[en] THis lecture deals with some basic subjects concerning reactor theory namely: - neutron multiplication, - material and geometrical buckling, - reflected reactors and reflector saving, and - reactor control including fuel depletion, fission products and temperature coefficients

[en] This lecture deals with the in core fuel management schemes beginning by batch loading, zone refuelling, scatter loading and ending by the checker board loading. The fuel grouping and region classification are discussed in detail. Also, a detail description of the shuffling algorithm is given. 5 fig

[en] The control of nuclear reactors, may be studied using several control methods, such as control by rod absorbers, by inserting or removing fuel rods (moderator cavities), or by changing reflector thickness. Every method has its advantage, the comparison between these different methods and their effect on the reactivity of a reactor is the purpose of this work. A computer program is written by the authors to calculate the critical radius and worth in any case of the three precedent methods of control

[en] The 'Nile code', for nuclear reactor cell and core parameters, burn-up and fission products concentrations calculations, is written as a combination of the 'EGCODE 2' are 'FLUPFISH 3' codes. The 'Nile code' is a general, code that begins with condensation of nuclear data into 76 group neutron cross sections, which are used as data for solving the integral transport equation inside the nuclear reactor lattice cell. This cods gives cell parameters as well as an averaged two group cross sections that are used for the solution of the diffusion equations inside the reactor core. It also gives reactor core parameters as well as an averaged one group cross section for burn-up calculations and accumulation of the fuel isotopes produced and fission products. This makes the 'Nile code' one of the international general codes for nuclear reactor design. The present code is applied to the Chernobyl reactor where cell and reactor parameters are calculated. The calculations of the fission products activity after the Chernobyl accident show that the activity after one year will drop to 5.5 % of its initial value (just after the Chernobyl accidents), 3.5 % after two years and 2.5 % after three years. This study is very important for the reactor safety analysis and can be applied to any other type of reactors using the 'Nile code'. The 'Nile code' results are also useful for the comparison with experimental non-destructive results of the isotopic composition of the fuel and fission product concentrations. 11 fig.,1 tab

[en] In this paper, a computer program solving the two group two dimensional diffusion equation is applied for a fast breeder reactor (∼ 250 MWe). The reactor size is divided into four zones. The first zone is the central zone with 19.2% PuO₂ to UO₂-PuO₂ fuel. The second zone is the radial one with PuO₂ percentage to fuel of the order of 27.1%. The third zone is the radial blanket zone with depleted uranium fuel. The fourth zone is similar in fuel as the third one but it corresponds to the axial blanket. A burn-up point model program is linked to the flux calculation program so as to calculate the fuel depletion as well as the Pu(Pu-239, Pu-240, Pu-241) accumulation inside each of the four zones of the fast breeder reactor. The flux calculations at the beginning of the reactor operation as well as after each step time are made and the results are drawn. The values of the four factors as well as the infinite multiplication factor are calculated at each time step and compared with French results. A comparison is also made between the results for this fast breeder reactor and a pressurized water reactor of approximately equal power

[en] This paper describes how the fictitious parameters of black bodies can be used to apply the continuity of thermal fluxes and currents at the surface of control rod. This leads to an easier method for describing the criticality condition of cylindrical core containing array of cylindrical control rods differ in material, location and size. Some special configurations for rods through bare or reflected core are tested to calculate the critical core radius, rod worth and flux distribution. Parameters of the ET-RR-1 reactor core are taken as a data for calculations, the rods are taken to be of boron carbide. A complete computer code is written on Fortran 4 language to calculate the above items

[en] The nuclear reactor lattice cell code 'EGCODE 2' is modified to be available to calculate boiling water reactor (BWR) cell parameters (5 zones) as well as pressurized water reactor (PWR) cell parameters (3 zones). This modified code calculates all cell parameters such as, neutron flux, reaction rates, nuclear temperature coefficients, fission products concentrations and activities and fuel isotropic concentrations at room temperature as well as at high temperatures for any void fractions. This more complicated new version of the code 'EGCODE 2' is tested throughout a comparison between our results and same of already published ones for five zones of the RBMK-1000 reactor cell type (Chernobyl reactor). This comparison leads to a conclusion that the 'EGCODE 2' is a good code for the cell parameters calculations for different reactor type. The results of calculations using this code show variations of cell nuclear parameters of the RBMK-1000 reactor at different temperatures, and the variation of these parameters with burn-up. They also show the variation of fuel isotopes concentrations with time of operation.4 fig., 3 tab

[en] A study of the fast breeder physical reactor parameters as well as comparison with those of a thermal pressurized water reactor of the same power is presented in this paper. The flux is calculated inside the four zones of the fast breeder reactor using the two dimensional two groups treatment. The fission products accumulation in each zone are calculated after several times of operations and shut down using a computer program. The activity of the radioactive isotopes is also calculated after several times of operation and shut down of the reactor and compared with that of the PWR. From this comparison the activity in the FBR is about one tenth of the activity of PWR which gives an advantage for the FBR from the metallurgical point of view as well as from the safety point of view. The activity just after shut down in a FBR is less than that of the PWR after one year which gives another advantage for the FBR with respect to the treatment of fuel waste. The activity of the very small half lives is 65% of the total activity. This comparisons show some advantages of the FBR over the PWR from the following points of view. 1. Waste management: since the activity after shut down of the FBR is 1/10 of that of the PWR. Also it is (after one week of shutdown) much less than that after one year of shut down of the PWR. 2. Safety: from the preceding point, if an accident happened, the FBR will be less dangerous than the PWR accident

[en] A proposed nondestructive method and its feasibility for the determination of U-235, Pu-239 and Pu-240 contents in an irradiated fuel is described. The method is based on the use of shape fit analysis of the Time-Of-Flight (TOF) neutron transmission data of the irradiated fuel for neutron energies below 3 eV. The neutron transmission experiment of the irradiated fuel is planned to carry out using one of the TOF spectrometers installed at ET-RR-1 reactor. The computer code SHAPE is adapted taking into account the known parameters of resonances of certain fissile and fission product nuclei to provide the fit analysis. The content of the gross-fissile and fission product isotopes are determined from the burn-up calculations of the fuel assembly of the ET-RR-1 reactor with defined history. The effect of both uncertainties in resonance parameters on the deduced contents of fissile nuclei and statistical accuracy of the TOF measurements are estimated

[en] The present work provides an optimized model for fuel loading patterns in pressurized water reactors. The developed model satisfies the following objectives: A) An improved power distribution throughout the reactor core at the begining of the core cycle. This model provides also an acceptable power profile throughout the whole core cycle. The minimum peak-to-average power ratio was chosen as the objective characteriatic of the optimum loading. b) Less management effort, as for instance simple computer programme or small computer time, for generating optimized loading patterns. This is justified by: 1. Using an improved fuel grouping and region classifications. 2. Excluding many of the failed shuffling iterations due to either using the loading priority sequence in order to perfor a good initial loading or using the local reactivity requirement for initially accepting or rejecting a shuffle iteration before power calculation is made. Because of the vast number of possible fuel assembly allocations, it is not possible to obtain a strictly optimum solution for a given criteria function. In this work the number of feasible patterns is greatly reduced through the use of a logical set of shuffling rules which utilizes the radial power and reactivity distributions of each shuffling iteration. In order to calculate the two dimensional power distribution, a two dimensional simulated 1.5 group coarse mesh diffusion theory model is used. The results obtained shown that the proposed simple modified algorithm adopted is efficient and applicable in determining the optimum loading patterns when compared to results from previous publications