ML14209A027

Issuance of Amendment No. 276, Request to Revise Updated Safety Analysis Report for Westinghouse Plant-Specific Leak-Before-Break Analysis

ML14209A027
Person / Time
Site:	Fort Calhoun
Issue date:	08/07/2014
From:	Lyon C Plant Licensing Branch IV
To:	Cortopassi L Omaha Public Power District
Lyon C
References
TAC MF2559
Download: ML14209A027 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 7, 2014 Mr. Louis P. Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station 9610 Power Lane, Mail Stop FC-2-4 Omaha, NE 68008

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 -ISSUANCE OF AMENDMENT RE:

PLANT-SPECIFIC LEAK-BEFORE-BREAK ANALYSIS (TAC NO. MF2559)

Dear Mr. Cortopassi:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 276 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1 (FCS). The amendment consists of changes to the FCS Updated Safety Analysis Report (USAR) in response to your application dated August 5, 2013, as supplemented by letter dated January 28, 2014.

The amendment revises the structural design basis for the reactor coolant system piping described in Section 4.3.6 of the FCS USAR, to include leak-before-break methodology.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 276 to DPR-40

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. DPR-40

1. The U.S. Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by the Omaha Public Power District (OPPD, the licensee), dated August 5, 2013, as supplemented by letter dated January 28, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changes as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-40 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance. In addition, the licensee shall include the revised information in the next Updated Safety Analysis Report update submitted to the NRC in accordance with 10 CFR 50.71 (e), as described in the licensee's application dated August 5, 2013, as supplemented by letter dated January 28, 2014, and evaluated in the staff's safety evaluation enclosed with this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-40 Date of Issuance: ,1\ u gust 7 , 20 14

ATTACHMENT TO LICENSE AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following page of the Renewed Facility Operating License No. DPR-40 with the attached revised page. The revised page is identified by amendment number and contains vertical lines indicating the areas of change.

License Page REMOVE INSERT (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or when associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is, subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not in excess of 1500 megawatts thermal (rate power).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276 are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.

C. Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan," submitted by letter dated May 19, 2006.

OPPD shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The OPPD CSP was approved by License Amendment No. 266.

Renewed Operating License No. DPR-40 Amendment No. 276

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION. UNIT NO. 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By letter dated February 1, 1984 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML031150562), the U.S. Nuclear Regulatory Commission (NRC) staff approved a generic leak-before-break (LBB) application for a consortium of nuclear power plants based on a bounding LBB analysis of a typical reactor coolant system (RCS) primary loop piping. This consortium included Fort Calhoun Station, Unit 1 (Fort Calhoun or the facility) and its licensee, Omaha Public Power District (OPPD or the licensee). The original LBB analysis was documented in Westinghouse Electric Company, LLC (Westinghouse) report, WCAP-9558, Revision 2, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack," May 1981 (proprietary).

By letter dated January 9, 2002, the licensee submitted its license renewal application (ADAMS Accession Nos. ML020180051 and ML020180054). By letter dated September 5, 2003, the NRC approved the Fort Calhoun license renewal application (ADAMS Accession No. ML032481233) and published its safety evaluation, NUREG-1782, in October 2003 (ADAMS Accession No. ML032481209). In the license renewal application, the licensee stated that it would perform a plant-specific LBB analysis, considering a 60-year life and thermal aging effects of the cast austenitic stainless steel (CASS) in the RCS primary loop piping. The NRC staff requested that the licensee consider, in its plant-specific LBB analysis, the impact of the potential for primary water stress-corrosion cracking (PWSCC) in nickel-based Alloy 82/182 welds in the RCS primary loop piping. In response, the licensee stated that prior to entering the period of extended operation, it would implement actions or perform analyses to confirm continued applicability of the original LBB evaluations for the period of extended operation. The licensee further stated that these actions or analyses will be consistent with those required to address the impact of PWSCC on the original LBB evaluations. The licensee committed to complete the plant-specific LBB analysis before the period of extended operation. The period of extended operation began on August 9, 2013.

