International Journal of Nuclear Energy Science and Technology (10 papers in press)
Burn/breed: a wavelet-based nuclear fuel burnup and decay heat code
by Hesham Nasif
Abstract: Burn/breed is a code designed to aid in the analysis, prediction and optimisation of fuel burnup performance and decay heat calculation in a nuclear reactor. The code uses the output parameters generated by the Monte Carlo neutronics codes to determine the isotopic inventory as a function of time and power density. Burn/breed directly uses the neutron absorption tally/reaction information generated by Monte Carlo code for each nuclide of interest to determine the nuclide inventory. When the isotope inventories have been calculated for a specific reactor operation and cooling period, the decay heat can be derived. Beside the U-235 and Pu-239 decay chains, the code includes the U-233 decay chain to manage the possible scenarios for using thorium in a nuclear fuel cycle. This paper describes the theoretical basis of this code, and shows the results of the code for two test cases. The results show good agreement with other codes for the first test case and with the experimental results for the second test case.
Keywords: burn/breed; burnup; decay heat; nuclear reactor; Monte Carlo; isotopic inventory; wavelet; U-233 decay chain.
A preliminary comparative study between oxide and metallic fueled ASTRID-like reactor under a B&B strategy
by Juan Luis Francois, Elías-Yammir García-Cervantes, Cecilia Martín-del-Campo
Abstract: Depleted uranium stockpiles have increased from enriching nuclear fuel of current commercial plants. This resource will not be fully useful unless nuclear technologies are developed for using it. For fast reactors, the breed and burn concept enables the maximisation of uranium usage by burning part of the bred material without the necessity of reprocessing fissile isotopes, which simplifies the nuclear fuel cycle with important economic advantages. The ASTRID nuclear reactor is a fourth-generation sodium-cooled fast reactor of 1500 MWth which considers an innovative design: the low void effect core. Our previous research on the ASTRID-like reactor involved the MCNP6 model assessment of an oxide-fuelled core and its comparison with a proposed metallic-fuelled core design. Considering the amount of fissile material (mainly 239Pu) produced after a first operating cycle, a reshuffling scheme is suggested in this work, which consists of changing, strategically, the position of the fuel assemblies, at the end of the operating cycle, when the reactivity drops near to the critical state. Then, the main objective of this paper is to study the reactivity behaviour and the isotopic fuel performance of an ASTRID-like reactor under a breed and burn strategy for both core designs, fuelled by oxide as well as metallic fuels. The proposed reshuffling scheme was simulated with MCNP6 and the JEFF-3.2 cross-sections library to extend the fuel life by two more cycles, for both core designs. Our findings showed that the implementation of the reshuffling scheme enhanced the fuel usage, obtaining cycle extensions of 805 days and 1305 days, over the first 365 days, for the oxide and metallic-fuelled designs, respectively. For the metallic-fuelled design the breeding of 239Pu achieved a total net production of 577.7 kg, which can be expressed as a production rate of 126.3 kg/EFPY. The conversion rate is 1.06 for the metallic-fuelled design and 0.96 for the oxide-fuelled design. As for the coolant void reactivity worth, it was observed that for the oxide-fuelled design this value becomes positive after the second fuel reshuffling, which is not the case for the metallic-fuelled design. In summary, results show a superior performance of the metallic-fuelled core design.
Keywords: fast reactors; sodium-cooled reactor; breed and burn; metallic fuel; ASTRID.
Semi-empirical formula for photon energy absorption buildup factors of elements and compounds
by H.C. Manjunatha, L. Seenappa, K.N. Sridhar
Abstract: We have formulated a simple semi-empirical formulae for photon energy absorption buildup factors in the energy region 0.015-15 MeV, atomic number range 1≤Z≤92 and for mean free path up to 40 mfp. The results produced by the present formulae agree well with the data available in the literature. This semi-empirical formula may be extended to any compounds/mixtures/biological samples. This semi-empirical formula finds importance in the calculation of buildup factors of any materials that are required for radiation shielding, nuclear engineering, radiotherapy and nuclear medicine.
