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:ANP-2899NPRevision 0.wFuel Design Evaluat ion forATRIUM TM 1OXM BWR Reloa d FuelApril 2010AREVA.INP. Inc.EVA
AREVA NP Inc.ANP-2899NPRevision 0Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload Fuel
AREVA NP Inc.ANP-2899NPRevision 0Copyright 2010AREVA NP Inc.All Rights Reserved
ANP-2899NPRevision 0Page iFuel Design Evaluation forATRIUMTm 1OXM BWR Reload FuelNature of ChangesItemPaneDescription and Justification1.AllThis is the initial release.AREVA NP Inc.
ANP-2899NPRevision 0Page iiFuel Design Evaluation forATRIUMTM 10XM BWR Reload FuelContents1.0Introduction2.0Design Description of the ATRIUM 10XM .2-12.1Overview .2-12.2Fuel Assembly .2-22.2.1Spacer Grid .2-22.2.2W ater Channel .2-22.2.3Lower Tie Plate (LTP) .2-32.2.4Upper Tie Plate (UTP) and Connecting Hardware . 2-42.2.5Fuel Rods .2-42.3Fuel Channel and Components .2-53.0Fuel System Design Evaluation .3.1Objectives .3.2Fuel Rod Evaluation .3.2.1Internal Hydriding .3.2.2Cladding Collapse .3.2.3Overheating of Cladding .3.2.4Overheating of Fuel Pellets .3.2.5Stress and Strain Limits .3.2.6Cladding Rupture .3.2.7Fuel Rod Mechanical Fracturing .3.2.8Fuel Densification and Swelling .3.3Fuel System Evaluation .3.3.1Stress, Strain, or Loading Limits on AssemblyComponents .3.3.2Fatigue .3.3.3Fretting Wear .3.3.4Oxidation, Hydriding, and Crud Buildup .3.3.5Rod Bow .3.3.6Axial Irradiation Growth .3.3.7Rod Internal Pressure .3.3.8Assembly Lift-off .3.3.9Fuel Assembly Handling .3.3.10 Miscellaneous Component Criteria .3.4Fuel Coolability .3.4.1Cladding Embrittlement .3.4.2Violent Expulsion of Fuel .3.4.3Fuel Ballooning .3.4.4Structural Deformations .4.0.Thermal and Hydraulic Design Evaluation .4.1Thermal-Hydraulic Design Criteria .4.1.1Hydraulic Compatibility .4.1.2Thermal Margin Performance .4.1.3Fuel Centerline Temperature .4.1.4Rod Bow .4.1.5Bypass Flow .AREVA NP 14-14-24-34-44-44-4
Fuel Design Evaluation forATRIUM TM IOXM BWR Reload Fuel4.24.34.44.5ANP-2899NPRevision 0Page Iii4.1.6Stability .Loss-of-Coolant Accident (LOCA) Analysis .Control Rod Drop Accident (CRDA) Analysis .ASME Overpressurization Analysis .Seismic/LOCA Lift-off .4-54-54-64-64-75-15-25-25-35-35.0Nuclear Design Evaluation .5.1Power Distribution .5.2Kinetic Parameters .5.3Stability .5.4Control Rod Reactivity .6.0Testing, Inspection and Surveillance .6.1Design Verification Testing .6.1.1Mechanical Testing .6.1.2Thermal-Hydraulic Testing .6.2Operating Experience .6.3Summary of Lead Test Assembly (LTA) Programs .6.3.1Initial Alloy 718 Spacer Grid Test Program .6.3.2ATRIUM 1OXP and ATRIUM 1OXM .6.4Poolside Examination Results .6.4.1Performance of Alloy 718 Spacer Grid Material .6.4.2Performance of 0.4047-Inch Fuel Rod .6.4.3Fuel Assembly Growth .6.4.4Visual Appearance of Water Channel Crowns and OtherComponents .6.5Conclusions c Design Criteria Evaluation for ATRIUM 1OXM .7-18.0References .8-1AREVA NP Inc.6-256-27
