advertisement

www.themegallery.com Company Logo بسم هللا الرحمن الرحيم ﴿ َقالُو ْا ُس ْب َحا َن َك لَ ِع ْل َم َل َنا إِلا َ ْ ْ َ اعلا ْم َت َنا إِ ان َك أنت ال َعلِي ُم ال َح ِِكي ُم َم َ صدق اهلل العظيم سورة البقرة اآلية )(۳۲ Thorium Fuel Performance In a Tight LWR Lattice Paper presented by • Sayed saeed abdelfattah 2012 Prof. Dr. Esmat A. Amin Prof. of reactor safety analysis ,Nuclear $Radiological , Regularity Authority,Cairo,Egypt Prof.Dr.Ibrahim I. Bashter Prof of nuclear physics ,physics department, faculty of Science,Zagazig university Sayed saeed abdelfattah Faculty of science,zagazig university Introduction Thorium fuel in PWR reactors Characteristics of thorium fuel:• Thorium is 3 to 4 more abundant than uranium ,widely distributed in nature as an easily exploitable resource in many countries . • Thorium oxide is chemically more stable and has higher radiation resistance than uranium oxide. • Thorium oxide has an extremely melting point which is about 3300 C Thorium fuel in PWR reactors • Th -232 is a fertile material like the isotope U-238 which can absorb neutron and become Th-233 which decays to Pa-233 by beta decay.Pa-233 decays by beta decay to U-233 which is a fissile material as U-235. • The absorption cross section for thermal neutrons of Th-232 equals 7.4 barn which is three times that of U-238 (2.7 barn). Thus Th-232 is a better fertile material than U-238 in thermal reactors. • The high thermal capture cross section of thorium will require more amount of initial fissile material as a compensatory where thorium will reduce neutron capture in moderator so, the neuron yield of fission probability of U-233 in thermal region is higher than that of U-235 and Pu-239 The aim of the work The aim of the work can be summarized in the following points • The aim of this work is studying the feasibility of thorium fuel in PWR reactors which takes up more 60 percent of the nuclear power plants in the world. • This can be performed by using one of the thorium fuels as thorium –plutonium oxide. • Pin cell model is carried out for (Th-Pu)OX and many parameters have been calculated such as k-infinity , fluxes, average energy per fission ,atom densities of fission products and actinides and the absorption and fission cross sections of the produced isotopes at different stages of burnup. Methods of calculation Methods of Calculations of pin cell model .The burnup calculation of this work is performed using MCNP5 and WIMSD-5 code .MCNP5 is computerized analysis tool used for designing the fuel rod of thorium-plutonium oxide .WIMSD-5 is a general lattice cell program which can determine k-infinity ,neutron flux, burnup of the fuel .wims5b library of 69 groups is used in these calculations. .the burnup calculations are performed at a constant specific power (37.7 W/gm HM) Parameters and dimensions of pin cell model of thoriumplutonium oxide Fuel pellet radius 0.47 cm Cladding radius 0.54 cm radius Water (moderator) 0.85 cm Temperature of fuel 1023 K Temperature of clad 923 K Temperature of water 583 K The initial atom densities of fuel isotopes (atoms/cm3) isotopes Zone 1 Th-232 2.11E+22 Pu-238 9.72E+18 Pu-239 5.99E+20 Pu-240 2.32E+20 Pu-241 7.69E+19 Pu-242 4.78E+19 Cr Zone 2 Zone 3 8.14E+19 Mn Fe 1.60E+20 Ni Zr 4.37E+22 C 2.68E+18 H 4.80E+22 1-K-inf versus burnup for pin cell of (Th-Pu)OX • k-infinity versus burnup is shown in the following graph (IAEA-TECDOC-1349) (k-inf by ORIGEN-2 code) The present work 1.15 1.10 1.05 k-inf 1.00 0.95 0.90 0.85 0.80 0 10 20 30 40 Burnup (GWd/ton) 50 60 **K-infinity values of the reference results and WIMSD5 results are tabulated as following:Burnup( GWD/ton) 0 30 40 60 k-infinity Ref ( TECDOC-1349) 1.12479 0.925198 0.887499 0.84756 k-inf ref by ORIREN-2code 1.112 0.889 0.851 0.822 The present work 1.127744 0.916267 0.873056 0.82698 k-infinity difference(Ref TECDOC1349 - our present work) -0.002954 0.008571 0.01444 0.02058 From the previous figure and results ,we can say that: **k-infinity for all cases decreases due to the depletion of fissile isotopes • and the production of fission products and poisons. The remarkable decrease in the beginning is due to the production of Xe-135 which has a high neutron capture cross section and it has impact on thermal utilization factor and thus multiplication factor .It is an important poison in the reactor operation. **The sharp decrease in the values of k-infinity at the end of burnup is • due to the production of fission products which have high absorption cross sections which are not taken in consideration in the results of the reference (IAEA-TECDO-1349) • **Our work represents the blue curve which is compared to the other curves of the reference results. The results obtained shows good agreement with the reference. this is indicated in the previous table. 