Volume 79, Issue 1

Recently the reactors of both units of Ignalina NPP in Lithuania were shutdown for decommissioning (in 2004 and 2009). According to the design, the spent fuel should be returned for reprocessing to Russia. However, up to now, all assemblies of spent fuel are still stored in the spent fuel pools and in the dry storage facility on-site of the Ignalina NPP. Thus, the safety justification for short and intermediate term spent fuel assemblies storage in Ignalina NPP is very important. The most important barrier, preventing release of radioactivity from the fuel matrix to the environment is the fuel rod cladding. The condition (integrity) of cladding at the end of intermediate storage of spent fuel assemblies may be evaluated by simulating processes in fuel rods during the whole “life cycle” of the fuel assembly: beginning from the first loading into the reactor core, until the end of dry intermediate storage in special casks. This paper presents the modelling of processes in fuel rods during normal operation of fuel assemblies in reactor, short term wet storage in spent fuel pool and intermediate storage in dry cask. The analysis was performed using FEMAXI-6 integral code for the analysis of processes in fuel rod. The behaviour of thermal (pressure inside fuel rod, temperatures of fuel pellets and cladding, etc.) and mechanical (change of the gap between pellets and cladding, stresses and plastic deformation in fuel pellets and cladding, etc.) parameters were calculated. The analysis performed demonstrates the possibility to describe the state of the fuel rods after normal operation and short and intermediate term fuel storage.

The diffusion of hydrogen in zircaloy under the influence of mechanical stresses is investigated. The governing equations are derived from the chemical potential of hydrogen including the contribution due to hydrostatic stresses. The obtained differential equation is converted to a system of algebraic equations by applying the finite element method and the weighted residual procedure. The equations are used to study the hydrogen diffusion in a fuel rod cladding tube with an axial crack and in a strip plate. It is demonstrated that the hydrogen tends to flow towards the areas of positive hydrostatic stress gradients where it precipitates in hydrides when the solubility limit is exceeded. The precipitation is itself a mechanism enhancing the transport to these areas by weakening the effect of the diffusion driven by concentration differences. The presented method can readily be applied to the diffusion of other species in metals.

In the first part of the present paper, for a 5 MW thermal pool-type research reactor, the fuel element is modelled for when undergoing both natural circulation of the coolant and forced convection of the coolant operational conditions. First, the required dimensionless groups were identified and then the pertinent similarity criteria were derived accordingly. The derived similitude laws were modified under the conditions of identical pressure, identical temperature difference and identical coolant and fuel cladding in the model and the prototype. These modifications were done for the system under both natural and forced convections. The effect of varying cladding materials under normal operating conditions of the research reactor were observed via coolant channel thickness. Also the effect of a wider coolant channel on the nature of the coolant fluid was observed. The results obtained indicate that it is not possible to conserve all the dimensionless groups between the model and the prototype and hence achieve an errorless outcome. Among all the liquids available, methanol is the only liquid which nearly satisfies the thermal-hydraulic similitude and must be used in place of ordinary coolant water. This in turn necessitates the coolant channel to be wider and as a consequence the traditional Aluminium cladding in research reactors should be replaced by Iron. The derived scale down criteria can be used for the design of fuel element for the out-of-pile testing.

Experiments have been carried out to study the heatup behavior of a submerged debris bed constituted of segmented reactor channels of Pressurized Heavy Water Reactors (PHWR). The study has been done for under fully submerged, partially submerged and nearly exposed conditions. This situation may arise from a severe accident scenario of PHWR where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure tube and calandria tube constitute each segmented reactor channels. Ten such reactor channels constitute the debris bed. Heatup of this debris bed is observed with respect to different water levels and at power levels of 10, 20 and 30 kW, typical to decay power levels of 0.25 %, 0.5 % and 0.75 % of reactor full power respectively. It has been observed from the set of experiments that fully submerged debris bed does not get heated up and fuel rod temperature remains at saturation temperature. Heatup is observed for exposed channels only, however the heatup rate is limited with steam and radiative cooling. For exposed debris bed, channels surrounded by neighboring channels get heated up as compared to channels at the periphery of the debris bed.

Rewetting of a hot surface is the process of establishing direct liquid contact with a large portion of the surface whose initial temperature exceeds that required to maintain film boiling for prescribed surface and flow conditions. The Advanced Heavy water Reactor (AHWR) is a natural circulation vertical pressure tube type boiling light water cooled and heavy water moderated reactor. In case of a loss of coolant accident, the clad surface temperature goes up very high and comes down due to coolant injection from the Emergency Core Coolant System (ECCS). The rewetting takes place in Boiling Water Reactors (BWR) due to top flooding and Pressurised Water Reactors (PWR) due to bottom flooding. But in AHWR, the emergency coolant enters into the reactor core in radial direction and after this, cross flow phenomenon takes place from one fuel pin ring to next. The study is being carried out on the effect of cross flow on rewetting of AHWR fuel bundle. This paper will discuss the modeling of the experimental setup having pressure tube, fuel cluster, steam generator, accumulator etc and study the effect of radial flow on rewetting of fuel pins. An analysis of the model, considering with and without cross flow, has been carried out and shows that the pick fuel temperature is sensitive to cross flow. The thermal hydraulic safety analysis code Relap5/3.2 is being used for modeling of experimental setup for rewetting study.

