ML18045A060

Issuance of Amendment a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in RG 1.183, Tale 3, Footnote 11 (CAC No. MF9740; EPID L-2017-LLA-0233)

ML18045A060
Person / Time
Site:	Harris
Issue date:	03/26/2018
From:	Martha Barillas Plant Licensing Branch II
To:	Hamilton T Duke Energy Carolinas
Barillas M DORL/LPL2-2 301-415-2760
References
CAC MF9740, EPID L-2017-LLA0233, RG-1.183
Download: ML18045A060 (23)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Ms. Tanya M. Hamilton Site Vice President Shearon Harris Nuclear Power Plant Duke Energy 5413 Shearon Harris Road M/C HNP01 New Hill, NC 27562-0165 March 26, 2018

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING A NEW SET OF FISSION GAS GAP RELEASE FRACTIONS FOR HIGH BURNUP FUEL RODS THAT EXCEED THE LINEAR HEAT GENERATION RATE LIMIT DETAILED IN REGULATORY GUIDE 1.183, TABLE 3, FOOTNOTE 11 (CAC NO. MF9740; EPID L-2017-LLA-0233)

Dear Ms. Hamilton:

The U.S. Nuclear Regulatory Commission (Commission) has issued Amendment No. 163 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1. This amendment changes the plant's licensing basis as described in the Updated Final Safety Analysis Report in response to your application dated May 22, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17142A411 ), as supplemented by letters dated October 30 and November 29, 2017 (ADAMS Accession Nos. ML17303A667 and ML17338A122).

The amendment changes the facility as described in the Updated Final Safety Analysis Report to provide gap release fractions for high-burnup fuel rods (i.e., greater than 54 gigawatt days per metric ton uranium that exceed the 6.3 kilowatt per foot linear heat generation rate limit detailed in Table 3 of Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0.

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

Docket No. 50-400

Enclosures:

1. Amendment No. 163 to NPF-63

2. Safety Evaluation cc: Listserv Sincerely, Martha Barillas, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 163 Renewed License No. NPF-63

1.

The U.S. Nuclear Regulatory Commission (Commission) has found that:

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated May 22, 2017, as supplemented by letters dated October 30 and November 29, 2017, 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 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.

2.

Accordingly, by amendment No. 163, the license is amended to authorize revision to the Updated Final Safety Analysis Report (UFSAR), as set forth in the application dated May 22, 2017, as supplemented by letters dated October 30 and November 29, 2017.

The licensee shall update the UFSAR to incorporate the change as described in the licensee's application dated May 22, 2017, as supplemented by letters dated October 30 and November 29, 2017, and the NRC staff's safety evaluation attached to this amendment, and shall submit the revised description authorized by this amendment with the next update of the UFSAR. Also, the license is amended by changes as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-63 is hereby amended to read as follows:

(2)

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. 163, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance. The UFSAR changes shall be implemented in the next periodic update to the UFSAR in accordance with 10 CFR 50.71 (e).

Attachment:

Changes to the Renewed License FOR THE NUCLEAR REGULATORY COMMISSION ii=p~e:~f Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 2 6, 201 8

ATTACHMENT TO LICENSE AMENDMENT NO. 163 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 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 marginal line indicating the area of change:

Remove Page 4 Insert Page 4

1 C.

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.

(1)

Maximum Power Level Duke Energy Progress, LLC, is authorized to operate the facility at reactor Core power levels not in excess of 2948 megawatts thermal (100 percent rated core power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan (3)

(4)

(5)

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. 163, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Antitrust Conditions Duke Energy Progress, LLC. 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 1 O 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 II (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 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.

• On April 29, 2013, the name "Carolina Power & Light Company" (CP&L) was changed to "Duke Energy Progress, Inc." On August 1, 2015, the name "Duke Energy Progress, Inc." was changed to "Duke Energy Progress, LLC."

Renewed License No. NPF-63 Amendment No. 163

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 163 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By letter dated May 22, 2017 (Reference 1 ), as supplemented by letters dated October 30 and November 29, 2017 (References 2 and 3), Duke Energy Progress, LLC (the licensee) submitted a request to change the Shearon Harris Nuclear Power Plant, Unit 1 (Shearon Harris or HNP),

licensing basis. The requested change would amend the facility as described in the Updated Final Safety Analysis Report (UFSAR) to provide gap release fractions for high-burnup fuel rods (i.e., greater than 54 gigawatt days per metric ton uranium (GWD/MTU)) that exceed the 6.3 kilowatt per foot (kW/ft) linear heat generation rate (LHGR) limit detailed in Table 3 of Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, July 2000 (Reference 4).

The supplements dated October 30 and November 29, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U. S. Nuclear Regulatory Commission (NRC) staffs initial proposed no significant hazards consideration determination as published in the Federal Register on August 29, 2017 (82 FR 41067).

