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N.A. Dollezhal Research and development institute of power engineeringRESEARCH AND DEVELOPMENT FORDEMONSTRATION OF FUEL PERFORMANCEIN THE BREST-OD-300 CORES.Yu. Bulkin, V.V. Lemekhov, A.G. Sila-Novitsky, V.S. Smirnov,A.A. Umansky, A.S. Firsov(OJSC NIKIET, Moscow, RF)IAEA TECHNICAL MEETINGDESIGN, MANUFACTURE AND IN-PILE BEHAVIOUR OF FAST REACTOR FUELInstitute of Physics and Power Engineering (IPPE)Russia, Obninsk, May 30 – June 3, 2011


BREST-OD-300Control rodstubeКорпусReactorVesselMain coolantpumpSteamgenerator2Core


In 2010 R&D activities included:• neutronic and thermal-hydraulic analysis of the core withtwo regions for radial power density profiling;• design studies of shrouded hexagonal fuel assemblieswith gas-bonded fuel rods made of high-density (U-Pu)N fueland fitted up with wire or rib spacers;• preparation of documents for an upgraded experimentaldismountable fuel assembly (ETVS-M) of the BOR-60 reactor,which allows a match between the performancecharacteristics of irradiated (U-Pu)N fuel and the testtemperatures;• improvements in the design of the independent leadcooledloop-channel, in which 3 mockup fuel elements ofBREST-OD-300 will be tested.3


In 2011, it is planned:• to develop the BREST-OD-300 core materials, withenergy-grade Pu dioxide from irradiated fuel of VVERs to beused as feed material;• to analyse the startup of a large lead-cooled reactor (≥ 1GWe) on uranium nitride fuel of moderate enrichment (with235U≤15%) and the conditions of changeover to (U-Pu-MA)Nto be regenerated from the reactor’s own irradiated fuel;• to fabricate mockup fuel assemblies and their dummycomponents;• to incorporate (U-Pu)N fuel rods into the ETVS-M1 fuelassembly for in-pile irradiation.4


One of the design objectives studied was the possibilityof radial flattening of power and coolant temperature gaindistributions by radial profiling of fuel loads between twocore regions with fuel rods of the same standard size. Thecentral part of the core accommodates not onlyconventional fuel assemblies but also assembliescomprising CPS components, such as automatic controlrods (ACR), shim rods (SR) and emergency protectionrods (EPR). The core periphery is made up of all-fuelassemblies and is surrounded by rows of replaceable leadreflector blocks.The fuel considered for the first core was uraniumplutoniumnitride.Calculations carried out by means of the Monte Carlomethod, using 3-D detailed FA and core models.5


For the case of reactor fuel based on low-backgroundplutonium coming, e.g., from BN-600 blankets, and partialrefueling unprovided for studies were carried out to assessthe feasibility of profiling power density distribution by usingdifferent Pu content in the fuel of the core centre and itsperiphery.In a U-Pu nitride core with high-background plutoniumand with operation in a closed cycle in the mode of partialrefueling, it will take only two campaigns for the reactor toreach a state where it can run on its own fuel of equilibriumcomposition. In this case power distribution can be flattenedby using fuel of the same plutonium content but of differentdensity in the core centre and at the periphery.6


A drawback of power distribution flattening by differentPu contents in the core centre and at the periphery lies in FApower redistribution during the campaign. At the beginning,power maximum is found in the peripheral fuel assemblieswith high Pu content, but will later drift to the core centre,where the CBR and the Pu breeding rate are greater thanthose at the periphery.In the case of power distribution flattening by using fuelof different density, appropriate choice of fuel density andplutonium percentage allows not only full reproduction offissionable nuclides, while retaining the reactivity margin, butalso flat power distribution constant over lifetime.7


Method of powerdistribution flatteningFuelRadial profiling of fuel loads between two core regionBy different of plutoniumcontent(U-Pu)N — with lowbackgroundplutoniumThe beginning (B) andend (E) max. values of K eff 1,019(B)/1,016(E)The beginning (B) and end(E) maximum valuesrelative power of FA1,22 (B) in periphery core,1,21 (E) in central core.The power peak wave willmove from the peripheryto the centre crossingRefueling type Complete PartialBy different of fuel density(U-Pu-MA)N — with highbackgroundplutonium/energy grade Pu1,020(B)/1,020(E)1,22(B)≈1,22(E) in centralcore,1,17(B)≈1,17(E) inperiphery;flat power distributionconstant over lifetime8


Figures 1 and 2 present distributions of maximumtemperatures (with regard to overheating factors) as shown bythe outer cladding surfaces and the coolant at the outlet of the700 MW (U-Pu)N core.The permissible reactor power was assumed to be limitedby the maximum values of the following operating parameters:650 С for the cladding temperature, 1500 С for the fueltemperature, and 2 m/s for the coolant flow rate. Coolanttemperature at the core inlet was taken to be 420 С, while itsaverage temperature at the core and reflector was set at540 С.9


Fig. 1. Distribution of outer cladding temperatures in the case of (U-Pu)N fuel withlow-background plutonium10


Fig. 2. Temperature distribution in the case of (U-Pu-MA)N fuel based on energygradePu (end of lifetime)11


