RA-18-0100, Request for License Amendment Regarding Application of Advanced Framatome Methodologies
| ML18284A395 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/11/2018 |
| From: | William Gideon Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-18-0100 | |
| Download: ML18284A395 (1154) | |
Text
October 11, 2018
Serial: RA-18-0100 10 CFR 50.90
U.S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, DC 20555-0001
Subject: Brunswick Steam Electric Plant, Unit Nos. 1 and 2
Renewed Facility Operating License Nos. DPR-71 and DPR-62
Docket Nos. 50-325 and 50-324
Request for License Amendment Regarding Application of Advanced Framatome
Methodologies
Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), hereby requests a
revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP),
Unit Nos. 1 and 2. The proposed license amendments revise TS 5.6.5.b to allow application of
Advanced Framatome Methodologies for determining core operating limits in support of loading
Framatome fuel type ATRIUM 11. The Enclosure to this letter provides a description and
assessment of the proposed change. Enclosure Attachment 4 provides a list of regulatory
commitments associated with the proposed license amendments.
Duke Energy has evaluated the proposed change in accordance with 10 CFR 50.91(a)(1), using
the criteria in 10 CFR 50.92(c), and determined that this change involves no significant hazards
consideration.
Approval of the proposed amendments is requested by February 20, 2020 to support reactor
startup following the Unit 1 refueling outage. Once approved, the Unit 1 amendment shall be
implemented prior to start-up from the 2020 Unit 1 refueling outage, and the Unit 2 amendment
shall be implemented prior to startup from the 2021 Unit 2 refueling outage.
The Enclosure to this letter contains Attachments considered proprietary to Framatome (i.e.,
Enclosure Attachments 5a, 6a, 7a, 8a, 9a, 10a, 11a, 12a, 13a, 14a, 15a, 16a, and 17a). Within
these Attachments, proprietary information has been denoted by brackets. As owner of the
proprietary information, Framatome has executed affidavits for each proprietary document,
which identify the information as proprietary, is customarily held in confidence, and should be
withheld from public disclosure in accordance with 10 CFR 2.390. Enclosure Attachments 5b,
6b, 7b, 8b, 9b, 10b, 11b, 12b, 13b, 14b, 15b, 16b, and 17b provide non-proprietary versions of
each proprietary Framatome document. Corresponding affidavits are provided in Enclosure
Attachments 5c, 6c, 7c, 8c, 9c, 10c, 11c, 12c, 13c, 14c, 15c, 16c, and 17c.
In accordance with 10 CFR 50.91, Duke Energy is providing a copy of the proposed license
amendments to the designated representative for the State of North Carolina.
Letter Enclosure Attachments 5a, 6a, 7a, 8a, 9a, 10a, 11a, 12a, 13a, 14a, 15a, 16a, and 17a
Contain Proprietary Information Withhold in Accordance with 10 CFR 2.390
U.S. Nuclear Regulatory Commission
Page 2 of 4
Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory
Affairs, at (910) 832-2487.
I declare, under penalty of perjury, that the foregoing is true and correct. Executed on
October 11, 2018. Si;;w
William R. Gideon
SBY/sby
Enclosure:
Description and Assessment of the Proposed Change
Attachments:
1: Technical Specification Mark-Ups - Unit 1
2: Technical Specification Mark-Ups - Unit 2
3: Technical Specification Bases Mark-Ups - Unit 1 (For Information Only)
4: List of Regulatory Commitments
Sa: ANP-370SP, Applicability of Framatome BWR Methods to Brunswick with
ATRIUM 11 Fuel [Proprietary Information -Withhold from Public Disclosure
in Accordance with 10 CFR 2.390]
Sb: ANP-370SNP, Applicability of Framatome BWR Methods to Brunswick with
ATRIUM 11 Fuel
Sc: Affidavit for ANP-370SP, Applicability of Framatome BWR Methods to Brunswick
with ATRIUM 11 Fuel
6a: ANP-3686P, Mechanical Design Report for Brunswick ATRIUM 11 Fuel
Assemblies [Proprietary Information - Withhold from Public Disclosure in
Accordance with 10 CFR 2.390]
6b: ANP-3686NP, Mechanical Design Report for Brunswick ATRIUM 11 Fuel
Assemblies
6c: Affidavit for ANP-3686P, Mechanical Design Report for Brunswick ATRIUM 11
Fuel Assemblies
7a: ANP-3643P, Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM 11
Fuel Assemblies [Proprietary Information - Withhold from Public Disclosure
in Accordance with 1 O CFR 2.390]
7b: ANP-3643NP, Brunswick Unit 1 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
7c: Affidavit for ANP-3643P, Brunswick Unit 1 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
8a: ANP-3644P, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM 11
Fuel Assemblies [Proprietary Information - Withhold from Public Disclosure
in Accordance with 10 CFR 2.390]
8b: ANP-3644NP, Brunswick Unit 2 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
U.S. Nuclear Regulatory Commission
Page 3 of 4
8c: Affidavit for ANP-3644P, Brunswick Unit 2 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
9a: ANP-3668P, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Brunswick
LAR [Proprietary Information – Withhold from Public Disclosure in
Accordance with 10 CFR 2.390]
9b: ANP-3668NP, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for
Brunswick LAR
9c: Affidavit for ANP-3668P, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation
for Brunswick LAR
10a: ANP-3661P, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle Design
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
10b: ANP-3661NP, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle Design
10c: Affidavit for ANP-3661P, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle
Design
11a: ANP-3667P, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear Fuel
Design Report [Proprietary Information – Withhold from Public Disclosure in
Accordance with 10 CFR 2.390]
11b: ANP-3667NP, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear Fuel
Design Report
11c: Affidavit for ANP-3667P, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear
Fuel Design Report
12a: ANP-3702P, Brunswick ATRIUM 11 Transient Demonstration [Proprietary
Information – Withhold from Public Disclosure in Accordance with 10 CFR 2.390]
12b: ANP-3702NP, Brunswick ATRIUM 11 Transient Demonstration
12c: Affidavit for ANP-3702P, Brunswick ATRIUM 11 Transient Demonstration
13a: ANP-3674P, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
13b: ANP-3674NP, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel
13c: Affidavit for ANP-3674P, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11
Fuel
14a: ANP-3694P, ATWS-I Analysis Methodology for Brunswick Using RAMONA5-FA
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
14b: ANP-3694NP, ATWS-I Analysis Methodology for Brunswick Using RAMONA5-
14c: Affidavit for ANP-3694P, ATWS-I Analysis Methodology for Brunswick Using
RAMONA5-FA
15a: ANP-3703P, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
15b: ANP-3703NP, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA
15c: Affidavit for ANP-3703P, BEO-III Analysis Methodology for Brunswick Using
RAMONA5-FA
16a: DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate
Enhanced Option-III [Proprietary Information – Withhold from Public
Disclosure in Accordance with 10 CFR 2.390]
