Volume 80, Issue 1

The displacement cross-sections were evaluated for main components of EUROFER at the energies of primary neutrons up to 150 MeV. The number of primary defects was calculated using an advanced binary collision approximation model applying results of molecular dynamics simulations. The cross-sections obtained were processed into ENDF and ACE formatted data and tested using the MCNP code.

India is currently operating two BWR built by General Electric Company. The design features of these reactors are similar to the Fukushima's BWR except some better containment features in Indian BWR. This paper discusses the enveloping scenario of station blackout of infinite duration with no operator action and no component failure. The paper describes the details of modelling the TAPS-BWR plant model including SCDAP modelling of reactor core in system code RELAP5 and further thermal hydraulic safety assessment of station blackout scenario. The analysis brought out effectively the response of the plant to this high-pressure severe accident scenario. The time line of the severe accident progression will give details of various stages of accident progression along with hydrogen generation, which will be useful in evolving suitable severe accident management guidelines.

The steady state and transient characteristics of a molten salt natural circulation loop (NCL) are obtained by 3D CFD simulations. The working fluid is a mixture of NaNO 3 and KNO 3 in 60:40 ratio. Simulation is performed using PHOENICS CFD software. The computational domain is discretized by a body fitted grid generated using in-built mesh generator. The CFD model includes primary side. Primary side fluid is subjected to heat addition in heater section, heat loss to ambient (in piping connecting heater and cooler) and to secondary side (in cooler section). Reynolds Averaged Navier Stokes equations are solved along with the standard k-∊ turbulence model. Validation of the model is done by comparing the computed steady state Reynolds number with that predicted by various correlations proposed previously. Transient simulations were carried out to study the flow initiations transients for different heater powers and different configurations. Similarly the “power raising” transient is computed and compared with in-house experimental data. It is found that, using detailed information obtained from 3D transient CFD simulations, it is possible to understand the physics of oscillatory flow patterns obtained in the loop under certain conditions.

The present work was developed within the frame of the IAEA Coordinated Research Project 1496 “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”. Three benchmark experiments were designed and performed to study the performance of the Egyptian Research Reactor ETRR-2. The experiments included steady state measurements as well as loss of flow transient (LOFT) and loss of power transient (SCRAM) conditions. The Code RELAP5/Mod3.4 was used to simulate the components of the ETRR-2 systems for the thermal hydraulic analysis of the reactor. The dimensions and elevations of the primary cooling components are based on real conditions (3-D configuration). RELAP5 results provided benchmark data which verified the experimental measurements taken from instrumentations installed for this experiment at several positions in the core. The predicted values for the coolant and clad surface temperature at different locations in the core showed a remarkable agreement with the experimental values, for both the steady state and transient conditions.

In this work an attempt is made to achieve characteristics curves for the Tehran Research Reactor (TRR) during stopping. The TRR pump speed during the event is time dependent. This is done for the TRR first quadrant operating characteristics. The obtained results indicate that the time needed for the coolant flow rate to reach zero is greater than the time necessary for the pump to come to complete stoppage.

In this study, salt-heavy metal mixtures consisting of 93–85% Li 20 Sn 80 + 5% SFG-PuO 2 and 2–10% UO 2 , 93–85% Li 20 Sn 80 + 5% SFG-PuO 2 and 2–10% NpO 2 , and 93–85% Li 20 Sn 80 + 5% SFG-PuO 2 and 2–10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A beryllium (Be) zone with a width of 3 cm was used for neutron multiplicity between the liquid first wall and the blanket. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The contributions of each isotope in the fluids to the nuclear parameters, such as tritium breeding ratio ( TBR ), energy multiplication factor ( M ), and heat deposition rate, of the fusion–fission hybrid reactor were calculated in the liquid first wall, blanket, and shield zones. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

PN approximation is known as a proper method to solve neutron transport equation when literature is taken into consideration. Generally, conventional scattering function is used to solve criticality and diffusion problems in Legendre polynomial approximation. In this study, instead of conventional scattering function, Henyey-Greenstein (HG) and Anlı-Gungor phase functions (AG) are used in slab geometry transport equation and some critical thicknesses of the slab are calculated as an application with Legendre polynomial (PN) approximation and Marshak boundary condition. Results obtained from HG and AG scattering functions are compared and the correlations and discrepancies between the two functions are presented in the tables.

In this study, production cross–section calculations of 32 P, 117 Sn, 153 Sm and 186,188 Re radionuclides used in bone pain palliation treatment produced by 30 Si(d,γ) 32 P, 118 Sn(γ,n) 117 Sn, 116 Sn(n,γ) 117 Sn, 150 Nd(α,n) 153 Sm, 154 Sm(n,2n) 153 Sm, 152 Sm(n,γ) 153 Sm, 186 W(d,2n) 186 Re, 187 Re(γ,n) 186 Re, 185 Re(n,γ) 186 Re and 187 Re(n,γ) 188 Re reactions have been investigated in the different incident energy range of 0.003–34 MeV. Two-component exciton and generalised superfluid models of the TALYS 1.6 and exciton and generalised superfluid models of the EMPIRE 3.1 computer codes have been used to pre-equilibrium (PEQ) reaction calculations. The calculated production cross–section results have been compared with available experimental results existing in the experimental nuclear reaction database (EXFOR). Except the 118 Sn(γ,n) 117 Sn, 150 Nd(α,n) 153 Sm and 185 Re(n,γ) 186 Re reactions, the two-component exciton model calculations of TALYS 1.6 code exhibit generally good agreement with the experimental measurements for all reactions used in this present study.

In this study, carbonated hydroxyapatite nanoparticles were used for thermoluminescence dosimetry. The nano-structure carbonated hydroxyapatite synthesized via hydrolysis of CaHPO 4 and CaCO 3 . The obtained nano powders were characterized by XRD technique and FTIR spectroscopy system. The carbonated hydroxyapatite samples were irradiated at different doses using 60 Co gamma rays, and were subjected to thermoluminescence measurement system, consequently. The TL glow curve exhibited two distinguishable peaks centered at around of 165°C and 310°C. The TL response of carbonated hydroxyapatite samples as a function of absorbed dose was linear in the range of 25–1000 Gy. Other dosimetric features of the carbonated hydroxyapatite nanoparticles including fading and reproducibility were also investigated.

The nuclear fuel cycle is investigated for the perspective of the nuclear non-proliferation. The random number generation of the Monte-Carlo method is utilized for the analysis. Five cases are quantified by the random number generations. These values are summed by the described equations. The higher values are shown in 52 nd and 73 rd months. This way could be a useful obligation in the license of the plant construction. The security of the nuclear fuel cycle incorporated with nuclear power plants (NPPs) is investigated using social impact theory. The dynamic quantification of the theory shows the non-secured time for act of terrorism which is considered for the non-secured condition against the risk of theft in nuclear material. For a realistic consideration, the meta-theoretical framework for modeling is performed for situations where beliefs, attributes or behaviors of an individual are influenced by those of others.

The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.