To satisfy its previous license renewal commitment, by application dated August 5, 2013 (ADAMS Accession No. ML13220A072), as supplemented by letter dated January 28, 2014 Enclosure 2

(ADAMS Accession No. ML14030A591 ), the licensee submitted for NRC review and approval a license amendment request regarding the continued exclusion of dynamic effects of pipe rupture from the plant's design basis in accordance with Title 10 of the Code of Federal Regulations (1 0 CFR}, Part 50, Appendix A, General Design Criterion (GDC) 4, "Environmental and dynamic effects design bases." In support of this license amendment request, the licensee submitted a plant-specific LBB analysis. This analysis is contained in Westinghouse report, WCAP-17262-P, Revision 1, "Technical Bases for Eliminating Large Primary Loop Piping Rupture as the Structural Design Basis for Fort Calhoun Unit 1," July 2013 (proprietary); a non-proprietary version, designated as WCAP-17262-NP, Revision 1, is available in ADAMS under Accession No. ML13220A074.

2.0 REGULATORY EVALUATION

The licensee has requested to amend its license for Fort Calhoun to permit continued exclusion during its period of extended operation of the dynamic effects associated with postulated pipe ruptures. The licensee's original application for the use of LBB methodology for RCS piping (primary loop) during its original licensing period was authorized by the NRC staff in a letter dated August 25, 1994 (ADAMS Legacy Library Accession No. 9409120225). The licensee's justification for continuing to exclude dynamic effects associated with piping ruptures is based on a plant-specific LBB analysis.

In Section 3.2, "Licensing Methodologies," of the August 5, 2013, submittal, the licensee stated that the application of the LBB methodology for nuclear power plant piping is provided for in GDC 4 of Appendix A to 10 CFR Part 50. The licensee noted that guidance for the application of the LBB methodology is provided in NUREG-1061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks,"

November 1984 (ADAMS Accession No. ML093170485), and in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP), Section 3.6.3, Revision 1, "Leak-Before-Break Evaluation Procedures" (ADAMS Accession No. ML063600396). The licensee stated that Westinghouse has followed the guidance of NUREG-1 061, Volume 3, in performing the analyses in WCAP-17262-P, Revision 1.

GDC 4 of Appendix A to Part 50 of 10 CFR 50 states, in part, that "[s]tructures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with" postulated accidents. Further, GDC 4 states, in part, that "[h]owever, dynamic effects associated with postulated pipe ruptures

... may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping."

The NRC staff notes, however, that Fort Calhoun was licensed for construction prior to May 21, 1971, and is committed to the draft GDC published for comment in the Federal Register on July 11, 1967 (32 FR 10213), in lieu of 10 CFR Part 50, Appendix A. The draft GDCs are contained in Appendix G of the Fort Calhoun Updated Safety Analysis Report (USAR) and are similar to the final GDCs in 10 CFR Part 50, Appendix A. The licensee stated that the draft GDC that are most applicable to LBB are Criteria 9, 16, 33, 34, 35, and 36. Despite the fact that Fort Calhoun is a pre-GDC plant, the NRC staff finds that this request to amend its license was

submitted in accordance with GDC 4 and, therefore, that evaluation of the request in accordance with GDC 4 is appropriate.

The NRC staff notes that the adequacy of a plant's LBB analysis and, therefore, whether the plant meets the requirements of GDC 4, is generally evaluated in accordance with SRP Section 3.6.3, Revision 1, "Leak-Before-Break Evaluation Procedures." This document provides guidance on screening criteria, safety margins, and analytical methods for the piping systems to be qualified for LBB. The technical basis for the LBB evaluation is documented in the NRC report, NUREG-1 061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks," dated November 1984.

