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Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies
	ML19183A108
Person / Time
Site:	Brunswick
Issue date:	07/02/2019
From:	William Gideon; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML19183A107	List: RA-19-0243, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies;
References
	RA-19-0243 ANP-3674NP, Rev. 2
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William R. Gideon Vice President Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 Enclosure 1 Contains Proprietary Information o: 910.832.3698 Withhold in Accordance with 10 CFR 2.390 July 2, 2019 Serial: RA-19-0243 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 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies

Reference:

1. Letter from William R. Gideon (Duke Energy) to the U.S. Nuclear Regulatory Commission Document Control Desk, Request for License Amendment Regarding Application of Advanced Framatome Methodologies, dated October 11, 2018, ADAMS Accession Number ML18284A395.

2. ANP-10332P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios, Revision 0, March 2019.

3. Letter from William R. Gideon (Duke Energy) to the U.S. Nuclear Regulatory Commission Document Control Desk, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, dated May 14, 2019, ADAMS Accession Number ML19135A029.

Ladies and Gentlemen:

By letter dated October 11, 2018 (i.e., Reference 1), Duke Energy Progress, LLC (Duke Energy), submitted a license amendment request (LAR) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendment revises Technical Specification (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.

As committed to in Reference 1, this submittal is being provided to update the LAR following approval of the AURORA-B Loss of Coolant Accident (LOCA) Topical Report (i.e., Reference 2). provides the updated LOCA analysis for ATRIUM 11 fuel (i.e., ANP-3674P). This revised analysis also incorporates corrections for the issues related to automated input generation discussed in Reference 3. In addition, ANP-3674P, Appendix A, addresses the limitations and conditions from the Reference 2 Safety Evaluation. contains information considered proprietary to Framatome. The proprietary information in this report has been denoted by brackets. As owner of the proprietary information,

U.S. Nuclear Regulatory Commission Page 3 of 3 cc (with all Enclosures):

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 11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Mr. Dennis J. Galvin 11555 Rockville Pike Rockville, MD 20852-2738 cc (without Enclosure 1):

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

RA-19-0243 Enclosure 2 ANP-3674NP, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel, Revision 2

Controlled Document Brunswick Units 1 and 2 LOCA ANP-3674NP Revision 2 Analysis for ATRIUM 11 Fuel May 2019

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page i Nature of Changes Section(s) or Item Page(s) Description and Justification

1. Page 1-1 and Updated the analysis to reference the approved topical report for Section 11.0 the AURORA-B LOCA Evaluation Model and resolve CRs described in Reference 2 that [ ]

2. Page 1-2 Removed or nuclear fuel design from the exposure discussion in the second to the last paragraph.

3. Page 2-1 PCT, oxidation, and total hydrogen values in the table and footnote are updated based on new analyses.

4. Page 3-1 Added the [ ]

to the last paragraph.

5. Page 4-2 Appended a sentence in Section 4.2 to clarify [

]

6. Page 5-1 Revised the first sentence in Section 5.1 to consider the availability of offsite power supplies.

7. Page 6-1 There are two instances where the rated core flow is updated based on new analyses.

8. Table 6.1 There are two instances where the rated core flow is updated based on new analyses.

9. Table 6.2 The PCT results are updated based on new analyses.

10. Table 6.3 Event times are updated based on new analyses.

11. Figures 6.1 The figures are updated based on new analyses.

through 6.15

12. Page 7-2 The limiting SLO axial power shape is updated based on new analyses.

13. Table 7.1 The limiting SLO axial power shape and PCT results are updated based on new analyses.

14. Page 9-1 Values in the last two paragraphs for PCT, maximum local MWR, hot rod MWR, CMWR, and time of PCT are updated based on new analyses and planes is changed to rods.

15. Table 9.1 The results in the table and footnote are updated based on new analyses.

16. Figure 9.1 The figure is updated based on new analyses.

17. Page 10-1 The rated core flow is updated based on new analyses in the first bullet and the reference to Appendix K is changed in the second bullet.

18. Appendix A Revised to discuss compliance with the Limitations and Conditions in the approved topical report.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0

SUMMARY

OF RESULTS ................................................................................. 2-1 3.0 LOCA DESCRIPTION ....................................................................................... 3-1 3.1 Accident Description .........................................................................................3-1 3.2 Acceptance Criteria...........................................................................................3-2 4.0 LOCA ANALYSIS DESCRIPTION ..................................................................... 4-1 4.1 Break Spectrum Analysis ..................................................................................4-1 4.2 Exposure Analysis ............................................................................................4-2 4.3 Plant Parameters ..............................................................................................4-2 4.4 ECCS Parameters ............................................................................................4-2 5.0 BREAK SPECTRUM ANALYSIS DESCRIPTION ............................................. 5-1 5.1 Limiting Single Failure .......................................................................................5-1 5.2 Recirculation Line Breaks .................................................................................5-2 5.3 Non-Recirculation Line Breaks ..........................................................................5-3 5.3.1 Main Steam Line Breaks .......................................................................5-4 5.3.2 Feedwater Line Breaks..........................................................................5-4 5.3.3 HPCI Line Breaks ..................................................................................5-5 5.3.4 LPCS Line Breaks .................................................................................5-5 5.3.5 LPCI Line Breaks ..................................................................................5-5 5.3.6 RWCU Line Breaks ...............................................................................5-5 5.3.7 Shutdown Cooling Line Breaks..............................................................5-6 5.3.8 Instrument Line Breaks..........................................................................5-6 6.0 TLO RECIRCULATION LINE BREAK SPECTRUM ANALYSES ...................... 6-1 6.1 Break Spectrum Analysis Results .....................................................................6-1 7.0 SINGLE-LOOP OPERATION LOCA ANALYSIS ............................................... 7-1 7.1 SLO Analysis Modeling Methodology ................................................................7-1 7.2 SLO Analysis Results .......................................................................................7-1 8.0 LONG-TERM COOLABILITY............................................................................. 8-1 9.0 EXPOSURE-DEPENDENT LOCA ANALYSIS DESCRIPTION AND RESULTS .......................................................................................................... 9-1

10.0 CONCLUSION

S .............................................................................................. 10-1

11.0 REFERENCES

................................................................................................ 11-1 Appendix A Limitations from the Safety Evaluation for LTR ANP-10332PA ........A-1

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page iii Tables Table 4.1 Initial Conditions ..................................................................................................4-4 Table 4.2 Reactor System Parameters ................................................................................4-5 Table 4.3 ATRIUM 11 Fuel Assembly Parameters...............................................................4-6 Table 4.4 High-Pressure Coolant Injection Parameters .......................................................4-7 Table 4.5 Low-Pressure Coolant Injection Parameters ........................................................4-8 Table 4.6 Low-Pressure Core Spray Parameters ................................................................4-9 Table 4.7 Automatic Depressurization System Parameters ...............................................4-10 Table 5.1 Available ECCS for Recirculation Line Break LOCAs...........................................5-7 Table 6.1 Break Spectrum Results for TLO Recirculation Line Breaks ................................6-2 Table 6.2 Summary of Break Spectrum [ ] for TLO Recirculation Line Breaks..........................................................................................................6-3 Table 6.3 Event Times for the [ ] from the TLO Recirculation Line Break Spectrum Analysis .....................................................................................6-4 Table 7.1 Single- and Two-Loop Operation PCT Summary .................................................7-3 Table 9.1 ATRIUM 11 Exposure-Dependent LOCA Analysis Results ..................................9-2

