BSEP 10-0118, ANP-2950NP, Revision 0, Atrium 10XM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 Reload BRK2-20.

ANP-2950NP, Revision 0, Atrium 10XM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 Reload BRK2-20.

ML102920553
Person / Time
Site:	Brunswick
Issue date:	10/12/2010
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
BSEP 10-0118, TAC ME3856, TAC ME3857, TAC ME3858, TAC ME3859, TSC-2010-01, TSC-2010-02 ANP-2950NP, Rev 0
Download: ML102920553 (25)

Text

BSEP 10-0118 Enclosure 3 AREVA Report ANP-2950NP, Revision 0, ATRIUM IOXM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 ReloadBRK2-20, dated October 2010

Contrc';I: Docurl 1, ANP-2950NP Revision 0 ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Octobeýr 2010 AREVA NP Inc. AREVA

Controlled Document AREVA NP Inc.

ANP-2950NP Revision 0 ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 Reload BRK2-20

Controiled Document AREVA NP Inc.

ANP-2950NP Revision 0 Copyright © 2010 AREVA NP Inc.

All Rights Reserved

Contro% Dpocument ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Paqe i Nature of Changes Item Page Description and Justification I All This is the initial release.

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ControIed AREVA NP Document ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Paqe ii Contents C o nte n ts .................................................................................................................................... ii 1.0 Introduction ............................................................................................................ 1-1 2.0 Summary and Conclusions ........................................................................................... 2-1 3.0 Fuel Rod Design Evaluation ......................................................................................... 3-1 4.0 References ................................................................................................................... 4-1 Tables Table 2-1 Summary of Fuel Rod Design Evaluation Results ................................................................... 2-2 Table 3-1 Key Fuel Rod Design Parameters .......................................................................................... 3-11 Table 3-2 RODEX4 Fuel Rod Results for Equilibrium Cycle Conditions ................................................ 3-12 Table 3-3 RODEX4 Fuel Rod Results for Brunswick Unit 2 Cycle 20 Operation .................................... 3-12 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses ....................................................... 3-13 .

Figures Figure 2-1 LHGR Limit (Normal Operation) .............................................................................................. 2-3 This document contains a total of 24 pages.

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Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel[Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page iii Nomenclature AOO anticipated operational occurrences ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel BOL beginning of life BWR boiling water reactor CRWE control rod withdrawal error CUF cumulative usage factor EOL end of life FDL fuel design limit ID inside diameter MWd/kgU megawatt days per kilogram of initial uranium LHGR linear heat generation rate NRC Nuclear Regulatory Commission, U. S.

OD outside diameter PCI pellet-to-cladding-interaction PLFR part length fuel rod ppm parts per million SRA stress relieved annealed S-N stress amplitude versus number of cycles UTL upper tolerance limit AREVA NP Inc.

Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical. Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 1-1 1.0 Introduction Results of the fuel rod thermal and mechanical analyses are presented to demonstrate that the applicable design criteria are satisfied. The analyses are for the AREVA NP* ATRIUMt 1OXM fuel that will be inserted for operation in Brunswick Unit 2 Cycle 20 as reload batch BRK2-20.

The evaluations are based on methodologies and design criteria approved by the U.S. NRC.

Equilibrium cycle conditions as well as Cycle 20 conditions are included in the analyses.

The analysis results are evaluated according to the generic fuel rod thermal and mechanical design criteria contained in ANF-89-98(P)(A) Revision 1 and Supplement 1 (Reference 1) along with design criteria provided in the RODEX4 fuel rod thermal-mechanical topical report (Reference 2).

The RODEX4 fuel rod thermal-mechanical analysis code is used to analyze the fuel rod for fuel centerline temperature, cladding strain, rod internal pressure, cladding collapse, cladding fatigue and external oxidation. The code and application methodology are described in the RODEX4 topical report (Reference 2). The cladding steady-state stress and plenum spring design methodology are summarized in Reference 1.

The following sections describe the fuel rod design, design criteria and methodology with reference to the source topical reports. Results from the analyses are summarized for comparison to the design criteria.

AREVA NP Inc. is an AREVA and Siemens company.

-t ATRIUM is a trademark of AREVA NP.