U.S. Nuclear Regulatory Commission
Page 4 of 4
16b: DPC-NE-1009, Brunswick Nuclear Plant Implementation of Best-estimate
Enhanced Option-III
16c: Affidavit for DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Bestestimate
Enhanced Option-III
17a: ANP-3714P, Brunswick ATRIUM 11 Control Rod Drop Accident Analyses with
the AURORA-B CRDA Methodology [Proprietary Information – Withhold from
Public Disclosure in Accordance with 10 CFR 2.390]
17b: ANP-3714NP, Brunswick ATRIUM 11 Control Rod Drop Accident Analyses with
the AURORA-B CRDA Methodology
17c: Affidavit for ANP-3714P, Brunswick ATRIUM 11 Control Rod Drop Accident
Analyses with the AURORA-B CRDA Methodology
cc (with Enclosure and all Enclosure Attachments):
U. S. Nuclear Regulatory Commission, Region II
ATTN: Ms. Catherine Haney, Regional Administrator
245 Peachtree Center Ave, NE, Suite 1200
Atlanta, GA 30303-1257
U. S. Nuclear Regulatory Commission
ATTN: Mr. Gale Smith, NRC Senior Resident Inspector
8470 River Road
Southport, NC 28461-8869
U. S. Nuclear Regulatory Commission
ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A)
11555 Rockville Pike
Rockville, MD 20852-2738
cc (with Enclosure and Non-Proprietary Enclosure Attachments):
Chair - North Carolina Utilities Commission (Electronic Copy Only)
4325 Mail Service Center
Raleigh, NC 27699-4300
swatson@ncuc.net
Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)
Radiation Protection Section
North Carolina Department of Health and Human Services
1645 Mail Service Center
Raleigh, NC 27699-1645
lee.cox@dhhs.nc.gov
Enclosure
Page 1 of 16
Description and Assessment of the Proposed Change
Subject: Request for License Amendment Regarding Application of Advanced Framatome
Methodologies
1. SUMMARY DESCRIPTION
2. DETAILED DESCRIPTION
2.1 System Design and Operation
2.2 Current Technical Specification Requirements
2.3 Reason for the Proposed Change
2.4 Description of Proposed Change
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria
4.2 No Significant Hazards Consideration Determination Analysis
4.3 Conclusions
5. ENVIRONMENTAL CONSIDERATION
6. REFERENCES
ATTACHMENTS:
1: Technical Specification Mark-Ups – Unit 1
2: Technical Specification Mark-Ups – Unit 2
3: Technical Specification Bases Mark-Ups – Unit 1 (For Information Only)
4: List of Regulatory Commitments
5a: ANP-3705P, Applicability of Framatome BWR Methods to Brunswick with
ATRIUM 11 Fuel [Proprietary Information – Withhold from Public Disclosure
in Accordance with 10 CFR 2.390]
5b: ANP-3705NP, Applicability of Framatome BWR Methods to Brunswick with
ATRIUM 11 Fuel
5c: Affidavit for ANP-3705P, Applicability of Framatome BWR Methods to Brunswick
with ATRIUM 11 Fuel
6a: ANP-3686P, Mechanical Design Report for Brunswick ATRIUM 11 Fuel
Assemblies [Proprietary Information – Withhold from Public Disclosure in
Accordance with 10 CFR 2.390]
6b: ANP-3686NP, Mechanical Design Report for Brunswick ATRIUM 11 Fuel
Assemblies
6c: Affidavit for ANP-3686P, Mechanical Design Report for Brunswick ATRIUM 11
Fuel Assemblies
7a: ANP-3643P, Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM 11
Fuel Assemblies [Proprietary Information – Withhold from Public Disclosure
in Accordance with 10 CFR 2.390]
7b: ANP-3643NP, Brunswick Unit 1 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
Enclosure
Page 2 of 16
7c: Affidavit for ANP-3643P, Brunswick Unit 1 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
8a: ANP-3644P, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM 11
Fuel Assemblies [Proprietary Information – Withhold from Public Disclosure
in Accordance with 10 CFR 2.390]
8b: ANP-3644NP, Brunswick Unit 2 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
8c: Affidavit for ANP-3644P, Brunswick Unit 2 Thermal-Hydraulic Design Report for
ATRIUM 11 Fuel Assemblies
9a: ANP-3668P, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Brunswick
LAR [Proprietary Information – Withhold from Public Disclosure in
Accordance with 10 CFR 2.390]
9b: ANP-3668NP, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for
Brunswick LAR
9c: Affidavit for ANP-3668P, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation
for Brunswick LAR
10a: ANP-3661P, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle Design
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
10b: ANP-3661NP, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle Design
10c: Affidavit for ANP-3661P, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle
Design
11a: ANP-3667P, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear Fuel
Design Report [Proprietary Information – Withhold from Public Disclosure in
Accordance with 10 CFR 2.390]
11b: ANP-3667NP, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear Fuel
Design Report
11c: Affidavit for ANP-3667P, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear
Fuel Design Report
12a: ANP-3702P, Brunswick ATRIUM 11 Transient Demonstration [Proprietary
Information – Withhold from Public Disclosure in Accordance with 10 CFR 2.390]
12b: ANP-3702NP, Brunswick ATRIUM 11 Transient Demonstration
12c: Affidavit for ANP-3702P, Brunswick ATRIUM 11 Transient Demonstration
13a: ANP-3674P, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
13b: ANP-3674NP, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel
13c: Affidavit for ANP-3674P, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11
Fuel
14a: ANP-3694P, ATWS-I Analysis Methodology for Brunswick Using RAMONA5-FA
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
14b: ANP-3694NP, ATWS-I Analysis Methodology for Brunswick Using RAMONA5-
14c: Affidavit for ANP-3694P, ATWS-I Analysis Methodology for Brunswick Using
RAMONA5-FA
Enclosure
Page 3 of 16
15a: ANP-3703P, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA
[Proprietary Information – Withhold from Public Disclosure in Accordance
with 10 CFR 2.390]
15b: ANP-3703NP, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA
15c: Affidavit for ANP-3703P, BEO-III Analysis Methodology for Brunswick Using
RAMONA5-FA
16a: DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate
Enhanced Option-III [Proprietary Information – Withhold from Public
Disclosure in Accordance with 10 CFR 2.390]
16b: DPC-NE-1009, Brunswick Nuclear Plant Implementation of Best-estimate
Enhanced Option-III
16c: Affidavit for DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Bestestimate
Enhanced Option-III
17a: ANP-3714P, Brunswick ATRIUM 11 Control Rod Drop Accident Analyses with
the AURORA-B CRDA Methodology [Proprietary Information – Withhold from
Public Disclosure in Accordance with 10 CFR 2.390]
17b: ANP-3714NP, Brunswick ATRIUM 11 Control Rod Drop Accident Analyses with
the AURORA-B CRDA Methodology
17c: Affidavit for ANP-3714P, Brunswick ATRIUM 11 Control Rod Drop Accident
Analyses with the AURORA-B CRDA Methodology
Enclosure
Page 4 of 16
1. SUMMARY DESCRIPTION
Duke Energy Progress, LLC (Duke Energy), is requesting a revision to the Technical
Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The
proposed license amendments revise TS 5.6.5.b to allow application of Advanced Framatome
Methodologies for determining core operating limits in support of loading Framatome fuel type
ATRIUM 11.