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page iv Figures Figure 1.1 Brunswick MELLLA+ Power / Flow Map............................................................1-3 Figure 2.1 MAPLHGR Limit for ATRIUM 11 Fuel ...............................................................2-3 Figure 2.2 [ ] for ATRIUM 11 Fuel..............................................2-4 Figure 4.1 S-RELAP5 Vessel Model ................................................................................4-11 Figure 4.2 S-RELAP5 Core Model ...................................................................................4-12 Figure 4.3 ECCS Schematic ............................................................................................4-13 Figure 4.4 Rod Average Power Distributions for 102%P and [ ] Mid- and Top-Peaked ....................................................................................................4-14 Figure 4.5 Rod Average Power Distributions for 102%P and [ ] Mid- and Top-Peaked ...........................................................................................................4-15 Figure 4.6 Rod Average Power Distributions for [ ] Mid- and Top-Peaked ...........................................................................................................4-16 Figure 4.7 Rod Average Power Distributions for [ ] Mid- and Top-Peaked ....................................................................................................4-17 Figure 6.1 [ ] from the TLO Recirculation Line Break Spectrum Analysis Upper Plenum Pressure......................................................................6-5 Figure 6.2 [ ] from the TLO Recirculation Line Break Spectrum Analysis Total Break Flow Rate ........................................................................6-5 Figure 6.3 [ ] from the TLO Recirculation Line Break Spectrum Analysis Core Inlet Flow Rate ...........................................................................6-6 Figure 6.4 [ ] from the TLO Recirculation Line Break Spectrum Analysis ADS Flow............................................................................................6-6 Figure 6.5 [ ] from the TLO Recirculation Line Break Spectrum Analysis HPCI Flow ..........................................................................................6-7 Figure 6.6 [ ] from the TLO Recirculation Line Break Spectrum Analysis LPCS Flow..........................................................................................6-7 Figure 6.7 [ ] from the TLO Recirculation Line Break Spectrum Analysis LPCI Flow ...........................................................................................6-8 Figure 6.8 [ ] from the TLO Recirculation Line Break Spectrum Analysis RDIV Flows .........................................................................................6-8 Figure 6.9 [ ] from the TLO Recirculation Line Break Spectrum Analysis Relief Valve Flow ................................................................................6-9 Figure 6.10 [ ] from the TLO Recirculation Line Break Spectrum Analysis Downcomer LOCA Water Level ..........................................................6-9 Figure 6.11 [ ] from the TLO Recirculation Line Break Spectrum Analysis Upper Plenum Liquid Level ...............................................................6-10

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page v Figure 6.12 [ ] from the TLO Recirculation Line Break Spectrum Analysis Hot Channel Liquid Level ..................................................................6-10 Figure 6.13 [ ] from the TLO Recirculation Line Break Spectrum Analysis Core Bypass Liquid Level .................................................................6-11 Figure 6.14 [ ] from the TLO Recirculation Line Break Spectrum Analysis Lower Plenum Liquid Level ...............................................................6-11 Figure 6.15 [ ] from the TLO Recirculation Line Break Spectrum Analysis Hot Channel Inlet Flow......................................................................6-12 Figure 9.1 Limiting [ ] PCT Exposure-Dependent LOCA Analysis ........................9-3

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page vi Nomenclature ADS automatic depressurization system ADSVOOS ADS valve out-of-service BOL beginning of life BWR boiling-water reactor CFR Code of Federal Regulations CMWR core average metal-water reaction DC direct current DEG double-ended guillotine DG diesel generator ECCS emergency core cooling system FHOOS feedwater heaters out-of-service HPCI high-pressure coolant injection ICF increased core flow ID inside diameter LHGR linear heat generation rate LOCA loss-of-coolant accident LPCI low-pressure coolant injection LPCS low-pressure core spray MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MELLLA+ maximum extended load line limit analysis plus MSIV main stream isolation valve MSIVOOS main steam isolation valve out-of-service MWR metal-water reaction NSSS nuclear steam supply system NRC Nuclear Regulatory Commission, U.S.

OD outside diameter PCT peak cladding temperature RDIV recirculation discharge isolation valve RHR residual heat removal RWCU reactor water cleanup

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page vii SF-BATT single failure of battery (DC) power SF-HPCI single failure of the HPCI system SF-LPCI single failure of an LPCI injection valve SLO single-loop operation TLO two-loop operation UFSAR updated final safety analysis report

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 1-1 1.0 Introduction The results of a loss-of-coolant accident (LOCA) break spectrum and emergency core cooling system (LOCA-ECCS) analyses for Brunswick Units 1 and 2 are documented in this report. 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 for normal (two-loop) operation.

Variation in the following LOCA parameters is examined:

Break location

Break type (double-ended guillotine (DEG) or split)

Break size

Limiting ECCS single failure

Axial power shape (top- or mid-peaked)

Initial statepoint

Fuel rod type The analyses documented in this report are performed with LOCA Evaluation Models developed by Framatome*, and approved for reactor licensing analyses by the U.S. Nuclear Regulatory Commission (NRC). The models and computer codes used by Framatome for LOCA analyses are collectively referred to as the AURORA-B LOCA Evaluation Model (References 1 - 4). 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 10 CFR 50.46.

Key model characteristics included in the report analyses are shown below. Other initial conditions used in the analyses are described in Section 4.0.

Operation in the MELLLA+ domain of Figure 1.1 is supported. [

]

Framatome Inc. formerly known as AREVA Inc.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 1-2

[

]

Analyses support operation with one automatic depressurization system (ADS) valve out-of-service (ADSVOOS) and the potential failure of one ADS valve during the LOCA.

[

]

The core is composed entirely of ATRIUM 11 fuel.

A 2.0% increase in initial core power to address the maximum uncertainty in monitoring reactor power, as per NRC requirements, is included.

[ ] were assumed to be at the MAPLHGR limit shown in Figure 2.1.

[

]

The limiting break characteristics from the break spectrum study are used in analyses to determine the MAPLHGR limit and [ ] versus exposure. Even though the limiting break will not change with exposure, the value of PCT calculated for any given set of break characteristics is dependent on exposure and the corresponding MAPLHGR and

[ ].

Single-loop operation (SLO) results are discussed in Section 7.0. Long term coolability is addressed in Section 8.0.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 1-3 Figure 1.1 Brunswick MELLLA+ Power

/ Flow Map

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 2-1 2.0 Summary of Results The LOCA break spectrum analysis results presented in this report are for Brunswick Unit 2, which are conservatively applicable to Unit 1. Analyses performed with AURORA-B LOCA confirm the PCT results of Unit 2 bound Unit 1. A more detailed discussion of results is provided in Sections 6.0 - 7.0.

The PCT and metal-water reaction (MWR) results, from the ATRIUM 11 fuel exposure-dependent analysis presented in Section 9.0, are presented below.

