ML15027A366

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Issuance of Amendments DPC-NE-3001-P, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology
ML15027A366
Person / Time
Site: Catawba, McGuire, Mcguire  Duke Energy icon.png
Issue date: 03/25/2015
From: Geoffrey Miller
Plant Licensing Branch II
To: Henderson K
Duke Energy Carolinas
Miller G
References
TAC MF3119, TAC MF3120, TAC MF3121, TAC MF3122
Download: ML15027A366 (26)
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Text

UNITED STATES ",

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 25, 2015 Mr. K. Henderson Site Vice President Duke Energy Carolinas, LLC 4800 Concord Road York, SC 297 45

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 (CATAWBA 1 AND 2) AND MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 (MCGUIRE 1 AND 2) -

ISSUANCE OF AMENDMENTS RE: DPC-NE-3001-P, MULTIDIMENSIONAL REACTOR TRANSIENTS AND SAFETY ANALYSIS PHYSICS PARAMETERS METHODOLOGY (TAC NOS. MF3119, MF3120, MF3121, AND MF3122)

Dear Mr. Henderson:

By letter dated November 14, 2013, as supplemented by letters dated June 27, and November 10, 2014, Duke Energy Carolinas, LLC (Duke, the licensee), submitted a license amendment request for NRC review and approval of a revision to DPC-NE-3001-P, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."

The U.S. Nuclear Regulatory Commission has issued the en,closed Amendment No. 274 to Renewed-Facility Operating License NPF-35 and Amendment No. 270 to Renewed Facility Operating License NPF-52 for Catawba 1 and 2 and Amendment No. 277 to Renewed Facility Operating License NPF-9 and Amendment No. 257 to Renewed Facility Operating License for McGuire 1 and 2, respectively.

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

K. Henderson If you have any questions, please call me at 301-415-2481.

s&jf,\

G. Edward Miller, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413, 50-414, 50-369, and 50-370

Enclosures:

1. Amendment No. 274 to NPF-35
2. Amendment No. 270 to NPF-52
3. Amendment No. 277 to NPF-9
4. Amendment No. 257 to NPF-17
5. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 274 Renewed License No. NPF-35

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility),

Renewed Facility Operating License No. NPF-35, filed by Duke Energy Carolinas, LLC (the licensee), dated November 14, 2013, as supplemented by letters dated June 27, and November 10, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 conduct~d without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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.

Enclosure 1

2. Accordingly; by Amendment No. 274, Renewed Facility Operating License No. NPF-35 is hereby amended to utilize DPC-NE-3001-P, Revision 1, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-35 and the Technical Specifications Date of Issuance: March 25, 2015

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. NPF-52

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility),

Renewed Facility Operating License No. NPF-52, filed by Duke Energy Carolinas, LLC (the licensee), dated November 14, 2013, as supplemented by letters dated June 27 and November 10, 2014, complies with the standards and requirements of the Atomic E.nergy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; D. The issuance of this amendmentwill not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, by Amendment No. 270, Renewed Facility Operating License No. NPF-52 is hereby amended to utilize DPC-NE-3001-P, Revision 1, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-52 and the Technical Specifications .

Date of Issuance: March 25, 2015

\

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 277 Renewed License No. NPF-9

1. T~e Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the McGuire Nuclear Station, Unit 1.(the facility),

Renewed Facility Operating License No. NPF-9, filed by Duke Energy Carolinas, LLC (the licensee), dated November 14, 2013, as supplemented by letters dated June 27 and November 10, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

Enclosure 3

2. Accordingly, by Amendment No. 277, Renewed Facility Operating License No. NPF-9 is hereby amended to utilize DPC-NE-3001-P, Revision 1, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-9 and the Technical Specifications Date of Issuance: March 25, 2015

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 257 Renewed License No. NPF-17

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the McGuire Nuclear Station, Unit 2. (the facility),

Renewed Facility Operating License No. NPF-17, filed by Duke Energy Carolinas, LLC (the licensee), dated November 14; 2013, as supplemented by letters dated June 27 and November 10, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

Enclosure 4

2. Accordingly, by Amendment No. 257, Renewed Facility Operating License No. NPF-17 is hereby amended to utilize DPC-NE-3001-P, Revision 1, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-17 and the Technical Specifications Date of Issuance: March 25, 2015

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 274 TO RENEWED FACILITY OPERATING LICENSE NPF-35; AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NPF-52; AMENDMENT NO. 277 TO RENEWED FACILITY OPERATING LICENSE NPF-9; AMENDMENT NO. 257 TO RENEWED FACILITY OPERATING LICENSE NPF-17; DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413AND 50-414 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated November.14, 2013 (Reference 1), and supplemented by letters dated June 27, and November 10, 2014 (References 2 and 3, respectively), Duke Energy Carolinas, LLC (Duke),

the licensee for the McGuire and Catawba Nuclear Stations (McGuire and Catawba), proposed to amend the McGuire, Units 1 and 2, and Catawba, Units 1 and 2, operating licenses based on DPC-NE-3001-P, Revision 1, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."

DPC-NE-3001-PA, Revision 0, was approved by the U.S. Nuclear Regulatory Commission (NRG)

  • in 1991, and was republished with minor editorial changes in 2000 (Reference 4). Subsequently, a number of accumulated changes, many of which had received NRG approval in other documents, were made by Duke to DPC-NE-3001-PA through the 10 CFR 50.59 process, resulting in DPC-NE-3001-P, Revision. Oa. The NRG staff review of this methodology includes only the changes made to DPC-NE-3001-P in Revision 1, which were documented in Attachment 1a to Reference 1.

DPC-NE-3001-P serves two primary functions. First, it provides a determination of the key physics parameters for each Updated Final Safety Analysis Report (FSAR) Chapter 15 transient.

