Volume 82, Issue 6

The Condensate and Feedwater System of the Chinshan BWR units is used to provide reliable and high-quality water to maintain the reactor water level during operation. If a Loss of Feedwater Heating (LFWH) event occurs, the core inlet subcooling increases and then induces corresponding power excursion and the reactor pressure rise. In the Chinshan Final Safety Analysis Report (FSAR), a loss of the feedwater temperature of 55.6°C (100°F) is conservatively assumed in the event. This study analyzes the integral reactor system response with RETRAN. Furthermore, a partial vessel model (PVM) of CFD is used to acquire the conditions of the fuel channel inlet to compensate the weakness of the RETRAN system model to generate detailed thermal-hydraulic conditions. The evaluation shows that the feedwater temperature drop is about 40°C which is lower than the FSAR value. In addition, the sensitivity study shows that the hot channel method underestimates the ΔCPR about 0.025, and there is no direct relation between ΔCPR and either of inlet subcooling or power fraction in transient, which is quite different from the conclusion of the hot channel method. Finally, the sensitivity study also proves the ΔT of 55.6°C (100°F) used in FSAR analysis conservative enough to cover the worst channel with a margin of 0.015 in ΔCPR.

Saturated boiling of two-phase natural circulation has been experimentally investigated based on a natural circulation device with narrow rectangular channels. When heating power reaches a certain range, it is possible to observe the phenomenon of saturated boiling and flow pattern transition in the system. The results show the heat transfer coefficient of saturated boiling decreases with the increasing of pressure, heating power and size of narrow rectangle channels. The buoyancy force causing mixed convection decreases the heat transfer coefficient. Finally, a dimensionless number is introduced, which reflects length to width ratio of rectangular narrow section and Rayleigh number, in order to revise the presented correlation. All errors fall within the range of ±15%.

Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k eff ) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

The neutron physics characteristic of Pebble Bed Reactor (PBR) allows a better incineration of plutonium (Pu). An optimized design of simple PBR might give a symbiotic solution of providing a safe energy source, effective fuel utilization shown by a higher burnup value, and incineration of Pu stockpiles. This study perform a preliminary neutronic design study of a 200 MWt Once Through Then Out (OTTO) cycle PBR with Pu-based fuel. The safety criteria of the design were represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. In this preliminary phase, the parametric survey is limited to the heavy metal (HM) loading per pebble and the average axial speed of the fuel. An optimum high burnup of 419.7 MWd/kg-HM was achieved in this study. This optimum design uses a HM loading of 2.5 g/pebble with average axial fuel velocity 0.5 cm/day.

Although radiation exposures in manned space missions are normally below the limits recommended to NASA by NCRP, in long-duration deep space exploratory missions astronauts may receive relatively high doses of ionizing radiation. Novel light polyethylene-based composites can be considered as effective radiation shields in space explorations. However, normally these composites cannot provide desired mechanical properties. Over the past several years our laboratories have focused on developing efficient methods for both physical and biological protection of the crew in long term space missions. In this study carbon nanotubes and either nano-sized or micro-sized boron carbide (B 4 C) fillers were incorporated into the continuous phase of low density polyethylene (LDPE). In the next phase, the mechanical characteristics of the composites as well as their neutron attenuation properties were studied. Findings of this study indicated enhanced mechanical properties accompanied by an enhanced shielding efficiency for neutrons at some specific weight fraction of the fillers.

After spent fuel assemblies have been discharged from the reactor, reactivity will fluctuate as the cooling period progresses because of changes in the number density of fissile nuclides and neutron absorber nuclides. The purpose of this project was (1) to quantify the contribution of each individual nuclide to the reactivity of the fissile system, (2) to identify nuclides that are responsible for the fluctuation in reactivity, and (3) to determine the effect of the number of nuclides on reactivity. This paper will present the results of the study of the behaviour of the k eff with respect to variation in the duration of the cooling period during storage.