Parameter ATRIUM 11*

Peak cladding temperature (°F) 1957

[ ]

Local cladding oxidation (max %) 4.75

[ ]

Total hydrogen generated

(% of total hydrogen possible) 0.41 The MAPLHGR limit was determined by applying the AURORA-B LOCA Evaluation Model for the analysis of the limiting LOCA event. The exposure-dependent MAPLHGR limit for ATRIUM 11 fuel is shown in Figure 2.1. Exposure dependent results with the [

] are presented in Section 9.0. The results of these calculations confirm that the LOCA acceptance criteria in the Code of Federal Regulations (10 CFR 50.46) are met for operation at or below these limits.

The LOCA analysis results (i.e., the limiting break characteristics and exposure analysis) presented in this report are applicable for a full core of ATRIUM 11 fuel as well as transition cores containing ATRIUM 11 fuel. [

]

[

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 2-2

[

]

The SLO LOCA analyses support operation with an ATRIUM 11 multiplier of 0.85 applied to the normal two-loop operation MAPLHGR limit. [ ]

The long-term coolability evaluation confirms that the ECCS capacity is sufficient to maintain adequate cooling in an ATRIUM 11 core for an extended period after a LOCA.

All analyses support operation with one ADSVOOS. All analyses also support the [

] All analyses were performed assuming nominal feedwater temperature. [

] Therefore, this LOCA analysis supports FHOOS operation.

The analysis supports operation in the MELLLA+ domain of the Brunswick power/flow map shown in Figure 1.1.

At Brunswick, operation with 1 MSIVOOS is limited to two-loop operation and power levels less than 70% of rated. For a given power level, 1 MSIVOOS can result in a higher reactor pressure at the initiation of a LOCA and a slightly higher break flow. [

] Therefore, this LOCA analysis supports operation with 1 MSIVOOS.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 2-3 Figure 2.1 MAPLHGR Limit for ATRIUM 11 Fuel

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 2-4 Figure 2.2 [ ]

for ATRIUM 11 Fuel

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 3-1 3.0 LOCA Description 3.1 Accident Description The LOCA is described in the Code of Federal Regulations 10 CFR 50.46 as a hypothetical accident that results in a loss of reactor coolant from breaks in reactor coolant pressure boundary piping up to and including a break equivalent in size to a double-ended rupture of the largest pipe in the reactor coolant system. There is not a specifically identified cause that results in the pipe break. However, for the purpose of identifying a design basis accident, the pipe break is postulated to occur inside the primary containment before the first isolation valve.

For a boiling water reactor (BWR), a LOCA may occur over a wide spectrum of break locations and sizes. Responses to the break vary significantly over the break spectrum. The largest possible break is a double-ended rupture of a recirculation pipe; however, this is not necessarily the most severe challenge to the ECCS. A double-ended rupture of a main steam line causes the most rapid primary system depressurization, but because of other phenomena, steam line breaks are seldom limiting with respect to the event acceptance criteria (10 CFR 50.46).

Because of these complexities, an analysis covering the full range of break sizes and locations is performed to identify the limiting break characteristics.

Regardless of the initiating break characteristics, the event response is conveniently separated into three phases: the blowdown phase, the refill phase, and the reflood phase. The relative duration of each phase is strongly dependent upon the break size and location. [

]

During the blowdown phase of a LOCA, there is a net loss of coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered. There is a rapid decrease in pressure during the blowdown phase. During the early phase of the depressurization, the exiting coolant provides core cooling. Consistent with the discussion presented in Reference 1, [

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 3-2 In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory. During this phase the core sprays provide core cooling and, along with low-pressure and high-pressure coolant injection (LPCI and HPCI), supply liquid to refill the lower portion of the reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decay heat rate and the fuel cladding temperature continues to increase during the refill phase.

In the reflood phase, the coolant inventory has increased to the point where the mixture level re-enters the core region. During the core reflood phase, cooling is provided above the mixture level by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolant eventually reaches the core hot node and the fuel cladding temperature decreases. [

]

3.2 Acceptance Criteria A LOCA is a potentially limiting event that may place constraints on fuel design, local power peaking, and in some cases, acceptable core power level. During a LOCA, the normal transfer of heat from the fuel to the coolant is disrupted. As the liquid inventory in the reactor decreases, the decay heat and stored energy of the fuel cause a heatup of the undercooled fuel assembly.

In order to limit the amount of heat that can contribute to the heatup of the fuel assembly during a LOCA, an operating limit on the MAPLHGR is applied to each fuel assembly in the core.

The Code of Federal Regulations prescribes specific acceptance criteria (10 CFR 50.46) for a LOCA event as well as specific requirements and acceptable features for Evaluation Models (10 CFR 50 Appendix K). The conformance of the AURORA-B LOCA Evaluation Models to Appendix K is described in Reference 1. The ECCS must be designed such that the plant response to a LOCA meets the following acceptance criteria specified in 10 CFR 50.46:

The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

The calculated local oxidation of the cladding shall nowhere exceed 0.17 times the local cladding thickness.

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, except the cladding surrounding the plenum volume, were to react.

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 3-3

After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

These criteria are commonly referred to as the PCT criterion, the local oxidation criterion, the hydrogen generation criterion, the coolable geometry criterion, and the long-term cooling criterion. A MAPLHGR limit is established for each fuel type to ensure that these criteria are met.

LOCA results are provided in Section 6.0 to identify the LOCA events which produce the highest PCT [ ] LOCA analysis results demonstrating that the PCT, local oxidation, and hydrogen generation (core wide oxidation) criteria are met are provided in Section 9.0. Compliance with these three criteria ensures that a coolable geometry is maintained. Long-term coolability criterion is discussed in Section 8.0.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-1 4.0 LOCA Analysis Description The Evaluation Model used for the break spectrum analysis is the AURORA-B LOCA analysis methodology described in Reference 1. The AURORA-B LOCA methodology employs two major computer codes to evaluate the system and fuel response during all phases of a LOCA.

These are the S-RELAP5 and RODEX4 computer codes. A [

] of the LOCA to determine the PCT and maximum local clad oxidation for [

]

A complete analysis starts with the specification of fuel parameters using RODEX4 (Reference 4). RODEX4 is used to determine the [

] The initial stored energy used in S-RELAP5 is [

]

4.1 Break Spectrum Analysis S-RELAP5 is used to calculate the thermal-hydraulic response during all phases of the LOCA using a [

] The reactor vessel nodalization is shown in Figure 4.1 and the core nodalization is shown in Figure 4.2 consistent with those in the topical report submitted to the NRC (Reference 1). The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and decay heat as required by Appendix K of 10 CFR 50. The clad swelling and rupture models from NUREG-0630 (Reference 2) have been incorporated into S-RELAP5.

The S-RELAP5 model is executed over a range of break locations, break sizes, break types, initial statepoints, axial shapes and assumed single-failures to determine the break that yields the highest PCT [

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-2 4.2 Exposure Analysis The [

] from beginning-of-life to end-of-life [ ] increments to determine an exposure-dependent MAPLHGR limit and [ ] Figures of merit including PCT, local cladding oxidation, and core-wide metal-water reaction are evaluated over the range of exposures to confirm the acceptability of the LOCA analysis with respect to 10 CFR 50.46 criteria. [

]

4.3 Plant Parameters The LOCA analysis is performed using the plant parameters provided by the utility. Table 4.1 provides a summary of reactor initial conditions used in the break spectrum analysis. Table 4.2 lists selected reactor system parameters.

