نشریه فناوری و انرژی هسته ای
سال دوم شماره 2 (پیاپی 6، تابستان 1402)

In this research, using Fortran programming language, a computational code is presented to investigate the effect of xenon concentration on the thermal conductivity of the gap between the fuel and clad. For this purpose, first the heat transfer model between the fuel and the coolant is simulated in which the coolant temperature is calculated using the enthalpy rise method and then the outer and inner surface temperature of clad, the gap temperature and the radial temperature distribution of fuel are calculated. After modeling the heat transfer, the thermal conductivity is investigated by adding xenon to the helium gas in the gap. The results of this study are highly consistent with FRAPCON 3.1 code.

Caramel UO2 fuel is a new type of fuel that is suitable for use in high power density research reactors with reduced fuel enrichment. Reducing the thickness and flatness of caramel fuels increases heat transfer and reduces the fuel temperature. The purpose of this study is to investigate the neutronic and thermal-hydraulics properties of caramel fuel. The MCNP nuclear code have been used to analysis the neutronic parameters of caramel fuel and compare it with the neutron parameters of Tehran research reactor. Also, ANSYS engineering software using finite element method (FEM), thermal-hydraulics and mechanical efficiency of caramel fuel has been evaluated. In this research, the important and measurable quantity of neutrons is the study of reactivity and effective multiplication factor of the core of the redesigned open reactor, and in the mechanical analysis of fuel, stress and strain phenomena are obtained. The results obtained from the open reactor designed with caramel fuel have been investigated with the results of Tehran reactor with U3O8-Al fuel. According to the results, it can be said that caramel fuels in terms of heat increase heat transfer and lower fuel temperature, and also in terms of neutrons with a lower percentage of enrichment than the fuel loaded in the core of the Tehran reactor to the desired and to arrives at the beginning of the cycle.

Success criteria analysis (SCA) is one type of deterministic analysis and in the case of the PSA level 1 framework, it is used in 3 areas. 1- determination of the end state for each sequence of the event tree, 2- deterministic confirmation of the fault tree analysis, 3- help to calculate the human error probability and identify its consequences. This research covers the 1st type of SCA. Success criteria as a deterministic analysis must be based on a realistic steady-state model of the plant, so the MELCOR code is used to model the plant which is a 2-loops Westinghouse PWR. When the steady-state model is achieved, the design basis accident i.e. MSLB is modeled. The event tree of the accident contains 12 sequences and it is needed to determine the end state for each sequence. Determining the end states through success criteria analysis is to express the exact condition of the NPP i.e., is it a safe end state or not, so two different criteria are used in order to identify the safe end state, the peak clad temperature must not exceed 1477 K or the pressure of the containment is not allowed to be increased more than the design pressure (0.4 Mpa). For mitigation of the MSLB accident consequences, two safety functions are required: 1- reactivity control and 2- heat removal by use of the main steam line isolation and emergency feed water in the secondary circuit. Both of the mentioned functions above are satisfied in the 1st sequence, but if there is a failure in the heat removal, the operator must start the primary feed and bleed process by use of the Safety Injection System (SIS) and Safety Depressurization Valve (SDV) in the primary circuit. According to the operation or failure of the required systems or functions, in some sequences, the above criteria are not met and it is not a safe end-state condition for the reactor. By implementing the success criteria analysis which is known as sequence analysis SCA, the end state for all sequences (except the 6th and 11th sequences that lead to Induced Steam Generator Tube Rupture (ISGTR) accident is determined.according to the results, none of the criteria above are violated in the 1st, 2nd, and 7th sequences and thus they lead to the safe end state. The design pressure criterion for the containment is not met in the 3rd and 8th sequences and there will be containment damage. The 4th, 5th, 9th, 10th, and 12th sequences lead to core damage as the peak clad temperature exceeds 1447 K

The various models have been presented by the most reliable space centers in the world to simulate the space radiation environment and its types of radiation on the earth. In addition, the various software, including SPENVIS and CREAME96, have been designed based on these models. In this article, a comprehensive review of the most important radiation models in the space environment is presented. These models are divided into three categories: trapped particles, solar particle events, and cosmic rays. The performance of each of these models has been compared and finally the best of them has been investigated to describe particles trapped in the radiation belt, particles originating from solar events and cosmic rays.

Experiments have been designed and performed using the IR-MPF-100 dense plasma focus device (40 kV , 144 µF , 115 kJ) . In these experiments, the charging voltage of the capacitors is 14 kV, the pressure of the deuterium working gas is 1.5 mbar. These experiments are designed to detect the emitted neutrons and deuterons and to measure the energy of neutrons and deuterons. By placing a plastic scintillator detector 2.5 m away from the pinch area and measuring time of flight of neutrons, the neutron beam energy was measured in the range of 2 – 2.52 MeV. In order to measure the energy of the deuterons, an aluminum target is placed in front of the anode. When ions collide to the target, x-ray is emitted and the emitted x-ray is detected. Time interval between the anode X-ray (first signal) and the target X-ray (second signal) results deuterons time of flight (TOF). In this research, ions energy was measured in the range of 56 – 462 keV.

Incident analysis is one of the most important and complex processes in assessing the safety of nuclear reactors. The purpose of this study is to evaluate the dynamic behavior of the Isfahan Research Miniature Neutron Source Reactor during the reactivity injection incident. In this paper, using the capabilities of the RELAP5 code, the core behavior of the reactor during the reactivity injection incident is evaluated. Finally, the results have been compared with references in order to verify the information with high accuracy. Results showed that due to the injection 3.6 (mK) reactivity, the power with a very high slope begins to increase so that after 268 seconds the power to about 100 (kW) (99.9 (kW)). Also, due to the injection of 6.05 (mK) reactivity a sudden rise in power will occur.