2. DETAILED DESCRIPTION
2.1 System Design and Operation
Core operating limits are established each operating cycle. These operating limits ensure that
the fuel design limits are not exceeded during any conditions of normal operation and in the
event of any Anticipated Operational Occurrence (AOO).
2.2 Current Technical Specification Requirements
The Core Operating Limits Report (COLR) is the unit specific document that provides cycle
specific parameter limits for the current reload cycle. These cycle specific limits are determined
for each reload cycle in accordance with TS 5.6.5.
TS 5.6.5.a lists the core operating limits required to be established for each cycle. The methods
used to determine the operating limits are those previously found acceptable by the U.S.
Nuclear Regulatory Commission (NRC) and listed in TS 5.6.5.b.
2.3 Reason for the Proposed Change
BSEP plans to transition to the Framatome fuel type ATRIUM 11. These proposed license
amendments to allow application of Advanced Framatome Methodologies are necessary for this
fuel transition. Duke Energy is pursuing the ATRIUM 11 fuel type due to the improved fuel cycle
economics and safety margins.
The ATRIUM 11 fuel type consists of an array of 11 by 11 fuel rods; whereas the current fuel
design (i.e., ATRIUM 10XM) consists of an array of 10 by 10 fuel rods. This increase in the
number of fuel rods significantly reduces Linear Heat Generation Rate (LHGR) and fuel duty,
thereby improving safety margin.
The ATRIUM 11 fuel type incorporates enhanced debris protection features which make the fuel
design less susceptible to debris related fuel failures. In addition, the channel design changes
incorporated with ATRIUM 11 make the fuel design less susceptible to channel bow and bulge.
2.4 Description of the Proposed Change
The following Methodologies will be removed from TS 5.6.5.b:
• XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient
Thermal-Hydraulic Core Analysis (i.e., Reference 6.24)
• ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water
Reactor Transient Analyses (i.e., Reference 6.10)
• NEDC-33075P-A, GE Hitachi Boiling Water Reactor, Detect and Suppress Solution -
Confirmation Density, Revision 8, November 2013 (i.e., Reference 6.11)
Enclosure
Page 5 of 16
The above methodologies are no longer applicable with addition of the Advanced
Methodologies described below.
The Advanced Methodologies that will be added to TS 5.6.5.b are listed below:
• ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors;
Application to Control Rod Drop Accident (CRDA), Revision 0, March 2018 (i.e.,
Reference 6.7)
• ANP-10300P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors;
Application to Transient and Accident Scenarios, Revision 1, January 2018 (i.e.,
Reference 6.1)
• ANP-3703P, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA,
Revision 0, August 2018 (i.e., Reference 6.2)
• DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate Enhanced
Option-III, Revision 0, September 2018 (i.e., Reference 6.3)
• BAW-10247P-A, Supplement 2P-A, Realistic Thermal-Mechanical Fuel Rod
Methodology for Boiling Water Reactors Supplement 2: Mechanical Methods,
Revision 0, August 2018 (i.e., Reference 6.4)
• ANP-10340P-A, Incorporation of Chromia-Doped Fuel Properties in AREVA Approved
Methods, Revision 0, May 2018 (i.e., Reference 6.5)
• ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, Revision 0, May 2018 (i.e.,
Reference 6.6)
• ANP-10332P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors;
Application to Loss of Coolant Accident Scenarios, Revision 0, [DATE]
The Final Safety Evaluation (SE) for Topical Report (TR) ANP-10332P, AURORA-B LOCA
(i.e., Reference 6.8) has not been issued. As a result, the Draft SE (i.e., Reference 6.9) along
with ANP-10332P are being referenced for this LAR submittal and the date for ANP-10332P-A
is being left blank above and in the TS markups (i.e., Attachments 1 and 2). All the limitations
and conditions presented in Reference 6.9 are addressed in this LAR. Following issuance of the
final SE for ANP-10332P, this LAR will be supplemented with reference to the approved
AURORA-B LOCA TR (i.e., ANP-10332P-A) and any changes in limitations and conditions will
be addressed. Reference 6.14 documents NRC concurrence with this approach and
Attachment 4 of this Enclosure documents the commitment to supplement this LAR.
