ML12076A103

Issuance of Amendment Regarding the Revision to Technical Specification Core Operating Limits Report References for Realistic Large Break Loss-of-Coolant Accident Analysis

ML12076A103
Person / Time
Site:	Harris
Issue date:	05/30/2012
From:	Billoch-Colon A T Plant Licensing Branch II
To:	Burton C Progress Energy Carolinas
Billoch-Colon, Araceli
References
TAC ME6999
Download: ML12076A103 (19)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 May 30,2012 Mr. Christopher L Burton Vice President Shearon Harris Nuclear Power Plant Progress Energy Carolinas, Inc. Post Office Box 165, Mail Code: Zone 1 New Hill, NC 27562-0165 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 -ISSUANCE OF AMENDMENT RE: THE REVISION TO TECHNICAL SPECIFICATION CORE OPERATING LIMITS REPORT REFERENCES FOR REALISTIC LARGE BREAK LOSS-OFCOOLANT-ACCIDENT ANALYSIS (TAC NO. ME6999)

Dear Mr. Burton:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 138 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). This amendment changes the HNP Technical Specifications (TSs) in response to your application dated August 22, 2011, as supplemented by letters dated February 23, March 20, and April 2, 2012. The amendment revises HNP TS 6.9.1.6, "Core Operating Limits Report," to add plant specific methodology ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA [Ioss-of-coolant accident]

Analysis," Revision 1, that implements AREVA's NRC-approved Topical Report EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Revision O. A copy of the related safety evaluation is enclosed.

A notice of issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, /" A. ,J ,. >"' t, ... Araceli T. Billoch Colon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. Amendment No. 138 to NPF-63 2. Safety Evaluation cc w/enclosures:

Distribution via ListServ UNITED NUCLEAR REGULATORY WASHINGTON.

D.C. 20555"()001 CAROLINA POWER & LIGHT COMPANY, ET AL. DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 138 Renewed License No. NPF-63 The Nuclear Regulatory Commission (the CommiSSion) has found that: The application for amendment by Carolina Power & Light Company (the licensee), dated August 22, 2011, as supplemented by letters dated February 23, March 20. and April 2, 2012, 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 Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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.

-2 Accordingly, the license is amended by changes to the Renewed Facility Operating License and Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-63 are hereby amended to read as follows: Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 138 , are hereby incorporated into this license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Douglas A. Broaddus, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed License, and Technical Specifications Date of Issuance:

May 30, 2012 ATTACHMENT TO LICENSE AMENDMENT NO. 138 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following page of the renewed facility operating license with the revised page. The revised page is identified by amendment number and contains a line in the margin indicating the area of change. Remove Insert Page 4 Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain lines in the margin indicating the areas of change. Remove Insert 6-24a 6-24a This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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. Maximum Power Level Carolina Power & Light Company is authorized to operate the facility at reactor core power levels not in excess of 2900 megawatts thermal (100 percent rated core power) in accordance with the conditions specified herein. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 138 , are hereby incorporated into this license. Carolina Power &Light Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. Antitrust Conditions Carolina Power & Light Company shall comply with the antitrust conditions delineated in Appendix C to this license. Initial Startup Test Program (Section 14)1 Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change. Steam Generator Tube Rupture (Section 15.6.3) Prior to startup following the first refueling outage, Carolina Power & Light Company shall submit for NRC review and receive approval if a steam generator tube rupture analysis, including the assumed operator actions, which demonstrates that the consequences of the design basis steam generator tube rupture event for the Shearon Harris Nuclear Power Plant are less than the acceptance criteria specified in the Standard Review Plan, NUREG-0800, at 15.6.3 Subparts 11(1) and (2) for calculated doses from radiological releases.