2.0 REGULATORY EVALUATION

The gap release analysis determines release fractions for a variety of volatile fission products in the gap between the pellet and cladding of a fuel rod. The computed release fractions correspond to a proposed increase in the Regulatory Guide 1.183 allowable fuel rod LHGR above 54 GWD/MTU burnup. The results of this analysis are used as isotopic inventory input to dose calculations for the fuel handling accidents. Shearon Harris has implemented the Alternative Source Term (AST) method in accordance with Regulatory Guide 1.183 (Reference 5).

Regulatory Guide 1.183 Table 3, provides gap release fractions for various volatile fission product isotopes and isotope groups, to be applied to non-loss-of-coolant-accidents (LOCAs).

This table limits the fuel rod LHGR to 6.3 kW/ft for rod burnups above 54 GWD/MTU, but a footnote to the table (Footnote 11) states that gap fractions calculated directly by the licensee may be considered on a case-by-case basis, if the calculations follow NRG-approved methodologies.

In recent years, experimental data have demonstrated that fuel pellets undergo significant thermal conductivity degradation (TCD) at high burnup, which increases interior fuel pellet temperatures. The NRC Information Notice 2009-23 titled Nuclear Fuel Thermal Conductivity Degradation (Reference 6), discusses this issue in more detail. Higher fuel temperatures will yield larger fission gas release fractions in the American Nuclear Society (ANS) 5.4 [1982] and

[2011] models (References 7 and 8), particularly in the high-burnup range. The ANS 5.4 [1982]

standard has been revised, and the updated ANS 5.4 [2011] acknowledges the conservatism of the previous version, based on additional experimental data after 1982. The revised standard mandates the use of an NRG-approved fuel performance code that accounts for TCD, in determining temperature inputs for the gap fraction computations.

The Shearon Harris license amendment request (LAR) proposed to revise the dose consequences for the facility, as described in the Shearon Harris UFSAR, to provide for non-LOCA gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit in Footnote 11 of Table 3 in Regulatory Guide 1.183, "Non-LOCA Fraction of Fission Product Inventory in Gap."

The increases proposed are as follows:

The values in Regulatory Guide 1.183, Table 3 will be tripled for 85Kr [krypton],

134Cs [cesium], and 137Cs.

These increased gap fractions allow LHGRs above 6.3 kW/ft for rod burnup above 54 GWD/MTU, as long as the LHGRs remain within the bounding power history evaluated in Section 3.1 of the LAR.

The regulatory requirements and guidance, which the NRC staff considered in its review of the LAR are as follows:

Title 10 of the Code of Federal Regulations (10 CFR) 50.67, "Accident source term" (b)(2) states:

The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i)

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem)1 total effective dose equivalent (TEDE).

(ii)

An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from 1 The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.

the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent {TEDE).

(iii)

Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent {TEDE) for the duration of the accident.

Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants,"

Criterion 19 - Control room, states:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. Footnote 11 states:

As an alternative [to the non-LOCA gap fractions in Table 3 and the limits of Footnote 11 ], fission gas release calculations performed using NRG-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load.

To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the Boiling-Water Reactor (BWR) rod drop accident and Pressurized-Water Reactor (PWR) rod ejection accident, the gap fractions are assumed to be 10 percent for iodines and noble gases.

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, July 2000 (Reference 9) provides guidance to the NRC staff for the review of alternative source term amendment requests. The SRP Section 15.0.1 states that the NRG reviewer should evaluate the proposed change against the guidance in Regulatory Guide 1.183.

"Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Re: Steam Generator Replacement and Power Uprate (TAC Nos. MB0199 and MB0782)," dated October 12, 2001 (Reference 5) used an AST methodology for analyzing the radiological consequences of eleven design-basis accidents (DBAs) using Regulatory Guide 1.183. The NRC staff also considered relevant information in Chapter 15 of the Shearon Harris UFSAR, which describes the DBAs and the evaluation of their radiological consequences.

Section 50.71 of 10 CFR, "Maintenance of records, making of reports," contains the requirements for updating a facility's FSAR. The regulations in 1 O CFR 50. 71 ( e) require that the FSAR be updated to include all changes made in the facility or procedures described in the FSAR and all safety evaluations performed by the licensee in support of requests for license amendments. The analyses required by 10 CFR 50.67 are subject to this 10 CFR 50.71(e) requirement. Therefore, the affected radiological analyses descriptions in the Shearon Harris UFSAR will be updated to reflect the changes included with this amendment.

3.0 TECHNICAL EVALUATION

For design basis non-LOCA events in which fuel damage is assumed or predicted to occur, the quantity of each isotope released from the fuel is important input to the dose calculation. To determine the quantity of each isotope released, gap fractions and the number of failed fuel rods are used to determine the source term of the accident. Gap fractions describe, within a fuel rod, the fraction of the total inventory of an isotope that resides in the void region of the fuel rod (i.e.,

gap between the fuel pellet and the fuel cladding and plenum). When fuel damage is assumed to occur, the gas in the gap, containing these isotopes, is assumed to be released from the fuel for the purpose of modeling and predicting dose.

3.1 Fission Gas Gap Fractions Precedent The NRC staff has previously reviewed and approved a similar amendment, with regards to fission gas gap fraction, for McGuire Nuclear Stations, Unit Nos. 1 and 2, Catawba Nuclear Station, Unit Nos. 1 and 2, and Oconee Nuclear Station, Unit Nos. 1, 2, and 3 (Reference 10).