With the addition of the methodologies listed above, BSEP is transitioning from the Detect and
Suppress Solution – Confirmation Density (DSS-CD) stability methodology to the Best Estimate
Enhanced Option-III with Confirmation Density Algorithm (BEO-III w/CDA) stability methodology
(i.e., Reference 6.2 and 6.3). As with DSS-CD, the CDA will remain the licensing basis trip, and
identical Oscillation Power Range Monitor (OPRM) setpoints will be used for BEO-III w/CDA
thereby minimizing the impact to BSEP. As a result of the transition from DSS-CD to BEO-III
w/CDA, TS Table 3.3.1.1-1 Note f will be removed from TS. This is a conditional note regarding
arming DSS-CD that was only applicable during the first reactor startup and shutdown following
DSS-CD Implementation; therefore, this note is no longer applicable and is being removed.
In addition to the methodology changes outlined above, with the transition from ATRIUM 10XM
fuel to ATRIUM 11 fuel, BSEP will transition to RAMONA5-FA for the licensing basis Anticipated
Transient Without Scram with Instability (ATWS-I) analysis. The current ATWS-I licensing basis
analysis is the TRACG ATWS-I evaluation performed with ATRIUM 10XM fuel for Maximum
Enclosure
Page 6 of 16
Extended Load Line Limit Analysis Plus (MELLLA+) which was approved for BSEP
September 18, 2018 (i.e., Reference 6.13).
Unit 1 and Unit 2 TS Mark-Ups are provided in Attachments 1 and 2 respectively. The Mark-Ups
demonstrate how BSEP plans to incorporate the proposed change into TSs. In addition, Unit 1
TS Bases Mark-Ups directly related to the proposed change are provided in Attachment 3 for
information.
3. TECHNICAL EVALUATION
Attachments 5a through 17a provide the detailed Technical Evaluation for the proposed change
outlined in Section 2.4. The information presented in these Attachments demonstrates
acceptable safety margin for the proposed change supporting operation of the new ATRIUM 11
fuel type in the currently approved operating domain. The currently approved operating domain
includes Extended Power Uprate (EPU) conditions, approved for BSEP in 2002 (i.e., Reference
6.12), as well as MELLLA+, approved for BSEP in 2018 (i.e., Reference 6.13). It should be
noted that within this LAR, MELLLA+ and Extended Power/Flow Operating Domain (EPFOD)
represent the identical operating domain for BSEP, and therefore may be used interchangeably.
The sections below provide a brief summary of what is included in the Attachments. The table
below is provided to correlate the Advanced Methodologies that will be added to TS 5.6.5.b with
the Attachments in which the methodology is applied. Note that the Attachments with the ‘a’
designation provide the full report, while the Attachments with the ‘b’ designation provide the
non-proprietary version of the full report (i.e., proprietary information is redacted). For ease of
reference throughout this LAR, only the Attachments with the ‘a’ designation are referenced in
the discussions.
Methodology Application
ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water
Reactors; Application to Control Rod Drop Accident (CRDA), Revision 0,
March 2018 (i.e., Reference 6.7)
Attachment 17a
ANP-10300P-A, AURORA-B: An Evaluation Model for Boiling Water
Reactors; Application to Transient and Accident Scenarios, Revision 1,
January 2018 (i.e., Reference 6.1)
Attachment 12a
ANP-3703P, BEO-III Analysis Methodology for Brunswick Using
RAMONA5-FA, Revision 0, August 2018 (i.e., Reference 6.2)
Attachment 15a
DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Bestestimate
Enhanced Option-III, Revision 0, September 2018
(i.e., Reference 6.3)
Attachment 16a
BAW-10247P-A, Supplement 2P-A, Realistic Thermal-Mechanical Fuel
Rod Methodology for Boiling Water Reactors Supplement 2: Mechanical
Methods, Revision 0, August 2018 (i.e., Reference 6.4)
Attachment 6a
ANP-10340P-A, Incorporation of Chromia-Doped Fuel Properties in
AREVA Approved Methods, Revision 0, May 2018 (i.e., Reference 6.5)
Attachment 9a
ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, Revision 0,
May 2018 (i.e., Reference 6.6)
Attachment 7a,
8a, 10a, and 12a
ANP-10332P-A, AURORA-B: An Evaluation Model for Boiling Water
Reactors; Application to Loss of Coolant Accident Scenarios, Revision 0,
[DATE]
Attachment 13a
Enclosure
Page 7 of 16
Attachment 5a: ANP-3705P, Applicability of Framatome BWR Methods to Brunswick with
ATRIUM 11 Fuel
ANP-2637P, “Boiling Water Reactor Licensing Methodology Compendium” is a compendium of
Framatome methodologies and design criteria, which are described in TRs that the NRC has
found acceptable for referencing in Boiling Water Reactor (BWR) licensing applications.
Framatome provided this document to the NRC for information by letter dated September 18,
2018 (i.e., Reference 6.27). This compendium provides a concise, organized source for BWR
TRs. It presents information about the application of each TR, the associated Safety Evaluation
Report (SER) and its conclusions and restrictions/limitations for each TR, the relationships
among the TRs, and, for certain methodologies, descriptions of their unique characteristics or
applications. Compliance with the SER restrictions/limitations is assured by implementing them
within the engineering guidelines or by incorporating them into the computer codes.
ANP-3705P demonstrates that the Framatome licensing methodologies presented in
ANP-2637P are applicable to the ATRIUM 11 fuel type and operation of BSEP in the currently
approved operating domain.
Attachment 6a: ANP-3686P, Mechanical Design Report for Brunswick ATRIUM 11 Fuel
Assemblies
ANP-3686P documents the successful completion of all licensing analyses and related testing
necessary to verify that the mechanical design criteria are met for the ATRIUM 11 Fuel
Assemblies supplied by Framatome for insertion into BSEP reactors. This report also provides a
description of the mechanical design and licensing methods for ATRIUM 11. The scope of this
report is limited to an evaluation of the mechanical design of the fuel assembly and fuel
channel. The fuel assembly design was evaluated according to the Framatome BWR generic
mechanical design criteria (i.e., Reference 6.16). The fuel channel design was evaluated to the
criteria given in the fuel channel TRs (i.e., References 6.17 and 6.18). The generic design
criteria have been approved by the NRC and the criteria are applicable to the subject fuel
assembly and channel design. Mechanical analyses for ATRIUM 11 have been performed using
NRC-approved design analysis methodology (i.e., References 6.4, 6.16, 6.17, and 6.18).