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ControixW ADpcument ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 2-1 2.0 Summary and Conclusions Key results are shown in Table 2-1 in comparison to each of the design criterion. Results are presented for the limiting cases. Additional RODEX4 results from different cases are given in Section 3.0.

The analyses support a maximum fuel rod discharge exposure of 62 MWd/kgU. However, the licensed rod average exposure is limited at the Brunswick units to 60 MWd/kgU (Reference 3).

Fuel rod criteria applicable to the design are summarized in Section 3.0. Analyses show the criteria are satisfied when the fuel is operated at or below the LHGR (linear heat generation limit) presented in Figure 2-1.

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ont olAd AREVADoR NP umen ANP-2950NP ATRIUM 10XM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Paae 2-2 Table 2-1 Summary of Fuel Rod Design Evaluation Results Criteria Section* Description Criteria Result 3.2 Fuel Rod Criteria 3.2.1 Internal hydriding (3.1.1) Cladding collapse (3.1.2) Overheating of fuel No fuel melting pellets margin to fuel melt > 0. 'F 3.2.5 Stress and strain limits (3.1.1) Pellet-cladding [

(3.1.2) interaction 3.2.5.2 Cladding stress ]

3.3 Fuel System Criteria (3.1.1) Fatigue (3.1.1) Oxidation, hydriding, [

and crud buildup (3.1.1) Rod internal pressure [

(3.1.2) 3.3.9 Fuel rod plenum Plenum spring to [

spring (fuel handling)

Numbers in the column refer to paragraph sections in the generic design criteria document, ANF 98(P)(A) Revision 1 and Supplement 1 (Reference 1). A number in parentheses is the paragraph section in the RODEX4 fuel rod topical report (Reference 2).

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Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 2-3 I

I Figure 2-1 LHGR Limit (Normal Operation)

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ControA AREc ument ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-1 3.0 Fuel Rod Design Evaluation Summaries of the design criteria and methodology are provided in this section along with analysis results in comparison to criteria. Both the fuel rod criteria and fuel system criteria as directly related to the fuel rod analyses are covered.

The fuel rod analyses cover normal operating conditions and AOOs (anticipated operational occurrences). The fuel centerline temperature analysis (overheating of fuel) and cladding strain analysis take into account slow transients at rated operating conditions.

Other fuel rod related topics on overheating of cladding, cladding rupture, fuel rod mechanical fracturing, rod bow, axial irradiation growth, cladding embrittlement, violent expulsion of fuel and fuel ballooning are evaluated as part of the respective fuel assembly structural analysis, thermal hydraulic analyses, or LOCA analyses and are reported elsewhere. The evaluation of fast transients and transients at off-rated conditions also are reported separate from this report.

3.1 Fuel Rod Design AREVA NP Inc.

Con lAREqA troe ocument P

ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-2 Table 3-1 lists the main parameters for the fuel rod and components.

3.2 Summary of Fuel Rod Design Evaluation Results from the analyses are listed in Table 3-2 through Table 3-4. Summaries of the methods and codes used in the evaluation are provided in the following paragraphs. The design criteria also are listed along with references to the sections of the design criteria topical reports (References 1 and 2).

The fuel rod thermal and mechanical design criteria are summarized as follows.

" Internal Hydriding. The fabrication limit [

] to preclude cladding failure caused by internal sources of hydrogen (Section 3.2.1 of Reference 1).

" Cladding Collapse. Clad creep collapse shall be prevented. [

] (Section 3.1.1 of Reference 2).

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Controlled' Docu ent AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-3

• Overheating of Fuel Pellets. The fuel pellet centerline temperature during anticipated transients shall remain below the melting temperature (Section 3.1.2 of Reference 2).

" Stress and Strain Limits. [

] during normal operation and during anticipated transients (Sections 3.1.1 and 3.1.2 of Reference 2).

Fuel rod cladding steady-state stresses are restricted to satisfy limits derived from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code (Section 3.2.5.1 of Reference 1).

• Cladding Fatigue. The fatigue cumulative usage factor for clad stresses during normal operation and design cyclic maneuvers shall be below [ ] (Section 3.1.1 of Reference 2).

" Cladding Oxidation, Hydriding and Crud Buildup. The maximum cladding oxidation shall be less than [ ] to prevent clad corrosion failure (Section 3.1.1 of Reference 2).