This review included an audit (Reference 11) of the licensee's methods for evaluating the fission gas gap fraction for rods that exceeded the limits in Footnote 11 of Regulatory Guide 1.183 (Reference 4). The review found the licensee's fission gas gap fractions acceptable. The NRC staff performed confirmatory calculations using FRAPCON 4.0 (Reference 12) to provide assurance that Duke Energy's proposed gap fraction multipliers were conservative and appropriate, and found them to be so. The NRC staff verified the licensee's use of the COPERNIC code (Reference 13) and the gapfrac macro to ensure the underlying engineering calculations were consistent with the ANS 5.4 standard. These were found to be acceptable.

The NRC staff found a similar license amendment request acceptable for H. B. Robinson Steam Electric Plant, also operated by Duke Energy, which used the same methodology to calculate fission gas gap fractions (Reference 14).

3.1.1 Fission Gas Gap Fractions Staff Evaluation In this LAR, the licensee is proposing new gap fractions for AST determination in rods that exceed the 6.3 kW/ft peak rod average power after achieving burnups over 54 GWD/MTU. The licensee has calculated new gap fractions to support these higher LHGRs using the COPERNIC fuel performance code methodology and the ANS 5.4 standard as updated in 2011 (Reference 8).

The NRC staff approved the use of this approach in a previous SE (Reference 14) for Duke Energy, as described in the previous section. This LAR makes a similar request with two significant changes. First, the previously approved LAR used both the ANS 5.4 ( 1982) and the updated ANS 5.4 (2011) standards for calculating fission gas gap fractions, and used the most conservative results for each isotope group. The Shearon Harris LAR only uses the 2011 standard. The NRC staff finds this acceptable, as the approach for calculating fission gas gap fractions in the ANS 5.4 (2011) standard and the additional margin added by the licensee, coupled with the use of a bounding LHGR profile, results in a conservative estimation of fission gas gap fractions.

The other notable change is the use of a different LHGR vs. burnup limit. The NRC staff determined this was acceptable, as these inputs remain within the scope of the methodology that the NRC staff has previously reviewed and approved (Reference 10).

For the Shearon Harris LAR, the licensee calculated the gap inventory of the various isotopes by plugging nodal fuel temperatures from COPERNIC into a Visual Basic for Applications macro named gapfrac. Duke Energy use of Gapfrac applies the ANS 5.4 (2011) standard that the NRC staff previously reviewed and found acceptable (Reference 10). The COPERNIC calculations take into account the burnup and LHGR rates described by the licensee. The resulting isotope gap fractions are the same or lower than Regulatory Guide 1.183, Table 3, with the exception of Kr-85, Cs-134, and Cs-137. For these isotopes, the ratio between the gap fractions calculated by the licensee and that in Regulatory Guide 1.183, Table 3 is shown in the table below, taken from the LAR, Enclosure I, Table 4 of the submittal. The bounding gap fraction is taken from the LAR, Enclosure I, Table 5.

Isotope RG 1.183 HNP HTP* fuel Ratio Bounding Gap Table 3 Value calculated Fraction maximum gap fraction Kr-85 0.10 0.212 2.12 0.30 Cs-134 0.12 0.300 2.50 0.36 Cs-137 0.12 0.300 2.50 0.36

• HTP = high thermal performance In order to provide an added degree of conservatism, the licensee has applied a multiplier of 3 to the Regulatory Guide 1.183 values for all three of these isotopes, as shown in the final column. The remaining isotopes remain consistent with the values in Regulatory Guide 1.183, Table 3.

The methods used by Shearon Harris in the calculation of the fission gas gap fractions have been previously reviewed and found acceptable for similar Duke Energy plants and use cases.

Since the parameters in this LAR remain within the scope of the approved methods, the NRC staff finds the fission gas gap fractions for fuel rods that exceed the burnup and LHGR limits listed in Regulatory Guide 1.183, Table 3, Footnote 11, acceptable for use within the LHGR and burnup limits as described in Table 3 of the LAR.

3.2 Design Basis Accidents Dose Consequence Analysis The non-LOCA gap fractions stated in Table 3 of Regulatory Guide 1.183 are applied to the non-LOCAs if fuel failure occurs during the accident. The following accidents at Shearon Harris assume fuel failure: fuel-handling accident (FHA), locked-rotor accident (LRA), and control rod ejection accident (CREA). In the LAR, the licensee stated that no non-LOCAs that may result in departure from nucleate boiling are considered (e.g., LRA, CREA) because the fuel cycles for Shearon Harris will be designed so that no fuel rod predicted to enter departure from nucleate boiling will have been operated beyond the current limit in Regulatory Guide 1.183 Footnote 11 for maximum LHGR. However, Duke Energy did not explain how this new design feature would be incorporated into the Shearon Harris licensing basis. The staff requested additional information for Duke Energy to explain how it will incorporate this new design feature into the Shearon Harris licensing basis or for Duke Energy to provide the revised radiological consequence analyses for the other DBAs that assume fuel failure to demonstrate that the regulatory limits will be met with the proposed gap fractions for high-burnup rods. In its Request for Additional Information (RAI) response dated October 30, 2017 (Reference 2), Duke Energy stated that to address this new fuel cycle configuration design requirement in the Shearon Harris licensing basis, Duke Energy will add the following paragraph in HNP UFSAR Section 15.0.9.1:

Inventory in the gap between the fuel pellet and fuel rod cladding For some design basis accidents, fuel damage as predicted in bounding analyses is limited to fuel cladding failure. For these design basis accidents, the radioactive source term is limited to the activity initially in the gap between the fuel pellets and the fuel rod cladding. The percentage of the core fission product inventory in this gap -- the gap fraction -- for these non-LOCA accidents is consistent with the values from Table 3 in Regulatory Guide 1.183, except for the fuel handling accidents in the Fuel Handling Building and in Containment. For the fuel handling accidents, gap fractions for Kr-85, Cs-134, and Cs-137 are increased by a factor of 3 from the Regulatory Guide 1.183, Table 3 values (Reference 15. 7.4.11 ). The fuel cycle design ensures that no fuel rod predicted to experience DNB [departure from nucleate boiling] in any other non-LOCA accidents will have operated beyond the power/burnup criteria of Footnote 11 in Regulatory Guide 1.183.

The licensee proposed to apply the following bounding gap fractions to HNP FHA in containment and the fuel-handling building (FHB):

Isotope or isotope group Gap fraction from Table 3 of Bounding Gap fraction RG 1.183, Revision 0 1-131 0.08 0.08 Kr-85 0.10 0.30 Other Noble Gases 0.05 0.05 Other HaloQens 0.05 0.05 Cs-134 (Alkali Metal) 0.12 0.36 Cs-137 (Alkali Metal) 0.12 0.36 Other Alkali Metals 0.12 0.12 In addition, the licensee proposed to lower the control room unfiltered in leakage from 500 cubic feet per minute (cfm) to 300 cfm.

3.2.1 Fuel-Handling Accident (FHA)

The licensee reevaluated the radiological consequences of a postulated FHA occurring inside and outside of the containment in the FHB. In its submittal, Duke Energy concluded that the release of fission products following an FHA will result in doses that are well within the dose criteria specified in 10 CFR 50.67 for the exclusion area boundary, low population zone, and control room operator. The licensee reached this conclusion on the basis of the assumptions and parameters provided in Table 6, HNP Fuel Handling Accident-Dose Consequence Analysis Inputs, of its LAR (Reference 1 ).

The FHA in the FHB involves dropping a recently discharged PWR assembly (including the handling tool) on top of another recently discharged PWR assembly in a fuel storage rack. The dropped assembly subsequently falls, over landing on BWR fuel assemblies (note that discharged Brunswick BWR fuel assemblies are currently being stored in the Shearon Harris spent fuel pools) in an adjacent storage rack. Fifty fuel rods are projected to fail in the impacted PWR assembly in storage and all of the rods (264) in the dropped assembly fail when the assembly falls over. Due to the upper bail handle of the BWR fuel assemblies extending above the top of the BWR storage racks, up to 52 BWR assemblies could be impacted when the dropped PWR assembly falls over. All of the rods in the impacted BWR assemblies are assumed to fail.

The FHA in containment involves dropping a recently discharged PWR assembly. It is assumed that all of the fuel rods in the equivalent of one fuel assembly (264 rods) are damaged to the extent that all their gap activity is released.

3.2.2 Source Term Following reactor shutdown, decay of short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The licensee takes credit for the normal decay of irradiated fuel and the analysis is performed assuming a decay period of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown for the PWR fuel rods and 4 years after shutdown for the BWR fuel assemblies. The radial peaking factor is 1.73 for the PWR fuel rods and 1.50 for the BWR fuel assemblies. The reactor power used in the analysis is 2958 megawatts.

The fission product inventory that constitutes the source term for this event is the gap activity in the fuel rods assumed to be damaged as a result of the postulated design basis FHA. Volatile constituents of the core fission product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod cladding during normal power operations. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released to the surrounding water as a result of the accident. Consistent with Regulatory Guide 1.183, the radionuclide isotopes considered are xenon, krypton, halogen, cesium and rubidium. The source term used for the Shearon Harris FHAs in this LAR is unchanged from the licensing basis source term that is stated in the Shearon Harris UFSAR Table 15.7.4-1, "Activity in the damaged fuel assemblies (Ci)," and Table 15. 7.4-3, "Activity in the damaged fuel assembly (Ci)." This source term accounts for radial peaking factors and decay period for the affected assemblies and is shown below.