The LOCA analysis is performed for a full core of ATRIUM 11 fuel. Some of the key fuel parameters used in the analysis are summarized in Table 4.3.

4.4 ECCS Parameters The ECCS configuration is shown in Figure 4.3. Table 4.4 - Table 4.7 provide the important ECCS characteristics assumed in the analysis. The ECCS is modeled as time-dependent junctions connected to the appropriate reactor locations: LPCS injects into the upper plenum, HPCI injects into the upper downcomer, and LPCI injects into the recirculation lines. Although HPCI is expected to be available, no analysis mitigation credit is assumed for the HPCI system in any of the analyses discussed in this report.

The flow through each ECCS valve is determined based on system pressure and valve position.

Flow versus pressure for a fully open valve is obtained by linearly interpolating the pump

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-3 capacity data provided in Table 4.4 - Table 4.6. No credit for ECCS flow is assumed until the ECCS injection valves are fully open and the ECCS pumps reach rated speed.

The ADS valves are modeled as a junction connecting the reactor steam line to the suppression pool. The flow through the ADS valves is calculated based on pressure and valve flow characteristics. The valve flow characteristics are determined such that the calculated flow is equal to the rated capacity at the reference pressure shown in Table 4.7. Only five ADS valves are assumed operable in the analyses to support operation with one ADSVOOS and the potential single failure of one ADS valve during the LOCA.

In the Framatome LOCA analysis model, ECCS initiation is assumed to occur when the water level drops to the applicable level setpoint. No credit is assumed for the start of LPCS or LPCI due to high drywell pressure. [

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-4 Table 4.1 Initial Conditions Reactor power (% of rated) 102 102 [ ]

[ ]

Reactor power (MWt) 2981.5 2981.5 [ ]

[ ]

Steam flow rate (Mlb/hr) 13.1 13.1 11.6 8.9 Steam dome pressure (psia) 1047.6 1047.4 1034.9 1014.3 Core inlet enthalpy (Btu/lb) 527.6 522.3 516.4 512.1

[ ]

Rod average power distributions Figure 4.4 Figure 4.5 Figure 4.6 Figure 4.7

[ ]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-5 Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 220.5 Number of fuel assemblies 560 Recirculation suction pipe area (ft2) 3.67 1.0 DEG suction break area (ft2) 7.33 Recirculation discharge pipe area (ft2) 3.67 1.0 DEG discharge break area (ft2) 7.33

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-6 Table 4.3 ATRIUM 11 Fuel Assembly Parameters

Does not include additional inner channel milling near the top of the channel.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-7 Table 4.4 High-Pressure Coolant Injection Parameters Parameter Value Coolant temperature (maximum) (°F) 140 Initiating Signals and Setpoints Water level (in)* 459 High drywell pressure (psig) Not used Time Delays Time for HPCI pump to reach rated speed and injection valve wide open (sec) 60 Delivered Coolant Flow Rate Versus Pressure Vessel to Flow Torus P Rate (psid) (gpm) 0 0 150 3825 1164 3825

Relative to vessel zero.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-8 Table 4.5 Low-Pressure Coolant Injection Parameters Parameter Value Reactor pressure permissive for opening valves - analytical (psia) 410 Coolant temperature (maximum) (°F) 160 Initiating Signals and Setpoints Water level (in)* 358 High drywell pressure (psig) Not used Time Delays Time for LPCI pumps to reach rated speed (maximum) (sec) 31.8 LPCI injection valve stroke time (sec) 37.5 Delivered Coolant Flow Rate Versus Pressure Flow Rate for Flow Rate for 1 Pump 2 Pumps Injecting Into Injecting Into Vessel to 1 Recirculation 1 Recirculation Torus P Loop Loop (psid) (gpm) (gpm) 0 8,690 14,420 20 7,000 12,000 202 0 0

Relative to vessel zero.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-9 Table 4.6 Low-Pressure Core Spray Parameters Parameter Value Reactor pressure permissive for opening valves - analytical (psia) 410 Coolant temperature (maximum) (°F) 160 Initiating Signals and Setpoints Water level (in)* 358 High drywell pressure (psig) Not used Time Delays Time for LPCS pumps to reach rated speed (maximum) (sec) 39.7 LPCS injection valve stroke time (sec) 14.0 Delivered Coolant Flow Rate Versus Pressure Vessel to Flow Rate for Torus P 1 Pump (psid) (gpm) 0 5100 113 3850 265 0

Relative to vessel zero.

The delivered LPCS flow rate is reduced by a conservative 150 gpm to account for additional core shroud leakage.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-10 Table 4.7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 7 Number of valves available* 5 Minimum flow capacity of 4.15 at available valves 1112.4 (Mlbm/hr at psig)

Initiating Signals and Setpoints Water level (in) 358 High drywell pressure (psig) Not used Time Delays ADS timer (delay time from initiating signal to time valves are open (sec) 121

Only 5 valves are assumed operable in the analyses to support 1 ADSVOOS operation and the potential single failure of 1 ADS valve during the LOCA.

Relative to vessel zero.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-11 Figure 4.1 S-RELAP5 Vessel Model

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-12 Figure 4.2 S-RELAP5 Core Model

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-13 (j) (i) (i) (j)

DG-3 DG-1 DG-2 DG-4 LPCI LPCI LPCS HPCI LPCS LPCI LPCI 1A 1C 1A 1B 1D 1B LPCI LPCI Injection Injection Valve 1A Valve 1B Discharge Discharge Valve 1A Valve 1B Loop-A Loop-B Figure 4.3 ECCS Schematic*

This ECCS schematic is functionally equivalent for both units as it relates to single failure and ECCS availability. The schematic is provided for Unit 1, but representative of both Units 1 and 2.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-14 Figure 4.4 Rod Average Power Distributions for 102%P and [ ]

Mid- and Top-Peaked

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-15 Figure 4.5 Rod Average Power Distributions for 102%P and [ ]

Mid- and Top-Peaked

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-16 Figure 4.6 Rod Average Power Distributions for [ ]

Mid- and Top-Peaked

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 4-17 Figure 4.7 Rod Average Power Distributions for [ ]

Mid- and Top-Peaked

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-1 5.0 Break Spectrum Analysis Description The objective of the LOCA break spectrum analyses is to ensure that the operating conditions, break location, break type, break size, and ECCS single failure which produce the maximum PCT [ ] are identified. The LOCA response scenario varies considerably over the spectrum of break locations. Potential break locations have been separated into two groups: recirculation line breaks and non-recirculation line breaks. The basis for the break locations and potentially limiting single failures analyzed in this report is described in the following sections.

5.1 Limiting Single Failure Regulatory requirements specify that the LOCA analysis consider availability of offsite power supplies and that only safety grade systems and components are available. In addition, regulatory requirements also specify that the most limiting single failure of ECCS equipment must be assumed in the LOCA analysis. The term "most limiting" refers to the ECCS equipment failure that produces the greatest challenge to event acceptance criteria. The limiting single failure can be a common power supply, an injection valve, a system pump, or system initiation logic. The most limiting single failure may vary with break size and location. The potential limiting single failures identified in the UFSAR (Reference 6) are shown below:

DC power ( i ) (SF-BATT)

DC power ( j )

Diesel generator ( i )

Diesel generator ( j )

LPCI injection valve (SF-LPCI)

High-pressure coolant injection system (SF-HPCI)

The single failures and the available ECCS for each failure assumed in these analyses are summarized in Table 5.1. Other potential failures are not specifically considered because they result in as much or more ECCS capacity.