Attachment 7a and 8a: ANP-3643P/ANP-3644P, Brunswick Unit 1/2 Thermal-Hydraulic Design
Report for ATRIUM 11 Fuel Assemblies
ANP-3643P and ANP-3644P present the results of BSEP Unit 1 and Unit 2 thermal-hydraulic
analyses which demonstrate that Framatome ATRIUM 11 fuel is hydraulically compatible with
the previously loaded ATRIUM 10XM fuel design. These reports also provide the hydraulic
characterization of the ATRIUM 11 and the coresident ATRIUM 10XM design for both Units.
The generic thermal-hydraulic design criteria applicable to the design have been reviewed and
approved by the NRC in Reference 6.16. In addition, thermal-hydraulic criteria applicable to the
design have also been reviewed and approved by the NRC in Reference 6.19.
Attachment 9a: ANP-3668P, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for
Brunswick LAR
ANP-3668P reports the results of thermal-mechanical analyses for the performance of
ATRIUM 11 fuel assemblies inserted into to an equilibrium cycle for the BSEP units and
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demonstrates that the design criteria relevant to the Thermal-Mechanical limits are satisfied.
These analyses assume the use of chromia additive in the fuel and assume operation in the
currently approved operating domain. Both the design criteria and the analysis methodology
used in this report have been approved by the NRC. The analysis results are evaluated
according to the generic fuel rod thermal and mechanical design criteria contained
Reference 6.16 along with design criteria provided in Reference 6.28. In addition, the approved
methodology for the inclusion of chromia additive in the fuel pellets (i.e., Reference 6.5) is also
used.
Attachment 10a: ANP-3661P, Brunswick ATRIUM 11 Equilibrium Cycle Fuel Cycle Design
In ANP-3661P, Framatome has performed an equilibrium fuel cycle design for BSEP Unit 1.
This design uses the ATRIUM 11 fuel assembly and the currently approved operating domain.
This analysis has been performed with the approved Framatome neutronic modeling
methodology (i.e., Reference 6.20). This analysis has also used the Reference 6.6 critical power
methodology. The CASMO-4 lattice depletion code was used to generate nuclear data including
cross sections and local power peaking factors. The MICROBURN-B2 three dimensional core
simulator code, combined with the ACE critical power correlation, was used to model the core.
The MICROBURN-B2 pin power reconstruction (PPR) model was used to determine the thermal
margins presented in the report. Design results including projected control rod patterns and
evaluations of thermal and reactivity margins are presented.
Attachment 11a: ANP-3667P, Brunswick Unit 1 ATRIUM 11 Equilibrium Cycle Nuclear Fuel
Design Report
ANP-3667P provides results of the neutronic design analyses performed by Framatome for
BSEP Unit 1 ATRIUM 11 equilibrium cycle fuel assemblies (i.e., used in Attachment 10a).
NRC-approved neutronic design criteria are provided in Reference 6.16, and the NRC-approved
neutronic design analysis methodology (i.e., Reference 6.20) was used to determine
conformance to design criteria. Pertinent fuel design information is given in Section 2.0 and in
Appendices A through D of Attachment 11a.
Attachment 12a: ANP-3702P, Brunswick ATRIUM 11 Transient Demonstration
ANP-3702P summarizes the results of a subset of limiting transient analyses performed to show
example BSEP results utilizing the Reference 6.1 and Reference 6.6 methodologies based
upon an equilibrium cycle of ATRIUM 11 fuel. The AURORA-B AOO methodology (i.e.,
Reference 6.1) is used to calculate the change in the minimum critical power ratio (ΔMCPR)
during the AOO. The ΔMCPR is combined with the safety limit MCPR (SLMCPR) to establish or
confirm the plant operating limits for MCPR. The AURORA-B AOO methodology is also used to
calculate the maximum reactor vessel pressure and the maximum dome pressure during the
ASME and ATWS events. The ACE/ATRIUM 11 critical power correlation (i.e., Reference 6.6) is
used to evaluate the thermal margin of the ATRIUM 11 fuel.
Attachment 13a: ANP-3674P, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel
ANP-3674P presents the results of a loss-of-coolant accident (LOCA) break spectrum and
emergency core cooling system (ECCS) analyses for BSEP Units 1 and 2. The analyses
documented in this report are performed with Framatome LOCA Evaluation Models for reactor
licensing analyses pending approval by the NRC (i.e., Reference 6.14). The models and
computer codes used by Framatome for LOCA analyses are collectively referred to as the
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AURORA-B LOCA Evaluation Model (i.e., References 6.8, 6.9, 6.21, and 6.28). The purpose of
the break spectrum analysis is to identify the break characteristics that result in the highest
calculated peak cladding temperature (PCT) during a postulated LOCA. The results provide the
maximum average planar linear heat generation rate (MAPLHGR) limit for ATRIUM 11 fuel as a
function of exposure. The calculations described in this report are performed in conformance
with 10 CFR 50 Appendix K requirements and satisfy the event acceptance criteria identified in
Attachment 14a: ANP-3694P, ATWS-I Analysis Methodology for Brunswick Using
RAMONA5-FA
ANP-3694P documents Anticipated Transient Without Scram with Instability (ATWS-I) analyses
for Brunswick with ATRIUM 11 fuel and the currently approved operating domain.
Attachment 14a presents a description of BWR instability transients that are not terminated by
scram, and thus power and flow oscillations grow to large amplitudes. This class of transients is
referred to as ATWS-I. This report documents a plant-specific, RAMONA5-FA based method for
evaluating the fuel-related aspects of these events. The method is applicable to the limiting
BSEP ATWS-I events covering the event from the initiation through inception of oscillations and
final suppression through operator action.
Attachment 15a: ANP-3703P, BEO-III Analysis Methodology for Brunswick Using
RAMONA5-FA
ANP-3703P presents a plant-specific, best-estimate methodology for the evaluation of the
Long-Term Stability Solution Enhanced Option-III (i.e., BEO-III). This report documents a
RAMONA5-FA based method for determining the core operating limit MCPR based on the
Option-III Period-Based Detection Algorithm (PBDA). The methodology defines how the two
manual Backup Stability Protection (BSP) Regions are generated. The methodology shows
compliance with 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 10 and 12.
Included in the report is the application of this BEO-III methodology to the BSEP Unit 1 reactor.
