appendix E, Design Certification Rule for the ESBWR Design [7]. ESBWR DCD, Section 3.6.2, states that pipe ruptures inside containment are in accordance with BTP 3-4 (see also ESBWR DCD, Section 3.6.1.1) [8]. In addition, it makes the following statement:

For the piping system with reactor water, if the environmental fatigue is included in accordance with Regulatory Guide 1.207, the fatigue usage limit should be 0.40 as the criterion instead of 0.10 for determining pipe break locations.

At the time this was submitted for review, the CUF criterion in BTP 3-4 was a CUF of less than 0.10. Therefore, this approach was not in accordance with BTP 3-4 as stated in the DCD; 14

instead, it was an exception to the CUF criterion. The ESBWR DCD offers no technical basis for this exception. However, when the effects of the environment on fatigue life are accounted for, raising a CUF criteria that was developed before environmental fatigue effects were fully recognized may be appropriate. The ESBWR DCD does not explicitly state this or any other justification for this approach. Section 3.6.2.3.1, Criteria Used To Define Pipe Break and Crack Locations and Configurations, of NUREG-1966, Final Safety Evaluation Report Related to the Certification of the Economic Simplified Boiling-Water Reactor Standard Design, Volume 1, issued April 2014 [9], describes the NRC staffs review of the criteria to define break and crack locations. This section does not mention that an exception was taken to the BTP 3-4 criterion.

Although based on the discussion in Section 4.3, the NRC had a precedent that allowed it to consider 0.4 as an appropriate CUF criterion when the CUF calculation includes EAF. Future efforts should clearly document the rationale behind the staffs position on this topic. As a final point of clarity, the applicant in this case chose to express the criterion in terms of conditions under which breaks are not postulated, which is in opposition to the convention that this document adopts.

4.7 Summary This section explored, in detail, one specific aspect of BTP 3-4: the CUF break criterion. The section summarized several applicable studies, including the NRC staffs statements on the public record, stakeholder critique, proposed probabilistic approaches, and regulatory decisions for new reactor designs. Stakeholders have criticized the CUF criterion for its lack of technical basis. They assert that operating experience does not support the idea that thermal fatigue is a significant contributor to pipe damage. However, there is no evidence to suggest that the criteria in BTP 3-4 are based on thermal fatigue alone. Stakeholders have proposed probabilistic approaches to generate less conservative pipe break criteria. These approaches involve key assumptions that may be worth investigating further. Finally, the NRC has demonstrated a willingness to accept a CUF of 0.40 when EAF is accounted for ([8]-[9]).

Although existing regulatory documents (e.g., ESBWR) do not provide an explicit basis for this decision, the roots of a CUF criterion of 0.4 can be traced back to Rodabaughs work [21].

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5. Conclusions and Recommendations 5.1 Summary The primary purpose of this document is to provide a historical summary for the CUF break criterion in BTP 3-4, including any known technical basis and stakeholder feedback. This document summarized the current BTP 3-4 and all previous revisions. The historical information began with Giambussos letter to NRC applicants in December 1972 and ended with the issuance of BTP 3-4, Revision 3, in July 2016. Two noteworthy revisions to the staffs position with respect to CUF criterion were noted:

(1) the change in the CUF criterion from being conditionally applicable in MEB 3-1, Revision 0, to unconditionally applicable in MEB 3-1, Revision 1, as shown in Table 3 (2) the addition of the following statement in BTP 3-4, Revision 3:

For new reactor design certification reviews, the staff has considered a CUF limit of 0.4 to be acceptable when the effects of environmental assisted fatigue (EAF) are considered in the piping design.

This report discussed the CUF criterion in detail and described the known technical basis supporting the CUF criterion. This report also summarized alternative approaches that have been developed by NRC stakeholders in an effort to revise the CUF criterion. EPRI [5] and Gosselin and Simonen [6] proposed probabilistic and risk-informed approaches to this issue.

Both studies suggested that operating experience does not support the concept that thermal fatigue is an important contributor to pipe damage in the field. Finally, this report summarized the staffs position on the ESBWR design certification ([7]-[9]). In this case, the applicant proposed that the CUF criterion be increased to 0.4 for locations where EAF has been accounted for. Although regulatory documents do not clearly justify this proposal, the staffs acceptance of the proposal may be related to work performed by Rodabaugh [21] in the 1980s.

5.2 Future Work This work has demonstrated that the technical basis behind the CUF criterion is unavailable.

This staffs position could be bolstered by developing a documented technical basis for the criterion. Development of a technical basis may even result in a justified change in the regulatory position. There is general consensus in the community that CUF is a conservative design approach. Efforts in the past have pointed to the idea that, if EAF is accounted for, a CUF criterion of 0.1 may be excessively conservative. Before engaging in an effort to update the basis behind BTP 3-4, the NRC staff should thoughtfully consider the points discussed below.

The work by EPRI [5] and Gosselin and Simonen [6] suggests that actual pipe degradation operating experience should be used to guide the CUF criterion choice. The NRC staff should consider whether a CUF criterion is appropriate based on existing operating experience data.

The basis behind the decision should be documented. In addition, the staff should document a basis for the CUF criterion, if it is maintained or revised.

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The industrys work implies that NRC should consider other damage mechanisms when considering pipe break criteria. The PFM code, Extremely Low Probability of Rupture (xLPR)

[27], contains degradation models for stress-corrosion cracking and fatigue. Development of additional degradation models and integration into a probabilistic framework can be a resource-intensive effort. On the other hand, if the NRC determines that a CUF criterion is appropriate, the xLPR code may be an appropriate tool for investigating rupture probabilities caused by fatigue crack growth. To address stakeholder concerns that thermal fatigue is not a major contributor to pipe damage in the field, the transients used as input to the xLPR code need not be limited to thermal fatigue events.