As indicated earlier, no analysis mitigation credit is assumed for the HPCI system in any of the LOCA analyses presented in this report. A review of Table 5.1 shows that the DC power (i) and LPCI injection valve failures are the two potentially limiting single failures as the other single

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-2 failures result in as much or more ECCS capacity. Only five ADS valves are assumed operable in the analyses to support operation with one ADSVOOS and the potential single failure of one ADS valve during the LOCA.

5.2 Recirculation Line Breaks The response during a recirculation line LOCA is dependent on break size. The rate of reactor vessel depressurization decreases as the break size decreases. The high-pressure ECCS and ADS will assist in reducing the reactor vessel pressure to the pressure where the LPCI and LPCS flows start. For large breaks, rated LPCS and LPCI flow is generally reached before or shortly after the time when the ADS valves open so the ADS system is not required to mitigate the LOCA. ADS operation is an important emergency system for small breaks where it assists in depressurizing the reactor system faster, and thereby reduces the time required to reach rated LPCS and LPCI flow.

The two largest flow resistances in the recirculation piping are the recirculation pump and the jet pump nozzle. For breaks in the discharge piping, there is a major flow resistance in both flow paths from the reactor vessel to the break. For breaks in the suction piping, both major flow resistances are in the flow path from the vessel to the pump side of the break. As a result, pump suction side breaks experience a more rapid blowdown, which tends to make the event more severe. For suction side breaks, the recirculation discharge isolation valve on the broken loop closes which allows the LPCI flow to fill the discharge piping and supply flow to the lower plenum and core. For discharge side breaks, the LPCI flow in the broken loop is assumed to exit the system through the break resulting in a decrease in available LPCI flow to the core, thereby increasing the severity of the event. Both suction and discharge recirculation pipe breaks are considered in the break spectrum analysis.

Two break types (geometries) are considered for the recirculation line break. The two types are the double-ended guillotine (DEG) break and the split break.

For a DEG break, the piping is assumed to be completely severed resulting in two independent flow paths to the containment. The DEG break is modeled by setting the break area (at both ends of the pipe) equal to the full pipe cross-sectional area and varying the discharge coefficient between 1.0 and 0.4. The range of discharge coefficients is used to cover uncertainty in the

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-3 actual geometry at the break. [

] The most limiting DEG break is determined by varying the discharge coefficient.

A split type break is assumed to be a longitudinal opening or hole in the piping that results in a single break flow path to the containment. Appendix K of 10 CFR 50 defines the cross-sectional area of the piping as the maximum split break area required for analysis.

Break types, break sizes, and single failures are analyzed for both suction and discharge recirculation line breaks.

Section 6.0 provides a description and results summary for breaks in the recirculation line.

5.3 Non-Recirculation Line Breaks In addition to breaks in the recirculation line, breaks in other reactor coolant system piping must be considered in the LOCA break spectrum analysis. Although the recirculation line large breaks result in the largest coolant inventory loss, they do not necessarily result in the most severe challenge to event acceptance criteria. The double-ended rupture of a main steam line is expected to result in the fastest depressurization of the reactor vessel. Special consideration is required when the postulated break occurs in ECCS piping. Although ECCS piping breaks are small relative to a recirculation pipe DEG break, the potential to disable an ECCS system increases their severity.

The following sections address potential LOCAs due to breaks in non-recirculation line piping.

Non-recirculation line breaks outside containment are inherently less challenging to fuel limits than breaks inside containment. For breaks outside containment, isolation or check valve closure will terminate break flow prior to the loss of significant liquid inventory and the core will remain covered. If high-pressure coolant inventory makeup cannot be reestablished, ADS actuation may become necessary. [

] Although analyses of breaks outside containment may be required to address non-fuel

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-4 related regulatory requirements, these breaks are not limiting relative to fuel acceptance criteria such as PCT.

5.3.1 Main Steam Line Breaks A steam line break [

] The break results in high steam flow out of the broken line and into the containment. Prior to MSIV closure, a steam line break also results in high steam flow in the intact steam lines as they feed the break via the steam line manifold. A steam line break inside containment results in a rapid depressurization of the reactor vessel. Initially the break flow will be high quality steam; however, the rapid depressurization produces a water level swell that results in liquid discharge at the break. For steam line breaks, the largest break size is most limiting because it results in the most level swell and liquid loss out of the break.

[

]

5.3.2 Feedwater Line Breaks

[

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-5

[

]

5.3.3 HPCI Line Breaks The HPCI injection line is connected to the feedwater line outside containment.

[

]

The HPCI steam supply line is connected to the main steam line inside containment.

[

]

5.3.4 LPCS Line Breaks A break in the LPCS line is expected to have many characteristics similar to [

] However, some characteristics of the LPCS line break are unique and are not addressed in other LOCA analyses. Two important differences from other LOCA analyses are that the break flow will exit from the region inside the core shroud and the break will disable one LPCS system. The LPCS line break is assumed to occur just outside the reactor vessel. [

]

5.3.5 LPCI Line Breaks The LPCI injection lines are connected to the larger recirculation discharge lines. [

]

5.3.6 RWCU Line Breaks The reactor water cleanup (RWCU) extraction line is connected to a recirculation suction line with an additional connection to the vessel bottom head. [ ]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-6

[

]

The RWCU return line is connected to the feedwater line; [

]

5.3.7 Shutdown Cooling Line Breaks The shutdown cooling suction piping is connected to a recirculation suction line and the shutdown cooling return line is connected to a recirculation discharge line. [

]

5.3.8 Instrument Line Breaks

[

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 5-7 Table 5.1 Available ECCS for Recirculation Line Break LOCAs Recirculation Recirculation Assumed Suction Break Discharge Break Failure

Systems Systems

, , § , § Remaining Remaining DC power ( i )

(SF-BATT) 1LPCS + 3LPCI + ADS 1LPCS + 1LPCI + ADS DC power ( j ) 2LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS Diesel generator ( i ) 1LPCS + 3LPCI + HPCI + ADS 1LPCS + 1LPCI + HPCI + ADS Diesel generator ( j ) 2LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS LPCI injection valve (SF-LPCI) 2LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS HPCI system (SF-HPCI) 2LPCS + 4LPCI + ADS 2LPCS + 2LPCI + ADS

Failure of either DC power ( i ) or diesel generator ( i ) will result in the loss of one diesel generator (DG-1 or DG-2). The loss of DC power ( i ) will also result in the loss of the HPCI. The loss of DC power ( j ) or diesel generator ( j ) will result in the loss of one diesel generator (DG-3 or DG-4). The ECCS availability presented is consistent with the Unit 1 schematic, Figure 4.3. Since the ECCS availability is functionally equivalent for both units, the information identified in Table 5.1 is representative of both Units 1 and 2.

Systems remaining, as identified in this table for recirculation suction line breaks, are applicable to other non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed for recirculation suction breaks, less the ECCS in which the break is assumed.

1LPCI (1 pump into 1 loop) means one RHR pump operating in one LPCI loop, 2LPCI (2 pumps into 1 loop) means two RHR pumps operating in one loop, 3LPCI (3 pumps into 2 loops) means three RHR pumps operating in two loops, 4LPCI (4 pumps into 2 loops) means four RHR pumps operating in two loops.

