Volume 77, Issue 1

When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions.

In this work a detailed study of the annular fuel has been carried out. A thermal hydraulics code, ANUFAN (Annular Fuel Analysis), based on the bundle average method, capable of modeling both internally and externally cooled annular fuel pins is developed. Code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. Heat transfer fraction difference between ANUFAN and RELAP was found about 1.7%. Analysis of a 54 – fuel rod assembly is carried out with 36 and 45 numbers of annular fuel pins keeping the same channel size and bundle power as of the solid fuel assembly. Fuel pin maximum temperature of the annular fuel is found much less than the solid fuel. MCHFR value for annular fuel is found much higher compared to that of the solid fuel of 54 – fuel rod assembly. The full paper covers the details of the computer code, the analysis carried out and the results obtained.

Various thermohydraulic codes (like WAHA, ATHLET, RELAP) work perfectly for pressurized water and steam, but fail to give reliable results for supercritical water. This might be a surprise, because theoretically Supercritical Water (SCW) should be a simpler system than normal water, due to the lack of phase transition and two-phase flow in the supercritical region. Some of the problem is caused by low accuracy of the fluid properties due to the presence of the pseudo-critical line. In this presentation we would like to address this pseudocriticality-related problem and to establish a method for the error-reduction.

In water cooled power reactors, significant quantities of hydrogen could be produced following a severe accident (loss-of-coolant-accident along with non availability of emergency core cooling system). A sound understanding of dispersion, stratification and diffusion of released hydrogen during severe accidents is, therefore, of practical importance and use to better understand the possibility of ignition, combustion and explosion of such releases within the context of containment safety. The presence of air and steam in the containment atmosphere also affects the hydrogen distribution as steam condensation takes place at containment walls in presence of non condensable and bulk of the mixture diffuses towards wall. The application of general purpose CFD codes for the analysis of the hydrogen behaviour within NPP containments during severe accidents has been increasing over past few years. The commercial CFD codes generally do not have built-in steam condensations models. In the present work, the adaptation of a commercial multi-purpose code to this kind of problem is explained, i.e. by the implementation of models for steam condensation onto walls in presence of non-condensable gases. Steam condensation was modeled using the Uchida correlation, which was originally developed to be used for “lumped” (volume-averaged) modeling of steam condensation in the presence of non-condensable gases. The Uchida correlation is based on experiments on natural convection from relatively small vertical plates. The present methodology has been validated against experimental data from the TOSQAN and COPAIN experimental facilities.

A successful design of high pressure hydraulic valves requires a thorough analysis of both velocity and pressure fields, with the aim of improving the geometry to avoid cavitation. Cavitation behavior prediction of hydraulic valves and its associated performance drop is of high interest for the manufacturers and for the users. The paper presents a CFD analysis of the flow inside a high pressure hydraulic valve. First, the analysis was carried out without using cavitation model (single phase). It was observed that absolute pressure was going below the vapor pressure. Hence, it was required to turn on the cavitation model. This model enables formation of vapor from liquid when the pressure drops below the vaporization pressure. Since the cavitation bubble grows in a liquid at low temperature, the latent heat of evaporation can be neglected and the system can be considered isothermal. Under these conditions the pressure inside the bubble remains practically constant and the growth of the bubble radius can be approximated by the simplified Rayleigh equation. For typical poppet valve geometry, ½ of computational domain is assumed, with pressure inlet and outlet boundary conditions, and a steady flow solution is computed. Because of the highly complex geometry of the hydraulic valve, the computational domain was meshed using unstructured grids using tetrahedral cells only. The paper presents a numerical investigation of the flow inside a hydraulic valve using commercial CFD code CFD-ACE. The aim of the study is to provide a good basis for future designing of the hydraulic valve. The result indicated the cavitation zones which in turn suggest needs of modification of present geometry.

A computational model was developed earlier for 2 units of Nuclear Power Plants (NPPs) operational at Kaiga Atomic Power Station (KAPS) to understand the thermal plume behaviour in the Kadra reservoir wherein the hot water from the plant condensers is discharged. The model was successfully validated against the site data. The same model has now been extended for analyzing the thermal plume bahaviour in case of 4 NPP units as well as 6 NPP units operational at the same site. The present paper briefly describes details of the studies along with the results of earlier study to understand the overall behavior of thermal plume in Kadra reservoir.