Enclosure 5

This enables a bounding analysis of each transient to be performed and confirmed to bound core reload designs and plant operating conditions on a cycle-by-cycle basis. Second, it provides methodologies for analysis of Chapter 15 transients for which multidimensional core physics effects are important: the control rod ejection (CRE) accident, the steam line break (SLB) transient, and the dropped rod transient.

The new version of the document, DPC-NE-3001-P, Revision 1, updates the methodology to cover the hot full power (HFP) SLB transient, new versions of several codes, and revisions to a

  • number of calculation methodologies. The licensee's letter also included an appendix to DPC-NE-3001-P, Revision 1, which analyzed and evaluated the impact of fuel thermal conductivity degradation (TCD) on the transient analysis methodologies.
2. 0 REGULATORY .EVALUATION The cur.rent design and licensing bases of Catawba and McGuire are documented in each plant's respective FSAR, as required of licensees by Title 10 of the Code of Federal Regulations ( 10 CFR)
  • Part 50.34. The analysis methodologies which Duke propqses to* revise in Reference 1 are used
  • to docu.ment .and generate these design bases. After NRC approval has been granted, updated safety analysis results will be incorporated into the FSARs as required by 10 CFR 50. 71 (e) and these methodologies may be utilized for subsequent revisions.

McGuire and Catawba both use the 10 CFR Appendix A General Design Criteria (GDCs) as their design basis. The GDCs ultimately provide the regulatory requirements that must be fulfilled by the licensee. The appropriate GDCs considered by the NRC staff in the review of DPC-NE-3001-P, Revision 1, are GDC 10, 11, 12, 13, 17, 20, 25, and 27; these requirements will be reviewed by the NRC staff in the context of the acceptance criteria provided in the Standard Review Plan (SRP), as documented in NUREG-0800 (Reference 5). The NRC staff found the following SRP sections to be relevant to the review of DPC-NE-3001-P, Revision 1 :.

  • SRP 15.0.2, "Review of Transient and Accident Analysis Methods," which provides staff guidance for the review of transient and accident analysis methodologies such as those provided in DPC-NE-3001-P, Revision 1 *
  • SRP 4.3, "Nuclear Design," which is relevant because the key safety analysis physics parameters determined in DPC-NE-3001-P; Revision 1 ultimately determine the constraints on the nuclear design *
  • SRP 15.4.8, "Spectrum of Rod Ejection Accidents (PWR)," which provides staff guidance on review of CRE accident analyses
  • SRP 15.1.5, "Steam System Piping Failures Inside and Outside of Containment (PWR),"

which provides guidance for staff review of SLBs

  • SRP4.2, "Fuel System Design," and SRP 4.4, "Thermal and Hydraulic Design," which provide staff guidance for the review of the fuel and thermal-hydraulic consequences of the transients discussed above The NRC staff will judge the ability of the methods presented to meet these requirements and acceptance criteria in the review of DPC-NE-3001-P, Revision 1.

Duke's methodology follows a process wherein bounding analyses are performed when significant changes are made to the plant. On a cycle-by-cycle basis, certain key parameters that are calculated each cycle are then checked against the analyses to ensure that the analysis of record remains bounding: The NRC staff's review will ensure that the changes to the methodologies will not impact their ability to bound expected operation. The review will also ensure that the cycle-specific evaluations are appropriately changed to reflect the methodology changes.

3.0 TECHNICAL EVALUATION

3.1 Control Rod Ejection Analysis Methodology The major changes proposed by Duke for their CRE analysis methodology involves updating the SIMULATE-3K code from Version 1 to Version 2 and moving from CASM0:.*3 to CASM0-4. A number of other models ~nd methodologies are modified. Additionally, the licensee considered a new rod ejection location.

The CRE accident is classified as a postulated accident or an ANS Condition IV event (i.e., an event for which fuel failure is allowed to occur, subject to certain offsite dose requirements). The acceptance criteria applied to CRE analyses are described in SRP 15.4.8, and require the analysis to demonstrate that offsite dose limits will be met and damage to the reactor coolant system pressure boundary is .not so severe as to impair core cooling capacity. Duke's methodology for CRE accident analysis as presented in DPC-NE-3001-P, Revision 1, does not contain the radiological consequence analysis methodologies necessary to support the offsite dose calculations, and these methodologies are therefore not subject to NRC staff review in the subject LAR. However, the offsite dose analysis for the CRE remains a criterion for acceptance.

The need for prior NRC review and approval should be evaluated by Duke against 10 CFR 50.59 ...

3.1.1 CRE Code Changes In Reference 1, SIMULATE-3K is updated from Version 1 to Version 2. This code change introduces a number of modei changes, including:

  • the ability to model subcritical neutron sources;
  • the ability to model the impact of coolant density changes on nuclear excore detectors
  • a 5-equation hydraulic channel model; and
  • burnup-dependent models for fuel gap conductance, fuel pellet thermal conductivity, and fuel pellet radial power profile.
  • Each of these models and model changes is considered. by Duke to be an enhancement to the existing models available in SIMULATE-3K Version 1. Duke has provided code documentation for the new models for SIMULATE-3K Version 2 in SSP-98/13, "SIMULATE-3K Models &