§ Although HPCI is expected to be available for some events, no accident analysis mitigation credit is assumed for this system.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-1 6.0 TLO Recirculation Line Break Spectrum Analyses The largest diameter recirculation system pipes are the suction line between the reactor vessel and the recirculation pump and the discharge line between the recirculation pump and the riser manifold ring. LOCA analyses are performed for breaks in both of these locations with consideration for both DEG and split break geometries. The break sizes considered included DEG breaks with discharge coefficients from 1.0 to 0.4 and split breaks with areas ranging between the full pipe area and [ ] ft2. As discussed in Section 5.0, the single failures considered in the recirculation line break analyses are SF-BATT and SF-LPCI.

[

]

6.1 Break Spectrum Analysis Results The break spectrum analyses demonstrate that the recirculation line break case with the highest PCT [ ] is the 1.0 DEG break in the pump suction piping with a single failure of SF-BATT and a top-peaked axial power shape when operating at 102%

rated core power and [ ] These two cases are presented in Table 6.1.

Table 6.2 provides a summary of the [ ] from the recirculation line break calculations for each of the single failures, state points, and axial power shapes. The event times for the 1.0 DEG break in the pump suction piping with a single failure of SF-BATT and a top-peaked axial power shape when operating at 102% rated core power and [

] are presented in Table 6.3 and plots of key parameters from the LOCA analyses of this case are provided in Figures 6.1 - 6.15.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-2 Table 6.1 Break Spectrum Results* for TLO Recirculation Line Breaks Break spectrum case resulting 1.0 DEG pump suction

[ ] SF-BATT Top-peaked axial 102%P/[ ]

Break spectrum case resulting 1.0 DEG pump suction

[ ] SF-BATT Top-peaked axial 102%P/[ ]

The cases identified in Table 6.1 from the TLO break spectrum analyses are further evaluated in Section 9.0 with exposure dependent analysis.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-3 Table 6.2 Summary of Break Spectrum [ ] for TLO Recirculation Line Breaks

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-4 Table 6.3 Event Times for the [ ] from the TLO Recirculation Line Break Spectrum Analysis

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-5 Figure 6.1 [ ] from the TLO Recirculation Line Break Spectrum Analysis Upper Plenum Pressure Figure 6.2 [ ] from the TLO Recirculation Line Break Spectrum Analysis Total Break Flow Rate

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-6 Figure 6.3 [ ] from the TLO Recirculation Line Break Spectrum Analysis Core Inlet Flow Rate Figure 6.4 [ ] from the TLO Recirculation Line Break Spectrum Analysis ADS Flow

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-7 Figure 6.5 [ ] from the TLO Recirculation Line Break Spectrum Analysis HPCI Flow Figure 6.6 [ ] from the TLO Recirculation Line Break Spectrum Analysis LPCS Flow

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-8 Figure 6.7 [ ] from the TLO Recirculation Line Break Spectrum Analysis LPCI Flow Figure 6.8 [ ] from the TLO Recirculation Line Break Spectrum Analysis RDIV Flows

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-9 Figure 6.9 [ ] from the TLO Recirculation Line Break Spectrum Analysis Relief Valve Flow Figure 6.10 [ ] from the TLO Recirculation Line Break Spectrum Analysis Downcomer LOCA Water Level

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-10 Figure 6.11 [ ] from the TLO Recirculation Line Break Spectrum Analysis Upper Plenum Liquid Level Figure 6.12 [ ] from the TLO Recirculation Line Break Spectrum Analysis Hot Channel Liquid Level

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-11 Figure 6.13 [ ] from the TLO Recirculation Line Break Spectrum Analysis Core Bypass Liquid Level Figure 6.14 [ ] from the TLO Recirculation Line Break Spectrum Analysis Lower Plenum Liquid Level

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 6-12 Figure 6.15 [ ] from the TLO Recirculation Line Break Spectrum Analysis Hot Channel Inlet Flow

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 7-1 7.0 Single-Loop Operation LOCA Analysis During SLO, the pump in one recirculation loop is not operating. A break may occur in either loop, but results from a break in the inactive loop would be similar to those from a two-loop operation break. If a break occurs in the inactive loop during SLO, the intact active loop flow to the reactor vessel would continue during the recirculation pump coastdown period and would provide core cooling similar to that which would occur in breaks during TLO. The system response would be similar to that resulting from an equal-sized break during two-loop operation.

A break in the active loop during SLO results in a more rapid loss of core flow and earlier degraded core conditions relative to those from a break in the inactive loop. Therefore, only breaks in the active recirculation loop are analyzed.

A break in the active recirculation loop during SLO will result in an earlier loss of core heat transfer relative to a similar break occurring during two-loop operation. This occurs because there will be an immediate loss of jet pump drive flow. Therefore, fuel rod surface temperatures will increase faster in an SLO LOCA relative to a TLO LOCA. Also, the early loss of core heat transfer will result in higher stored energy in the fuel rods at the start of the heatup. The increased severity of an SLO LOCA can be reduced by applying an SLO multiplier to the two-loop MAPLHGR limit.

7.1 SLO Analysis Modeling Methodology

[

]

7.2 SLO Analysis Results

[

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 7-2

[

]

The SLO analyses are performed with a 0.85 multiplier applied to the two-loop MAPLHGR limit resulting in an SLO MAPLHGR limit of [ ] kW/ft. [

] The analyses are performed at maximum stored energy fuel conditions. The limiting SLO LOCA is the 1.0 DEG break in the pump suction piping with a single failure of SF-LPCI and a mid-peaked axial power shape when operating at [

]

A comparison of the limiting SLO and the limiting two-loop results is provided in Table 7.1. The results in Table 7.1 show that the two-loop LOCA results bound the limiting SLO results when a 0.85 multiplier is applied to the two-loop MAPLHGR limit. [

]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 7-3 Table 7.1 Single- and Two-Loop Operation PCT Summary Operation Limiting Case PCT (°F)

Single-loop 1.0 DEG pump suction mid-peaked SF-LPCI [ ]

Two-loop 1.0 DEG pump suction top-peaked SF-BATT [ ]

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 8-1 8.0 Long-Term Coolability Long-term coolability addresses the issue of reflooding the core and maintaining a water level adequate to cool the core and remove decay heat for an extended time period following a LOCA. For non-recirculation line breaks, the core can be reflooded to the top of the active fuel and be adequately cooled indefinitely. For recirculation line breaks, the core will initially remain covered following reflood due to the static head provided by the water filling the jet pumps to a level of approximately two-thirds core height. Eventually, the heat flux in the core will not be adequate to maintain a two-phase water level over the entire length of the core. Beyond this time, the upper third of the core will remain wetted and adequately cooled by core spray.

Maintaining water level at two-thirds core height with one core spray system operating is sufficient to maintain long-term coolability as demonstrated by the NSSS vendor (Reference 7).