In preparing their analysis Carolina Power & Light Company will not assume that operators will complete corrective actions within the first thirty minutes after a steam generator tube rupture. 1 The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-63 Amendment No. 138 ADMINISTRATIVE 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued) XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing," approved version as specified in the COLR. (Methodology for Specification 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor). EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs," approved version as specified in the COLR. (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis," I Revision 1, as approved by NRC Safety Evaluation dated May 30, 2012. (Methodology for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR. (Methodology for Specification 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 Control Bank Insertion Limits, and 3.2.2 -Heat Flux Hot Channel Factor). SHEARON HARRIS -UNIT 1 Amendment No. 138 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 138 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 CAROLINA POWER & LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated August 22, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 1123BA077), as supplemented by letters dated February 23,2012 (ADAMS Accession No. ML 120670403), and March 20, 2012 (ADAMS Accession No. ML 120BOA21B), and April 2, 2012 (ADAMS Accession No. ML 12102A016)

Carolina Power & Light Company (the licensee), doing business as Progress Energy Carolinas Inc., submitted a license amendment request (LAR) to revise Technical Specifications (TSs) to the Renewed Facility Operating License No. NPF-63 for Shearon Harris Nuclear Plant, Unit 1 (HNP). The amendment revises HNP TS 6.9.1.6, "Core Operating Limits Report [COLR]," by replacing "EMF-20B7(P)-A, "SEM/PWR-09B:

ECCS Evaluation Model for PWR LBLOCA Applications" with plant specific methodology ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA [Ioss-of-coolant accident]

Analysis," Revision 1. ANP-3011 (P) implements AREVA's NRC-approved evaluation model (EM) Topical Report EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Revision O. The NRC staff refers to the EM interchangeably as EMF-2103(P)(A)

Rev. 0 or Realistic Large Break LOCA (RLBLOCA).

The February 23, and March 20 and April 2, 2012, supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's initial proposed no significant hazards consideration, as published in the Federal Register.

2.0 REGULATORY EVALUATION

The requirements related to the emergency core cooling system (ECCS) performance are contained in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.46 "ECCS." Specifically, 10 CFR 50.46(a)(1)(i) includes the following: The ECCS performance must be calculated using an acceptable evaluation model. Enclosure The evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system performance during a LOCA. A number of postulated LOCAs of different sizes, locations, and other properties must be analyzed, to provide assurance that the most severe postulated LOCAs have been calculated. Uncertainty in the analysis method and inputs must be identified and assessed. The uncertainty must be accounted for so that, when the results are compared to the criteria contained in 10 CFR 50.46(b), there is a high level of probability that the acceptance criteria would not be exceeded.

General Design Criterion (GDC) 35: "Emergency Core Cooling" requires a system to provide abundant emergency core cooling. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. The proposed TS changes will be evaluated to ensure continued compliance with the requirements of 10 CFR 50.36 "Technical Specifications" Section(c)(2)(ii).

Compliance with this regulation requires a licensee to maintain a list of approved analytical methods used to establish potential cycle-specific core operating limits, per Generic Letter (GL) 88-16, "Removal of Specific Parameter Limits from Technical Specifications." The acceptance criteria in 10 CFR 50.46(b) for the results of the ECCS include the following, to which the AREVA realistic EM-2103 is intended to show conformance: The calculated maximum fuel element cladding temperature, peak cladding temperature, (PCT) shall not exceed 2200°F [degrees Fahrenheit]. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

If cladding rupture is calculated to occur, the inside surfaces of the cladding shall be included in the oxidation, beginning at the time of rupture. The calculated total amount of hydrogen generation from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. The licensee proposed to incorporate some features in addition to the generically approved, or accepted, evaluation model described in EMF-2103(P)(A)

Rev.0.1 These include corrections to The words "approved:

and "accepted," may be used interchangeably.

Accepted is used because it is consistent with the language in the regulation, referring to the use of an acceptable evaluation model. The word approved is also used because one way to identify an acceptable evaluation model is by its description in a licensing topical report that has received NRC review and approval.

1

-incorporate more realistic modeling of the fuel mechanical effects expected at the peak temperatures predicted for the HNP application.

The NRC staff reviewed the additional model features in consideration of the guidance contained in NRC Regulatory Guide (RG) 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance." 3.0 TECHNICAL EVALUATION 3.1 Background The August 22, 2011 LAR is a re-submittal of a prior LAR dated March 23, 2010 (ADAMS Accession No. ML 100890594), that the licensee subsequently withdrew on March 28, 2011 (ADAMS Accession No. ML 110950063).