Methodology," Revision 6 (Reference 6), which was produced by Studsvik, the developer of SlMULATE-3K. Verification and validation of these models for LWR reactivity initiated transients was provided in SSP-04/443, "LWR Core Reactivity Transients, SIMULATE-3K Models and Assessment," Revision 2 (Reference 7). This document includes several benchmarks, showing good agreement with reference cases for both steady and transient analyses. The NRC staff has

also previously reviewed the use of SIMULATE-3K Version 2 in DPC-NE-3005-PA, "Oconee Nuclear Station UFSAR Chapter 15 Accident Analysis Methodology," Revision 3b, which was contained in an October 19, 2009, submittal (Reference 8) requesting the use of Gadolinia integral burnable absorbers in fuel at the Oconee Nuclear Station (Oconee). This revision was approved by the NRC in a July 21, 2011 letter (Reference 9), which therefore found SIMULATE-3K Version 2 acceptable at Oconee, for which Duke is also the licensee. The Oconee CRE analysis methodology is substantially similar to that proposed for, and already in use at, Catawba and McGuire. In a request for additional information (RAI), provided in a May 22, 2014 letter (Reference 10), the NRC staff asked Duke to provide benchmarks comparing the two versions of SIMULATE-3K. In their RAI response (Reference 2), Duke provided a benchmark that compared the calculated core response to a CRE accident at both end-of-cycle (EOC) HFP and hot zero power (HZP) using Version 1 and Version 2 of SIMULATE-3K. The NRC staff reviewed the results of the benchmarking and determined that the impact of the changes is minor. Based on the extensive validation, benchmarking, and the previous approval of the code, the NRC staff has determined that the change from Version 1 to Version 2 of SIMULATE-3K is acceptable.

Reference 1 also proposes a change from CASM0-3 to CASM0-4 to provide the cross sections for the transient SIMULATE-3K analysis. Duke has already obtained NRC approval for the use of CASM0-4 for steady-state nuclear design analyses in DPC-NE-1005-PA, "Nuclear Design Methodology Using CASM0-4/SIMULATE-3 MOX," Revision 1 (Reference 11 ). Current nuclear design for Catawba and McGuire is performed using CASM0-4 and SIMULATE-3, the steady-state nodal core simulator on which SIMULATE-3K is based. When SIMULATE-3K is executed in the steady mode, it performs the same calculations as SIMULATE-3, and in the revised method SIMULATE-3K transient calculations will read in a CASM0-4 cross-section library and be based on a CASM0-4/SIMULATE-3 restart file. The NRC staff determined that the update from CASM0-3 to CASM0-4.will improve the consistency between the steady and transient calculations and it is thus acceptable.

3.1.2 Thermal-Hydraulic Analysis Methodology Changes A number of minor changes were made to the thermal-hydraulic analysis method provided by Duke for the CRE. A new VIPRE model was added to perform certain calculations. In response to an NRC staff RAI, Duke described the VIPRE model and its purpose. The NRC staff reviewed the licensee's RAI response, which defined the limited scope of the new model to compute the response of the peak-power fuel pin to the CRE. The model proposed by the licensee is appropriate to this application and will maintain the conser\tatism that exists in the previous VIPRE model. The NRC staff therefore determined that the new model is acceptable.

Additionally, the fuel thermal conduction model for the departure from nucleate boiling ratio

.(DNBR) evaluation has been modified to change the gas gap closure time. The proposed gap closure time is more physically realistic than the one assumed previously, while maintaining conservatism. As well, a new method was proposed to determine the initial gas gap conductance, while retaining the original one. This new option will provide a conservative means of determining the gap conductance by matching fuel temperatures, similar to the original method. The NRC staff has reviewed both of these methodology changes and determined that they are acceptable because they maintain the conservatism of the original methods while providing some additional realism.

3.1.3 SIMULATE-3K Analysis Changes Several changes to the SIMULATE-3K initial and boundary conditions were. proposed. The original methodology included a modeling approach that was intended to introduce conservatism by adjusting certain parameters. The original idea bel)ind the approach was to ensure that the CRE analysis would be bounding for future core designs. Duke proposes to remove this adjustment in Revision 1 to DPC-NE-3001-P, because additional experience has shown that it does not always produce conservative results and frequently has negligible impact.

Conservatism in the methodology will be maintained through selection of certain other conservatively bounding key parameters. The NRG staff has determined that removal of the adjustment originally included in the methodology is reasonable provided that the CRE analysis of record continues to be demonstrated each cycle to be bounding with respect to post ejected possible power shapes and distributions.

An alternative approach is added for modeling the negative reactivity insertion associated with the reactor trip. This method explicitly models the reactor trip, with both top and bottom peaked xenon distributions, rather than using a more conservative trip reactivity curve. Conservatism in the new method is maintained by assuming that the ejected rod and an adjacent rod do not fall into the core, and that the negative reactivity added by the rods that do fall into the core is reduced by adjusting cross sections. The NRC staff has determined that this is reasonable, provided the assumed xenon distributions and negative reactivity insertion rates from the control rods are confirmed to bound expected operation each cycle.

Additionally, an option is added for adjusting the moderator temperature coefficient (MTC) to a conservative value at the beginning of the cycle (BOC) by changing the boron concentration.

Decre(lsing the boron concentration would make the BOC MTC more conservative, and would also be conservative from a neutronics standpoint. The HZP *rod ejection, from fully inserted to fully withdrawn, takes place over the course of 0.1 seconds; the HFP method is updated such that the control rod ejects in a time proportional to that 0.1 seconds based on the rod insertion as a fraction of the total insertion length. This gives the licensee flexibility in changing rod insertion limits relative to the top of active fuel. A rod ejection from a different location, (i.e., H-08, the center control rod) is added as a case to be analyzed without removing the 0.:.12 rod ejection analysis. The HFP initial power uncertainty is modified to accommodate measurement uncertainty recapture (MUR) uprates. As well, the high flux positive rate trip is now credited in the analysis. The NRC staff reviewed these changes and determined that they maintain conservatism in the method and are, therefore, acceptable. ,

3.1.4 RETRAN-02 Analysis Changes Two minor changes are proposed for modeling the pressurizer safety valves in the RETRAN-02 analysis methodology for the CRE. First, the valve blowdown value - the pressure at which the*

valve re-closes - is decreased from 5% below the valve lift setpoint to 1%. This adds conservatism to the analysis by limiting the possible depressurization allowed by the valve.