Activity in the Activity in the damaged fuel damaged fuel Isotope assemblies assemblies (curies) for FHA (curies) for FHA in FHB in Containment lodine-131 7.21 ES 6.06E5 lodine-133 7.59E4 6.38E4 lodine-135 5.57E1 4.68E1 Krypton-85 1.41 ES 8.82E3 Xenon-131m 9.06E3 7.61 E3 Xenon-133m 1.77E4 1.49E4 Xenon-133 1.19E6 9.97E5 Xenon-135 2.41E2 2.03E2 The licensee assumed that all Shearon Harris HTP fuel rods involved in the accident could exceed the maximum LHGR limit of 6.3 kW/ft for burnups exceeding 54 GWD/MTU. The gap release fractions for rods that exceed the 6.3 kW/ft LHGR limit above 54 GWD/MTU would increase only for krypton (Kr)-85, cesium (Cs)-134, and Cs-137, which are tripled from Regulatory Guide 1.183. All other isotopes maintain the release fractions outlined in Table 3 of Regulatory Guide 1.183. Regulatory Guide 1.183 Appendix B, Regulatory Position 3, states that the particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity. Therefore, the licensee assumed that all radionuclide groups other than iodine and noble gases remain in nonvolatile form, and are not released from the water.

Fission products released from the damaged fuel are decontaminated by passage through the overlaying water in the reactor cavity or spent fuel pool depending on their physical and chemical form. Following the guidance in Regulatory Guide 1.183, Appendix B, Regulatory Position 1.3, the licensee assumes: (1) that the chemical form of radioiodine released from the fuel to the spent fuel pool consists of 95 percent cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide, (2) the Csl released from the fuel completely dissociates in the pool water, and (3) because of the low pH of the pool water, the Csl re-evolves and releases elemental iodine. This results in a final iodine distribution of 99.85 percent elemental iodine and 0.15 percent organic iodine. The licensee assumes that the release to the pool water and the chemical redistribution of the iodine species occurs instantaneously.

The iodine decontamination factor used for the Shearon Harris FHAs in this LAR is unchanged from the licensing basis decontamination factor, which is 200 for a water depth of 21 feet in the FHB analysis and 22 feet in the containment analysis. Consistent with Regulatory Guide 1.183, the licensee credits an infinite decontamination factor for the remaining particulate forms of the radionuclides contained in the gap activity and did not credit decontamination from water scrubbing for the noble gas constituents of the gap activity.

The licensee used conservative assumptions to evaluate the FHA source term, and therefore, the NRC staff finds the source term to be acceptable for the AST FHA analysis.

3.2.3 Transport 3.2.3.1 FHA in FHB Section 15. 7.4.2.2, "Postulated Fuel Handling Accident in the FHB," of the Shearon Harris UFSAR contains the radiological consequence analysis of a postulated FHA in the FHB. In this analysis, credit is taken for removal of iodine by filters during the operation of the spent fuel pool ventilation system. The spent fuel pool area ventilation system is a part of the FHB Heating Ventilating and Air Conditioning System. The UFSAR 6.5.1.2.1, "FHB emergency exhaust system," states:

... Following a fuel handling accident radioactivity released from fuel rods will be detected by the radiation monitors located around the fuel pools. These radiation monitors will then signal the switchover from the normal to the emergency ventilation and filtration system. The switchover time is 30 seconds for the emergency ventilation and filtration system to become fully operational. The isolation of the normal ventilation system is accomplished in ::s; 10 seconds.

Either train may then be manually de-energized from the Control Room and placed on standby. Negative pressure is established at 1/8 in. wg. [inches of water gauge] by continuously exhausting air from the operating floor....

Assuming the maximum distance traveled by the FHA radioactive gas is half the total distance to the exhaust register, the resulting travel time is 13.65 seconds.

This is the longest time considered possible before the radiation detectors initiate an alarm signal to isolate...

The Shearon Harris Technical Specification (TS} Limiting Condition of Operation (LCO) 3/4.9.12, "Fuel Handling Building Emergency Exhaust System," requires two independent fuel handling building emergency exhaust system trains to be operable whenever irradiated fuel is in a storage pool. TS LCO 3.9.12 has a footnote that states, "The Fuel Handling Building Emergency Exhaust System boundary may be opened intermittently under administrative controls."

During its review, the NRC staff identified the footnote to TS LCO 3.9.12 is not consistent with the licensing basis radiological consequence analysis of the postulated FHA in the FHB for Shearon Harris, which assumes the FHB emergency exhaust system is established with a negative internal pressure within 1 minute after occurrence of a FHA in the FHB. The footnote allows the FHB emergency exhaust system boundary to be open for an indefinite length of time, in addition to its unlimited use. Therefore, the NRC staff requested additional information for the licensee to provide a revised radiological consequence analysis of the FHA in the FHB that supports the FHB emergency exhaust system boundary being open for the duration of the event and has radiological dose results that meet the limits in 10 CFR 50.67 and General Design Criterion (GDC) 19 of 10 CFR Part 50, Appendix A or to provide a proposed change to TS 3.9.12 that is consistent with the design basis as reflected in the Shearon Harris UFSAR, Sections 15.7.4.2.2 and 6.5.1.2.1. In the licensee's RAI response dated October 30, 2017 (Reference 2) the licensee stated that in order to maintain operational flexibility as stated in TS LCO 3.9.12, including its footnote, a revised radiological consequence analysis of the FHA in the FHB was performed that conservatively removes all filtration credit for the FHB emergency exhaust system. Tables 6 and 7, as provided in the original LAR (Reference 1 ), were updated (References 2 and 3) to reflect the revised radiological consequence analysis as seen in Tables 1 and 2 below. Control room parameters used in the FHA radiological analysis were provided and can be found in Table 3 below. The methodology and technical basis for the revised analysis remain the same as discussed in Section 3.2 of the original LAR. The revised doses continue to satisfy the limits in 10 CFR 50.67 and 1 O CFR Part 50, Appendix A, GDC 19.