LBB analyses, as described in SRP Section 3.6.3, are based on six concepts. First, through the use of experimental data and/or modeling, it is possible to determine the rate at which a through-wall crack of a given length will leak. Second, using fracture mechanics, finite element analyses, and experimental data, it is possible to determine the length of a through-wall crack at which crack growth becomes unstable (i.e., at which the component fails). Third, at some plants, the materials of construction of the RCS piping are sufficiently tough and the capability to detect leaks from the RCS piping is sufficiently sensitive that leakage will be detected long before unstable crack growth occurs. Fourth, that no active degradation mechanisms are present. Fifth, that fatigue will not result in a through-wall crack. Sixth, when combined, the first five criteria provide reasonable assurance that the probability of rupture of a pipe is extremely low.

In the present case, a license amendment under 10 CFR 50.90 is required to permit the continued exclusion of the dynamic effects associated with pipe rupture as a result of the licensee commitment to submit a new, plant-specific LBB analysis prior to entering the period of extended operation.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request to amend its operating license and for the NRC to approve the license amendment request.

3.0 TECHNICAL EVALUATION

3.1 System Description The licensee's plant-specific LBB analysis, including a description of the applicable aspects of the RCS, is contained in the application and the topical report WCAP-17262-P, Revision 1.

These documents indicate that the RCS consists of two loops. Each loop contains one 38.5-inch outside diameter (OD) hot leg between the reactor vessel and the steam generator, two 29-inch OD cold legs between the reactor coolant pumps and the reactor vessel, and two 29-inch OD crossover legs between the steam generator and the reactor coolant pumps. The nominal pipe wall thickness for these pipes ranges from 1.938 to 2.688 inches. Each loop contains 24 welds. These components are all classified as American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 1 piping. The RCS primary loop piping is fabricated from CASS base metal. The dissimilar metal butt welds are fabricated from Alloy 82/182.

3.2 NRC Staff Evaluation Approach The NRC staff evaluated the licensee's plant-specific LBB analysis using the acceptance criteria in SRP Section 3.6.3, Revision 1, and GDC 4. The NRC staff's evaluation is divided into five major areas: the scope of LBB application, bounding components for analysis, fatigue crack growth analysis, LBB evaluation of the CASS piping components, and LBB evaluation of Alloy 82/182 welds.

3.3 Scope of LBB Application SRP Section 3.6.3.11.2 specifies that LBB should only be applied to high energy, ASME Code Class 1 or 2 piping or the equivalent. All the piping under consideration for this LBB evaluation is ASME Code Class 1 piping. Therefore, the NRC staff concludes that the scope of the licensee's application is acceptable.

3.4 Bounding Components for Analysis In its LBB analysis, the licensee screened all locations covered by the analysis for those locations which were least likely to be acceptable for LBB (i.e., those locations where the applied loads/stresses are the highest and/or the properties of the materials of construction are the lowest). As previously stated, the system under consideration consists of two loops of piping each consisting of base metal, similar metal welds, and dissimilar metal welds. Analysis of loads on base metal near to welds as compared to far from welds indicated that only locations adjacent to welds needed to be considered. Additionally, analysis of stresses/loads that cause circumferential cracks, as opposed to axial cracks, indicated that circumferential cracks are limiting and are considered for LBB calculations. The presence of symmetry between the loops permits the licensee to consider only one loop for the analysis. Similarity between welds within a single loop permitted reducing the number of pipe locations to be considered for analysis from 24 to 14. Based on further analysis of the 14 pipe locations, the licensee selected Locations 1 and 6 for further analysis of cracking in the CASS base metal.

The licensee also selected Location 1 for further analysis of cracking in the Alloy 82/182 weld metal.

The NRC staff reviewed the licensee's concept of establishing bounding components for further evaluation and finds the concept to be valid. The NRC staff also evaluated the method by which these components were selected. The NRC staff concludes that the process used by the licensee and the results of the process are consistent with accepted procedures for stress analysis and the estimation of material properties and are, therefore, acceptable.