ANP-2899NPRevision 0Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload FuelPage ivTablesTable 2.1ATRIUM 1OXM Key Design Parameters .Table 4.1BWR/4 Sample Problem Thermal-Hydraulic Design Conditions . 4-8Table 4.2BWRI4 Sample Problem Thermal-Hydraulic Results at RatedConditions (100%P / 100%F) for Transition to ATRIUM 1OXM Fuel . 4-9Table 4.3BWR/4 Sample Problem Thermal-Hydraulic Results at Off-RatedConditions (60%P /45%F) for Transition to ATRIUM 1OXM Fuel . 4-10Table 4.4BWR/4 Sample Problem First Transition Core Thermal-HydraulicResults at Rated Conditions (100%P/ 100%F) .4-11Table 4.5BWR/4 Sample Problem First Transition Core Thermal-HydraulicResults at Off-Rated Conditions (60%P / 45%F) .2-64-12Table 6.1Irradiation Experience for Materials Used in U.S. ATRIUM Designs . 6-6Table 6.2AREVA Experience with Advanced ATRIUM Designs .Table 6.3Comparison of Shadow Corrosion Obtained by Eddy CurrentMeasurem ents and Metallography .6-17G eneric Design C riteria .7-1Table 7.1AREVA NP Inc.6-7
ANP-2899NPRevision 0Fuel Design Evaluation for1OXM BWR Reload FuelATRIUM TMPage vFiguresFigure 2.1ATRIUM 1OXM Fuel Assembly (not to scale) .2-8Figure 2.2ATRIUM IOXM ULTRAFLOW Spacer Grid .2-9Figure 2.3ATRIUM 1OXM FUELGUARD LTP .2-10Figure 2.4ATRIUM 1OXM UTP and Locking Hardware .2-11Figure 2.5ATRIUM 1OXM Fuel Rods .2-12Figure 2.6ATRIUM 1OXM (or ATRIUM-10) Fuel Channel .2-13Figure 2.7ATRIUM 1OXM (or ATRIUM-1 0) Fuel Channel Fastener .2-14Figure 3.1Example LHGR Limit for the ATRIUM I0XM Design .3-14Figure 6.1ATRIUM 10XP/XM Surveillance Program .6-8Figure 6.2Visual Aspect of Alloy 718 Spacer Grids .6-11Figure 6.3Shadow Corrosion in Vicinity of Alloy 718 Spacer Grid .6-12Figure 6.4Typical Appearance of Spacer Spring Contact Points (4 cycles) . 6-12Figure 6.5Measurement Traces for Shadow Corrosion Evaluation .Figure 6.6Shadow Corrosion Database of ATRIUM Fuel Rods (lift-offmeasurement including the enhancement by shadow corrosion) . 6-14Figure 6.7Circumferential Shadow Corrosion (at spacer elevation) .Figure 6.8Comparison of Hydride Distribution between Mid-Span and at SpacerP ositio n . 6-16Figure 6.9Evaluation of Shadow Corrosion on Two-Cycle ATRIUM 10XP Fuel Rod . 6-17Figure 6.10Spacer-Induced Shadow Corrosion Inside Fuel: Channels .6-18Figure 6.11Corrosion Lift-Off Database .6-20Figure 6.12Uniform Corrosion (obtained by metallography) .6-21Figure 6.13Hydrogen Concentration Database .6-22Figure 6.14Impact of Fuel Rod Diameter on Fuel Rod Growth .6-22Figure 6.15Fuel Rod Diameter Change .6-23Figure 6.16Fission Gas Release Database .6-24Figure 6.17UTP Inspection .6-25Figure 6.18Bottom Nozzle and Seal Spring Inspection .6-26Figure 6.19Water Channel Crown Inspection .6-27This document contains a total of 88 pages.AREVA NP Inc.6-136-15
ANP-2899NPRevision 01Page viFuel Design Evaluation forATRIUMTm 10XM BWR Reload FuelNomenclatureAOOASMEanticipated operational occurrencesAmerican Society of Mechanical EngineersB&PVBOLBtuBWRBoiler and Pressure Vesselbeginning of lifeBritish thermal unitsboiling water reactorcal/gmCFRCHFCPRCRDACRWECUFcalories per gramCode of Federal Regulationscritical heat fluxcritical power ratiocontrol rod drop accidentcontrol rod withdrawal errorcumulative usage factorECCSEFPDEOLemergency core cooling systemeffective full power daysend of lifeFAFCFDLfuel assemblyfuel channelfuel design limitGdGWd/MTUgadoliniagigawatt days per metric ton of initial uraniumIDinner diameterklbmpounds mass*1000IbmLHGRLOCALTA (LUA)LTPLTPpounds masslinear heat generation rateloss- of- coolant accidentlead test (use) assemblylower tie platelow temperature m average planar linear heat generation rateminimum critical power ratiomillion pounds mass per hourmixed oxidemoderator temperature coefficientMegawatt-days per kilogram of Uraniummetal-water reactionMega-Watt thermalAREVA NP Inc.