2-fluxes of fuel,clad,moderator versus burnup • The following figure shows flux of fuel versus burnup (IAEA-TECDOC-1349) The present work flux n/cm2.sec 4.40E+014 4.20E+014 4.00E+014 3.80E+014 3.60E+014 3.40E+014 3.20E+014 3.00E+014 2.80E+014 0 10 20 30 40 burnup (GWD/ton) 50 60 **The total fluxes of fuel,clad,moderator as a function of burnup are recorded in the following table: Burnup(GWD/ton) 0 30 40 60 Fuel (Ref values) 2.913131E+14 3.5005688E+14 3.6624260E+14 3.8775357E+14 Fuel by WIMSD5 3.04E+14 3.82E+14 4.01E+14 4.250E+14 Clad(Ref values) 2.925591E+14 3.5062147E+14 3.6663592E+14 3.8785461E+14 Clad by WIMSD5 3.05E+14 3.83E+14 4.02E+14 4.2531E+14 Moderator(Ref values) 2.930024E+14 3.5120984E+14 3.6725799E+14 3.8851187E+14 Moderator by WIMSD5 3.06E+14 3.84E+14 4.03E+14 4.2537E+14 **from the previous figure and table it is obvious:- The total neutron flux of thorium-plutonium oxide increases with burnup because the macroscopic fission cross section decreases mainly due to the depletion of fissile nuclides this is a direct consequence of the constant linear power assumed. The reference flux results begins from 2.9E+14 to 3.8E+14 which are near to our WIMSD5 results which begin from 3.04E+14 to 4.2E+14 n/(cm2.sec) 3-The average energy per fission versus burnup • Energy per fission for (Th-pu)OX: energy / fission energy per fission( IAEA-TECDOC-1349 ) energy / fission (The present work) 212 Mev/fission 211 210 209 208 207 206 205 204 203 202 0 10 20 30 40 burnup (GWd/ton) 50 60 **The following table shows the energy per fission of reference results and that obtained by wimsd5 Burnup(GWD/ton) 0 30 40 60 Energy per fission (Mev/fission) Ref values 207.891 205.775 204.411 202.009 Energy per fission (MEV) by WIMSD5 code 211.439 207.312 205.433 202.34 As shown in the previous figure ,the average energy per fission decreases with burnup this is due to the change of the fissile nuclides where the smooth transition from plutonium fissioning to U-233 causes the decrease in the average energy per fission .this is due to the fact that the fissile plutonium isotopes release about 200 Mev thermal energy per fission and U-233 only release about 190 Mev thermal energy per fission. 4-the atom densities of fission products and actinides (atoms/barn x cm) **From the previous table we can see that there is a change in the concentration of fission products and actinides and this will be illustrated in the following notes: • 1-for plutonium isotopes,Pu-239 decreases with burnup, where its amount nearly has been burnt up at 60 MWD/kg HM. • Pu-241 increases and then decreases, this is due to that Pu-241 is produced via the capture of Pu-240 and on the other hand [Pu-241 is depleted due to its fission. 2-for protactinium and uranium isotopes, the fissile nuclide U-233 increases with burnup due to neutron capture in Th-232 and subsequent decay of Th-232 to Pa-233 then to U-233 which contributes more and more to the power. Pa-233 increases slightly with burnup due to increasing the total neutron flux. For U-234,U-235,U-232:U234 also increases with burnup due to neutron capture in U-233 and capture in Pa-233 and subsequent decay of Pa-234. For U-235 it increases slightly due to neutron capture in U-234. For U-232 it increases because it is formed by (n,2n) reaction For the concentration of major actinides,Am-243 and Cm-244 are the most abundant in the fuel.Cm-244 increases with burnup due to the neutron capture in Am-243 and subsequent decay of Am-244 Am-243 is produced by neutron capture of Pu-242 and subsequent decay of Pu243 5-the absorption and fission cross section of fission products and actinides conclusion **we can summarize the conclusion in the following points 1-MCNP5 and WIMSD5 are used for the calculation of a PWR fuel pin taken from 17X17 assembly in pressurized water reactor under specific conditions of its • operation. • 2-pin cell model is carried out for (Th-Pu) OX and many parameters are calculated and the obtained results are compared with the results announced in the reference. (Potential of thorium based fuel cycles to constrain plutonium and reduce long lived waste toxicity) and a good agreement is found between the two results. 3-the parameters calculated in this work are neutron multiplication factor, fluxes, • average energy per fission, fuel compositions, absorption and fission cross sections versus burnup which are important parameters in the reactor operation. • 4-for( Th-Pu) OX, Pu isotopes decreases with burnup .but uranium isotopes increases with burnup and actinides as Am-243, cm-244 increases with burnup this is explained previously. Thank you