ML13200A199
| ML13200A199 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/31/2013 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML13200A185 | List: |
| References | |
| L-MT-13-055 ANP-3221NP, Rev 0 | |
| Download: ML13200A199 (27) | |
Text
Enclosure 23 AREVA Report ANP-3221(NP)
Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Revision 0 26 pages follow
uontroiieci uoeurent ANP-3221 NP Revision 0 Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 May 2013 A
AREVA NP Inc. AREVA
uontrolled Uocument AREVA NP Inc.
ANP-3221NP Revision 0 Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28
Lontrolled uocument AREVA NP Inc.
ANP-3221NP Revision 0 Copyright © 2013 AREVA NP Inc.
All Rights Reserved
Uontrolled Vocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page i Nature of Changes Item Page Description and Justification 1 All This is the initial release.
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uontrollea uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page ii Contents 1.0 Introduction.............................1-1 2.0 Sum m ary and Conclusions........................................................................................... .2 -1 3.0 Fuel Rod Design Evaluation ........................................................................................... 3-1 3.1 Fuel Rod Design .............................................................................................................. 3-1 3.2 Sum mary of Fuel Rod Design Evaluation ........................................................................ 3-2 3.2.1 Internal Hydriding ............................................................................................. 3-3 3.2.2 Cladding Collapse ............................................................................................ 3-4 3.2.3 Overheating of Fuel Pellets .............................................................................. 3-4 3.2.4 Stress and Strain Lim its ................................................................................... 3-7 3.2.5 Fuel Densification and Swelling ....................................................................... 3-8 3 .2 .6 F a tig ue .............................................................................................................. 3 -8 3.2.7 Oxidation, Hydriding, and Crud Buildup ........................................................... 3-9 3.2.8 Rod Internal Pressure .................................................................................... 3-10 3.2.9 Plenum Spring Design (Fuel Assem bly Handling) ......................................... 3-10 4.0 References ..................................................................................................................... 4-1 Tables Table 2-1 Sum mary of Fuel Rod Design Evaluation Results .................................................................... 2-2 Table 3-1 Key Fuel Rod Design Parameters .......................................................................................... 3-12 Table 3-2 RODEX4 Fuel Rod Results for Equilibrium Cycle Conditions ................................................ 3-13 Table 3-3 RODEX4 Fuel Rod Results for Monticello Cycle 28 ............................................................... 3-14 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses ....................................................... 3-15 Figures Figure 2-1 LHGR Lim it (Normal Operation) .............................................................................................. 2-3 This document has a total of 26 pages.
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uJontrolied Vocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page iii Nomenclature AOO anticipated operational occurrences AREVA AREVA NP Inc.
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 EPU extended power uprate FDL fuel design limit ID inside diameter MELLLA Maximum extended load line limit analysis MPa Megapascal MWd/kgU megawatt days per kilogram of initial uranium LHGR linear heat generation rate LTP lower tie plate NRC U.S. Nuclear Regulatory Commission OD outside diameter PCI pellet -cladding-interaction PLFR part-length fuel rod ppm parts per million psia pounds per square inch absolute S-N stress amplitude versus number of cycles AREVA NP Inc.
uontrollea Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 1-1 1.0 Introduction Results of the fuel rod thermal-mechanical analyses are presented to demonstrate that the applicable design criteria are satisfied. The analyses are for the AREVA NP Inc. (AREVA)
ATRIUM M T
1OXM fuel that will be inserted for operation in Monticello Cycle 28 as reload batch MON 1-28. The evaluations are based on methodologies and design criteria approved by the U. S. Nuclear Regulatory Commission (NRC). Equilibrium cycle conditions, as well as Cycle 28 conditions, are included in the analyses. The analyses take into account extended power uprate (EPU) and maximum extended load line limit analysis (MELLLA) operating conditions for Cycle 28.
The analysis results are evaluated according to the generic fuel rod thermal and mechanical design criteria contained in Reference 1, along with design criteria provided in the RODEX4 fuel rod thermal-mechanical topical report (Reference 2). The cladding external oxidation limit was reduced according to a regulatory commitment made to the NRC when RODEX4 was first implemented (Reference 3).
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 fuel rod design is very similar to the ATRIUM 1OXM design currently supplied in reload quantities to two U.S. boiling water reactor BWR/4 units, except the fuel column length is shorter by 4.76 inches for compatibility with a BWR/3 core height. The ATRIUM 1OXM fuel rod design is based on the ATRIUM-10 design in a way that preserves the nearly 20 years of extensive operating experience and performance history of the ATRIUM-10 rod design.
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.
ATRIUM is a trademark of AREVA NP.
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uontrole:: Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Paqe 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 analysis methodology supports a maximum fuel rod discharge exposure of 62 MWd/kgU.
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 linear heat generation rate (LHGR) presented in Figure 2-1.
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uontrolled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 10XM Fuel Assemblies, Cycle 28 Page 2-2 Table 2-1 Summary of Fuel Rod Design Evaluation Results Criteria Section* Description Criteria Result, Margin t or Comment 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. 'C 3.2.5 Stress and strain limits (3.1.1) Pellet-cladding (3.1.2) interaction (PCI) 3.2.5.2 Cladding stress [ ]
3.3 Fuel System Criteria (3.1.1) Fatigue [ J (3.1 .1 Oxidation, hydriding, [ ]
and crud buildup (3.1.1) Rod internal pressure [ ]
(3.1.2) 3.3.9 Fuel rod plenum spring Plenum spring to [
(fuel handling)
_____________________________________]
Numbers in the column refer to paragraph sections in the generic design criteria document (Reference 1). A number in parentheses is the paragraph section in the RODEX4 fuel rod topical report (Reference 2).