3.5 Fatigue Crack Growth Analysis To determine the sensitivity of the RCS to the presence of small cracks, the licensee analyzed fatigue crack growth at Location 1. The licensee used a finite element model to analyze stresses of the reactor vessel outlet nozzle safe end region from thermal transients and mechanical loading. The licensee used design transient cycles that are applicable for the life of the plant, including the period of extended operation. The licensee combined the thermal and mechanical stresses with welding residual stresses at the stainless steel weld and Alloy 82/182 weld to calculate the fatigue crack growth of postulated flaws. For the postulated

circumferentially oriented surface flaws at Location 1, the licensee modeled an initial flaw depth of 10 percent through-wall thickness at the safe end-to-pipe stainless steel weld, safe end-to-nozzle Alloy 82/182 weld, and the ferritic steel nozzle. The licensee reported that fatigue crack growth is very small and concluded that the postulated surface flaw will not become a through-wall flaw during the period of extended operation.

The NRC staff reviewed this information and concludes that the licensee used the appropriate initial flaw size, fatigue crack growth rate, transient cycles, and crack growth model to calculate the fatigue crack growth for 60 years. The NRC staff concludes that the fatigue crack growth is not likely to cause pipe rupture during the period of extended operation and, therefore, satisfies SRP 3.6.3.111.10.

3.6 LBB Evaluation of the CASS Piping Components 3.6.1 Evaluation of Active Degradation Mechanisms SRP Sections 3.6.3.1 and 3.6.3.111 specify that the candidate piping should not experience active degradation mechanisms such as fatigue, water hammer, corrosion, creep, or cleavage failure.

Based on the licensee's evaluation of industry and plant-specific operating experience, the licensee stated that fatigue, water hammer, creep, corrosion, and cleavage are not the active degradation mechanisms for the CASS material of the RCS primary loop piping.

The NRC staff concludes that the CASS material of the RCS primary loop piping at Fort Calhoun has not experienced these degradation mechanisms. The NRC staff further concludes that the CASS material of the RCS primary loop piping satisfies the screening criteria of SRP Sections 3.6.3.1 and 3.6.3.111 (i.e., no active degradation mechanisms are present which affect the CASS piping components within the scope of the LBB analysis).

3.6.2 Load Combinations SRP Section 3.6.3.111.11.C specifies the manner in which normal loads (e.g., internal pressure, dead weight, and thermal expansion) are combined with accident loads (e.g., safe shutdown earthquake) to establish the loads which are to be used in the LBB analysis. This SRP section also specifies safety margins for the analysis.

The NRC staff evaluated the manner in which the licensee derived the loads for use in the LBB analysis for the CASS material associated with Locations 1 and 6 and concludes that the method used is consistent with the guidance contained in SRP Section 3.6.3.111.11.C and is, therefore, acceptable.

The NRC staff notes that the normal and fault loads used in the current analysis for Location 1 (21 ,287 and 23,806 in-kips, respectively) were less than the loads used in the previous LBB analysis (45,600 in-kips total) indicating that the current analysis for this location is bounded by the previous analysis.

3.6.3 Material Properties SRP Sections 3.6.3.111.11.A and 3.6.3.111.11.8 specify that material specifications and material properties should be identified.

The licensee stated that RCS primary loop piping elbows are CASS A-351-65 CF8M and the piping segments are CASS A-451 CPF8M. The safe end is stainless steel SA 182 F316. The reactor vessel nozzle is low alloy steel A-508-64, Class 2. Required tensile strengths for the CF8M and CPF8M materials, which are significant to the current LBB analysis, are contained in the ASME standards for the materials and in Section II Part D of the ASME Code. Through the inclusion of Certified Materials Test Reports in its application, the licensee demonstrated that the materials used in the piping subject to the current LBB analysis met these criteria when new.

In addition to the as-new material properties of the CASS material, the licensee noted in its application that CASS material is subject to thermal aging. Thermal aging will, over time, reduce the ductility, impact strength, and fracture toughness of the material. The extent of reduction of these properties due to thermal aging is a function of time, temperature, and chemical composition of the material. The licensee notes that for a given material at a given temperature, after a prolonged period of exposure, continued exposure fails to cause any further reduction in material properties of the material. At this point the material is said to be fully aged.