Regarding the low thermal conductivity of nat Te element, the provision of an effective cooling system is one of the critical issues in cyclotron targetry to prevent melting of target matter during the irradiation to 124 I production via nat Te(p,xn) 124 I reaction. Heat transfer on Te target and efficiency of cooling fluid in the solid target system have been simulated based on a Finite Element Analysis (FEA) code for the thermal behavior of the target during the irradiation and under different beam currents, coolant flow rates, substrate matters and target geometry. The results on the routinely used solid target in Radiation Application Research School (RARS) cyclotron showed that in a 3 m/s coolant flow rate, by using a fined-cooling system and a nickel substrate coated on copper backing plate, the irradiation beam current can be raised up to 180 μA without any risk of melting. The cooling flow rates greater than 3 m/s do not noticeably improve the heat dispersion of target layer. As expected, a linear increase was observed for the temperature and temperature gradient of plates in the beam currents of 100–300 μA.

The radiation hazards of radionuclides in the air arising from the storage room of nuclear devices to the operators cannot be ignored. A new ultrasonic-resonator-combined method for purifying nuclear aerosol particles is introduced. To remove particles with diameters smaller than 0.3 μm, an ultrasonic chamber is induced to agglomerate these submicron particles. An apparatus which is used to purify the nuclear aerosol particles is described in the article. The apparatus consists of four main parts: two filtering systems, an ultrasonic chamber and a high-pressure electrostatic precipitator system. Finally, experimental results demonstrated the effectiveness of the implementation of the ultrasonic resonators. The feasibility of the method is proven by its application to the data analysis of the experiments.

In this paper we present a new method of calculating reactivity with fluctuation reduction. First we propose the generalized predictor and corrector using the fourth order Adams-Basforth-Moulton (ABM) for the numerical solution of the point kinetic equations for the calculation of reactivity, without using the history of nuclear power. Due to the nature of point kinetic equations, we use modifiers of the different predictors to increase precision in the approximation obtained. Secondly, we use the filter known as Savitzky-Golay (SG), which permits the reduction of the fluctuation in reactivity. It is known that the SG filter smoothing without diminishing the value of the nuclear power, irrespective of its form; this guarantees the reduction of Gaussian random noise levels, which is distributed around the average value of the nuclear power of up to σ = 0.1, with a time step h = 0.01 s. This formulation uses the Gram polynomial approximation, with a degree d = 2 the results show better values in the maximum difference in reactivity in comparison with that reported in literature.

This article describes the need for ontology matching and describes the methods to achieve the same. Efforts are put in the implementation of the semantic web based knowledge management system for nuclear domain which necessitated use of the methods for development of ontology matching. In order to exchange information in a distributed environment, ontology mapping has been used. The constraints in matching the ontology are also discussed. Pareto based ontology matching algorithm is used to find the similarity between two ontologies in the nuclear reactor domain. Algorithms like Jaro Winkler distance, Needleman Wunsch algorithm, Bigram, Kull Back and Cosine divergence are employed to demonstrate ontology matching. A case study was carried out to analysis the ontology matching in diversity in the nuclear reactor domain and same was illustrated.

ATHLET-CD is the severe accident module of the code system AC 2 that is designed to simulate the core degradation phenomena including fission product release and transport in the reactor circuit, as well as the late phase processes in the lower plenum. In case of a severe accident degradation of the reactor core occurs, the fuel assemblies start to melt. The evolution of such processes is usually accompanied with the failure of the core support plate and relocation of the molten core to the lower plenum. Currently, the criterion for the failure of the support plate applied by ATHLET-CD is a user-defined signal which can be a specific time or process variable like mass, temperature, etc. A new method, based on FEM approach, was developed that could lead in the future to a more realistic criterion for the failure of the core support plate. This paper presents the basic idea and theory of this new method as well as preliminary verification calculations and an outlook on the planned future development.

The infinite medium Green function can be written by using the jump condition, found Case's eigenfunctions. Thus, any reactor theory problem which is inplane geometry such the criticality problem as can be investigated by using the proper boundary conditions and suggested flux definitions. By using the criticality equation the critical thicknesses can be calculated as numerically. The selected numerical results can be tabulated.