Second, an option is added to allow the valves to pop open rather than open as a linear ramp.

This approach was previously reviewed and approved by the NRC for use at Catawba and McGuire in DPC-NE-3002-PA, "FSAR Chapter 15 System Transient Analysis Methodology,"

Revision 2 (Reference 12), and its use in DPC-NE-3001-P will add both realism and consistency with other analyses, without negatively impacting the results. The NRC staff therefore concludes that these changes are acceptable.

3.1.5 Other Changes The limit to the number of fuel rods allowed to experience DNB .is revised. It is now no longer included as a fixed limit in DPC-NE-3001-P; rather, replaced indirectly by the offsite dose consequences. The methods used to analyze the offsite dose consequences were not part of the current review effort. However, a limit on the number of fuel rods allowed to experience DNB that "maintains offsite dose consequences within applicable regulatory limits" would, by definition, maintain the dose-related acceptance criterion for CRE analyses. This change is therefore determined to be acceptable by the NRC staff.

Finally, the key parameters to be used in the cycle-specific evaluation of the CRE analysis were modified. These key parameters are analyzed each cycle to ensure that the cycle-specific core design will remain bounded by the analysis of record. Rather than providing both the ejected' rod worth in percent change millirho (pcm) and the effective delayed neutron fraction(~), Duke

.proposes that these parameters be consolidated into a single parameter, ejected rod worth in dollars, which Duke argues would better define the transient response than either of its constituent pieces. The NRC staff finds that consolidation of multiple parameters into a single parameter will still account for the appropriate variables and is, therefore, acceptable.

3.2 Steam Line Break Analysis Methodology The majority of the changes proposed to DPC-NE-3001-P Chapter 5, "Steam Line Break Analysis," are made to add methods to analyze the HFP SLB: However, some changes are made to the HZP SLB methodology and a number of changes are made to clarify whether a method is used for the HZP or HFP analysis. Additionally, the centerline fuel melt (CFM) limit is now evaluated as part of the SLB analyses. Due to the number of changes made in Chapter 5, the entire chapter was submitted for review with the changes incorporated.

  • The SLB accident is classified a postulated accident or an ANS Condition IV event (i.e., an event for which fuel failure' is allowed to occur, subject to certain offsite dose requirements): According to SRP 15.1.5, "Steam System Piping Failures Inside and Outside of Containment," the SLB analysis must demonstrate that fuel failure will be sufficiently limited to ensure the core remains in place with no loss of cooling capability. It must be assumed, absent a specific justification using an acceptable fuel damage model, that fuel will fail if the calculated DNBR falls below the DNBR limit. The radiological consequences of the accident are then evaluated based on a number of failed fuel pins that bounds the SLB analysis results, and the acceptance criteria for the radiological analysis are based on meeting offsite dose limits. Duke stated that the SLB analysis meets the core damage and coolability criteria by maintaining the DNBR above its approved DNBR limit and maintaining the fuel centerline temperature below the approved CFM limit, subject to the acceptance criteria provided in SRP 4.4 and 4.2, respectively. Radiological analyses of the consequences of the SLB are not included in DPC-NE-3001-P, Revision 1, and are therefore not a part of the NRC staff's current review effort. The. NRC staff considers Duke's overall approach to the SLB analysis to be acceptable.
  • 3.2.1 Syst.em Thermal-Hydraulic Analysis Changes The HFP SLB contains two additional modifications to the base RETRAN-02 model for each plant.

The first modification was the result of operating experience in preliminary applications of the HFP SLB model, and is necessary to maintain conservatism in the model forthe HFP SLB analysis. In

response to an NRC staff RAI regarding this change (Reference 9), Duke stated that the change would be conservative with respect to modeling the long-term cooling capabilities of the steam generator. Additionally, the Duke RAI response discussed how the modification would be applied differently to the different plants due to physical differences between the plants. The NRC staff reviewed the information provided by the licensee and determined.that the proposed modification to the base RETRAN-02 model is appropriate and conservative for the case when the model is appliedjo the HFP SLB. The change is therefore acceptable.

The second modification was also developed based on operating experience in preliminary applications of the RETRAN-02 model to the HFP SLB. This modification corrects an error that would occur when the model was used to evaluate the HFP case. The licensee clarified that this modification, which changes how and where the RETRAN-02 enthalpy transport model is applied, is necessary to correctly model reversed flow without failing to calculate a solution. The licensee also explained how it will have a negligible impact on the solution accuracy because it results in saturated steam conditions on the secondary side of the steam generator. Because the modification is necessary to compute the plant response to the HFP SLB but will have a negligible effect on the primary-to-secondary heat transfer, the NRC staff was able to determine that the change is acceptable.

3.2.2

  • Nuclear Analysis Changes The methodology had previously relied on specification of both the Doppler temperature coefficient (DTC) and a k-effective versus moderator temperature curve. This curve implicitly includes the reactivity effects of both DTC and MTC, effectively combining them (along with other effects) into a single curve. It is computed at a condition corresponding to the end of the cycle with the technical specification minimum shutdown margin and, as discussed in a Duke response to RAls for the original version of the topical report (ADAMS Accession No. rvll010080284),

includes the effects of pressure, temperature, and the highest worth stuck rod in its fully withdrawn position. However, in Revision 1, the HZP core physics parameters section is now modified such that the DTC is no longer specified directly in the report. Instead, the reactivity insertion provided by following the k-effective versus temperature curve through the cooldown is verified to be conservative on a cycle-specific basis by comparing the reactivity insertion calculated by the cycle-specific SIMULATE-3 analysis to that of the RETRAN-02 analysis of record at each RETRAN-02 statepoint. Because the reactivity curve is calculated using conservative assumptions for the HZP SLB transient; and because the change retains the cycle-specific verification of the analysis of record, the NRC staff determined that it is acceptable.