" Rod Internal Pressure. The rod internal pressure is limited [

] to assure that significant outward clad creep does not occur and unfavorable hydride reorientation on cooldown does not occur (Section 3.1.1 of Reference 2).

• Plenum Spring Design (Fuel Handling). The rod plenum spring must maintain a force against the fuel column stack [ ](3.3.9 of Reference 1).

The cladding collapse, overheating of fuel, cladding transient strain, cladding cyclic fatigue, cladding oxidation, and rod pressure are evaluated [ ] Cladding stress and the plenum spring are evaluated on a design basis.

3.2.1 Internal Hydridinq The absorption of hydrogen by the cladding can result in cladding failure due to reduced ductility and formation of hydride platelets. Careful moisture control during fuel fabrication reduces the potential for hydrogen absorption on the inside of the cladding. The fabrication limit [

] is verified by quality control inspection during fuel manufacturing.

3.2.2 Claddingq Collapse Creep collapse of the cladding and the subsequent potential for fuel failure is avoided in the design by limiting the gap formation due to fuel densification subsequent to pellet-clad contact.

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Controlled AREVA NP Document ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-4 The size of the axial gaps which may form due to densification following first pellet-clad contact shall be less than [ I The evaluation is performed using RODEX4. The design criterion and methodology are described in Reference 2. RODEX4 takes into account the [

]. A brief overview of RODEX4 and the statistical methodology is provided in the next section.

Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.

3.2.3 Overheatinq of Fuel Pellets Fuel failure from the overheating of the fuel pellets is not allowed. The centerline temperature of the fuel pellets must remain below melting during normal operation and AQOs. The melting point of the fuel includes adjustments for gadolinia content. AREVA establishes an LHGR limit to protect against fuel centerline melting during steady-state operation and during AQOs.

Fuel centerline temperature is evaluated using the RODEX4 code (Reference 2) for both normal operating conditions and AQOs. A brief overview of the code and methodology follow.

RODEX4 evaluates the thermal-mechanical responses of the fuel rod surrounded by coolant.

The fuel rod model considers the fuel column, gap region, cladding, gas plena and the fill gas and released fission gases. The fuel rod is divided into axial and radial regions with conditions computed for each region. The operational conditions are controlled by [

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Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-5 The heat conduction in the fuel and clad is [

Mechanical processes include [

As part of the methodology, fuel rod power histories are generated [

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Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-6 I.

Since RODEX4 is a best-estimate code, uncertainties [

]. Uncertainties taken into account in the analysis are summarized as:

P

• Power measurement and operational uncertainties -[

Manufacturing uncertainties - [

. Model uncertainties - [

Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.

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ControNEJAt cument ANP-2950NP ATRIUM IOXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Pacje 3-7 3.2.4 Stress and Strain Limits 3.2.4.1 Pellet/CladdingInteraction Cladding strain caused by transient-induced deformations of the cladding is calculated using the RODEX4 code and methodology as described in Reference 2. See Section 3.2.3 for an overview of the code and method. [

1.

Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.

3.2.4.2 Cladding Stress Cladding stresses are calculated using solid mechanics elasticity solutions and finite element methods. The stresses are conservatively calculated for the individual loadings and are categorized as follows:

Category Membrane Bending Primary Secondary Stresses are calculated at the cladding outer and inner diameter in the three principal directions for both beginning of life (BOL) and end of life (EOL) conditions. At EOL, the stresses due to mechanical bow and contact stress are decreased due to irradiation relaxation. The separate stress components are then combined, and the stress intensities for each category are compared to their respective limits.

The cladding-to-end cap weld stresses are evaluated for loadings from differential pressure, differential thermal expansion, rod weight, and plenum spring force.

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Controt ed Document AREVA NP ANP-2950NP ATRIUM I0XM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-8 The design limits are derived from the ASME (American Society of Mechanical Engineers)

Boiler and Pressure Vessel (B&PV) Code Section III (Reference 4) and the minimum specified material properties.

Table 3-4 lists the results in comparison to the limits for hot, cold, BOL and EOL conditions.

3.2.5 Fuel Densification and Swellincq Fuel densification and swelling are limited by the design criteria for fuel temperature, cladding strain, cladding collapse, and rod internal pressure criteria. Although there are no explicit criteria for fuel densification and swelling, the effect of these phenomena are included in the RODEX4 fuel rod performance code.