The licensee withdrew the March 23, 2010 LAR because the NRC staff identified issues with the licensee's application of the generically approved RLBLOCA EM. In summary, the model did not account for the following phenomena: The generic model, which uses input from the RODEX3A fuel performance code, did not account for the thermal conductivity degradation issue described in NRC Information Notice (IN) 2009-23, "Nuclear Fuel Thermal Conductivity Degradation" (ADAMS Accession No. ML091550527). The generic model did not incorporate a fuel clad swellil1g and rupture model. The generic model did not account for fuel relocation following a predicted rupture of the fuel cladding.

Generally, the NRC has approved requests to implement RLBLOCA, provided that the predicted results are lower than a PCT threshold value of about 1850°F. Above this threshold value, the above phenomena are expected to contribute more significantly to the cladding temperature transient.

Therefore, if the predicted results exceed 1850°F, it may not be possible to use RLBLOCA to conclude that the reactor system performance is realistically modeled if an analytic treatment for the above phenomena is not included in the evaluation model. The August 22, 2011, LAR submittal, included the model enhancements to address the above concerns.

The licensee consulted with AREVA to develop a more realistic evaluation model. AREVA has incorporated, into this EMF-21 03(P)(A) Rev. 0 application, elements from other NRC-approved evaluation models that address these phenomena.

The NRC staff review of the licensee's application of the RLBLOCA EM for HNP considered the following based on the above issues: Use of a generically approved evaluation model. HNP conformance to the 10 CFR 50.46 acceptance criteria. Updates to EMF-2103(P)(A)

Rev. 0 to address previously expressed staff concerns with RLBLOCA. The presently proposed model improvements.

-4 3.2 EMF-2103(P)(A)

Rev. 0 Generic Approval The RLBLOCA is generically approved to evaluate the ECCS performance during postulated large break LOCAs in 3-loop and 4-loop Westinghouse plants, as well as in Combustion Engineering plants. HNP is a 3-loop Westinghouse plant and so is appropriately analyzed using RLBLOCA. The NRC staff safety evaluation approving EMF-2103(P)(A), Rev. 0 for generic use includes a number of conditions and limitations on its plant-specific use. The licensee addressed these conditions and limitations as described in Table 3-4 "[Safety Evaluation Report] SER Conditions and Limitations" of ANP-3011 (P) (ADAMS Accession No. ML 11238A078).

The NRC staff found the licensee responses in Table 3-4 to the SER conditions and limitations acceptable.

However, Item 3 in Table 3-4 related to the need for a blowdown clad rupture model required additional evaluation.

Item 3 indicates that the current HNP calculation set contains 4 cases for which the initial fuel pin rupture occurred more than 2 seconds prior to the beginning of core recovery, which is considered a reflood rupture. To clarify the information in Item 3, in its February 23, 2012 letter, the licensee compared the results of ANP-3011 (P) with sensitivity studies characterizing swelling and rupture with no relocation for fuel fragments and simulating swelling and rupture with relocation.

The licensee provided the results of the expanded sensitivity studies in a range of packing fractions from 0 to 80 percent, which increased the predicted PCT to account for the effects of fuel rupture. The NRC staff finds the analysis for the fuel cladding rupture provided by the licensee acceptable, and that the limitation in Item 3 is satisfied.

Based on the NRC staff review of the licensee's discussion of the conditions and limitations for the use of EMF-21 03(P)(A), Rev. 0, the NRC staff finds the HNP's RLBLOCA plant-specific implementation acceptable.

3.3 HNP Conformance to Acceptance Criteria As shown in Table 1: "Conformance with Acceptance Criteria" below, the results indicate that HNP meets the 10 CFR 50.46 acceptance criteria, and hence the results are acceptable.