'The core physics parameters for HFP SLB are chosen in a different manner, because the values of the limiting parameters for the HFP SLB change over the spectrum of possible SLB sizes and times in core life. Duke discussed this effect in th.eir response to RAI 6 (Reference 2). Duke stated that break sizes are iterated on from smallest to largest, and MTC is iterated on from a least negative BOC value to a most negative EOC value. The value of MTC is correlated to the time in cycle, and a conservative DTC is selected for the time in cycle corresponding to the chosen MTC.

These parameters are varied in order to determine the limiting break sizes with respect to centerline fuel melt limits or DNBR limits. Because this method will determine the most limiting combination of MTC, DTC, and break size throughout the c.ore life, in conformance with the SRP Acceptance Criteria in SRP 15.1.5, the NRC staff concludes that this addition to the method is acceptable.

The RETRAN-02 limiting statepoint is input to SIMU.LATE-3 to calculate the power distributions for both the HZP and HFP SLB analyses. In both cases, the SIMULATE-3-calculated power distributions are used as inputs to the CFM evaluation. Duke provided additional detail on the interface between the power distributions and the CFM evaluation in their response to RAI 7 (Reference 2). The response explained how SIMUL.,ATE-3 is used to calculate the axial peaking factors (Fa), which are compared against CFM linear heat generation rate limits generated by an appropriate fuel performance code. As discussed in a subsequent Duke RAI response (Reference 3), the core power level uncertainty is applied to the predicted power distribution before comparison to CFM linear heat generation rate limits. Because both the SIMULATE-3 power distributions and the CFM linear heat generation limits they will be compared to have been determined by the NRC staff to be conservative, and because the core power level uncertainty will be appropriately applied to the predicted power distribution prior to this comparison, the NRC staff determined that this change is acceptable.

  • 3.2.3 Core Thermal-Hydraulic Analysis Changes For the HFP SLB analysis, the thermal hydraulic analysis is performed with a VIPRE-01 model as 1 described in DPC-NE-3000-PA, 'Thermal-Hydraulic Transient Analysis Methodology," Revision 3 (Reference 13). The model is then used along with Duke's statistical core design (SCD)

, methodology, described in DPC-NE-2005-PA, 'Thermal-Hydraulic Statistical Core Design Methodology," Revision 4a (Reference 14) to statistically convohJte many of the plant operating condition uncertainties into the DNBR limit. The critical heat flux correlations used are the WRB-2M correlation, with an SCD DNBR limit of 1.30, and the BWU-N correlation, which is used below the first mixing vane. The way the SCD methodology is applied and the DNBR limit to be used were clarified in the Duke response to RAI 8 (Reference 2). The approved topical report DPC-NE-2009-A, "Duke Power Westinghouse Fuel Transition Report," Revision 2, allows the use of WRB-2M and BWU-N correlations for RFA fuel with VIPRE-01 at Catawba and McGuire.

Because of the previous approval of the codes, methods, models, and limits, the NRC staff concludes that this approach is acceptable.

For the HZP SLB analysis, use of the Wl,-OP correlation is proposed. WLOP will be used as the primary correlation. Duke clarified that W-3S is left in DPC-NE-3001-P as a backup correlation to be used only if operating outside of the WLOP range of applicability. The WLOP correlation will be used with a DNBR limit of 1.18, and the range of applicability to be used is consistent with that of the NRC Safety Evaluation (SE) approving the WLOP topical. report. Based on use consistent with previous NRC approval, the NRC staff finds this change to the HZP SLB analysis to be acceptable.

3.2.4 Transient Analysis The HFP transient analysis is initialized from nominal operating conditions. For minimum DNBR analyses, this is consistent with the SCD methodology as discussed in Section 3.2.3 of this SE, and is therefore considered by the NRC staff to be acceptable. The HZP initial conditions remain unchanged.

The method proposed in DPC-NE-3001-P, Revision 1, for determining the excore neutron flux detector error due to overcooling is employed for the HFP SLB analysis. This method uses the SAS2H/ORIGEN-S modules of the SCALE code, as well as MCNP, to characterize the relative impact of Reactor Coolant System temperature on excore detector response. This method for

calculating excore flux detector error is identical to the one approved by the NRG in Revision 2 of DPC-NE-3005, "UFSAR Chapter 15 Transient Analysis Methodology," for use at Oconee. Overall, it is written in such a manner that it is applicable to Oconee as well as Catawba and McGuire. The NRG staff has therefore determined that it is acceptable.

The reactor trips credited in the HFP SLB analysis are the high flux, high positive flux rate, and overpower!::.T trips, which are appropriate for the analysis. Other changes to the boundary conditions for the transient analysis include the assumptions that rod control is in manual, turbine control is in automatic, that a loss of offsite power occurs coincident with the turbine trip, and that the main feedwater pump speed and control valves are in manual to maximize the feedwater flow rate. These modifications, all made for the HFP SLB analysis, are considered to be conservative and, therefore, acceptable.