Since fuel temperatures during long-term cooling are low relative to the PCT and are not significantly affected by fuel design, this conclusion is applicable to ATRIUM 11 fuel. This LOCA analysis assesses conditions from the time of the initiation of the break to the time when long term cooling conditions can be established as demonstrated in Reference 7.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 9-1 9.0 Exposure-Dependent LOCA Analysis Description and Results Exposure-dependent LOCA results for ATRIUM 11 fuel are obtained by repeated analyses based on the cases identified in Table 6.1 from the break spectrum analysis [

]

Table 9.1 shows the exposure-dependent LOCA analysis results for the ATRIUM 11 fuel. The S-RELAP5 model is applied to obtain these results as described in Section 4.2. The analysis is performed at [

] which ensures appropriate limits are applied up to the monitored maximum assembly average and rod average exposure limits. The MAPLHGR input is consistent with the data in Figure 2.1. [

] Exposure-dependent fuel rod data is provided from RODEX4 results [

] The impact of thermal conductivity degradation is addressed with RODEX4.

The ATRIUM 11 limiting PCT is 1957°F at [ ] exposure for the 1.0 DEG break in the pump suction piping with a single failure of SF-BATT and a top-peaked axial power shape when operating at 102% rated core power and [ ]. The maximum local MWR of 4.75% occurred at [ ] exposure, [ ]

Analysis results show that the hot rod average MWR is 0.41%. Since all other rods in the core are at lower power, the core average metal water reaction (CMWR) will be significantly less than 0.41%.

Figure 9.1 shows the cladding temperature of the ATRIUM 11 PCT rod as a function of time for the limiting PCT result from the exposure-dependent LOCA analysis. The maximum temperature of 1957°F occurs at [ ]. These results demonstrate the acceptability of the ATRIUM 11 MAPLHGR limit shown in Figure 2.1.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 9-2 Table 9.1 ATRIUM 11 Exposure-Dependent LOCA Analysis Results CMWR is < 0.41% at all exposures.*

The rod average MWR for the hot rod is 0.41% which supports the conclusion that the CMWR is less than 0.41%.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 9-3 Figure 9.1 Limiting [ ] PCT Exposure-Dependent LOCA Analysis

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 10-1 10.0 Conclusions The AURORA-B LOCA Evaluation Model was applied to confirm the acceptability of the ATRIUM 11 MAPLHGR limit and [ ] for Brunswick Units 1 and 2. The following conclusions were made from the analyses presented in this report.

The limiting PCT is obtained from Section 9.0 based on a recirculation line break of 1.0 DEG break in the pump suction piping with a single failure of SF-BATT and a top-peaked axial power shape when operating at 102% of rated core power and [

].

The limiting break analysis identified above satisfies all the acceptance criteria specified in 10 CFR 50.46. The analysis is performed in accordance with 10 CFR 50 Appendix K requirements.

The multiplier applied to the MAPLHGR limit for SLO is 0.85 for ATRIUM 11 fuel. [

] This multiplier ensures that a LOCA from SLO is less limiting than a LOCA from two-loop operation.

The acceptance criteria of the Code of Federal Regulations (10 CFR 50.46) are met for operation at or below the ATRIUM 11 MAPLHGR limit given in Figure 2.1 [

].

Peak PCT < 2200oF.

Local cladding oxidation thickness < 17%.

Total hydrogen generation < 1%.

Coolable geometry, satisfied by meeting peak PCT, local cladding oxidation, and total hydrogen generation criteria.

Core long-term cooling, satisfied by concluding core flooded to top of active fuel or core flooded to the jet pump suction elevation (Reference 1).

The MAPLHGR limit and [ ] are applicable for ATRIUM 11 full cores as well as transition cores containing ATRIUM 11 fuel.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page 11-1 11.0 References

1. ANP-10332(P)(A) Revision 0, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios, Framatome, March 2019.

2. ANP-3772(P), Revision 0, CR Supplement Report on Brunswick LAR Analyses, Framatome, May 2019.

3. XN-NF-82-07(P)(A) Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982.

4. BAW-10247(P)(A) Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Framatome, February 2008.

5. Safety Evaluation by the Office of Nuclear Reactor Regulation, Licensing Topical Report NEDC-33006P, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, General Electric Hitachi Nuclear Energy America, LLC, October 2008 (ML081130008).

6. Updated FSAR Brunswick Steam Electric Plant, Units 1 and 2, Revision 25.

7. NEDO-20566A, General Electric Company Analytical Model for Loss of Coolant Analysis in Accordance with 10CFR50 Appendix K, September 1986.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-1 Appendix A Limitations from the Safety Evaluation for LTR ANP-10332PA Compliance to the limitations and conditions from Section 5 of the safety evaluation in ANP-10332PA, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios" (Reference 1) is discussed in the following table.

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-2 Appendix A (Continued)

Limitation and Limitation and Condition Description Disposition/Discussion Condition Number The AURORA-B LOCA evaluation model shall be supported by MICROBURN-B2 and the underlying cross section an approved nodal core simulator and lattice physics generation code, CASMO-4, are used for the nodal core methodology. Plant-specific licensing applications referencing simulator and lattice physics methodology from the the AURORA-B LOCA evaluation model shall identify the nodal following NRC-approved TR: EMF-2158(P)(A) Revision 0, 1

core simulator and lattice physics methods supporting the Siemens Power Corporation Methodology for Boiling AURORA-B LOCA analysis and reference an NRC-approved TR Water Reactors: Evaluation and Validation of CASMO-4 /

confirming their acceptability for the intended application. MICROBURN-B2, Siemens Power Corporation, October 1999.

The full, stand-alone version of the RODEX4 code shall be used The stand-alone version of RODEX4 is used to supply in accordance with an approved methodology to supply steady- steady-state fuel thermal-mechanical input in accordance state fuel thermal-mechanical inputs to the AURORA-B LOCA with the following NRC-approved methodology: BAW-2 evaluation model. 10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., February 2008.

The AURORA-B LOCA evaluation model may not be used to The analyses are within the limits of the TRs, SEs, code perform analyses that result in any of its constituent components manuals and plant-specific licensing applications.

or supporting codes (i.e., S-RELAP5, RODEX4 kernel, RODEX4, 3

core simulator and lattice physics methods) being operated outside approved limits documented in their respective TRs, SEs, code manuals, and plant-specific licensing applications.

TR ANP-10332P [ [

] LOCA report.

4

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-3 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number As discussed above in Section 2.1, the conclusions of this SE The analyses only apply regulatory requirements in effect apply only to the use of the AURORA-B LOCA evaluation model at the time the NRC staffs review was completed. They for the purpose of demonstrating compliance with relevant [ ].

5 regulatory requirements in effect at the time the NRC staffs technical review of ANP-10332P was completed (i.e., as of December 31, 2018).

This SE does not constitute [ The evaluation model [

].

6

] of the evaluation model.

[ The [

7 ].

].

[ The [

] in the analyses.

8

].

Safety analyses performed with the AURORA-B LOCA evaluation [

model [

].

9

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-4 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number To ensure adequate conservatism in future plant-specific safety A[

analyses, absent specific NRC staff approval for higher values, ].

10 this SE limits [

].

Plant-specific licensing applications referencing the AURORA-B BWR fuel rods are [

LOCA evaluation model [

].

11

].

The Appendix K lockout preventing the return to nucleate boiling The analyses [

[

12

]. ].

[ A[

13

]. ].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-5 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number Plant-specific licensing applications referencing the AURORA-B Analyses [

LOCA evaluation model [

].

14

].

[ The [

].

15

].