We here present a method for the production of Iodine-124 using a cyclotron. With this method, it will be possible to produce clinically relevant amounts of 124 I with high chemical and radionuclidic purity for both diagnostic and therapeutic applications. The electroplated TeO 2 targets were irradiated with 29.5 MeV protons at current of 39.5 μA for 3.8 h. The separation of the produced Iodine activity from the bulk target material was performed by ion exchange chromatography method. 124 I positron emitter excitation functions via 124 Te(p,n) 124 I, 125 Te(p,2n) 124 I, 126 Te(p,3n) 124 I, 124 Te(d,2n) 124 I, 121 Sb(a,n) 124 I and nat Te(p,xn) 124 I reactions were calculated by ALICE/ASH 0.1 (GDH Model & Hybrid Model) and TALYS-1.2 codes and compared to existing data. Theoretical calculation of production yield and calculation of target thickness requirement were obtained by TALYS-1.2 and SRIM codes for each reaction.

In this study, ground state properties for some medical radionuclides used widely in medical applications have been investigated using Skyrme-Hartree-Fock (SHF) method. We calculated the proton, neutron and charge density RMS (root-mean-square) radii of some medical radionuclides such, as 42 K 45 Ca, 81 Kr, 81 Rb, 82 Rb, 85 Sr, 110 In, 111 In, 123 Cs, 201 Tl and 201 Pb using the SHF calculations with the Skyrme force parameters. The calculated results are compared with the available experimental values for charge density RMS radii. The proton, neutron, charge and mass density distributions for all nuclei used in this study have also been calculated. The calculated conclusions can be used in estimating nuclear properties of the medical nuclei and for the calculations of pre-equilibrium reactions cross section data for the production of new medical radioisotopes.

The successive order scattering approximation method is used to study the radiative transfer equation. By separating the coherent component of the scattered flux, the transport equation is represented in terms of each order scattering flux and a simplified solution is obtained with this approach. The method is then used to calculate the reflected flux and the transmitted flux with three different phase functions. A new boundary condition has been added for the first order forwardly scattered flux. Thus it becomes possible to study the transmittance. We compare our results with the available data for a biological medium.

In this work, the F N method with different numerical approximations is used to solve the constant source problem for the quadratic anisotropic scattering kernel. Case's eigenfunctions and the orthogonality relations of this scattering are obtained analytically for one-speed and homogenous transport equation. The aim is to find the albedo values in plane geometry for the half-space when the medium boundary contains a plane source.

It is well known that when spent fuel is removed from the core, it is still composed of considerable amount of radioactive elements with significant half-lives. Most actinides, in particular plutonium, fall into this category, and have to be safely disposed of. One solution is to store the long-lived spent fuel as it is, by encasing and burying it deep underground in a stable geological formation. This implies estimating the transmutation of these radioactive elements with time. Therefore, we describe in this paper the application of the Laplace transform technique in matrix formulation to analytically solve initial value problems that mathematically model radioactive decay series. Given the initial amount of each type of radioactive isotopes in the decay series, the computer code generates the amount at a given time of interest, or may plot a graph of the evolution in time of the amount of each type of isotopes in the series. This computer code, that we refer to as the LTRad L code, where L is the number of types of isotopes belonging to the series, was developed using the Scilab free platform for numerical computation and can model one segment or the entire chain of any of the three radioactive series existing on Earth today. Numerical results are given to typical model problems to illustrate the computer code efficiency and accuracy.

The theory of a “Standing Wave Reactor” is developed whereby fertile fuel elements are continuously shuffeled from inward to outward in a cylindrical reactor core. The temporal and spacial stiffness of the S-shaped fluence and the M-shaped flux distributions are typical for such a fertile fuel consuming fast system.

The radial M-shaped flux distribution of a fast Standing Wave Reactor (SWR) with inward → outward fuel shuffling is derived and presented in a polynom-approximation. It is the supplement to the S-shaped radial fluence distribution characterizing this new type of a fertile fuel consuming breeder system.