Fuel Design Evaluation forATRIUMM 1OXM BWR Reload FuelNRCU.S. Nuclear Regulatory CommissionODOLMCPRouter diameteroperating limit minimum critical power cladding interactionpeak cladding temperaturepercent flowpercent powerPortable Hydraulic Test Facilitypost-irradiation examinationpart-length fuel rodsparts per millionpounds per square inchpounds per square inch absolutepounds per square inch differenceQCquality controlRPFRXAradial peaking factorfully recrystallized annealedSLMCPRSRASRPSSTsafety limit minimum critical power ratiostress relieved annealedStandard Review Planstainless steelU0 2uranium dioxideupper tolerance limitupper tie plateUTLUTPAREVA NP Inc.ANP-2899NPRevision 0Page vii
Fuel Design Evaluation forATRIUMTM 10XM BWR Reload Fuel1.0ANP-2899NPRevision 0Page 1-1INTRODUCTIONThis report provides the results of evaluations performed for the ATRIUMtM* 1OXM fuel designto demonstrate compliance with U. S. Nuclear Regulatory Commission (NRC) approved fuellicensing criteria defined in Reference 1. With Reference 1, the NRC approved a set of genericacceptance criteria to be satisfied by AREVA NP Inc. (AREVA) for new boiling water reactor(BWR) fuel designs. In accordance with the process described in Reference 1, new fueldesigns or fuel design changes satisfying the generic acceptance criteria do not require explicitstaff review. Satisfaction of the acceptance criteria is sufficient for approval by reference to theacceptance criteria.Reference 1 (as clarified by References 2 and 3) requires that AREVA NP provide the NRC withan informational summary of the evaluation of the design against the NRC-approvedacceptance criteria for a generic evaluation that is independent of plant and cycle.Documentation of analyses performed to demonstrate compliance with any criterion for aspecific plant and/or cycle is provided to the Licensee as part of the normal reload licensingdocument package. AREVA's standard practice is to demonstrate compliance to the NRCapproved acceptance criteria on a plant- and cycle-specific basis.The fuel design licensing process described in References 1 through 3 is the same processused to introduce and license AREVA's current ATRIUM-10 design and the previous ATRIUM-9design. The evaluation results for the ATRIUM-10 and ATRIUM-9 designs were provided to theNRC in Reference 4 for information.This report contains a detailed description of the ATRIUM 1OXM design (Section 2.0); theATRIUM 1OXM evaluation results (Sections 3.0, 4.0, 5.0, and 7.0); a summary of the operatingexperience and the post-irradiation examination (PIE) results supporting the ATRIUM IOXMdesign features (Section 6.0); and a listing of the NRC-approved methods applicable to theATRIUM IOXM design (Section 8.0).ATRIUM is a trademark of AREVA NP Inc.AREVA NP Inc.