The HZP SLB analysis continues to use the RETRAN-02 point kinetics model, which provides a conservative value for the positive reactivity insertion induced by the overcooling event. However, Duke proposed a modification to the method that allows the analysis to account for spatial reactivity effects, reducing some of the conservatism of the point-kinetics model. In response to an NRG staff RAI (Reference 3), Duke clarified the manner in which the temperature feedback will be adjusted and how it will be confirmed to be conservative. The RAI response also clarified how SIMULATE-3 will be used to perform reload checks to ensure that the analysis of record will remain bounding. The NRG staff reviewed the proposed method and determined, based on the supplemental information provided in the RAI response, that it will remain suitably conservative and that this conservatism will be demonstrated on a cycle-specific basis as part of the reload design process. The NRG staff therefore concludes that this change is acceptable.

The HFP SLB analysis also uses the RETRAN-02 point kinetics model with conservative inputs such as minimum OTC. Because the model itself is conservative and is used with conservative inputs, the NRG staff determined that this approach to modeling the HFP SLB core kinetics is acceptable.

3.2.5 Analysis Results and Cycle-Specific Evaluations The NRG staff reviewed the HFP SLB analysis results presented in DPC-NE-3001-P, Revision 1 Section 5.5 and determined that they were reasonable and consistent with the assumptions made in the analysis methodology. Changes to the HZP SLB results section only included updates to the DNBR limits and a reference to the CFM limit check now performed in the analysis. The NRG staff verified that both the typical HFP SLB results and the typical HZP SLB results meet the acceptance criteria discussed earlier in the section. The NRG staff therefore concludes that these typical results are acceptable.

The cycle-specific evaluation methodology for the HZP SLB case is modified to reflect the changes to the reactivity feedback model discussed in Section 3.2.4 of this SE. A check against CFM limits is also added as part of the cycle-specific evaluation. The HFP SLB cycle-specific reload check method compares cycle-specific values of MTG and OTC against those used in the analysis of record. Core power distributions are compared against maximum allowable radial peaking factor (MARP) limits determined in the VIPRE-01 thermal hydraulic analysis. Because the key physics parameters and power distributions are evaluated and determined to be bounded by the analysis of record on a cycle-specific basis, the NRG staff determined that the proposed methods for cycle-specific evaluation of the HZP and HFP SLB cases are acceptable.

3.3 Dropped Rod Analysis Methodology The main proposed changes to the dropped rod analysis methodology include clarifications that the dropped rod could potentially be in Control Bank D, reducing the positive reactivity insertion when Bank D is withdrawn on the resulting power mismatch. Other changes are proposed based on increased experience with the existing method, as well as changes to reflect physical plant modifications.

  • In the dropped rod event, negative reactivity is inserted into the core by a dropped rod, immediately decreasing the excore nuclear detector response. If the rod control system is in automatic, the power mismatch between the excore nuclear detedor and the turbine power causes the rod control system to begin withdrawing the other control rods. The power mismatch between the turbine and the core may result in a reduced moderator temperature, because more power is being drawn out of the secondary system than is being produced by the core. This would result in a further increase in nuclear power in the parts of the core not directly affected by the droppe.d rod. Analysis of this event must include both the thermal~hydraulic effects and the fuel thermal-mechanical effects of the transient According to SRP 15.4.3, the thermal-hydraulic analysis performed for the dropped rod event is subject to the DNBR limit discussed in SRP 4.4, while fuel therm.al-mechanical analysis is subject to the fuel centerline melt and uniform cladding strain limits discussed in SRP 4.2.

3.3.1 Initial and Boundary Condition Changes The main change to the initial conditions is to use the minimum average fuel temperature rather than the maximum. The NRC staff asked Duke to justify use of the minimum rather than the maximum. In response, Duke explained that, based on operating experience, that the competing effects of fuel temperature have made it more conservative to use minimum temperatures for

  • DNBR evaluations. Considering the operating experience from Duke, the NRC staff finds the change to be acceptable.

Duke proposed clarifications to the boundary conditions to account for the consideration of dropped rods in Control Bank D. If the dropped rod(s) were to be in Bank D, which is the first bank withdrawn from the core, the bank worth available for withdrawal in the event would be lower. The report was, modified to account for this fact. As an additional clarification, the effect of the dropped rod on Fe.His noted to be a function of the dropped rod location. The NRC staff considers this approach to be realistic while maintaining appropriate conservatism. Both of these changes are minor clarifications and are therefore acceptable.

Additionally, an option is also added to use post-drop axial and radial power shapes for each dropped rod combination as input to the DNBR analysis. This approach represents a move away from the use of enveloping Fe.H curves and a bounding axial power shape, and has become more practical over time with increased computational power and efficiency. All post-drop power distributions, for all possible combinations of dropped rod locations, will now be calculated on a cycle-specific basis. Because of this, the cycle-specific evaluation for the rod drop event was also updated to use post-drop power distributions rather than the initial Fe.H and axial flux shape.

Post-drop excore tilt is also added to the cycle-specific check. Because the main consideration in the dropped rod transient is the withdrawal of control rods and eventual return to power, post-drop values for power distributions and excore tilt are reasonable parameters to evaluate. Thus,

Post-drop power distributions are also appropriate to check against the power distribution limits.

The NRC staff has therefore determined that this set of changes is acceptable.

Duke proposed that excore detectors can now be characterized as a function of the dropped rod worth and number/location of the dropped rods. Weighting factors will be assigned to fuel assemblies to account for their relative importance in generating an excore detector response. As discussed in Duke's response to RAI 13 (Reference 2), the method for calculating nodal weighting factors for excore detector response is part of the SIMULATE-3K methodology. The RAI response also included a benchmarking of the SIMULATE-3K excore detector response, which showed very good agreement with the reference values. Therefore, the NRC staff finds that the new model will provide sufficient fidelity in the method and is acceptable.