Table 1: Conformance with Acceptance Criteria Acceptance Criterion Regulatory Limit HNP Result Reference Peak Clad Temperature (PCT) 2200°F 2071°F Table 3-1 "HNP PCT Rackup" of ANP-3011 Q1 (P): "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis" (ADAMS Accession No. ML 12067A180)

Maximum Local Oxidation 17% 4.38% (M5ŽCladding) 6.15% (Zircaloy, Zr-4) Table 3-5 "Summary of Results for the Limiting PCT Case" of ANP-3011 (P) Total Core Oxidation Percent 1% 0.05%

-3.4 Departures from EMF-2103(P)(A)

Rev. 0 Related to Transition Package Subsequent to the approval of EMF-2103(P)(A), Rev. 0, the NRC staff found that certain modeling assumptions and constitutive relationships contained in RLBLOCA are not suitable for demonstrating compliance with the 10 CFR 50.46(b) acceptance criteria, as described in a draft safety evaluation dated April 3.2007. This safety evaluation was never formally issued, and the vendor withdrew the topical report revision that it supported.

Therefore.

there is no publicly available copy of this report. Accordingly, the NRC staff convides to the conditions and limitations as discussed below for requests to implement EMF-21 03(P)(A), Rev. O. as an interim review approach until the staff completes its review of EMF-21 03(P), Revision 2. This approach is known as the "Transition Package." 3.4.1 Assumed Power Level The power assumed in the analyses.

2958 megawatt thermal (MWt) , is 2.0 percent higher than the current licensed thermal power (CL TP) level of 2900 MWt to account deterministically for measurement uncertainties.

This departure from the previously approved methodology is acceptable because it is conservative in that the previously approved methodology permitted ranging the assumed power level, meaning that some cases could have initiated at a power level less than 2958 MWt. It is also acceptable because it is consistent with the NRC staff's position in RG 1.157 that parametrically ranging the assumed initial power level is inconsistent with 10 CFR 50.46 requirements, whereas deterministically including uncertainty in the assumed initial power level is acceptable.

Note that the above discussion applies to HNP operating at current licensed power level of 2900 MWt. Section 5.1 "Reactor Power," of ANP-3011 (P) states that the analysis is also applicable at a power level proposed for a measurement uncertainty recapture (MUR) power u prate , which is currently under NRC staff review. At the proposed uprated power level of 2948 MWt. the analysis is applicable because the power level measurement uncertainty will be reduced from 2.0 percent to 0.34 percent. Based on the fact that the analyzed power level includes bounding uncertainty sufficient to cover both CL TP level and MUR power level, the NRC staff determined that the analysis is applicable at both CL TP and MUR conditions.

3.4.2 Blowdown and Refill Heat Transfer and Liquid Flow The RLBLOCA analysis was performed with a version of S-RELAP5 that requires both (1) the void fraction to be less than 0.95 and (2) the clad temperature to be less than the minimum temperature for film boiling before the rod is allowed to quench. During its review of EMF-2103(P), Revision 1, the NRC staff determined that the S-RELAP5 evaluation model could allow rod quench to occur once the temperature drops below the minimum film boiling temperature regardless of the void fraction in the channel. Contrarily, NUREG-0915 "A Criterion for the Onset of Quench for Low Flow Reflood" demonstrated that the void fraction must also be less than 0.95 for rod rewet to occur. To address this concern for HNP, the HNP specific analyses include this departure from the approved methodology.

The NRC staff finds this acceptable because the departure provides for analytic predictions that are not only more consistent with observed data, but also more conservative than predictions obtained using methodology previously approved by the NRC. The Forslund-Rohsenow heat transfer correlation is documented in "Thermal Non-Equilibrium in Dispersed Flow Film Boiling in a Vertical Tube," Forslund, Robert P., and Warren M. Rohsenow, Department of Mechanical Engineering, Massachusetts Institute of Technology, November, 1996. The correlation models heat transfer of two-phase nitrogen in a vertical tube. In S-RELAP5, it is used to predict the heat transfer behavior of two-phase water in a vertical cylindrical array. To compensate for the differences between nitrogen and water, and for the differences between tube flow and cylindrical array flow, the correlation is adapted by benchmarking to full-scale tests. Through this benchmarking process, the correlation's applicability for use in the code is determined.

The correlation and its benchmarking are widely accepted, within certain limitations, within the nuclear industry.

