ML15251A385
| ML15251A385 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 08/31/2015 |
| From: | AREVA |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15251A394 | List:
|
| References | |
| RS-15-237 ANP-3324NP, Rev 1 | |
| Download: ML15251A385 (25) | |
Text
Attachment 10 Fuel Rod Thermal-Mechanical Design Report (Non-Proprietary Version)
Controiled Document ARE VA 1/4 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design AN P-3324N P Revision 1 Licensing Report August 2015 AREVA Inc.
(c) 2015 AREVA Inc.
Copyright © 2015 AREVA Inc.
All Rights Reserved
Controlled Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensina Report P~np~
. ii i
i Nature of Changes Item 1
Section(s) or Page(s)
Page 1-1 Description and Justification Added Proprietary brackets to two values in Section 1.0.
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AN P-3324N P Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Report Pagqe ii Contents Page
1.0 INTRODUCTION
............................................................................. 1-1 2.0
SUMMARY
AND CONCLUSIONS.......................................................... 2-1 3.0 FUEL ROD DESIGN EVALUATION........................................................ 3-1 3.1 Fuel Rod Design..................................................................... 3-1 3.2 Summary 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 Limits..................................................... 3-7 3.2.5 Fuel Densification and Swelling............................................ 3-8 3.2.6 Fatigue....................................................................... 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 Assembly Handling)....................... 3-10
4.0 REFERENCES
............................................................................... 4-1 List of 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 Quad Cities Unit 2 Cycle 24......................... 3-13 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses............................ 3-14 List of Figures Figure 2-1 LHGR Limit (Normal Operation)........................................................ 2-3
Controlled Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensinq Report Paqe iii Nomenclature Acronym Definition 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 CUE cumulative usage factor EOL end of life FDL fuel design limit ID inside diameter MWd/kgU megawatt dlays per kilogram of initial uranium LHGR linear heat generation rate NRC Nuclear Regulatory Commission, U. S.
OD outside diameter OOS Equipment Out Of Service PCI pellet-to-cladding-interaction PLFR part length fuel rod ppm parts per million S-N stress amplitude versus number of cycles SRS Single Rod Sequence
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ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensin~q Report Pa~qe 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 Inc. (AREVA)
ATRIUM TM 10XM* fuel that will be inserted for operation in Quad Cities Unit 2 Cycle 24 as reload batch QCI2-24. 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 24 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 cladding external oxidation limit defined by Reference 2 is [
],
however the reduced value of [
] was used to match a regulatory commitment made to the NRC when RODEX4 was first implemented in the U.S.
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 10XM design currently supplied in reload quantities to two U.S. BWR/4 units except the fuel column length is shorter by 4.76 inches for compatibility with a BWR/3 core height. The ATRIUM 10XM 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 Inc.
CotrlldDoum n
AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 1OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Rep~ort
.Pagqe 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 LHGR (linear heat generation rate) limit presented in Figure 2-1.
Controlled Docurnent AREVA Inc.
ANP-3324N P Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensincl Report Paqie 2-2 Table 2-1 Summary of Fuel Rod Design Evaluation Results Criteria1 Section*
Description Criteria JResult, Margint 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 3.2.5.2 Cladding stress
[
]
3.3 Fuel System Criteria (3.1.1)
Fatigue
[
]
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, 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).
t Margin is expressed as (limit - result)
- The cladding external oxidation limit is restricted to the reduced value of [
J pm.
§ All values except the oxidation margin are taken from calculations which conservatively assume that the oxidation limit of [
] pm will be reached at EOL. The oxidation margin is taken from standard runs and represents the actual expected margin for EOL oxidation.
Controled Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM lOXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensing Report Pacqe 2-3 Figure 2-1 LHGR Limit (Normal Operation)
Contfl~eid Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Report 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
] plenum spring on the upper end of
AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensina Report Paae 3-2 the fuel column I" 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).
ContrledD Doum ent.
AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Report 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 3.2.5.2 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. 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 in the U.S.
- 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 Hydriding 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.
Controlled Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensin~q Report Paaqe 3-4 3.2.2 Cladding 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 [
].
The evaluation is performed using RODEX4. The design criterion and methodology are described in Reference 2. RODEX4 takes into account the [
J. 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 Overheating 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 [
o~k~d La ~mern.
AREVA Inc.
AP-qe23-5 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensina Report Pa~?L5 I
I The heat conduction in the fuel and clad is ['
1.