Because the downcomer temperature decreases due to the negative reactivity insertion of the dropped rod, excore detector response is also modified by a conservative attenuation factor. This attenuation factor is calculated using the same method as is proposed for the HFP SLB. As

.discussed above, this method for calculating excore flux detector error is identical to the one approved by the NRC in Revision 2 of DPC-NE-3005, "UFSAR Chapter 15 Transient Analysis Methodology," for use at Oconee. Because this methodology is written in such a way that it is applicable to both events, was already determined to be acceptable by the NRC, and is applicable to Oconee, Catawba, and McGuire, the NRC staff has determined that its use for the dropped rod event is acceptable.

Additional trip functions, including the high flux, over-temperature .6.T, and over-power .6.T trips, are now credited for the dropped rod event. In an RAI response, the licensee stated that the limiting DNB cases do not result in a reactor trip and that the trips added to the analysis may become limiting in the future as core designs or trip setpoints change. Because the additional trip functions are valid for the dropped rod event and because their addition does not affect the conservatism of the methodology, the NRC staff concludes that they are acceptable.

Subsequent to the NRC review and approval of DPC-NE-3001-PA, Revision 0, a new control system, the Distributed Control System (DCS), was installed at Catawba and McGuire. Because of the changes to the physical plant, several changes have been made to the control system in the model. These changes result in a modification to the single failure assumed in the dropped rod analysis. This change is based on the conversion of the turbine power signal from impulse pressure to inlet pressure and the conversion of the auctioneered high nuclear instrumentation (NI) flux and primary loop Tave signals from auctioneered high signals to median select signals.

The NRC staff considers the new single worst control system failure to be reasonable in light of the changes to the control system described by Duke. The licensee's method for implementing this failure assumes a conservative nuclear instrument signal to the Rod Control System that bounds possible signals generated in response to dropped rods. This method is also acceptable and will be verified to be conservative on a cycle-specific basis, because the worth of dropped rod combinations will vary from cycle to cycle. Therefore, the NRC staff considers the proposed rod control system model changes to be acceptable.

Other changes to the boundary conditions are made based on additional operating experience in applying the methodology. The licensee stated that allowing the operation of the pressurizer heaters, which were previously assumed not to function, was slightly more conservative for some

.dropped rod cases. The wording was changed to state that pressurizer heaters are not credited as a benefit - so their operation will be dependent on what is more conservative for the dropp.ed

rod case in question. Additionally, the licensee stated that operating experience with the method*

  • has shown that either automatic or manual turbine control can be conservative depending on the dropped rod case. The method is therefore changed to allow operation in either manual or automatic, depending on which would be more conservative for the case being analyzed. The NRC staff determined that both of these changes are acceptable, as they will enhance the realism of the methodology while-maintaining conservatism.
  • 3.4 Key Safety Analysis Physics Parameters and Core Reload Design Process DPC-NE-3001-P, Chapter 2.0, identifies all of the FSAR Chapter 15 transients and the reactor physics parameters that are important in their analyses, providing a discussion section and a table at the end of the chapter. This information is not limited to the transients discussed later in the report, where detailed neutronics calculations are necessary for the analysis, but instead cover all FSAR Chapter 15 transients.

The main changes to this section are to add the HFP SLB description and key physics parameters to the discussion section and table. These changes are aligned with the changes that were reviewed and found previously acceptable in Section 3.2 of this SE and are therefore determined by the NRC staff to be acceptable. Another change, in the moderator boron dilution accident description, allows conservatism to be maintained in the conversion of shutdown margin to a value in parts per million of boron (ppmb) using best estimate boron worth rather than a conservatively large boron worth. Conservatism in the method is maintained by applying a conservative allowance to the initial boron concentration. Because this method will maintain conservatism in the calculation of the Initial boron concentration, the NRC staff has determined it is acceptable.

3.5 Thermal Conductivity Degradation A new Appendix B was added to DPC-NE-3001-P as part of Revision 1 to discuss the impact of TCD on Duke's non-loss-of-coolant accident (LOCA) safety analyses. TCD is the phenomenon by which the thermal conductivity of the nuclear fuel - which governs the ability of the fuel to transfer heat through the fuel and gas gap, and eventually in to the coolant - degrades as the fuel is irradiated in the core. The appendix breaks down the impacts of TCD into sections on core design and neutronics analyses, fuel design and performance, and FSAR Chapter 15 transient analyses.

Duke states in their evaluation that core design and neutronics analyses will not be impacted by TCD, although the thermal-hydraulic and systems codes used in DPC-NE-3001-P, Revision 1, do not explicitly account for TCD. The inputs are verified to be conservative through comparison to codes that do account for TCD, therefore demonstrating that operation with TCD is bounded.

Therefore, the NRC staff finds this to be acceptable.

Fuel. mechanical analysis is performed by Duke using methodologies. licensed from the fuel vendor .. Duke's evaluation indicates that the effects of TCD on the fuel mechanical analyses

. using these licensed codes may be offset with existing margin to the fuel rod design acceptance criteria. As part of the transient analyses in the proposed methodology, the fuel mechanical analysis methodology is adjusted to account for impact of TCD through the use of a burnup.,.dependent linear heat rate limit. The NRC staff finds use of a burnup-dependent linear heat rate limit to account for the impact of TCD to be appropriate.

Duke also addressed the impact of TCD on their FSAR Chapter 15 transient analyses and stated that they were unaffected. Inclusion of TCD will produce results that will be bounded by equivalent analyses without TCD, because TCD degrades the heat transfer out of the fuel, increasing the Doppler feedback and margin to the critical heat flux, respectively, for the two events. The one case where it would be potentially more conservative to include TCD impacts is for fuel enthalpy calculations during a rod ejection. However, as described in Section B2.3.2 for the control rod ejection analysis, Duke has determined that the maximum fuel enthalpy calculation analyses including TCD will remain bounded by the BOC case, which by definition does not include TCD.