As documented in NUREG-0915, and also in "The Development of a Non-Equilibrium Dispersed Flow Film Boiling Heat Transfer Modeling Package," Ph.D Dissertation, Michael J. Meholic, Pennsylvania State University, August 2011, the correlation can predict behavior that is unsupported by more relevant experimental testing than that from which the Forslund-Rohsenow correlation was derived. The applicability of the correlation must be limited, therefore, to conditions appropriate for its use. Otherwise, as may be the case with S-RELAP5 if the limitation described in the body text is not applied, the heat transfer behavior may be over-predicted, and the resulting peak fuel element cladding temperature may be under-predicted.

The RLBLOCA analysis was performed with a version of S-RELAP5 that limits the contribution of the Forslund-Rohsenow model to no more than 15 percent of the total heat transfer at and above a void fraction of 0.9. This departure from the approved methodology accounts for experimentally observed phenomena that appear to inhibit droplet contact with heated fuel rods at high void fractions.

Thus, this departure is conservative relative to the approved methodology because it corrects for any potential to over-predict heat transfer through conduction to entrained droplets, which experimental observations have shown not to come in contact with the fuel rods at such high void fractions.

The net effect of this conservatism would serve to increase the overall predicted PCT compared to evaluations performed using methodology previously approved by the NRC and it is therefore, acceptable to the NRC staff. The analyses ranged in the area between the minimum break area and an area of twice the size of the broken pipe. The licensee stated that the minimum area was calculated to be 26 percent of the double-ended guillotine break area. This information demonstrates that the total number of sampled cases is appropriate because the phenomenology dominating the limiting aspects of the event for all sampled break areas is consistent.

That is, a certain number of sampled cases is appropriate, because the limiting results of the accident for pipe ruptures ranging from about 20 to 100 percent of the double-ended pipe rupture size are all limited by dispersed flow film boiling ahead of the quench front. If the sampled break area included a greater range, i.e., break sizes less than 20 percent of the double-ended guillotine rupture, additional phenomenology would govern the limiting events, and additional cases would be required to provide the necessary high level of statistical confidence that a bounding upper tolerance limit had been identified.

-7 3.4.3 GOC 35: "Emergency Core Cooling System" Conformance The analyses addressed the availability of offsite power by ranging each case separately.

This is acceptable because it satisfies GOC 35 of 10 CFR Part 50, Appendix A, in that each distribution type has been accounted for separately with its own set of cases, thereby addressing possible concerns associated with the mixing of two separate statistical spectra. The NRC staff finds this treatment acceptable because it is consistent with GOC 35 as noted in ANP-3011 (P). 3.4.4 Oowncomer Two-Phase Conditions The NRC staff has historically identified differences in results between staff confirmatory calculations and those produced using the AREVA RLBLOCA EM, attributable to downcomer boiling modeling, that cause significant differences in peak cladding temperature results. The NRC staff's confirmatory models predict the PCT on the order of 400°F higher than those obtained using the AREVA model. As discussed in Section 1.0 of ANP-3011 (P), AREVA attributes this in part to an underprediction of cold leg condensation efficiency.

To correct this, AREVA has identified appropriate multipliers to force fluid entering the downcomer to saturated conditions following the deployment of the safety injection tanks. The NRC staff finds this departure

'from the previously approved methodology acceptable because (1) the artificially saturated fluid conditions will conservatively reduce both the downcomer driving head and the core flooding rate, which becomes conducive to portions of the fuel remaining in a vapor-cooled environment, thus presenting a greater challenge to clad surface cooling, and (2) conditions in the downcomer following safety injection tank discharge are expected to be slightly subcooled, meaning that assuming fully saturated conditions is conservative.