The present paper deals with evaluation of organic base nuclear grade anion exchange resins Amberlite IRN-78 and Purolite NRW-8000 by application of radioactive tracer isotopes 131 I and 82 Br. The evaluation was made on the basis of their performance in iodide and bromide ion-isotopic exchange reactions carried out under different experimental conditions like temperature and ionic concentration. It was observed that during iodide ion-isotopic exchange reaction at a constant temperature of 30.0 °C, as the concentration of labeled iodide ion solution increases from 0.001 mol/L to 0.004 mol/L, the percentage of iodide ions exchanged increases from 70.10 % to 77.10 % for Amberlite IRN-78, which was higher than an increase of 63.90 % to 71.00 % as obtained for Purolite NRW-8000 resins. Also at a constant temperature of 30.0 °C, using 1.000 g of ion exchange resins and 0.001 mol/L labeled iodide ion solution, the values of specific reaction rate (min −1 ), amount of iodide ion exchanged (mmol), initial rate of iodide ion exchange (mmol/min) and log K d were calculated to be 0.280, 0.175, 0.049 and 11.4 respectively for Amberlite IRN-78 resin, which was higher than the values of 0.258, 0.160, 0.041 and 10.6 respectively as that obtained by using Purolite NRW-8000 resins. The identical trend was observed for the two resins during bromide ion-isotopic exchange reaction. The results of present investigation also indicate that during the two ion-isotopic exchange reactions, for both the resins, there exists a strong positive linear correlation between amount of ions exchanged and concentration of ionic solution; and strong negative correlation between amount of ions exchanged and temperature of exchanging medium. The overall results indicate that under identical experimental conditions, Amberlite IRN-78 resins show superior performance over Purolite NRW-8000 resins.

In this study, the effect of spent fuel grade plutonium content on 239–243 Pu was investigated in a designed hybrid reactor system. In this system, the fluids were composed of a molten salt, heavy metal mixture with increased mole fractions 99 – 95 % Li 20 Sn 80 -1 – 5 % SFG-Pu, 99 – 95 % Li 20 Sn 80 -1 – 5 % SFG-PuF 4 , 99 – 95 % Li 20 Sn 80 -1 – 5 % SFG-PuO 2 . Beryllium (Be) is a neutron multiplier by (n,2n) reactions. Thence, a Be zone of 3 cm thickness was used in order to contribute to fissile fuel breeding between the liquid first wall and a 9Cr2WVTa ferritic steel blanket which is used as structural material. The production of 238–242 Pu(n,γ) 239–243 Pu was calculated in liquid first wall, blanket and shielding zones. Three-dimensional nucleonic calculations were performed by using the most recent version MCNPX-2.7.0 Monte Carlo code and nuclear data library ENDF/B-VII.0.

The nuclear reactions of the deuteron particles with light nuclei have been extensively investigated in the history of nuclear physics. In this work, the cross-sections of deuteron reactions 6 Li(d,n) 7 Be, 6 Li(d,t) 5 Li, 7 Li(d,2n) 7 Be, 7 Li(d,p) 8 Li, 7 Li(d,t) 6 Li, 9 Be(d,p) 10 Be, 9 Be(d,t) 8 Be, 19 F(d,n) 20 Ne and 19 F(d,p) 20 F were studied for the investigation of structural fusion materials. Excitation functions of these reactions were determined by using ALICE-ASH code considering equilibrium (EQ) and pre-equilibrium (PEQ) effects. The calculated cross-sections, experimental results and TENDL-2011 library data are presented in graphical form.

A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve.

The health impacts and corresponding damage costs of radioactive emissions of Yeniköy and Kemerköy lignite-fired power plants in Muğla have been assessed by using the simplified impact pathway approach. Radiation dose and risk calculations have been carried out by the code CAP88-PC around the power plants. Specific isotopes, 226 Ra, 232 Th, 40 K and 238 U in the flying ash samples are considered as radioactive sources. The estimated total collective doses around Yeniköy and Kemerköy power plants are 3.15 × 10 −4 man Sv/year and 3.77 × 10 −4 man Sv/year. Health effects and the corresponding damage costs around the power plants due to radioactive emissions from the power plants are negligible.