The primary difference between Unit 1 and 2 that directly affects stability is the inlet pressure
drop. Since BSEP Unit 1 has a much lower inlet pressure drop due to larger inlet orifices, the
fuel assemblies will be hydraulically less stable and therefore bound the results for Unit 2.
Attachment 16a: DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate
Enhanced Option-III
DPC-NE-1009-P provides the Duke Energy methodology that will be used to implement the
CDA stability algorithm within the plant specific BEO-III analysis framework (i.e., BEO-III
w/CDA). BSEP is not seeking generic approval of the BEO-III w/CDA stability methodology;
application of the BEO-III w/CDA methodology will be specific to BSEP. This report, in
conjunction with Attachment 15a, replaces Reference 6.11 as the Long-Term Stability Solution
for BSEP Units 1 and 2.
The CDA was installed as the licensing basis algorithm with the implementation of MELLLA+ at
BSEP (i.e., Reference 6.13), and was incorporated in TS, TS Bases, plant procedures, and
training programs. The CDA is contained in the OPRM function of the Nuclear Measurement
Analysis and Control, Average Power Range Monitor hardware which is part of the Power
Range Neutron Monitoring System. The CDA is able to detect oscillations at the onset of
instability before there is a significant reduction in margin to the SLMCPR. For these reasons, it
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was decided to maintain the CDA trip function as the primary stability protection mechanism for
BSEP instead of the Period Based Detection Algorithm that serves as the licensing basis for
BEO-III as described in Attachment 15a. The PBDA, along with the other Option III algorithms
(i.e., Growth Rate and Amplitude Based) will be maintained for defense-in-depth functions and
will not be credited in the SLMCPR margin determination.
DPC-NE-1009-P shows the implementation of BEO-III with CDA with sample Brunswick
ATRIUM 11 results from the currently approved operating domain. In addition, DPC-NE-1009-P
describes the methodology for the Automated BSP and BSP Boundary which are defined in the
COLR and are utilized when the OPRM Upscale trip function is inoperable.
Attachment 17a: ANP-3714P, Brunswick ATRIUM 11 Control Rod Drop Accident Analyses with
the AURORA-B CRDA Methodology
ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to
Control Rod Drop Accident Scenarios (i.e., Reference 6.7) is the Framatome methodology to
analyze the BWR control rod drop accident (CRDA). The methodology includes the use of a
nodal three-dimensional kinetics solution with both thermal-hydraulic and fuel temperature
feedback. These models provide more precise localized neutronic and thermal conditions than
previous methods.
The Framatome methodology for the CRDA evaluation includes both generic evaluations and
cycle-specific analysis. Generic studies are used to address at power conditions and system
pressurization. The cycle specific analysis includes the determination of candidate control rods
that could challenge fuel failure criteria and the subsequent evaluation of these candidate rods
with a three-dimensional neutron kinetics and thermal-hydraulics code system.
This methodology has been developed to support recent changes in the CRDA acceptance
criteria and evaluation process as reflected in the Interim Acceptance Criteria and Guidance of
Appendix B of NUREG-0800 Section 4.2 (i.e., Reference 6.15).
ANP-3714P provides the initial application demonstration of the new CRDA methodology (i.e.,
Reference 6.7). This CRDA analysis is performed using the ATRIUM-11 Equilibrium Cycle
Design (i.e., Attachment 10a). Though not part of the BSEP licensing basis, the criteria used for
the Brunswick initial application demonstration are based upon Draft Regulatory Guide DG-1327
(i.e., Reference 6.22) which was also used in the Reference 6.23 responses to NRC request for
additional information.
ATRIUM 11 Fuel Design and Cycle Specific Reports
The NRC approved the use of Framatome fuel and core design methodologies to determine
BSEP core operating limits with the issuance of License Amendments 246 and 274 for BSEP
Units 1 and 2, respectively. Framatome licensing TR ANF-89-98(P)(A) Revision 1 and
Supplement 1 is one of these NRC-approved methodologies. ANF-89-98(P)(A) Revision 1 and
Supplement 1, as clarified by a Siemens Power Corporation letter dated October 12, 1999,
(i.e., Reference 6.25) and an NRC letter dated May 31, 2000, (i.e., Reference 6.26) requires
that a summary of the evaluation of the ATRIUM 11 design against the NRC-approved generic
design criteria be provided to the NRC for information. Framatome provided this evaluation to
the NRC for information by letter dated September 18, 2018, which transmitted Framatome
document ANP-3653P Revision 0, “Fuel Design Evaluation for ATRIUM 11 BWR Reload Fuel”
(i.e., Reference 6.27). In accordance with the process described in ANF-89-98(P)(A) Revision 1
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and Supplement 1, new fuel designs or fuel design changes satisfying the ANF-89-98(P)(A)
design criteria do not require explicit NRC review and approval (i.e., satisfaction of the design
criteria is sufficient for approval by reference to the criteria).
ANP-3653P identifies fuel design criteria, specified in ANF-89-98(P)(A) Revision 1 and
Supplement 1, which are evaluated on a cycle specific basis. Reports summarizing the results
of analyses performed to demonstrate BSEP compliance with the cycle specific criteria are
provided by Framatome to Duke Energy as part of the normal reload licensing document
package. This type of information is not available until later in the reload licensing process.
Consistent with the process described in ANF-89-98(P)(A) Revision 1 and Supplement 1 (i.e.,
as clarified by References 6.25 and 6.26), Duke Energy will provide the BSEP Unit 1 Cycle 23
reload reports outlined in the table below to the NRC for information. The reports will be
provided in supplemental letters as documented in Attachment 4. The anticipated schedule is
presented in the table below.
Report Estimated Transmittal Date
Fuel Cycle Design Report March 2019
Nuclear Fuel Bundle Design Report March 2019
Safety Limit MCPR Report May 2019
Fuel Rod Design Report October 2019
Reload Safety Analysis Report October 2019
ANP-3653P also identifies fuel design criteria, specified in ANF-89-98(P)(A) Revision 1 and
Supplement 1, that are evaluated on a plant-specific basis. The key differences in system
configuration between BSEP Unit 1 and Unit 2 are in the core inlet region and the Turbine
Bypass System. The central orifice diameter in Unit 2 is smaller than Unit 1; 2.09 inches
compared to 2.43 inches, and the Turbine Bypass System for Unit 2 has 10 valves whereas
Unit 1 has 4 valves. Differences in neutronic design and operation are minimal since both units
operate on 24 month fuel cycles.