The licensee also discussed benchmarking to the Upper Plenum Test Facility (UPTF) tests, a full-scale simulation of a four-loop German pressurized-water reactor (PWR) which is similar to a 4-loop Westinghouse, as well as nodalization sensitivity studies performed on a three-loop Westinghouse plant to investigate the validity of the S-RELAP5 model and its downcomer noding sensitivity with respect, in particular, to downcomer boiling. The licensee investigated the sensitivity to downcomer boiling by re-nodalizing the S-RELAP5 downcomer both axially and azimuthally and executing reflood PCT cases, i.e., time of PCT equals approximately 90 seconds. The comparisons illustrated mild effect on downcomer liquid levels during the refill period, but no significant effects during the reflood. The PCT was unaffected by the changes associated with the study. The NRC staff finds that the licensee has addressed downcomer boiling acceptably because (1) the study examined the effects of increased downcomer nodalization, (2) the original model was benchmarked against UPTF data, and (3) the sensitivity studies considered PCTs as high as that predicted for the HNP limiting cases. 3.5 Additional Issues with EMF-2103(P)(A), Revision 0 3.5.1 Transient and Pre-Transient Oxidation On August 3, 1998, the NRC issued Information Notice 98-29: "Predicted Increase in Fuel Rod Cladding Oxidation, n expressing concern that predicted cladding total oxidation (including both

-pre-accident and accident oxidation) resulting from a postulated LOCA could for some plants exceed the 17 percent oxidation limit, and that LOCA methodologies were not addressing that concern. In the letter dated November 8, 1999 (ADAMS Accession No. ML993270252), to the Nuclear Energy Institute, the NRC provided its position that both pre-accident and accident oxidation must be estimated, citing several references, including the Opinion of the Commission dated December 28, 1973. To address the concern about pre-accident and accident oxidation, the licensee analyzed transient oxidation using the generic EMF-21 03(P){A) Rev. 0 model, and also listed the predicted pre-transient oxidation.

The results were added together.

This approach ensures that the transient oxidation calculation is conservative, because more oxidation will form on the surface of unoxidized fuel. The approach also ensures that pre-transient oxidation is included in the calculation.

Since the approach includes pre-transient oxidation and a conservative calculation of transient oxidation, the NRC staff finds the treatment acceptable.

Therefore, the combined oxidation values provided in Table 3-5 "Summary of the Results for the Limiting PCT Case" of AN P-3011 (P) demonstrate that there is a high level of probability that the oxidation would not exceed 17 percent. 3.5.2 Nuclear Fuel Thermal Conductivity Degradation IN 2009-23 describes an evolving identified issue concerning the ability of legacy thermal-mechanical fuel modeling codes to predict the exposure-dependent degradation of fuel thermal conductivity accurately.

This phenomenon is known as thermal conductivity degradation (TCD). Some legacy codes, including RODEX3A, non-conservatively over-predict fuel thermal conductivity at higher burnups. To correct this issue, AREVA has applied a polynomial transformation, which is used to bias the fuel pellet centerline temperature based on empirical data collected supporting the more recent RODEX4 fuel performance code. In ANP-3011(P), Section 6.0, the licensee provided information stating that the fuel pellet centerline temperature correction has a minor impact on the cladding surface temperature.

Additional information provided by the licensee on February 23, 2012, also demonstrated that, for the PCT-limiting case, the pellet centerline and cladding surface temperatures tend to converge rapidly -on the order of seconds -during the transient.

For the limiting HNP cases, the PCT occurs at approximately 100 seconds. Figure 6-6 of ANP-3011 (P) shows that the fuel centerline, surface, and average temperatures have converged by the beginning of reflood at 26.4 seconds. Based on (1) the licensee's analytic correction to fuel centerline temperatures, which increases the fuel stored energy, and (2) the time difference between the limiting PCT and the time required for the fuel and cladding temperatures to converge, the staff finds that TCD is acceptably addressed for the HNP RLBLOCA implementation request. 3.5.3 Decay Heat Consistent with the NRC RG 1.157 for realistic ECCS evaluations, the licensee uses the ANSIIANS-5.1-1979, "American National Standard for Decay Heat Power in Light Water Reactors" (ANSIIANS 1979) standard.

In the analysis provided, the uncertainty for the decay heat parameter is set to zero and no sampling is done on this parameter, resulting in the decay heat being used at its nominal value. This is a change from the EMF-2103(P)(A)

Rev. 0 evaluation model. The decay heat in the analysis proposed for implementation, ANP-3011 (P), is always the ANSI/ANS 1979 standard for decay heat from Uranium-235 with fully saturated decay chains, corresponding to infinite operation, assuming 200 MeV [million electron volts]. The licensee compared the decay heat resulting from this approach to several finite operation decay models that included 2-sigma uncertainty treatments.