3.2.6 Fatigue

.. The CUF (cumulative usage factor) is summed for all of the axial regions of the fuel rod using Miner's rule. The axial region with the highest CUF is used in the subsequent [

] is determined. The maximum CUF for the cladding must remain below [ ]to satisfy the design criterion.

Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.

3.2.7 Oxidation, Hydridinq, and Crud Buildup Cladding external oxidation is calculated using RODEX4. Section 3.2.3 includes an overview of the code and method. The corrosion model includes an enhancement factor that is derived from poolside measurement data to obtain a fit of the expected oxide thickness. An uncertainty on the model enhancement factor also is determined from the data. The model uncertainty is included as part of the [

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Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-9 In the event abnormal crud is discovered or expected for a plant, a specific analysis is required to address the higher crud level. An abnormal level of crud is defined by a formation that increases the calculated fuel average temperature by 250C above the design basis calculation.

The formation of crud is not calculated within RODEX4. Instead, an upper bound of expected crud is input by the use of the crud heat transfer coefficient. The corrosion model also takes into consideration the effect of the higher thermal resistance from the crud on the corrosion rate. A higher corrosion rate is therefore included as part of the abnormal crud evaluation. A similar specific analysis is required if a plant experiences higher corrosion instead of crud.

The maximum oxide on the fuel rod cladding shall not exceed [ J. The limit is evaluated such that greater than [ I Currently, there is no hydrogen limit and no hydrogen uptake is reported.

Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.

3.2.8 Rod Internal Pressure Fuel rod internal pressure is calculated using the RODEX4 code and methodology as described in Reference 2. Section 3.2.3 provides an overview of the code and method. The maximum rod pressure is calculated under steady-state conditions and also takes into account slow transients. Rod internal pressure is limited to [ I. The expected upper bound of rod pressure [ ]

is calculated for comparison to the limit.

Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.

3.2.9 Plenum Sprinq Desigqn (Fuel Assembly Handling)

The plenum spring must maintain a force against the fuel column to [

]. This is accomplished by designing and verifying the spring force in relation to AREVA NP Inc.

Controlled ARaVA Document NEu ANP-2950NP ATRIUM I10XMV Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-10 the fuel column weight. The plenum spring is designed such that the [

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Controlled Document AREVA NP ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Paae 3-11 Table 3-1 Key Fuel Rod Design Parameters Characteristic Material or Value I

i

+

_____________________________1*

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ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 3-12 Table 3-2 RODEX4 Fuel Rod Results for Equilibrium Cycle Conditions Table 3-3 RODEX4 Fuel Rod Results for Brunswick Unit 2 Cycle 20 Operations Criteria Topic Limit Steady-State [ ] [

Note that Cycle 20 results are provided up to the end of Cycle 20.

Fatigue result is extrapolated to three cycles of operation based on the Cycle 20 result.

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Controlled AREVADocument NP ANP-2950NP ATRIUM 10XM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Paqe 3-13 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses Description, stress category ICriteria Result Cladding stress ]

Pmn (primary membrane stress)[]

Pm + Pb (primary membrane + bending) []

P + Q (primary + secondary)[]

Cladding-End Cap stress PM + Pb []

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ControlqaA pcument ANP-2950NP ATRIUM 1OXM Fuel Rod Thermal and Mechanical Evaluation Revision 0 for Brunswick Unit 2 Cycle 20 Reload BRK2-20 Page 4-1 4.0 References

1. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic MechanicalDesign Criteriafor BWR Fuel Designs,Advanced Nuclear Fuels Corporation, May 1995.

2. BAW-1 0247PA Revision 0, Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors, February 2008.

3. Letter, U.S. Nuclear Regulatory Commission to CP&L, "Issuance of Amendment No. 153 to Facility Operating License No. DPR-62, Brunswick Steam Electric Plant, Unit 2, Regarding Fuel Cycle No. 8 - Reload Extended Burnup Fuel (TAC No. 66155),"

September 20, 1988 (38-9061815-000).

4. ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components," 1977.

5. O'Donnell, W.J., and B. F. Langer, "Fatigue Design Basis for Zircaloy Components,"

Nuclear Science and Engineering, Vol. 20, 1964.

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