The licensee modeled releases from the FHA in FHB via the spent fuel pool ventilation system with no credit taken for filtration of the iodine. Prior to the FHA in the FHB, the normal FHB Heating Ventilating and Air Conditioning (HVAC) system air supply and exhaust systems maintains ventilation above the spent fuel pool. Upon receipt of high radiation signal, the radiation monitors located along the walls of each fuel pool will cause isolation dampers in the common ducts to close before the contaminated air rising from the surface of the pools reaches the ducts; the normal FHB HVAC system will be de-energized. These radiation monitors will also signal the switchover from the normal FHB HVAC system to the FHB emergency ventilation and filtration system. This is consistent with the current licensing basis that is discussed in the Shearon Harris UFSAR Sections 6.5.1 and 9.4.2. The FHB emergency ventilation and filtration system charcoal filter efficiency is not credited. Consistent with Regulatory Guide 1.183, the FHA in FHB is released over a 2-hour period.

The licensee used conservative assumptions to evaluate the FHA transport in the FHB. The NRC staff finds the licensee's assumptions for the FHA in FHB analysis are also consistent with Section 4 of Regulatory Guide 1.183, Appendix B, and therefore are acceptable.

3.2.3.2 FHA in containment For the FHA in containment, releases from the containment are through the containment openings such as the equipment door and the personnel airlock door. No credit is taken for isolation of the containment during the FHA in containment. The Shearon Harris TS 3.9.4 allows at least three different configurations during core alterations and during movement of irradiated fuel within containment. Duke Energy submitted a revised FHA in containment analysis that did not specify the configuration of the equipment door and each airlock to demonstrate the most bounding condition. The NRC staff did not find a discussion or analysis provided in the LAR submittal that explains that the resultant radiological doses provided in the LAR for the FHA in containment bounds all the possible configurations. It appears, depending on the plant building configuration, that the open airlocks could allow a pathway for activity to migrate from the open containment airlocks into the adjacent building and eventually into the control room. Therefore, the NRC staff requested that the licensee specify all the possible configurations of the equipment door and each airlock, and describe how the analysis provided bounds all the possible configurations. In addition, the NRC staff requested that the licensee explain if a pathway for activity to migrate from the open containment airlock into the adjacent building and eventually into the control room during ingress/egress of personnel is possible at Shearon Harris following a FHA event in containment. In its RAI response dated October 30, 2017 (Reference 2), the licensee stated the possible configurations of the personnel airlock, emergency airlock and equipment hatch during refueling activities and the resultant releases of activity following a FHA in containment are as follows:

1) One or both airlocks may be open while the equipment hatch is closed during refueling activities.

2) The equipment hatch may be open while both airlocks are closed during refueling activities.

3) The equipment hatch and one or both airlocks could be open during refueling activities.

In its RAI response, Duke Energy described the release path for each of these configurations.

The locations of releases of activity to the environment following a DBA are important to the calculations of control room atmospheric dispersion factors (x/Qs) and with them the calculations of control room doses for these accidents. Control room x!Qs were calculated for these release locations and show that the values taken for the control room x!Q currently used in the AST analysis of the FHA in containment, as provided in the LAR, produce the bounding control room dose for this scenario and continue to meet the 10 CFR 50.67 radiation dose limit.

The licensee evaluated the potential for activity released from an open containment airlock to the Reactor Auxiliary Building (RAB) and subsequently to the control room after a FHA in containment. The RAB HVAC system and RAB Emergency Exhaust System are designed to exhaust activity released to the RAB to the plant vent stack where it is monitored prior to release to the environment. In addition, the control room is separated from the remainder of the RAB by a wall whose only penetration is a sealed pipe chase. As such, the licensee concluded there is no potential for activity released from an open containment airlock to the RAB to migrate to the control room.

The NRC staff finds the licensee's response to be conservative and consistent with Section 5 of Regulatory Guide 1.183 Appendix B. Consistent with Regulatory Guide 1.183, the FHA in containment release rate to the environment from the airlocks and the equipment hatch is assumed to occur linearly over the 2-hour period. No credit is taken for mixing or filtration prior to being released to the environment. The licensee used assumptions that are consistent with Regulatory Guide 1.183 Appendix B to evaluate the FHA transport in containment and therefore, the NRC staff finds them to be acceptable.