The two-sigma uncertainty evaluation for the finite operation curves included a separate uncertainty value for each of the decay groups, meaning the uncertainty as a fraction of the total decay heat was a dependent value. Two seconds following shutdown, the infinite chain was bounding of the other chains. During the first two seconds, when the infinite chain is not bounding, it was determined that the difference in energy deposited between the infinite chain and the bounding finite chain would have an insignificant effect on clad temperature.

The NRC staff finds that this treatment of decay heat is acceptable because the decay heat behavior is modeled assuming that the fuel content will deliver more energy to the coolant over a longer period of time than models that employ more realistic assumptions and include an uncertainty treatment.

3.5.4 Clad Rupture and Swelling; Fuel Relocation The licensee has augmented the S-RELAP5-based EM with cladding strain and rupture correlations from Topical Report BAW-10227(P)-A: "Evaluation of Advanced Cladding and Structural Material (M5Ž) in PWR Reactor Fuel." This NRC-approved licensing topical report describes the fuel mechanical models that are used in the ECCS EM that is approved for use in the Babcock and Wilcox (B&W)-designed nuclear steam supply systems. While the B&W EM framework is not necessarily applicable to other PWR systems, the use of fuel correlations is acceptable because the fuel is fabricated from similar materials, and the post-LOCA phenomena existing in PWRs is expected to be sufficiently similar as not to cause one to question the applicability of the fuel mechanical models. These correlations were used to predict the hot pin and assembly cladding structural performance under the LOCA conditions.

Then, to simulate fuel relocation, the licensee augmented the fuel pellet to cladding heat transfer prescriptively.

The relocated fuel was assumed to be in contact with the strained and ruptured cladding surface, with a heat source term adjusted to reflect a prescribed packing fraction.

Based on the licensee's review of the available data concerning fuel relocation, the licensee initially proposed to assume pellet to cladding heat transfer characteristics associated with relocating fuel with an approximately 50 percent packing fraction.

This selection was based on a review of data, and a conclusion that the best quality of data came from post-irradiation examination of high burnup rods. Since 10 CFR 50.46(a)(1)(i) requires that comparisons be made to experimental data to quantify uncertainty, the staff believes that the limited database should also include gamma scan and post experiment analytical data, some of which indicated fuel relocation to a packing fraction as high as 80 percent. Although the NRC staff agrees that the limited data available are widely varied, the NRC staff finds that it is necessary to consider all available data in the absence of a large database that could significantly reduce the variance and uncertainty of available information concerning pellet relocation.

The licensee performed sensitivity studies on those cases that were predicted to experience cladding rupture to determine the change in PCT associated with assuming an 80

-percent packing fraction as stated in ANP-3011-Q1 (P). The NRC staff finds that this approach is prudent and conservative in recognition of the limitations of the available data concerning fuel pellet relocation and is acceptable.

The licensee also proposed to incorporate droplet shattering effects via a Weber number-based correlation to modify the interfacial heat transfer based on core geometry changes associated with hot assembly fuel pin strain and rupture. Based on the NRC staff's concern that the licensee's approach would require substantial review effort to determine whether this change to the S-RELAP5 model realistically modeled dispersed flow film boiling, the licensee performed additional sensitivity studies to compensate for the effect of this modeling feature and effectively disable it. The NRC staff finds that the licensee's sensitivity studies quantify the impact of the droplet shattering model, so that a PCT can be calculated that does not benefit from the droplet shattering model. The NRC staff finds the licensee's approach acceptable.

The licensee's sensitivity studies determined that, in order to express a PCT that is not based on the use of a droplet shattering model, and that includes consideration of fuel relocation to 80 percent packing fraction following cladding rupture, the predicted PCT would increase by 138°F to 2071 OF. The predicted fuel element PCT is within the 2200°F acceptance criteria in 10 CFR 50.46(b)(1).