Fuel Design Evaluation forATRIUM TM 1OXM BWR Reload Fuel2.02.1ANP-2899NPRevision 0Page 2-1DESIGN DESCRIPTION OF THE ATRIUM 1OXMOverviewA summary of the ATRIUM 1OXM mechanical design that will be used in reload applications isgiven in this section. The ATRIUM 1OXM design described herein shares many of the sameproven design features of AREVA's ATRIUM-10 and ATRIUM-9 fuel designs that are in broaduse in BWR plants. All materials used in the ATRIUM 1OXM design have significant irradiationexperience in the ATRIUM-10 and ATRIUM-9 designs.In general, the design changes introduced with the ATRIUM 1OXM design are evolutionary innature and represent a []. The ATRIUM 1OXM fuel bundle shares the same basic geometry asthe current ATRIUM-10 fuel assembly design. This geometry consists of a 10xl fuel latticewith a square internal water channel that displaces a 3x3 array of rods. Relative to theATRIUM-10 fuel, the ATRIUM 1OXM incorporates the following key design features:[].Table 2.1 lists the key design parameters of the ATRIUM I0XM fuel assembly and comparesthem to the current ATRIUM-10 design.tULTRAFLOW is a trademark of AREVA NP Inc.AREVA NP Inc.
Fuel Design Evaluation forATRIUM TM 1OXM BWR Reload Fuel2.2ANP-2899NPRevision 0Page 2-2Fuel AssemblyThe ATRIUM IOXM fuel assembly consists of a lower tie plate (LTP) and upper tie plate (UTP),91 fuel rods, 9 spacer grids, a central water channel with [], andmiscellaneous assembly hardware. Of the 91 fuel rods, 12 are PLFRs. The structural membersof the fuel assembly include the tie plates, spacer grids, water channel, and connectinghardware. The structural connection between the LTP and UTP is provided by the central waterchannel. The lowest of the nine spacer grids is located just above the LTP to restrain the lowerends of the fuel rods.The fuel assembly is accompanied by a fuel channel, as described later in this section.Table 2.1 lists the main fuel assembly attributes, and an illustration of the fuel bundle assemblyis provided in Figure 2.1.2.2.1Spacer Grid[,Table 2.1 lists the main spacer grid attributes, and an Illustration of the spacer grid is provided inFigure 2.2.2.2.2[AREVA NP Inc.Water Channel
Fuel Design Evaluation forATRIUMTm 1OXM BWR Reload FuelTable 2.1 lists the main water channel attributes and are illustrated in Figure 2.1.2.2.3Lower Tie Plate (LTP)[FUELGUARD is a trademark of AREVA NP Inc.AREVA NP Inc.ANP-2899NPRevision 0Page 2-3
Fuel Design Evaluation for1OXM BWR Reload FuelATRIUMTMANP-2899NPRevision 0Page 2-4Table 2.1 lists the main LTP attributes, and Figure 2.3 provides an illustration of the LTP.2.2.4Upper Tie Plate (UTP) and Connectinq Hardware[]1.Table 2.1 lists the main UTP attributes, and Figure 2.4 provides an illustration of the UTP andlocking components.2.2.5Fuel Rods[AREVA NP Inc.
ANP-2899NPRevision 0Page 2-5Fuel Design Evaluation forATRIUMTm 1OXM BWR Reload Fuel.Table 2.1 lists the main fuel rod attributes, and Figure 2.5 provides an illustration of the fuel rodcomponents.2.3Fuel Channel and Components[I.The fuel channel and fuel channel fastener are depicted in Figure 2.6 and Figure 2.7,respectively.AREVA NP Inc.
ANP-2899NPRevision 0Page 2-6Fuel Design Evaluation forATRIUMTm IOXM BWR Reload FuelTable 2.1ATRIUM 1OXM Key Design Parameters[11F-44AREVA NP Inc.I
ANP-2899NPRevision 0Page 2-7Fuel Design Evaluation forATRIUMTm 10XM BWR Reload FuelTable 2.1ATRIUM IOXM Key Design Parameters (Continued)IAREVA NP Inc.
Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload FuelANP-2899NPRevision 0Page 2-8[IFigure 2.1AREVA NP Inc.ATRIUM IOXM Fuel Assembly (not to-scale)
ANP-2899NPRevision 0Page 2-9Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload FuelIIFigure 2.2AREVA NP Inc.ATRIUM 1OXM ULTRAFLOW Spacer Grid
ANP-2899NPRevision 0Page 2-10Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload FuelIIFigure 2.3AREVA NP Inc.ATRIUM IOXM FUELGUARD LTP
Fuel Design Evaluation forATRIUMTM IOXM BWR Reload FuelANP-2899NPRevision 0Page 2-11IIFigure 2.4AREVA NP Inc.ATRIUM IOXM UTP and Locking Hardware
ANP-2899NPRevision 0Page 2-12Fuel Design Evaluation forATRIUMTm 10XM BWR Reload Fuel[IFigure 2.5AREVA NP Inc.ATRIUM IOXM Fuel Rods
Fuel Design Evaluation forATRIUMTM I0XM BWR Reload FuelANP-2899NPRevision 0Page 2-13FsjmOO905Figure 2.6AREVA NP Inc.ATRIUM 1OXM (or ATRIUM-10) Fuel Channel
Fuel Design Evaluation forATRIUMTm 1OXM BWR Reload FuelANP-2899NPRevision 0Page 2-14IIFigure 2.7AREVA NP Inc.ATRIUM 1OXM (or ATRIUM-10) Fuel Channel Fastener
Fuel Design Evaluation forATRIUMTM IOXM BWR Reload FuelANP-2899NPRevision 0Page 3-1FUEL SYSTEM DESIGN EVALUATION3.03.1ObjectivesThe objectives of building fuel assemblies (systems) to specific design criteria are to provideassurance that: The fuel assembly (system) shall not fail as a result of normal operation and anticipatedoperational occurrences (AOOs). The fuel assembly (system) dimensions shall bedesigned to remain within operational tolerances, and the functional capabilities of thefuels shall be established to either meet or exceed those assumed in the safety analysis. Fuel assembly (system) damage shall never prevent control rod insertion when it isrequired. The number of fuel rod failures shall be conservatively estimated for postulatedaccidents. Fuel coolability shall always be maintained. The mechanical design of fuel assemblies shall be compatible with co-resident fuel andthe reactor core internals." Fuel assemblies shall be designed to withstand the loads from In-plant handling andshipping.The first four objectives are those cited in the Standard Review Plan (SRP). The latter twoobjectives are to assure the structural integrity of the fuel and the compatibility with the existingreload fuel. To satisfy these objectives, the criteria are applied to the fuel rod and the fuelassembly (system) designs. Specific component criteria are also necessary to assurecompliance. The criteria established to meet these objectives Include those given InChapter 4.2 of the SRP.3.2Fuel Rod EvaluationThe detailed fuel rod design evaluation entails such parameters as pellet diameter and density,cladding-pellet diametral gap, fission gas plenum size, and rod pre-pressurization level. Thedesign evaluation also considers effects and physical properties of fuel rod components, whichvary with bumup. The integrity of the fuel rods is ensured by designing to prevent excessivefuel temperatures, excessive rod internal pressures, and excessive cladding stresses andstrains. This end is achieved by designing the fuel rods to satisfy the design criteria duringnormal operation and AQOs over the fuel lifetime. Summaries of the methods and codes usedAREVA NP Inc.