3.2.4 Control Room Habitability assumptions for the FHA The licensee evaluated control room habitability for the FHA assuming that the control room HVAC system begins in normal mode. When radioactive material from the accident reaches the radiation monitors in the control room intakes it causes a high radiation signal almost immediately, and the radiation monitors initiate the automatic isolation of the control room emergency filtration system (CREFS). Within 15 seconds after event initiation, the post-accident recirculation mode of CREFS is entered. When this happens the normal flow rate of 1050 cfm shifts to 4000 cfm of filtered recirculation. The control room operators will manually initiate the pressurization mode of CREFS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the isolation signal which maintains the control room envelope at a positive pressure relative to the surrounding area. While in CREFS pressurization mode 400 cfm of makeup air is drawn from one of the two emergency intakes.

The CREFS has charcoal and high efficiency particulate air (HEPA) filters that have an efficiency of 99 percent for all forms of iodine. Additionally, during CREFS pressurization mode, part of the control room flow is recirculated and filtered through the charcoal and HEPA filters at a flow rate of 3600 cfm. In-leakage to the control room is 300 cfm.

The licensee evaluated the radiological consequences resulting from a postulated FHA and concluded that the radiological consequences at the exclusion area boundary (EAB), low population zone (LPZ), and control room are within the radiological dose guidelines provided in 1 O CFR 50.67 and accident-specific dose criteria specified in SRP Section 15.0.1 and Regulatory Guide 1.183. The NRC staff's review has found that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 above. The licensee's calculated dose results are given in Table 1 below, and the assumptions found acceptable to the NRC staff are presented in Tables 2, 3, and 4 below. To verify the licensee's analyses, the NRC staff performed confirmatory radiological consequence dose calculations and compared the results to those calculated by the licensee. The radiological consequences calculated by the NRC staff for the EAB, LPZ, and control room are consistent with those calculated by the licensee. Moreover, the radiological consequences calculated by both the licensee and the NRC staff are well within the radiation dose criteria set forth in 10 CFR 50.67. The NRC staff finds that the EAB, LPZ, and control room doses estimated by the licensee for the FHA meet the applicable accident dose criteria and are, therefore, acceptable.

Table 1 Radiological Dose Results (rem TEDE)

Current Licensee Dose SRP 15.0.1 and Accident RG 1.183 Dose UFSAR Results Acceptance Criteria Fuel-Handling Accident in Containment Exclusion Area Boundary 2.03 2.02 6.3 Low Population Zone 0.46 0.46 6.3 Control Room 1.39 0.88 5.0 Fuel Handling Accident in Fuel-Handling Building Exclusion Area Boundary 0.34 2.42 6.3 Low Population Zone 0.077 0.55 6.3 Control Room 0.12 1.05 5.0 Table 2 Parameters and Assumptions Used in the Fuel-Handling Accident Radiological Consequence Calculations Parameter Value Reactor Power with uncertainty 2958 megawatts thermal Number of fuel rods damaged In containment 264 rods In fuel handling building 314 rods plus 52 BWR fuel assemblies Radial peaking factors PWR fuel assemblies 1.73 BWR fuel assemblies in spent fuel pool 1.5 Fission product decay period PWR fuel assemblies 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> BWR fuel assemblies in spent fuel pool 4 years Number of high-burnup (above 54 GWD/MTU) 264 in containment rods t.hat can exceed 6.3 kW/ft 314 in fuel-handling building Fuel gap fission product inventory Krypton-85 30 percent lodine-131 8 percent Other Halogens and Noble Gas nuclides 5 percent Cesium 134 and 137 36 percent Alkali metals 12 percent Fuel pool water depth In containment 22 feet In fuel-handling building 21 feet Fuel pool decontamination factors Iodine 200 Noble gases 1

Table 2 Parameters and Assumptions Used in the Fuel-Handling Accident Radiological Consequence Calculations Parameter Value Iodine Filter efficiencies (elemental, organic, particulate)

In containment none assumed In fuel-handling building none assumed Iodine chemical form in release to atmosphere Elemental 70 percent Organic 30 percent Particulate 0

Isolation of release In containment none assumed In fuel-handling building none assumed Duration of fission product release 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Table 3 Control Room Parameters Used for FHA Radiological Analysis Volume 71,000 cubic feet Normal Ventilation Flow Rates Filtered Makeup Flow Rate 0

Filtered Recirculation Flow Rate 0

Unfiltered Makeup Flow Rate 1,050 cubic feet per minute Unfiltered In Leakage 300 cubic feet per minute Post-Accident Recirculation Flow Rates Filtered Makeup Flow Rate 0