Margin is expressed as (limit - result)
The cladding external oxidation limit is restricted to [ ] pm by Reference 3.
AREVA NP Inc.
uontrolled Vocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 2-3 I
I Figure 2-1 LHGR Limit (Normal Operation)
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luontrolled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 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 anticipated operational occurrences (AOOs). 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 loss-of-coolant accident analyses and are reported elsewhere. The evaluation of fast transients and transients at off-rated conditions also are reported separately from this report.
3.1 Fuel Rod Design I
AREVA NP Inc.
uontrolled uocument ANP-3221 NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-2 I
] plenum spring on the upper end of the fuel column [
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).
AREVA NP Inc.
Uontrolled uocument ANP-3221 NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-3
" Cladding Collapse. Clad creep collapse shall be prevented. [
] (Section 3.1.1 of Reference 2).
- 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 3.2.5.2 of Reference 1).
" Cladding Fatigue. The fatigue CUF 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. Section 3.1.1 of Reference 2 limits the maximum cladding oxidation to less than [ ] pm to prevent clad corrosion failure. The oxidation limit is further reduced to [ ] pm consistent with a regulatory commitment made to the NRC during the first application of the RODEX4 methodology (Reference 3).
- 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 [ ] (Section 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 Hydridincq 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 [ I AREVA NP Inc.
uontrollea uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-4
[ ] is verified by quality control inspection during fuel manufacturing.
3.2.2 Claddinq 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.
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 AOOs. 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 AOOs.
Fuel centerline temperature is evaluated using the RODEX4 code (Reference 2) for both normal operating conditions and AOOs. 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 the [ ]
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Lontrolled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Paqe 3-5
[
[
The heat conduction in the fuel and clad is [
Mechanical processes include [
]e.
As part of the methodology, fuel rod power histories are generated [
I AREVA NP Inc.
uontrolled Uocument ANP-3221 NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Paqe 3-6 I
I I
I Since RODEX4 is a best-estimate code, uncertainties [
]. Uncertainties taken into account in the analysis are summarized as:
0 Power measurement and operational uncertainties - [
Manufacturing uncertainties - [
0 Model uncertainties - [
AREVA NP Inc.
uontrolled uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Paqe 3-7 I
Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.
3.2.4 Stress and Strain Limits 3.2.4.1 Pellet/Cladding Interaction 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. [
I.
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:
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Uontrolled uocument ANP-3221NP FuelRod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-8 Category Membrane Bending Primary Secondary Stresses are calculated at the cladding outer and inner diameter in the three principal directions for both 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.
The design limits are derived from the ASME B&PV Code (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 Swellinq 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 is summed for all of the axial regions of AREVA NP Inc.
Lontrolled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-9 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, Hydridin.q, 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 [ ].
1.
In the event abnormal crud is observed 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 25°C 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 current water chemistry conditions at the Monticello plant, along with past operating history, indicate normal, low crud levels. The fuel rod analyses were performed with the assumption that normal, current crud conditions continue through Cycle 28.
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Uontrolled uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Paae 3-10 The maximum calculated oxide on the fuel rod cladding shall not exceed [ ] pm. Previously, a [ ] pIm limit was approved as part of the RODEX4 methodology (Reference 2). Concerns were raised on the effect of non-uniform corrosion, such as spallation, and localized hydride formations on the ductility limit of the cladding. As a result, a regulatory commitment was made to reduce the limit to [ ] pm (Reference 3).
Currently, there is [ ]. However, as mentioned above, the ' ] pm was established, in part, as a means of [
The oxide limit is evaluated such that greater than [
I 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 [
]. 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 Spring Design (FuelAssembly 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 the fuel column weight. The plenum spring is designed such that the I I AREVA NP Inc.
Gontroled uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-11 I
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uontroiled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-12 Table 3-1 Key Fuel Rod Design Parameters Characteristic Material or Value II I
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(Jontrolled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-13 Table 3-2 RODEX4 Fuel Rod Results for Equilibrium Cycle Conditions Margin to Limit Criteria Topic Limit Steady-State I [
[ ]
]
C I C I C I Margin is defined as (limit - result).
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Uontrolled uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-14 Table 3-3 RODEX4 Fuel Rod Results for Monticello Cycle 28*
Margin to Limit Criteria Topic Limit Steady-State [ [
I I
[ I
[ ]
Note that cycle-specific results are provided up to the end of cycle.
1 Fatigue result is extrapolated to three cycles of operation.
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Uontrolled uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 3-15 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses Description, Stress Category Criteria Result Cladding stress [
Pm (primary membrane stress) [ ]
Pm + Pb (primary membrane + bending) [ ]
P + Q (primary + secondary) [ ]
Cladding-End Cap stress Pm + Pb[ ]
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uontrolled Uocument ANP-3221NP Fuel Rod Thermal-Mechanical Design for Revision 0 Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28 Page 4-1 4.0 References
- 1. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 2. BAW-1 0247PA Revision 0, Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors,AREVA NP Inc., February 2008.
- 3. Letter from Farideh E. Saba (NRC) to Michael J. Annacone (CP&L), "BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING ADDITION OF ANALYTICAL METHODOLOGY TOPICAL REPORT TO TECHNICAL SPECIFICATION 5.6.5 (TAC NOS. ME3858 AND ME3859),
ML11101A043," NRC 1109968, dated April 8, 2011.
- 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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