The material properties used by the licensee in this LBB analysis were the fully aged properties for Locations 1 and 6. These properties were determined in accordance with procedures contained in Argonne National Laboratory reports, 0. K. Chopra and W. J. Shack, "Assessment of Thermal Embrittlement of Cast Stainless Steels," NUREG/CR-6177, NRC, May 1994 (not publicly available), and 0. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems," NUREG/CR-4513, Revision 1, NRC, August 1994 (ADAMS Accession No. ML052360554).

The NRC staff concludes that the use of fully aged material properties and the process used to obtain these properties for the CASS material of the RCS primary loop piping are in accordance with accepted procedures and, therefore, acceptable.

3.6.4 Leakage Crack Calculations SRP Section 3.6.3.111.11.C specifies that a postulated leakage crack size be calculated at the pipe location with the worst material property and that the flaw size is sufficiently large so that the estimated leak rate during normal operation is 10 times greater than the minimum RCS leakage detection system capability. This SRP section further states that the normal operating loads are to be combined based on the algebraic sum of individual values to derive the leakage flaw size.

The licensee stated that the Fort Calhoun RCS leak detection system meets the intent of Regulatory Guide 1.45, Revision 1, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage," dated May 2008 (ADAMS Accession No. ML073200271 ), and its leak detection capability is 1 gallon per minute (gpm) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Thus, to satisfy the margin of 10 on the leak rate, the licensee calculated the size of a crack which would cause leakage at a rate of 10 gpm. In the CASS material at Locations 1 and 6, the length of this leakage crack size is 5.28-inch and 6.64-inch, respectively.

The NRC staff evaluated this information and concludes that the results obtained and the processes used to obtain them are in accordance with the NRC guidance and, therefore, acceptable.

3.6.5 Crack Stability Analysis SRP Section 3.6.3.111.11.C.iv specifies that a fracture mechanics stability analysis or a limit load analysis should be used to determine the critical crack size for a postulated through-wall flaw.

The critical crack size is the length of a crack at which it transitions from slow, stable, predictable growth to very rapid unstable growth. Loads used for this analysis should include both normal loads plus safe shutdown earthquake loads. For the limit load calculation, a Z factor should be applied to the fault load at welds made using shielded metal arc welding (SMAW) or submerged arc welding processes. The analysis should demonstrate that the critical crack size is at least twice as long as the leakage crack size.

The licensee initially evaluated the critical crack length through the use of elastic plastic fracture mechanics (i.e., J-integral). The licensee's J-integral analysis found that for both Locations 1 and 6, applied J (Japplied) was less than J at crack initiation (JIG). This indicates that for the given material and loading conditions at these two locations, a crack would neither initiate nor grow.

Under these conditions, a flaw could not be long enough to be considered critical.

The licensee also conducted a limit load evaluation of the CASS material at Locations 1 and 6.

The welds at these locations were made using SMAW. As a result, the licensee calculated and applied a Z factor to the fault loads. Based on the limit load method, the licensee calculated critical crack sizes of 35.27-inch and 34.16-inch at Locations 1 and 6, respectively.

The NRC staff evaluated both the licensee's elastic plastic fracture mechanics analysis and the limit load analysis and concludes that they were conducted in accordance with the guidance provided in SRP Section 3.6.3.111.11.C. The NRC staff also evaluated the results of these calculations relative to the calculated length of the leakage crack. The NRC staff found that in both analyses, the length of the critical crack exceeded two times the length of the leakage crack. Therefore, in accordance with SRP Section 3.6.3.111.11.C, the NRC staff concludes that these results are acceptable.

3.7 LBB Evaluation of Alloy 82/182 Welds 3.7.1 Evaluation of Active Degradation Mechanisms One of the LBB acceptance criteria contained in SRP Section 3.6.3 is the absence of an active degradation mechanism. The NRC staff notes, however, that Alloy 82/182 welds are subject to PWSCC. The NRC staff also notes that the potential for PWSCC in Alloy 82/182 welds constitutes the presence of an active degradation mechanism in these components. Due to the presence of six unmitigated Alloy 82/182 welds in the components considered in this LBB analysis, the NRC staff concludes that the acceptance criteria of SRP Section 3.6.3 are not met with respect to these welds.