Plant-specific licensing applications referencing the AURORA-B [

LOCA evaluation model [

].

16

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-6 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number To assure satisfaction of GDC 35 (or similar plant-specific design The [

criterion), [

limiting, then 17

].

Safety analyses performed with the AURORA-B LOCA evaluation [ ].

18 model [

].

Safety analyses for [ The application of AURORA-B LOCA [

]. ].

19

[

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-7 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number Approximately [

].

Simulations supporting plant safety analyses [ Simulations [

].

20

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-8 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number As discussed in Section 3.3.5.7, Framatome used a [ The [ ].

21

].

The NRC staff has not specifically reviewed any plant parameters The licensee [

in ANP-10332P or deemed them acceptable for use in plant safety analyses. Therefore, [

].

22

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-9 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number Safety analyses performed with the AURORA-B LOCA evaluation A[

model shall include justification that [

].

23

].

[ [

].

24

].

Plant-specific licensing applications referencing the AURORA-B The [

LOCA evaluation model [

25

]

].

Controlled Document Framatome Inc. ANP-3674NP Revision 2 Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel Page A-10 Limitation and Limitation and Condition Description Disposition/Discussion Condition Number Plant-specific licensing applications referencing the AURORA-B The [

LOCA evaluation model [

].

26

].

As discussed in Section 4.3 of this SE, new or modified The analyses [

Framatome [ ].

27

].

RA-19-0243 Enclosure 3 Affidavit for ANP-3674P, Brunswick Units 1 and 2 LOCA Analysis for ATRIUM 11 Fuel, Revision 2

RA-19-0243 Enclosure 4 Technical Specification Mark-Ups - Unit 1

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)

c. Neutron FluxHigh 1 3(c) F SR 3.3.1.1.2 118.7% RTP SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13

d. Inop 1,2 3(c) G SR 3.3.1.1.5 NA SR 3.3.1.1.11

e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.2 NA SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17

f. OPRM Upscale 18% RTP(f) 3(c) I SR 3.3.1.1.2 (d)

SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18

3. Reactor Vessel Steam Dome Pressure 1,2 2 G SR 3.3.1.1.2 1077 psig High SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

4. Reactor Vessel Water LevelLow Level 1 1,2 2 G SR 3.3.1.1.2 153 inches SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

5. Main Steam Isolation ValveClosure 1 8 F SR 3.3.1.1.5 10% closed SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

6. Drywell PressureHigh 1,2 2 G SR 3.3.1.1.2 1.8 psig SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 (continued)

(c) Each APRM channel provides inputs to both trip systems.

(d) See COLR for OPRM Confirmation Density Algorithm (CDA) setpoints.

(f) Following DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region.

Brunswick Unit 1 3.3-10 Amendment No. 285

ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident Reporting Requirements (CRDA), Revision 0, March 2018. 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-19(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2.

ANP-10300P-A, AURORA-B: An 9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Evaluation Model for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Boiling Water Reactors; Summary Description.

Application to Transient and Accident 10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.

Scenarios, Revision 1, January 2018.

11. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, June 2011.

12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.

17. EMF-2292(P)(A), ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis -

Assessment of STAIF with Input from MICROBURN-B2.

19. NEDC-33075P-A, GE Hitachi Boiling Water Reactor, Detect and Suppress Solution - Confirmation Density, Revision 8, November 2013.

(continued)

Brunswick Unit 1 5.0-21 Amendment No. 285 ANP-3703P, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA, Revision 0, August 2018.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

Insert A

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Oscillation Power Range Monitor (OPRM) Report When a report is required by Condition I of LCO 3.3.1.1, "RPS Instrumentation,"

a report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.

Brunswick Unit 1 5.0-22 Amendment No. 285

Insert A

22. DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate Enhanced Option-III, Revision 0, September 2018

23. BAW-10247P-A, Supplement 2P-A, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 2: Mechanical Methods, Revision 0, August 2018

24. ANP-10340P-A, Incorporation of Chromia-Doped Fuel Properties in AREVA Approved Methods, Revision 0, May 2018

25. ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, Revision 0, May 2018

26. ANP-10332P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios, Revision 0, March 2019

RA-19-0243 Enclosure 5 Technical Specification Mark-Ups - Unit 2

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)

c. Neutron FluxHigh 1 3(c) F SR 3.3.1.1.2 118.7% RTP SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13

d. Inop 1,2 3(c) G SR 3.3.1.1.5 NA SR 3.3.1.1.11

e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.2 NA SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17

f. OPRM Upscale 18% RTP(f) 3(c) I SR 3.3.1.1.2 (d)

SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18

3. Reactor Vessel Steam Dome Pressure 1,2 2 G SR 3.3.1.1.2 1077 psig High SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

4. Reactor Vessel Water LevelLow Level 1 1,2 2 G SR 3.3.1.1.2 153 inches SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

5. Main Steam Isolation ValveClosure 1 8 F SR 3.3.1.1.5 10% closed SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17

6. Drywell PressureHigh 1,2 2 G SR 3.3.1.1.2 1.8 psig SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 (continued)

(c) Each APRM channel provides inputs to both trip systems.

(d) See COLR for OPRM Confirmation Density Algorithm (CDA) setpoints.

(f) Following DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region.

Brunswick Unit 2 3.3-10 Amendment No. 313

ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident Reporting Requirements (CRDA), Revision 0, March 2018.

5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-19(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2.

ANP-10300P-A, AURORA-B: An 9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Evaluation Model for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Boiling Water Reactors; Summary Description.

Application to Transient and Accident 10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code Scenarios, Revision 1, for BWR Transient Thermal-Hydraulic Core Analysis.

January 2018.

11. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, June 2011.

12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.

17. EMF-2292(P)(A), ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis -

Assessment of STAIF with Input from MICROBURN-B2.

19. NEDC-33075P-A, GE Hitachi Boiling Water Reactor, Detect and Suppress Solution - Confirmation Density, Revision 8, November 2013.

(continued)

Brunswick Unit 2 5.0-21 Amendment No. 313 ANP-3703P, BEO-III Analysis Methodology for Brunswick Using RAMONA5-FA, Revision 0, August 2018.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Insert A Methodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Oscillation Power Range Monitor (OPRM) Report When a report is required by Condition I of LCO 3.3.1.1, "RPS Instrumentation,"

a report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.

Brunswick Unit 2 5.0-22 Amendment No. 313

Insert A

22. DPC-NE-1009-P, Brunswick Nuclear Plant Implementation of Best-estimate Enhanced Option-III, Revision 0, September 2018

23. BAW-10247P-A, Supplement 2P-A, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 2: Mechanical Methods, Revision 0, August 2018

24. ANP-10340P-A, Incorporation of Chromia-Doped Fuel Properties in AREVA Approved Methods, Revision 0, May 2018

25. ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, Revision 0, May 2018

26. ANP-10332P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios, Revision 0, March 2019

v t e Letter Sequence Supplement
EPID:L-2018-LLA-0273, Plant, Units 1 and 2 -Proprietary Request for Additional Information (Approved, Closed)
Initiation Request, Request Acceptance... Supplement, Supplement, Supplement Administration Questions 18 October 2019 9 October 2019 Answers 29 May 2019 18 June 2019 17 October 2019 Results Approval Other:
MONTHYEAR2019-10-09 [Table View]

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