Filtered Recirculation Flow Rate 4,000 cubic feet per minute Unfiltered In Leakage 300 cubic feet per minute Pressurization Mode Flow Rates Filtered Makeup Air Flow Rate 400 cubic feet per minute Filtered Recirculation Flow Rate 3,600 cubic feet per minute Unfiltered In Leakage 300 cubic feet per minute Iodine Filter Efficiencies Elemental 99 percent Organic 99 percent Particulate 99 percent Control Room Radiation Monitor Sensitivity 3.0E-6 micro curies per millimeter for Xenon-133 Control Room Radiation Monitor Location Emergency and Normal Air Intakes Delay to Initiate Switchover to Post-Accident 15 seconds Recirculation HVAC mode after radiation signal Operator Action Time to Switch to Pressurization 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Mode Breathing Rate for Duration of the Event 3.SOE-4 cubic meter per second Occupancy Factors 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 1 - 4 days 0.60 4 - 30 days 0.40 Table 3 Control Room Parameters Used for FHA Radiological Analysis Control Room Atmospheric Dispersion Factors 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.08E-3 seconds per cubic meter 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.16E-3 seconds per cubic meter 1 to 4 days 3.25E-4 seconds per cubic meter 4 to 30 days 1.23E-5 seconds per cubic meter Table 4 Offsite Atmospheric Dispersion Factors Exclusion Area Boundary 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.17E-4 seconds per cubic meter Low Population Zone Distance 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.4E-4 seconds per cubic meter 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0E-4 seconds per cubic meter 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 5.9E-5 seconds per cubic meter 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 2.4E-5 seconds per cubic meter

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on January 25, 2018. The State official had no comments.

5.0 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 amendmentinvolves no significant hazards consideration, and there has been no public comment on such finding (82 FR 41067). 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.

7.0 REFERENCE LIST

1.

Hamilton, T. M., Duke Energy Progress, LLC, letter to Document Control Desk, U.S.

Nuclear Regulatory Commission, "License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods that Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11,"

May 22, 2017 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML17142A411 ).

2.

Hamilton, T. M., Duke Energy Progress, LLC, letter to Document Control Desk, U.S.

Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding License Amendment Request Proposing a ~ew Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods that Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 (CAC No. MF9740}," October 30, 2017 (ADAMS Accession No. ML17303A667).

3.

Hamilton, T. M., Duke Energy Progress, LLC, letter to Document Control Desk, U.S.

Nuclear Regulatory Commission, "Supplement to License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods that Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 (CAC No. MF9740)," November 29, 2017 (ADAMS Accession No. ML17338A122).

4.

U. S. Nuclear Regulatory Commission, Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, July 2000 (ADAMS Accession No. ML003716792).

5.

Kalyanam, N., U.S. Nuclear Regulatory Commission, letter to Scarola, J., Carolina Power & Light Company, "Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Re: Steam Generator Replacement and Power Uprate," October 12, 2001 (ADAMS Accession No. ML012830516).

6.

U.S. Nuclear Regulatory Commission, Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," October 8, 2009 (ADAMS Accession No. ML091550527)

7.

ANSI/ANS 5.4-1982, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel," American National Standard published by the American Nuclear Society, November 1982.

8.

ANSI/ANS 5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel," American National Standard published by the American Nuclear Society, May 2011.

9.

U. S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,"

Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms,"

Revision 0, July 2000 (ADAMS Accession No. ML003734190).

10.

Hall, 'J. R., U.S. Nuclear Regulatory Commission, letter to Repko, R. T., Duke Energy, "Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; and Oconee Nuclear Station, Units 1, 2, and 3 - Issuance of Amendments Regarding Request to Use an Alternate Fission Gas Gap Release Fraction," July 19, 2016 (ADAMS Accession No. ML16159A336).

11.

Miller, G. E., U. S. Nuclear Regulatory Commission, letter to Repko, R. T., Duke Energy Carolinas LLC, "Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; and Oconee Nuclear Station, Units 1, 2, and 3, Plan for the Regulatory Audit Regarding License Amendment Request for Alternate Fission Gas Gap Fraction, March 21, 2016 (ADAMS Accession No. ML16067A291 ).

12.

Pacific Northwest National Laboratory, PNNL-19418, Volume 1, Revision 2, "FRAPCON-4.0: A Computer Code for the Calculation of Steady-State,

'Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burn up,"' and Volume 2, Revision 2, "FRAPCON-4.0: Integral Assessment," September 2015 (ADAMS Accession Nos. ML16118A427 and ML16118A434).

13.

BAW-10231 P-A, Revision 1, "COPERNIC Fuel Rod Design Computer Code," Framatone ANP (now AREVA), January 2004 (ADAMS Accession No. ML042930240).

14.

Galvin, D. J., U. S. Nuclear Regulatory Commission, letter to Kapopoulos, E., Jr., Duke Energy Progress LLC, H. B. Robinson Steam Electric Plant, Unit 2 - Issuance of Amendment Regarding Request to Modify the Licensing Basis Alternate Source Term (CAC No. MF8378), September 29, 2017 (ADAMS Accession No. ML17205A233).

Principal Contributors: K. Bucholtz J. Whitman Date: March 26, 2018

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING A NEW SET OF FISSION GAS GAP RELEASE FRACTIONS FOR HIGH BURNUP FUEL RODS THAT EXCEED THE LINEAR HEAT GENERATION RATE LIMIT DETAILED IN REGULATORY GUIDE 1.183, TABLE 3, FOOTNOTE 11 (CAC NO. MF9740; EPID L-2017-LLA-0233)

DATED March 26, 2018 DISTRIBUTION:

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