Keywords: photon energy absorption buildup; mean free path; radiation shielding; nuclear engineering; radiotherapy; nuclear medicine.
Feasibility study for productions of 99Mo and 99mTc by the neutron activation of 98Mo in the MNSR reactor
by K. Khattab, George Saba
Abstract: The calculated weekly specific activities of 99Mo and 99mTc produced from the irradiation of the MoO3 targets in the Miniature Neutron Source Reactor (MNSR) are presented in this paper. The productions of the 99Mo and 99mTc is modelled and calculated through a set of differential equations using the Mathcad program. The resonance self-shielding factor (Gres) was calculated using the MATSSF and MCNP4C codes. The effects of the physical parameters, such as the neutron flux and irradiation time, on the weekly specific activities of 99Mo and 99mTc are analysed. It is found that the optimum irradiation scheme was achieved when the MNSR was operated for an extended period of 5 hours a day for 5 days a week at the neutron flux of 7.5
Keywords: Mo production; neutron flux; irradiation time; specific activity; MNSR.
Fuel loading, criticality and control rod worth calculations of the Triga Mark III reactor using Serpent and MCNP
by Jaime Hernandez-Galeana, Juan Galicia-Aragon, Armando Miguel Gómez-Torres, Carlos Filio-López
Abstract: In this work, a model of the TRIGA MARK III reactor core (RTMIII) of the National Institute for Nuclear Research (ININ) located in Mexico was developed with the Monte Carlo codes Serpent and MCNP, and verified and validated by means of the experiments carried out as part of the starting tests to change the mixed fuel (High Enrichment Uranium, HEU, and Low Enrichment Uranium, LEU) to low enrichment fuels (LEU) of the TRIGA reactor core. A serious verification and validation (V&V) process is essential in order to have trustworthy tools not only to calculate and analyse the physical parameters of the reactor core to evaluate the optimal and safe operation of the reactor core, but also to extend the experimental capabilities of the Triga Mark III reactor. The reactor data used in the V&V process consisted of fuel loading
measurements, simulating the different stages of loading of fuel elements to the
core to reach the reactor core criticality and the additional loading to achieve the reactivity excess to operate the reactor, as well as the evaluation of the shutdown
margin reactivity and the control rods worth. Since the differences obtained for the multiplication factor are acceptable, it is found that with the computational models developed, the calculations and neutron analysis required during the operation of the reactor can be performed. According to the results of the reactivity values obtained, it is inferred that a comprehensive analysis of the experimental measurements and the inclusion of an uncertainty analysis for the calculations performed with the model, are required. The most important result of this paper is that the technical personnel
responsible for the operation of the TRIGA MARK III reactor will have at their
disposal a computational tool that models in detail the reactor core in order to
calculate and analyse the most important neutronic core parameters during the
reactor operation, as well as to have the numerical capabilities for performing fuel
utilisation studies and analysis for extension of experimental capabilities of the
Keywords: Triga Mark III reactor; Monte Carlo methods; criticality and fuel loading; control rod worth.
Study of Grodzins product in collective nuclear structure
by Harish Mohan Mittal, Amit Bindra
Abstract: Systematic dependence of Grodzins product 〖E(2〗_1^+)*B(E2)↑ on the nucleonic promiscuity factor, i.e. P-factor = (NpNn)/(Np+Nn), is studied in the Z = 50-82 N = 82-126 major shell space. The anomalous behaviour is noticed for N = 88-90, where the Grodzins product 〖E(2〗_1^+)*B(E2)↑ fails to maintain the constancy corresponding to P≥4. The P≥4 is recognised for enhanced collectivity, where p-n interaction of the order of ~250 keV begins to dominate the pairing interaction of ~1 MeV. The P-factor also resolves the anomaly of not obtaining a smooth universal Casten curve in NpNn scheme. The approximate constancy in Grodzins product 〖E(2〗_1^+)*B(E2)↑ is noticed for well deformed nuclear region. We have studied, for the first time, the relation between Grodzins product 〖E(2〗_1^+)*B(E2)↑ and the P-factor.
Keywords: Grodzins product; nuclear structure; P-factor.