ANP-2899NPRevision 0Page 3-2Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload Fuelin the evaluation, along with the design criteria, are provided in the following sections. Details ofthe criteria and evaluation methodology can be found by consulting the referenced documents.3.2.1Internal HydridinqThe absorption of hydrogen by the cladding can result in cladding failure due to reduced ductilityand formation of hydride platelets. Careful moisture control during fuel fabrication reduces thepotential for hydrogen absorption on the inside of the cladding. The fabrication limit EJ is verifiedby qualitycontrol (QC) inspection during fuel manufacturing.3.2.2Claddinq CollapseCreep collapse of the cladding and the subsequent potential for fuel failure is avoided in theAREVA fuel system design by limiting the gap formation due to fuel densification subsequent topellet-clad contact. The size of the axial gaps that may form due to densification following thefirst pellet-clad contact shall be less than [].The evaluation Is performed using RODEX4. The design criterion and methodology aredescribed in Reference 5. RODEX4 takes into account []. A brief overview of RODEX4 and the statistical methodology isprovided in Section 3.2.4.3.2.3Overheatinq of CladdingThe design basis to preclude fuel rod cladding overheating is that 99.9% of the fuel rods shallnot experience transition boiling. Prevention of potential fuel failure from overheating of thecladding is accomplished by minimizing the probability of exceeding thermal margin limits onlimiting fuel rods during normal operation and Aoos. Compliance with this criterion Is confirmedas part of the plant- and cycle-specific reload thermal-hydraulics analysis. An experimentallybased, ATRIUM 1OXM design-specific CHF correlation, which has been accepted by the NRC,is used In this evaluation (see Section 4.1.2).AREVA NP Inc.
ANP-2899NPRevision 0Page 3-3Fuel Design Evaluation forATRIUMTM IOXM BWR Reload Fuel3.2.4Overheating of Fuel PelletsFuel failure from the overheating of the fuel pellets is not allowed. The centerline temperature ofthe fuel pellets must remain below melting during normal operation and AQOs. The meltingpoint of the fuel includes adjustments for gadolinia (Gd) content. AREVA establishes the linearheat generation rate (LHGR) limit for each fuel system, which protects against fuel centerlinemelting during steady-state operation and during AQOs.Fuel centerline temperature is evaluated using the RODEX4 (Reference 5) for both normaloperating 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 fill gas; andreleased fission gases. The fuel rod is divided into axial and radial regions, with conditionscomputed for each region. The operational conditions are controlled by [T.The heat conduction in the fuel and clad is [.Mechanical processes include [AREVA NP Inc.
ANP-2899NPRevision 0Page 3-4Fuel Design Evaluation forATRIUMM 1OXM BWR Reload FuelAs part of the methodology, fuel rod power histories are generated [1.Since RODEX4 is a best-estimate code, uncertainties []. Uncertainties taken into account in theanalysis are summarized as:0Power measurement and operational uncertainties: [.0Manufacturing uncertainties: [I.AREVA NP Inc.
Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload Fuel ANP-2899NPRevision 0Page 3-5Model uncertainties: [1.3.2.53.2.5.1Stress and Strain LimitsPellet/Cladding InteractionCladding strain caused by transient-induced deformations of the cladding is calculated using theRODEX4 code and methodology, as described in Reference 5. See Section 3.2.4 for anAREVA NP Inc.
ANP-2899NPRevision 0Page 3-6Fuel Design Evaluation forATRIUMTm IOXM BWR Reload Fueloverview of the code and method. [.3.2.5.2Cladding StressCladding stresses are calculated using solid mechanics elasticity solutions and finite elementmethods. The stresses are conservatively calculated for the individual loadings and arecategorized as tresses are calculated at the cladding outer and inner diameter (ID) in the three principaldirections for both beginning of life (BOL) and end of life (EOL) conditions. At EOL, the stressesdue to mechanical bow and contact stress are decreased due to irradiation relaxation. Theseparate stress components are then combined, and the stress intensities for each category arecompared to their respective limits.The end cap weld stresses are evaluated for loadings from differential pressure, differentialthermal expansion, rod weight, and plenum spring force.The design limits are based on the American Society of Mechanical Engineers (ASME) Boilerand Pressure Vessel (B&PV) Code Section III and the minimum specified material properties.3.2.6Cladding RuptureAccording to Code of Federal Regulations 10 CFR 50 Appendix K, the cladding rupture mustnot be underestimated when analyzing a loss-of-coolant accident (LOCA). NRC-approvedAREVA NP Inc.