Based on the minimal differences between Units 1 and 2, the information that is included in this
LAR, and the information in the table above which will be provided for Unit 1 Cycle 23, limited
information needs to be provided for Unit 2. Therefore, Duke Energy will include, for information,
the Unit 2 Cycle 25 Reload Safety Analysis Report with transmittal of the COLR prior to startup
from the Unit 2 Cycle 25 refueling outage (i.e., March 2020 timeframe) which will load the first
reload batch of ATRIUM 11 fuel into the Unit 2 reactor core. Attachment 4 of this Enclosure
documents this commitment to provide this report.
4. REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria
10 CFR 50.36(c)(5) states that the TSs will include administrative controls that address the
provisions relating to organization and management, procedures, record keeping, review and
audit, and reporting necessary to assure operation of the facility in a safe manner. The COLR is
required as a part of the reporting requirements specified in the BSEP TSs Administrative
Controls section. The TSs require the core operating limits to be established prior to each reload
cycle, or prior to any remaining portion of a reload cycle, and to be documented in the COLR. In
addition, it requires the analytical methods used to determine the core operating limits to be
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those that have been previously reviewed and approved by the NRC, and specifically to be
those described in TS 5.6.5.b. The proposed amendments ensure that these requirements are
met.
10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water power
reactors," establishes the acceptance criteria for the design basis LOCA. Paragraph (b)(1)
requires the calculated maximum fuel element cladding temperature (i.e., PCT) to not exceed
2200°F. 10 CFR Part 50, Appendix K, establishes required and acceptable features of
evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA. The use
of the proposed analytical methods to determine core operating limits will continue to ensure
that fuel performance during normal, transient, and accident conditions complies with these
requirements. Specific Average Planar Linear Heat Generation Rate (APLHGR) limits will be
determined in conformance with 10 CFR 50 Appendix K requirements and documented in the
COLR to ensure compliance with 10 CFR 50.46(b)(1).
The BSEP design was reviewed for construction under the General Design Criteria for Nuclear
Power Plant Construction, issued for comment by the AEC in July 1967, and is committed to
meet the intent of the GDC, published in the Federal Register on May 21, 1971, as Appendix A
to 10 CFR Part 50. GDC 10, 12, and 28 are applicable to this review.
GDC 10 requires that the reactor core and associated coolant, control, and protection systems
be designed with appropriate margin to assure that specified acceptable fuel design limits are
not exceeded during any condition of normal operation, including the effects of AOOs. Duke
Energy will use the proposed analytical methods to perform plant-specific analyses for
APLHGR, MCPR, and LHGR. The limits on the APLHGR are specified to ensure that the PCT
during the postulated design basis LOCA does not exceed the limits specified in 10 CFR 50.46.
The SLMCPR ensures that sufficient conservatism exists in the operating limit MCPR such that,
in the event of an AOO, there is a reasonable expectation that at least 99.9 percent of the fuel
rods in the core will avoid boiling transition for the power distribution within the core including all
uncertainties. Limits on the LHGR are specified to ensure that fuel thermal-mechanical design
limits are not exceeded anywhere in the core during normal operation, including AOOs.
Therefore, compliance with GDC 10 of 10 CFR 50, Appendix A is maintained.
GDC 12 requires that the reactor core and associated coolant, control, and protection systems
be designed to assure that power oscillations which can result in conditions exceeding specified
acceptable fuel design limits are not possible or can be reliably and readily detected and
suppressed. As part of the methodology changes outlined above, Duke Energy will transition
from DSS-CD to BEO-III w/CDA stability methodology. With BEO-III w/CDA the CDA will
continue to provide the licensing basis trip function, and the Automated and Manual BSP
functions will be maintained. The CDA provides reliable, automatic detection and suppression of
stability related power oscillations. The Automated and Manual BSP functions assure that power
oscillations which can result in conditions exceeding specified acceptable fuel design limits are
not possible. Therefore, compliance with GDC 12 of 10 CFR 50, Appendix A is maintained.
GDC 28 requires the reactivity control systems be designed with appropriate limits on the
potential amount and rate of reactivity increase to assure that the effects of postulated reactivity
accidents can neither (1) result in damage to the reactor coolant pressure boundary greater
than limited local yielding nor (2) sufficiently disturb the core, its support structures or other
reactor pressure vessel internals to impair significantly the capability to cool the core. These
postulated reactivity accidents include consideration of rod ejection (i.e., unless prevented by
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Page 13 of 16
positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and
pressure, and cold water addition. The use of the proposed analytical methods for the CRDA
calculations will continue to demonstrate compliance with GDC 28.
10 CFR 50.62, “Requirements for reduction of risk from anticipated transients without scram
(ATWS) events for light-water-cooled nuclear power plants,” defines an ATWS as an AOO
followed by the failure of the reactor trip portion of the protection system specified in GDC 20.
During an ATWS the potential exists for thermal-hydraulic instability to develop. The analyses
presented in this LAR demonstrate ATWS regulatory criteria are satisfied for BSEP, including
those specifically applicable to ATWS-I (i.e., demonstrating core coolability is maintained).
Duke Energy has determined that the proposed change does not require any exemptions or
relief from regulatory requirements, other than the TSs, and does not affect conformance with
the intent of any GDC differently than described in the Updated Final Safety Analysis Report.
4.2 No Significant Hazards Consideration Determination Analysis
Duke Energy Progress, LLC (Duke Energy), is requesting a revision to the Technical
Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The
proposed license amendments revise TSs to allow application of Advanced Framatome
Methodologies for determining core operating limits in support of loading Framatome fuel type
ATRIUM 11.
Duke Energy has evaluated whether a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance
of amendment, as discussed below:
1. Does the proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The probability of an evaluated accident is derived from the probabilities of the individual
precursors to that accident. The proposed amendments revise the list of NRC-approved
analytical methods used to establish core operating limits. The change does not require
any physical plant modifications, physically affect any plant components, or entail changes
in plant operation. Since no individual precursors of an accident are affected, the
proposed amendments do not increase the probability of a previously analyzed event.