Calculation for gamma ray buildup factor for aluminium, graphite and lead
by Hiwa Qadr
Abstract: The purpose of this work was to investigate and quantify the linear attenuation coefficient and the buildup factor for different materials. The measurement of the linear attenuation coefficient of absorber materials such as graphite was (0.097 cm-1), whereas it was observed to be (0.136 cm-1) for aluminium, and (0.596 cm-1 ) for lead. By using the gamma radiation energies emitted from 60Co source with 1.25 MeV, the attenuation coefficients were measured by using counts of good geometry and bad geometry. The results show that the linear attenuation is higher for lead and better radiation shielding compared with graphite and aluminium. Furthermore, the buildup factor decreases with increasing thickness of the absorber material.
Keywords: sodium iodide detector; gamma attenuation; buildup factor; good geometry; graphite.
A multigroup extended linear discontinuous method for fixed-source discrete ordinates problems in slab geometry
by Iram B. Rivas-Ortiz, Dany Dominguez, Carlos R. Garcia Hernández, Susana M. Iglesias
Abstract: At present, neutron density calculation in non-multiplying media is relevant in many areas of engineering and science. In this paper, we propose the Extended Linear Discontinuous (ELD) method in multigroup discrete ordinates formulation, originally formulated for one-energy group fixed-source problems with isotropic scattering source in slab geometry. The proposed auxiliary equations are uncoupled on angular directions and combine the linear discontinuous approximation of the finite element method and the quasi-analytical general solution of the spectral nodal method. Thus, we can implement an efficient and simple algorithm using the conventional source iteration scheme for the sweeping equations. Numerical results for benchmark problems are presented to illustrate the accuracy and computational performance of the ELD method. The work shows that the main advantages of the proposed method are that the numerical scheme is stable for coarse meshes, and its numerical results are more accurate than those generated by the Diamond Difference (DD) and Linear Discontinuous (LD) methods.
Keywords: multigroup transport problems; discrete ordinate formulation; fixed-source problems; linear discontinuous.
Nuclear power plant in India: achieving clean and green energy
by Usha Shukla, Rishabh Bajpai
Abstract: With the increasing demand in electricity and rising temperature of the Earth owing to global warming, nuclear power plants can address the current needs. Development in the realm of nuclear energy has become a necessity in order to fulfill the present need. The present paper will summarise the basic knowledge regarding the nuclear power plant and current status of nuclear energy in India. Moreover, the paper presents some limitations to nuclear energy. This review paper will be helpful for the beginners in the field of nuclear power plants.
Keywords: nuclear power plant; nuclear energy; nuclear programs; India's nuclear energy future.
Surface diffuseness parameter with quasi-elastic scattering for some heavy-ion systems
by Qasim J. Tarbool, Khalid S. Jassim, Ali Abojassim
Abstract: Itemised studies of the surface properties of the inter-nucleus potential in heavy-ion reactions have been achieved using large-angle quasi-elastic scattering at sub-barrier energies close to the Coulomb barrier height for (_3^6)Li +(_30^64)Zn and (_3^7)Li + (_30^64)Zn systems. In this piper, the nuclear potentials have been expressed using the Woods-Saxon (WS) formula. The effect of rotational deformation was included for the nucleus (_30^64)Zn with ground state rotational band up to the 4+ states. The single-channel (SC) and coupled-channels (CC) calculations have been carried out to elicit the diffuseness parameter of the nuclear potential as well as the potential depth. These calculations have been performed by using (CQEL) program, which is considered the latest version of computer code (CCFULL). The chi square method χ2 played an important role in determining the best fitted value of the diffuseness parameter. Through CC calculations with inert projectile and rotational target for (_3^6)Li + (_30^64)Zn and (_3^7)Li + (_30^64)Zn systems, we got full compatibility of diffuseness parameter with the standard value 0.63 fm with χ2 = 0.130 and χ2 = 0.163, respectively, while the SC calculations give 0.64 fm and 0.65 fm, respectively.
Keywords: coupled-channels calculations; heavy-ion fusion reactions; deep sub-barrier energies; quasi-elastic scattering; Woods-Saxon potential.