Fuel Design Evaluation forATRIUMTM 1OXM BWR Reload FuelANP-2899NPRevision 0Page 3-7cladding ballooning and rupture models are used by AREVA in the evaluation of claddingrupture. The specific models are those presented in NUREG-0630. There is no explicit limit onthe deformation. However, the calculations with the deformation models must satisfy the eventcriteria given in 10 CFR 50.46. This analysis is performed as part of the reload licensing and isevaluated for each plant reload on a cycle-specific basis. (See Section 4.2.)3.2.7Fuel Rod Mechanical FracturingA mechanical fracture refers to a defect in a fuel rod caused by an externally applied force, suchas loads due to earthquakes and postulated pipe breaks. These externally applied forcestherefore Include hydraulic loads and loads derived from core-plate motion. See Section 3.4.4for a discussion of this accident evaluation.3.2.8Fuel Densification and SwellingFuel densification and swelling are limited by the design criteria for fuel temperature, claddingstrain, cladding collapse, and internal rod pressure criteria. Although there are no explicitcriteria for fuel densification and swelling, the effect of these phenomena are included in theRODEX4 fuel rod performance code.3.3Fuel System EvaluationThe detailed fuel system design evaluation is performed to ensure the structural integrity of thedesign under normal operation, AOO, faulted conditions, handling operations, and shipping.The analysis methods are based on fundamental mechanical engineering techniques--oftenemploying finite element analysis, prototype testing, and correlations based on In-reactorperformance data. Summaries of the ma
given in this section. The ATRIUM 1OXM design described herein shares many of the same proven design features of AREVA's ATRIUM-10 and ATRIUM-9 fuel designs that are in broad use in BWR plants. All materials used in the ATRIUM 1OXM design have significant irradiation experience in the ATRIUM-10 and ATRIUM-9 designs.
Bruksanvisning för bilstereo . Bruksanvisning for bilstereo . Instrukcja obsługi samochodowego odtwarzacza stereo . Operating Instructions for Car Stereo . 610-104 . SV . Bruksanvisning i original
10 tips och tricks för att lyckas med ert sap-projekt 20 SAPSANYTT 2/2015 De flesta projektledare känner säkert till Cobb’s paradox. Martin Cobb verkade som CIO för sekretariatet för Treasury Board of Canada 1995 då han ställde frågan
service i Norge och Finland drivs inom ramen för ett enskilt företag (NRK. 1 och Yleisradio), fin ns det i Sverige tre: Ett för tv (Sveriges Television , SVT ), ett för radio (Sveriges Radio , SR ) och ett för utbildnings program (Sveriges Utbildningsradio, UR, vilket till följd av sin begränsade storlek inte återfinns bland de 25 största
Hotell För hotell anges de tre klasserna A/B, C och D. Det betyder att den "normala" standarden C är acceptabel men att motiven för en högre standard är starka. Ljudklass C motsvarar de tidigare normkraven för hotell, ljudklass A/B motsvarar kraven för moderna hotell med hög standard och ljudklass D kan användas vid
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Fuel transfer pump (35) is mounted on the back of unit injector hydraulic pump (1). The fuel transfer pump pushes pressurized fuel out of the outlet port and the fuel transfer pump draws new fuel into the inlet port. Fuel is drawn from fuel tank (12) and flows through two micron fuel filter (11) . Fuel flows from fuel filter (11) to the inlet .
5-6 FUEL SYSTEM AND THROTTLE BODY FUEL TANK LIFT-UP Remove the front seat. ( 7-4) Remove the fuel tank mounting bolts. Lift and support the fuel tank with its prop stay. FUEL TANK REMOVAL Lift and support the fuel tank with its prop stay. (L7above) Disconnect the fuel pump lead wire coupler 10. Pla
Fuel Pressure: Fuel Pressure Regulator and System Pressure. Fuel System: Pumps, Relays . significant volume of fuel may come out. Be ready to catch all the gas in the filter . The 3 main things to check in the fuel circuit are the fuel pump relay, and the 2 fuel pumps. 1. Fuel Injection Relay Test