3.5.5 Changes to Technical Specifications GL 88-16 provides the regulatory framework for establishing a COLR, and for including a list of references to be used in the generation of such a report in the facility TSs. A parenthetical element of guidance appearing in the Enclosure to GL 88-16 states, " ... the individual specifications that address (core) operating limits may be referenced

.... " The HNP TS references listed in TS 6.9.1.6 include the TS requirement consistent with GL 88-12. The supplement provided by the HNP licensee on April 2, 2012 is consistent with this formatting, and with the guidance contained in GL 88-16. 3.6 Conclusion Based on its review, the NRC staff concludes that the proposed implementation of RLBLOCA is acceptable.

The NRC staff determined the following with respect to the analysis proposed for implementation: The analysis includes an acceptable PCT offset to account for cladding rupture and fuel relocation, The analysis includes a model for decay heat that provides appropriate treatment of uncertainties, The concerns identified in NRC IN 2009-23 are adequately addressed, The analysis considers both pre-existing and aCCident-generated cladding oxidation, The licensee has incorporated elements of the EMF-2103(P)(A)

Rev. 0 Transition Package, and The licensee's results indicate that HNP would not exceed 10 CFR acceptance Based on these considerations, the NRC staff finds that the HNP ECCS evaluation shows that

-11 there is a high level of probability that the 10 CFR 50.46(b) acceptance criteria would not be exceeded under postulated large break LOCA conditions, consistent with the requirement at 10 CFR 50.46(a)(1

)(i). On this basis, the NRC staff finds the licensee's proposed implementation of ANP-3011 (P) Rev. 1 acceptable.

4.0 REGULATORY COMMITMENTS In the letter dated February 23, 2012 the licensee made the following regulatory commitment:

CP&L commits to apply a 138 degree Fahrenheit conservative adder to peak cladding temperatures calculated using the plant-specific methodology that implements AREVA 's NRC-approved topical report EMF-21 03(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors, " Rev. O. The 138 degree Fahrenheit conservative adder will be reflected in reports of peak cladding temperature submitted in accordance with 10 CFR 50.46 (a)(3). In February 23,2012, letter, the licensee clarified that the commitment only applies to the use of the Topical Report EMF-2103(P)(A), Rev. O. However, the licensee stated, in letter dated April 2, 2012, that CP&L rescinded its request for use of EMF-2103, Revision 2 and higher upon approval of the specific revision by the NRC. Therefore, the NRC's approval as described in this safety evaluation only applies to the plant-specific methodology of ANP-3011 (P), Revision 1, that implement EMF-2103(P)(A), Revision O.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment.

The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to 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 (January 10, 2012; 77 FR 1516). 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.

7.0 CONCLUSION

The NRC 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

-12 conducted in compliance with the NRC'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 Contributors:

Benjamin Parks Leonard Ward Date: May 30,2012 May 30,2012 Mr. Christopher L Burton Vice President Shearon Harris Nuclear Power Plant Progress Energy Carolinas, Inc. Post Office Box 165, Mail Code: Zone 1 New Hill, NC 27562-0165 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 -ISSUANCE OF AMENDMENT RE: THE REVISION TO TECHNICAL SPECIFICATION CORE OPERATING LIMITS REPORT REFERENCES FOR REALISTIC LARGE BREAK LOSS-OFCOOLANT-ACCIDENT ANALYSIS (TAC NO. ME6999)

Dear Mr. Burton:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 138 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). This amendment changes the HNP Technical Specifications (TSs) in response to your application dated August 22, 2011, as supplemented by letters dated February 23, March 20, and April 2, 2012. The amendment revises HNP TS 6.9.1.6, "Core Operating Limits Report," to add plant specific methodology ANP-3011 (P),

Nuclear Plant Unit 1 Realistic Large Break LOCA [Ioss-of-coolant accident]

Analysis," Revision 1, that implements AREVA's NRC-approved Topical Report EMF-21 03 (P)(A) , "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Revision O. A copy of the related safety evaluation is enclosed.

A notice of issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, IRA by Farideh Saba forI Araceli T. Billoch Col6n, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. Amendment No. 138 to NPF-63 2. Safety Evaluation cc w/enclosures:

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