Characteristics of fuel optionsCharacteristicShroudlesstetrahedralFAShrouded hexahedral FAFuel composition (U-Pu-MA)N (U-Pu-MA)N (U-Pu)N UN (U-Pu)O 2Thermal power, MW 700 700 700 600 500Core diameter, mm 2300 2650 2650 2650 2650Core height, mm 1250 1100 1100 1100 1100Fuel load mass, t 19.0 25.5 27.1 32.6 25.9(Pu+MA+U) mass, t 2.60 3.62 2.74 3.95 2.77239+241Pu mass, t 1.75 2.43 2.60 --- 2.63Lifetime, eff.days 1500/1800 1500/1800 1500/1800 1500/1800Average burnup6.0/7.2 4.6/5.5 4.3/5.2 3.1/3.7(maximum), %(11.0) (8.3) (7.8) (5.6)1500/18003.2/3.8(5.7)12


Upgraded experimental dismountable fuel assembly (ETVS-M)In the course of irradiation ETVS provides:• Supervise the orientation of loaded rods;• maintain a maximum linear power of fuel elements in thelayout range 35 ÷ 38 kW / m, the maximum temperatureof the shell - 610±15 C.Rather stringent requirements on the stability of irradiationparameters (linear power and temperature of the shell)and the dependence of the reactor power on time ofyear (winter - 55 MW, summer - 50 MW) and reducingpower ETVS with fuel burnup necessitates periodicoverload assembly in the cell core with more suitableparameters.13


View of assembly ETVS (RIAR, old)Fuel rods14


Independent lead-cooled loop-channel (RIAR, old)15


Fast reactors, which would open the way for large-scalenuclear power development unconstrained by available fuelresources, can be started both on Pu separated from irradiatedfuel of thermal reactors and on Pu mixed with enriched U, andeven on enriched U alone with gradual conversion to U-Pu fuelof equilibrium composition in the process of U-235 burning andPu-239 production. Moreover, considering consumption ofnatural U and its separation work, the option of using enrichedU for the first core of fast reactors is much (to ~5 times) moreprofitable than use of Pu from irradiated fuel of thermalreactors.16


Using reactor BREST-1000, as an example, considerationwas given to the possibility of starting it on enriched U nitride,with gradual changeover to (U+Pu)N fuel. Choice of the fuelwith ~12 % enrichment in 235 U and geometry of the first corewas dictated by the safety requirement of a small reactivitymargin, commensurable with β eff, and just as small reactivityvariations in reactor operation during refueling intervals.17


The feasibility of converting BREST-1000 from the startingUN charge to operation on (U-Pu)N fuel may be illustrated by avery simple and clear scheme in which the reactor operateswithout intermediate refueling for a period of 5 times fuellifetime (1500 eff.days). At the end of each campaign and aftercooling in the core, all fuel is retrieved and subjected toregeneration which amounts basically to removal of fissionproducts (FP). The resulting fuel mixture, with its masssomewhat reduced, will have an appropriate amount ofdepleted U added to it and will be refabricated into a newcharge.18


Fig. 3 shows variations in the mass of 235 U, 239 Pu and allisotopes of Pu over 3 campaigns without adjusting fuel loadsduring refueling intervals. Fig. 4 depicts reactivity variationsover 3 five-year periods under the assumption that, duringrefueling and regeneration, fission products were removedfrom the irradiated fuel and replaced with depleted U, but theloaded fuel mass was not adjusted to reduce the reactivitymargin.19


Fig.3 – Variations in isotope composition.Fig. 4 – Reactivity variations.20


Changes in the fuel characteristics during transition for UN to (U+Pu)N fuelOperationtime, eff.daysMass of loaded/unloaded nuclides, t∑U235U239+241Pu ∑Pu235U+PuPu____( 235 U+Pu)K effBOL/EOL0 – 1500 77.6/71.3 9.30/6.01 0/2.57 0/2.69 9.30/8.70 0 / 0.31 1.014/1.0121500 – 3000 68.9/63.3 5.54/3.50 2.35/4.02 2.47/4.44 8.01/7.94 0.31/0.56 1.012/1.0093000 – 4500* 63.3/58.3 3.34/2.08 3.81/4.88 4.20/5.62 7.54/7.69 0.56/0.73 1.014/1.0114500 – 6000** 55.4/50.8 1.83/1.10 4.29/4.95 4.94/5.98 6.77/7.08 0.73/0.84 1.012/1.009* Fuel density reduced from 13.2 to 12.8 g/cm 3 .** Core height is 12% smaller and fuel density increased to 13.0 g/cm 3 .These changes reflect gradual replacement of 235 U with the Puconverted from 238 U. Though 235 U burnup is not fully compensated by239Pu accumulation, the higher physical worth of 239 Pu as compared to235U stabilises the reactivity variations.21


The State Atomic Energy Corporation ROSATOMAddress: 119017 Moscow, Bolshaya Ordynka Str., 24Phone: +7 499 949-4535Fax: +7 499 949-4679E-mail: info@rosatom.ruWebsite : www.rosatom.ruN.A. Dollezhal Research and development institute of power engineeringAddress: 101000 Moscow, PO Box 788Phone: +7 499 263-7388Fax: +7 499 788-2052E-mail: nikiet@nikiet.ruWebsite : www.nikiet.ru

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