Previous probabilistic work in this area assumes that a rupture event will be preceded by a through-wall crack. This assumption neglects the probability of rupture from a crack that does not leak, such as a relatively long subsurface crack. Because the appropriate computational tools exist, future probabilistic approaches to this problem should focus on calculating total rupture probability. Future studies could also compare leakage probabilities with rupture probabilities as a check on the previous work.

Simpler approaches to this problem may also provide insight. The staff could make an assessment of the appropriate margins against rupture by fatigue relative to that already provided by the ASME Code approach. This assessment should involve deterministic fatigue crack growth studies under a variety of realistic inputs. The work could be further related to CUF by considering typical design inputs for piping systems.

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6. References

[1] U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy.

[2] U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 3.6.3, Leak-Before-Break Evaluation Procedures, Rev. 1, March 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML063600396).

[3] U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Rev. 3, December 2016 (ADAMS Accession No. ML14230A035).

[4] U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Branch Technical Position 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Rev. 3, December 2016 (ADAMS Accession No. ML16085A315).

[5] Electric Power Research Institute, Technical Report 1022873, Improved Basis and Requirements for Break Location Postulation, Palo Alto, CA, October 2011.

[6] Gosselin, S.R., and F.A. Simonen, A Risk-Informed Approach to Fatigue Break Criterion for ASME Class 1 High Energy Piping, in Proceedings of the 20th International Conference on Nuclear Engineering, July 30-August 3, 2012, Anaheim, CA, ICONE20-54534.

[7] U.S. Code of Federal Regulations, Licenses, Certifications, and Approvals for Nuclear Power Plants, Appendix E, Design Certification Rule for the ESBWR Design, Part 52, Chapter I, Title 10, Energy.

[8] GE Hitachi Nuclear Energy, ESBWR Design Control Document, Chapter 3, Design of Structures, Components, Equipment, and Systems, Section 3.1, Conformance with NRC General Design Criteria, through Section 3.8, Seismic Category I Structures, Rev. 10, April 2014 (ADAMS Accession No. ML14100A506).

[9] U.S. Nuclear Regulatory Commission, NUREG-1966, Final Safety Evaluation Report Related to the Certification of the Economic Simplified Boiling-Water Reactor Standard Design, Vol. 1, April 2014 (ADAMS Accession No. ML14099A519).

[10] U.S. Nuclear Regulatory Commission, 10 CFR Part 50; Modification of General Design Criterion 4 Requirements for Protection against Dynamic Effects of Postulated Pipe Ruptures, Federal Register, Vol. 52, No. 207, 27 October 27, 1987, pp. 41288-41294.

[11] American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, 2017 edition,Section III, Rules for Construction of Nuclear Facility Components, Division 1, New York, NY, July 2015.

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[12] American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, 2017 edition,Section II, Material Specifications, Part D, Subpart 1, New York, NY, July 2015.

[13] U.S. Nuclear Regulatory Commission, Branch Technical Position 3-3, Protection against Postulated Piping Failures in Fluid Systems Outside Containment, Rev. 3, March 2007 (ADAMS Accession No. ML070800027).

[14] The Public Health and Welfare, 42 U.S.C. § 5801, 88 Stat. 1233, Energy Reorganization Act of 1974, October 11, 1974.

[15] U.S. Nuclear Regulatory Commission, NUREG-75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, September 1975 (ADAMS Accession No. ML081510817).

[16] U.S. Nuclear Regulatory Commission, Branch Technical Position MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Rev. 1, July 1981 (ADAMS Accession No. ML19137A333).

[17] U.S. Nuclear Regulatory Commission, Branch Technical Position MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Rev. 2, June 1987 (ADAMS Accession No. ML19137A335).

[18] U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Rev. 2, March 2007 (ADAMS Accession No. ML070660494).

[19] U.S. Nuclear Regulatory Commission, NUREG-1803, Safety Evaluation Report Related to the License Renewal of the Edwin I. Hatch Nuclear Plant, Units 1 and 2, Chapter 4, Time Limited Aging Analyses, through Appendix E, Requests for Additional Information, December 2001 (ADAMS Accession No. ML020020301).

[20] U.S. Nuclear Regulatory Commission, Public Meeting Summary Report, Category 2 Public MeetingNuclear Power Plant Piping Fatigue Issues, January 12, 2012 (ADAMS Accession No. ML120120028).

[21] The Light Company Letter, Use of Increased Cumulative Usage Factor for Pipe Rupture Postulation, dated February 28, 1986 (ADAMS Accession No. ML18284A024).

[22] American National Standards Institute/American Nuclear Society (ANSI/ANS)-58.2-1988, Design Basis for Protection of Light Water Nuclear Power Plants against the Effects of Postulated Pipe Rupture, January 1, 1988, La Grange Park, IL.

[23] U.S. Nuclear Regulatory Commission, NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Rev. 1, March 2014 (ADAMS Accession No. ML14087A068).

[24] U.S. Nuclear Regulatory Commission, Regulatory Guide 1.207, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components due 19

to the Effects of the Light-Water Reactor Environment for New Reactors, March 2007 (ADAMS Accession No. ML070380586).

[25] U.S. Nuclear Regulatory Commission, NUREG/CR-6674, Fatigue Analysis of Components for 60-Year Plant Life, June 2000 (ADAMS Accession No. ML003724215).

[26] U.S. Nuclear Regulatory Commission, NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the Elicitation Process, Vol. 1, April 2008 (ADAMS Accession No. ML082250436).

[27] Rudland, D., C. Harrington, and R. Dingreville, Development of the Extremely Low Probability of Rupture (xLPR) Version 2.0 Code, in Proceedings of the ASME Pressure Vessels and Piping Conference, Boston, MA, July 2015.

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