In assessing the failure of the Alloy 82/182 welds to meet the screening criteria of SRP Section 3.6.3, the NRC staff has three observations:

1) the licensee is conducting and will continue to conduct inspections of these welds in accordance with ASME Code Case N-770-1;

2) the NRC has taken the position in Regulatory Issue Summary (RIS) 2008-25, "Regulatory Approach for Primary Water Stress Corrosion Cracking of Dissimilar Metal Butt Welds in Pressurized Water Reactor Primary Coolant System Piping,"

dated October 22, 2008 (ADAMS Accession No. ML081890403), that inspections contained in Materials Reliability Program (MRP) 139, "Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline," (now ASME Code Case N-770-1, as implemented and conditioned by 10 CFR 50.55a(g)(6)(ii)(F)) "provide adequate protection of public health and safety for addressing PWSCC in butt welds"; and

3) the NRC Office of Nuclear Regulatory Research (RES) staff has recently independently calculated PWSCC crack growth in welds such as those under consideration here and has found that in all normal operating conditions the time interval between pipe leakage and pipe rupture exceeds 1 year (ADAMS Accession No. ML14141A302).

The NRC staff believes that these observations, in conjunction with the analysis below, provide reasonable assurance that degradation of the subject piping will be detected sufficiently in advance of a pipe rupture that the presence of an active degradation mechanism will not likely increase the probability of rupture of the subject piping beyond that of piping in which no active degradation mechanism exists. Therefore, with respect to the presence of an active degradation mechanism, the NRC staff concludes that, while the Alloy 82/182 welds do not meet the criteria of SRP Section 3.6.3, they do meet the criteria of GDC 4 (i.e., that the "probability of fluid system piping rupture is extremely low ... ").

3.7.2 Load Combinations The loads that the licensee applied to the Alloy 82/182 welds at Location 1 were the same as those used for the CASS material. As stated above, the NRC staff concludes that the loads used to analyze the cracks at Location 1 for the CASS material are acceptable; therefore, the loads used to analyze the cracks for Alloy 82/182 are also acceptable.

3.7.3 Material Properties Except in the case of short-term operability assessments, where certified material test reports specific to the piping under consideration are acceptable, the NRC staff requires that material properties (e.g., allowable stresses) used in pipe stress analyses be in accordance with the ASME Code.Section II of the ASME Code contains tabulations of allowable stresses for piping base materials.Section II of the ASME Code does not contain similar tabulations for allowable stresses of weld materials. For flaw evaluations, the ASME Code addresses welds through the use of stress intensity factors and the requirement that the material properties of the weld materials be compatible with that of the piping base material.

Given that the present analysis is specific to the Alloy 82/182 weld and that neither an ASME Code allowable stress nor a certified material test report for the "as welded" weld is available, an alternate approach to identifying acceptable material properties (i.e., yield and ultimate strength) is required. To meet this requirement, the licensee selected material properties which it considers to be proprietary.

The NRC staff evaluated the material properties selected by the licensee with respect to values that the NRC uses in its independent analysis of modeling cracking in Alloy 82/182 welds (i.e.,

yield strength of 55.5 ksi and ultimate strength of 84.6 ksi). Based on this evaluation, the NRC staff concludes that the material properties used by the licensee are acceptable.

3.7.4 Leakage Crack Calculations The licensee calculated the length of the leakage crack in the Alloy 82/182 welds via a computer model. The licensee considers both the model and its output to be proprietary. The NRC staff notes that the output of this model (i.e., crack length versus leak rate) for PWSCC cracks has not been fully validated. However, the NRC staff also notes that the uncertainty associated with the model decreases with increasing leak rate and/or crack length. Thus, for a leak rate of 10 gpm, as is critical in this analysis, the NRC staff concludes that both the modeling approach and the output of the licensee's model are reasonable.

