Volume 78, Issue 3

The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated.

Severe accident analysis of a reactor helps in emergency planning and evolution of Severe Accident Management Guidelines (SAMG). Actions recommended in the SAMG aim at arresting accident progression and limiting significant radioactive release. However, success of these SAMG actions needs to be assessed with respect to the evolution of accident. Analysis of consequences of injection of water into the reactor pressure vessel from bottom only as a SAMG action has been carried out for VVER-1000 (V320) reactor. The analysis shows that the success of this SAMG action depends not only on the state of core degradation at the time of injection, but also on the highest temperature reached in the reactor core at the time of injection as well as the availability of steam in the RPV.

Natural circulation systems offer simplicity, enhanced safety and reliability and thus are advantageous over their forced circulation counterpart. However, natural circulation is susceptible to flow instabilities. These instabilities are undesirable for various reasons. Literature suggests the use of orifices at the inlet to suppress instability. However, orificing introduces pressure drop which limits the flow rate and hence the heat carrying capacity of the fluid. In the present study, investigations have been carried out with different mechanical gadgets to control the natural circulation flow instabilities. Extensive experiments have been carried out in a single-phase rectangular natural circulation loop to study the effect of these mechanical gadgets on instability. The paper brings out the results of experimental investigations carried out on the role of mechanical gadgets in controlling instability.

In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232 Th and mixed 233 U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period.

In this study, 98–90% Li 2 BeF 4 -2–10% ThF 4 , 98–90% Li 2 BeF 4 -2–10% UF 4 , 98–90% Li 2 BeF 4 -2–10% UO 2 and 100% Li 2 BeF 4 molten salt-heavy metal was used as fluid. The fluids were used in the liquid first wall, liquid second wall and shield zones of the designed hybrid reactor system. A steel wall of 4 cm thickness is used as structural material. Proton, deuterium, tritium, He-3 and He-4 gas production rates are the parameters of radiation damage. In the study, the effect of liquid second wall thicknesses (20 cm, 30 cm, 40 cm, 50 cm) on the neutron flux distribution and the parameters of radiation damage according to neutron energy spectrum in the structural material were investigated for the selected fluids. A three-dimensional analysis was done by using the most recent version of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.

The radionuclide 55 Co (T 1/2 = 17.5 h, E β+ = 1.5 MeV, 76% β + decay) is an important β + emitting radioisotope. For production of 55 Co via nat Fe( p, xn ) 55 Co reaction, an iron layer was deposited on a copper substrate by means of electro-deposition method which could be irradiated by 29.5 MeV protons at 100 μA. No-carrier-added (n.c.a.) 55 Co was separated from the iron target via an anion exchange column (Dowex 1 · 8). The achieved production yield was 31.25 MBq/μAh. Also, excitation functions for the 55 Co radionuclide via nat Fe(p, xn) 55 Co, 56 Fe(p, 2n) 55 Co and 54 Fe(d, n) 55 Co reactions were calculated by TALYS-1.4 code and TENDL-2011 database and compared with previous published data.

Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV.

Cu–Al bimetallic nanoparticles were synthesized by gamma irradiation technique in aqueous solutions containing metal chlorides as precursors, polyvinyl alcohol (PVA) as a capping agent, isopropanol as a radical scavenger, and distilled water as a solvent. The Cu–Al bimetallic nanoparticles were characterized by transmission electron microscopy (TEM), UV-visible absorption spectrometry, powder X-ray diffractometer (XRD), and Energy-dispersive X-ray spectroscopy (EDX). The TEM, XRD, EDX, and absorption analyses confirmed the formation of core-shell structure of Cu–Al bimetallic nanoparticles at lower Cu 2+ /Al 3+ mole ratio, and the formation of Cu–Al alloy nanoparticles at higher Cu 2+ /Al 3+ mole ratio. The TEM analysis for particle size and size distribution revealed that the average particle size of Cu–Al bimetallic nanoparticles decreased with the increase of absorbed dose. It may be explained due to the competition between nucleation and aggregation processes in the formation of metallic nanoparticles under irradiation.

A numerical method has been proposed for resonance integral calculations and a cubic fit based on least squares approximation to compute the optimum Bell factor is given. The numerical method is based on the discretization of the neutron slowing down equation. The scattering integral is approximated by taking into account the location of the upper limit in energy domain. The accuracy of the method has been tested by performing computations of resonance integrals for uranium dioxide isolated rods and comparing the results with empirical values.

An analytical numerical method applied to three different types of monoenergetic neutral particle inverse transport problems in the discrete ordinates (S N ) formulation is presented: (a) boundary condition estimation; (b) interior source estimation; and (c) effective slab length estimation. These three types of inverse problems governed by the linear integrodifferential transport equation in S N formulation are related respectively to medical physics; nuclear waste storage; and non-destructive testing in industry. Numerical results and a brief discussion are given to conclude the paper.

Criticality and diffusion problems have been widely studied in neutron transport theory using polynomial expansion techniques in different geometries. In this study, instead of conventional scattering functions, two different phase function will be used in slab geometry transport equation by using the T N approximation and Marshak boundary conditions. Critical half thicknesses of the slab will be calculated. The numerical results obtained will be presented in tables.

U 1 and P 1 approximations are used for the calculation of asymptotic relaxation length in one-dimensional neutron transport equation. The approximation methods are applied to anisotropic neutron transport equation with backward and forward scattering. The methods are based on the series expansion of the neutron angular flux in terms of the Chebyshev polynomials of second kind and Legendre polynomials. By applying the first order approximation to the transport equation, asymptotic relaxation lengths are calculated. Numerical results obtained from both methods are compared with each other.

The critical thickness for one-speed neutrons in a uniform finite slab with anisotropic scattering is investigated using Marshak boundary condition. The angular part of the neutron angular flux is expanded in terms of the Chebyshev polynomials of first kind. Numerical results for the critical thickness of the slab are calculated for various values of the collision and the anisotropy parameters and they are given in the tables together with the ones obtained by Legendre polynomials approximation and the ones available in literature for comparison. The results obtained by the present method are in good accordance with the literature values.

The half-space albedo problem and the constant source problem have been solved for a combination of the linearly anisotropic scattering and İnönü’s scattering functions. The linear transport equation for extremely anisotropic scattering kernel can be converted into an equivalent equation with a linearly anisotropic scattering kernel and the modified F N method can be used for albedo calculations.