Though the effect of TCD on enthalpy deposition in fuel during the rod ejection is non-negligible, it can be accommodated in the existing margin to the limits. Because DPC-NE-3001-P, Revision 1, does not include a LOCA analysis methodology, the NRC staff did not assess the potential impacts of TCD on core stored energy and LOCA analysis initialization.

The NRC staff concludes based on the analysis above, that Duke's approach to accounting for TCD in fuel performance analyses at Catawba and McGuire is acceptable.

3.6 Common Changes Throughout DPC-3001-P, Revision 1, Duke introduced changes that are intended to allow the methodologies to be used without directly specifying a particular fuel performance code. Instead,.

the document has been revised such that it specifies the use of an NRG-approved fuel performance code appropriate to the fuel in use at the given plant, rather than specifying the actual code to be used in the topical report. These changes are acceptable because they do not change the intent of the methodology. NRC oversight will remain because changes in the fuel design and/or vendor will prompt further review.in other areas, allowing the agency to make the determination as to which fuel performance code(s) will be applicable to the fuel design(s) in use at the plant.

3. 7 References
1. Henderson, K., Duke, to NRC Document Control Desk (DCD), "License Amendment Request for Methodology Report DPC-NE-3001-P, Revision 1, Multidimensional Reactor Transients and Safety'Analysis Physics Parameters Methodology (Proprietary)," with attachments, November 14, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML13325B142 and ML13325B143).
2. Henderson, K., Duke, to NRC DCD, "License Amendment Request for Methodology Report DPC-NE~3001-P, Revision 1, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology (Proprietary) Response to NRC Request for Additional Information (RAI) (TAC Nos. MF3119, MF3120, MF3121, and MF3122)," June 27, 2014 (ADAMS Accession No. ML14183B259).
3. Henderson, K., Duke, to NRC DCD, "License Amendment Request for Methodology Report DPC-NE-3001-P, Revision 1, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology (Proprietary) Response to NRC Request for Additional Information (RAI) (TAC Nos. MF3119, MF3120, MF3121, and MF3122),'j November 10, 2014 (ADAMS Accession No. ML14321A531 ).

. 4. DPC-NE-3001-A, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology," Duke Power Company, November 1991 (republished December 2000) (ADAMS Accession No. ML010080284).

5. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (ADAMS Accession Nos. ML070820123, ML070740003, ML070550014, ML070550006, ML063600415, and ML070740002).
6. SSP-98113, "SIMULATE-3K Models and Methodology," Revision 6, Studsvik Scandpower,
7. SSP-04/443, "LWR Core Reactivity Transients. SIMULATE-3K Models and Assessment,"

Revision 2, Studsvik Scandpower, May 2006 (ADAMS Accession No. ML14321A532).

8. Jones, RA., Duke, to NRG DCD, "Proposed License Amendment Request to Revise the Technical Specifications Pursuant to the Use of Gadolinia Integral Burnable Absorber,"

October 19, 2009 (ADAMS Accession No. ML092960626).

9. Stang, J., NRG, to P. Gillespie, Duke, "Oconee Nuclear Station, Units 1, 2, and 3 -

Issuance of Amendments Regarding Approval for the use of Gadolinia as an Integral Burnable Absorber (TAC Nos. ME2504, 2505, and 2506)," July 21, 2011 (ADAMS Accession No. ML11137A150) ..

10. Miller, G.E., NRG, to K. Henderson and S.D. Capps, Duke, "Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station, Units 1 and 2 - Request for Additional Information Regarding License Amendment Request Re: Methodology Report DPC-NE-3001-P, Revision 1, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology (TAC Nos. MF3119, MF3120, MF3121, and MF3122),"

May 22, 2014 (ADAMS Accession No. ML14128A352). -

11. DPC-NE-1005-PA, "Nuclear Design Methodology Using CASM0-4/SIMULATE-3 MOX,"

Revision 1, Duke Power Company, May 2007 (ADAMS Accession No. ML071370091 ).

12. DPC-NE-3002-PA, "UFSAR Chapter 15 System Transient Analysis Methodology,"

Revision 2, Duke Power Company, May 2005 (ADAMS Accession No. ML053540177).

13. DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," Revision 3, Duke Power Company, September 2004 (ADAMS Accession No. ML050680319).
14. DPC-NE-2005-PA, "Thermal-Hydraulic Statistical Core Design Methodology," Revision 3, Duke Power Company, September 2002 (ADAMS Accession Nos. ML023090183 and ML023090223).
15. Miller, G.E., NRG, to K. Henderson and S.D. Capps, Duke, "Catawba Nulcear Station, Units 1 aDd 2, anp McGuire Nuclear Station, Units 1 and 2 ...;. Requesffor Additional
  • Information Regarding License Amendment Request Re: Methodology Report DPC-NE-3001-P, Revision 1, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology (TAC Nos. MF3119, MF3120, MF3121, and MF3122),"

October 14, 2014 (ADAMS Accession No. ML14276A561).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina and South Carolina State officials were notified of the proposed issuance of the amendments. The States officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined iri 10 CFR Part 20. The NRG staff has

.determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on February 27, 2014 (79 FR 11147). Accordingly, the amendments meet 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: R. Anzalone Date: March 25, 2015

ML15027A366 OFFICE DORL/LPL2-1 /PM DORL/LPL2-1 /LA DSS/SNPB/BC DSS/SRXB/BC OGC DORL/LPL2-1 /BC NAME GEMiller SFigueroa JDean CJackson DRoth INLO RPascarelli DATE 03/11 /15 01 /30/15 02/11/15 03/11/15 03/13/15 03/25/15

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