3.7.5 Crack Stability Analysis The licensee calculated the critical crack length for the Alloy 82/182 weld at Location 1 using the proprietary material properties and the limit load approach. The licensee then compared the resultant critical crack to the leakage crack. The NRC staff notes that the ratio of the lengths of the critical crack and the leakage crack exceeds the requirement of 2:1. Therefore, the Alloy 82/182 weld has satisfied the margin of 2 for the crack size as provided in SRP Section 3.6.3.

Additionally, the NRC staff notes that had the licensee chosen to use ASME Code allowable material properties for the base material adjacent to the weld in calculating the length of the critical crack in the weld, the critical crack would have been: (a) shorter than the crack calculated using the material properties for the weld, (b) identical in length to the critical crack calculated above for the CASS material (35.27 inches}, and (c) still maintain the margin of 2 with respect to the leakage crack calculated for the weld.

Based on these analyses, the NRC staff concludes that the guidance of SRP Section 3.6.3 have been met.

3.8 Conclusion On the basis of information submitted, the NRC staff has determined that, with the exception of the unmitigated Alloy 82/182 welds, the RCS primary loop piping satisfies SRP Section 3.6.3, Revision 1, and, therefore, GDC 4. The NRC staff concludes that the Alloy 82/182 welds do not satisfy SRP Section 3.6.3 because they are susceptible to PWSCC.

In assessing the failure of the Alloy 82/182 welds to meet the screening criteria of SRP Section 3.6.3, the NRC staff makes four observations: (a) the Alloy 82/182 welds meet all of the guidance of SRP Section 3.6.3 other than the presence of an active degradation mechanism; (b) the licensee is conducting, and will continue to conduct, inspections of these welds in accordance with ASME Code Case N-770-1; (c) The NRC has taken the position in NRC RIS 2008-25 that inspections contained in MRP 139 (now ASME Code Case N-770-1, as implemented and conditioned by 10 CFR 50.55a(g)(6)(ii)(F)) "provide adequate protection of public health and safety for addressing PWSCC in butt welds"; and (d) the NRC staff has recently performed an independent analysis of PWSCC crack growth in welds such as those under consideration here and has found in all normal operating condition cases modeled that the time interval between pipe leakage and pipe rupture exceeds 1 year.

The NRC staff believes that these observations provide reasonable assurance that degradation of the subject piping will be detected sufficiently in advance of a rupture that the presence of an active degradation mechanism will not likely increase the probability of rupture of this piping beyond that of piping in which no active degradation mechanism exists. Therefore, the NRC staff concludes that, while the nickel alloy welds do not meet the criteria of SRP Section 3.6.3, they do meet the criteria of GDC 4 (i.e., that the "probability of fluid system piping rupture is extremely low ... ").

Despite the NRC staffs current position regarding the acceptability of an active degradation mechanism in Alloy 82/182 welds, the cracking of these welds remains a topic of NRC interest and research and the NRC staff is continuing to review the generic implications of PWSCC on LBB approvals.

Based on the above, the NRC staff concludes that the Fort Calhoun RCS primary loop piping meets the requirements of GDC 4 and, therefore, it is acceptable that the Fort Calhoun operating license be amended to permit the exclusion of dynamic effects associated with pipe rupture for the plant's period of extended operation.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on April 8, 2014 (79 FR 19400). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, {2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: J. Tsao, NRRIDE/EPNB Date: August 7, 2014

August 7, 2014 Mr. Louis P. Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station 9610 Power Lane, Mail Stop FC-2-4 Omaha, NE 68008

SUBJECT:

FORT CALHOUN STATION, UNIT NO.1 -ISSUANCE OF AMENDMENT RE:

PLANT-SPECIFIC LEAK-BEFORE-BREAK ANALYSIS (TAC NO. MF2559)

Dear Mr. Cortopassi:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 276 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1 (FCS). The amendment consists of changes to the FCS Updated Safety Analysis Report (USAR) in response to your application dated August 5, 2013, as supplemented by letter dated January 28, 2014.

The amendment revises the structural design basis for the reactor coolant system piping described in Section 4.3.6 of the FCS USAR, to include leak-before-break methodology.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, IRA/

Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 276 to DPR-40

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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