Mechanical processes include ['
As part of the methodology, fuel rod power histories are generated ['
I
Controlled Document AREVA Inc.
ATRIUM I0XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensina Report APa-3324-P Paae 3-6 Since RODEX4 is a best-estimate code, uncertainties [
]. Uncertainties taken into account in the analysis are summarized as:
Power measurement and operational uncertainties - [
Manufacturing uncertainties - [
1, aModel uncertainties - [
I
CotrlldDoumn AREVA Inc.
ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensina Report ANP-3324NP Revision 1 Paale 3-7 C
,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. [
]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
[
Cotroiledi Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Report Paaqe 3-8 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.
The design limits are derived from the ASME (American Society of Mechanical Engineers)
Boiler and Pressure Vessel (B&PV) Code Section III (Reference 3) 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 Swelling 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
- 1. The CUE (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 r remain belo [
to satisfy the design criterion.
Controc~e Docu~ment AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Report Pagqe 3-9 Table 3-2 and Table 3-3 list the results for equilibrium and cycle-specific conditions, respectively.
3.2.7 Oxidation, Hydriding, 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 maximum calculated oxide on the fuel rod cladding shall not exceed [
1 Ipm. Previously, a [
i pm 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 for the first U.S. application to reduce the limit to [
] pmo. While not formally required, this reduced value has been conservatively adopted for Quad Cities analyses.
The current measurements of crud at Quad Cities Unit 2 indicate normal low crud levels.
However, in order to address the potential impact of changing water chemistry conditions, the
Controlled Diocument AREVA Inc.
ANP-3324NP Revision 1 ATRIUM I0XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensin~q Report Parae 3-10 input parameters have been conservatively selected in order to generate corrosion in excess of the current operating experience at Quad Cities. The values given in this report are the results from these conservative calculations, and as shown, meet all design criteria.
Currently, there is [
]. However, as mentioned above, the [
1 p~m was established, in part, as a means of [
1.
The oxide limit is evaluated such that greater than ['
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 methodology. The maximum rod pressure is calculated under steady-state conditions and also takes into account slow transients. Rod internal pressure is limited to ['
- 1. 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 (Fuel Assembly Handling)
The plenum spring must maintain a force against the fuel column to ['
- 1. 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" 1.
Contiot~d Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensing Rep~ort Pagqe 3-11 Table 3-1 Key Fuel Rod Design Parameters
[
I
ControIled Docurnent AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design I icn*.~ina Re~rnrt P~riam 5-12 Table 3-2 RODEX4 Fuel Rod Results for Equilibrium Cycle Conditions t
Margin is defined as (limit - result).
]
Controlled Document AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licen~ina Renort Paae 3-13
,I'1 *.yr.-
i-i..
Table 3-3 RODEX4 Fuel Rod Results for Quad Cities Unit 2 Cycle 24*
Margint to Limit Criteria Topic Limit Steadye
[
[
]
I=t t
Note that cycle-specific results are provided up to the end of cycle.
Margin is defined as (limit - result).
Note that Quad Cities Unit 2 intends to operate cycle 24 with a single rod sequence (SRS) that limits the number of control rods inserted in the core. For this cycle none of the ATRIUM-I10XM fuel assemblies reside in a SRS control cell. As such, control rod withdrawal error transients are not part of the scope of this calculation.
AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10XM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design LicensinQ Report Paae 3-14 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses Description, Stress Category Criteria*
Result
[
Cladding stress]
Pmn (primary membrane stress)
[
]
Pm + Pb (primary membrane + bending)
[]
P + Q(primary +secondary)
[]
Cladding-End Cap stress t1 In all analyzed conditions, the value of Su is less than twice the value of Sy,. As such, the Su criteria are more limiting than the Sy, criteria and only the stress result ratios to Su have been reported.
Con '.rolled Documnent AREVA Inc.
ANP-3324NP Revision 1 ATRIUM 10OXM Fuel Rod Thermal-Mechanical Design for Quad Cities Unit 2 Cycle 24 Representative Fuel Cycle Design Licensingq Report Paqe 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-Mechanical Fuel/Rod Methodology for Boiling Water Reactors, AREVA NP Inc., February 2008.
- 3.
ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components," 1983.
- 4.
O'Donnell, W.J., and B. F. Langer, "Fatigue Design Basis for Zircaloy Components,"
Nuclear Science and Engineering, Vol. 20, 1964.