The consequences of an evaluated accident are determined by the operability of plant
systems designed to mitigate those consequences. The proposed amendments revise the
list of NRC-approved analytical methods used to establish core operating limits. The
changes in methodology do not alter the assumptions of accident analyses. Based on the
above, the proposed amendments do not increase the consequences of a previously
analyzed accident.
Therefore, the proposed amendments do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Enclosure
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2. Does the proposed change create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No
Creation of the possibility of a new or different kind of accident requires creating one or
more new accident precursors. New accident precursors may be created by modifications
of plant configuration, including changes in allowable modes of operation. The proposed
amendments revise the list of NRC-approved analytical methods used to establish core
operating limits. The proposed amendments do not involve any plant configuration
modifications or changes to allowable modes of operation thereby ensuring no new
accident precursors are created.
Therefore, the proposed amendments do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No
The proposed amendments revise the list of NRC-approved analytical methods used to
establish core operating limits. The proposed change will ensure that the current level of
fuel protection is maintained by continuing to ensure that the fuel design safety criteria are
met.
Therefore, the proposed amendments do not result in a significant reduction in the margin
of safety.
Based on the above, Duke Energy concludes that the proposed amendments present no
significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of "no significant hazards consideration" is justified.
4.3 Conclusion
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance
that the health and safety of the public will not be endangered by operation in the proposed
manner, (2) 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.
5. ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with
respect to installation or use of a facility component located within the restricted area, as defined
in 10 CFR 20, or would change an inspection or surveillance requirement. However, the
proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant
change in the types or a significant increase in the amounts of any effluents that may be
released offsite, or (iii) a significant increase in individual or cumulative occupational radiation
exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical
exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no
environmental impact statement or environmental assessment need be prepared in connection
with the proposed amendment.
Enclosure
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6. REFERENCES
1. ANP-10300P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors;
Application to Transient and Accident Scenarios, Revision 1, January 2018.
2. ANP-3703P, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA,
Revision 0, August 2018.
3. DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate Enhanced
Option-III, Revision 0, September 2018.
4. BAW-10247P-A, Supplement 2P-A, Realistic Thermal-Mechanical Fuel Rod
Methodology for Boiling Water Reactors Supplement 2: Mechanical Methods,
Revision 0, August 2018.
5. ANP-10340P-A, Incorporation of Chromia-Doped Fuel Properties in AREVA Approved
Methods, Revision 0, May 2018.
6. ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, Revision 0, May 2018.
7. ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors;
Application to Control Rod Drop Accident (CRDA), Revision 0, March 2018.
8. ANP-10332P, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application
to Loss of Coolant Accident Scenarios, Revision 0, February 2014.
9. Project No. 728 / Docket No. 99902041, Draft Safety Evaluation for Framatome Inc.
Topical Report ANP-10332P, Revision 0, AURORA-B: An Evaluation Model for Boiling
Water Reactors; Application to Loss of Coolant Accident Scenarios, (CAC No. MF3829 /
EPID: L-2014-TOP-0004) US NRC, August 2018.
10. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water
Reactor Transient Analyses.
11. NEDC-33075P-A, GE Hitachi Boiling Water Reactor, Detect and Suppress Solution -
Confirmation Density, Revision 8, November 2013.
12. Letter from the U.S. Nuclear Regulatory Commission to Mr. John S. Keenan, Issuance of
Amendment Re: Extended Power Uprate, dated May 31, 2002, ADAMS Accession
Number ML021430551.
13. Letter from the U.S. Nuclear Regulatory Commission to Mr. William R. Gideon,
Brunswick Steam Electric Plant, Units 1 and 2 – Issuance of Amendment Regarding
Core Flow Operating Range Expansion (MELLLA+), dated September 18, 2018, ADAMS
Accession Number ML18172A258.
14. Letter from the U.S. Nuclear Regulatory Commission to Mr. Bryan B. Wooten, Brunswick
Steam Electric Plant, Units 1 and 2 – Request for One-Time Exception to Nuclear
Reactor Regulation Office Instructions LIC-109, LIC-101, and LIC-500 Acceptance
Review Criteria, dated September 21, 2018, ADAMS Accession Number ML18239A309.
15. NUREG-0800, Section 4.2, Revision 3, FUEL SYSTEM DESIGN, Standard Review
Plan: LWR Edition, US NRC: Washington, DC, March 2007.
16. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for
BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
17. EMF-93-177(P)(A) Revision 1, Mechanical Design for BWR Fuel Channels, Framatome
ANP Inc., August 2005.
Enclosure
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18. EMF-93-177P-A Revision 1, Supplement 1P-A, Revision 0, Mechanical Design for BWR
Fuel Channels Supplement 1: Advanced Methods for New Channel Designs, AREVA NP
Inc., September 2013.
19. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water
Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear
Company, June 1986.
20. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water
Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power
Corporation, October 1999.
21. XN-NF-82-07(P)(A) Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and
Rupture Model, Exxon Nuclear Company, November 1982.
22. US NRC Draft Regulatory Guide DG-1327, Pressurized Water Reactor Control Rod
Ejection and Boiling Water Reactor Control Rod Drop Accidents, November 2016,
ADAMS Accession Number ML16124A200.
23. ANP-10333Q1P Revision 0, AURORA-B: An Evaluation Model for Boiling Water
Reactors: Application to Control Rod Drop Accident (CRDA) – Responses to NRC
Request for Additional Information, AREVA Inc., April 2017.
24. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient
Thermal-Hydraulic Core Analysis.
25. Letter from Siemens Power Corporation to the NRC Document Control Desk, Revisions
to Attachment 1 of Letter NRC:99:030, Request for Concurrence on SER Clarifications,
dated October 12, 1999.
26. Letter from the U.S. Nuclear Regulatory Commission to Siemens Power Corporation,
Siemens Power Corporation Re: Request for Concurrence on Safety Evaluation Report
Clarifications (TAC No. MA6160), dated May 31, 2000, ADAMS Accession Number
27. Letter from Framatome Inc. to the NRC Document Control Desk, Informational
Transmittal of ANP-3653P Revision 0, “Fuel Design Evaluation for ATRIUM 11 BWR
Reload Fuel” and ANP-2637P Revision 7, “Boiling Water Reactor Licensing
Methodology Compendium”, dated September 18, 2018.
28. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for
Boiling Water Reactors, AREVA NP Inc., February 2008.