[en] In this project, a Neutron Flux-Trap intended for 5MW Tehran Research Reactor was designed. Fuel conversion from HEU to LEU in research reactors usually deprives the core from the high neutron flux. Therefore one has to look for a remedy in such situations otherwise radioisotope production, especially for those neutron demanding ones, falls down dramatically. The initiations of Neutron Flux-Trap idea comes true to face up this problem and provide an appropriate place inside or outside the core with sufficient neutron flux higher than the normal average level. To implement such a design, a number of codes and calculational tools have been used. At preliminary stage, WIMSD and EXTERMINATOR-II were used and then at later stages CITATION and MCNP codes were used for final design. Furthermore, SAND-II and ORIGEN were also employed for spectral analysis and radioisotope production calculations. Good agreements resulted with experiments and especially the case that Neutron Flux-Trap filled with ordinary water

[en] Separation of ¹³¹I from natural uranium fission product mixtures has been accomplished by sorbing the ¹³¹I on special platinum-charcoal sorbent and desorbing by buffer solution (NaHCO₃+Na₂S₂O₃). High radiochemical and chemical purity is obtained by this method. Important parameters such as temperature, distillation rate, sorbing and desorbing rates and ¹³¹I separation yields have been investigated

[en] Neutron transmutation doping of silicon (NTD) is one of the important applications in industrial utilization of research reactors. In this paper, the proposed irradiation system based on using a reflecting environment around the irradiation channel of the NTD facility was designed and simulated for irradiation of silicon in the Tehran research reactor (TRR) by using codes such as MCNP, WIMS and CITATION. This system is very simple and does not require conventional methods, such as continuous movement and use of absorbing materials and windows, for flattening of neutron flux in the radial and longitudinal direction. The influence of type and dimensions of reflectors on the irradiated silicon crystal was investigated and also the optimum conditions were determined for TRR for silicon ingots with diameter up to 5 cm and up to 10 cm

[en] The purpose of this paper is to develop a nonlinear model to investigate the instabilities of a two-phase natural circulation loop. The natural circulation loop consists of several components, i.e. a channel test section, a steam separator, an upper horizontal section and an adiabatic downcomer. Moreover, each channel consists of a lower horizontal section, single-phase standpipe, heated section and riser. We obtain a stability map to explore the unstable regions of this natural circulation loop. The results show that the considered loop has two unstable regions, instability type-I in the low power region and instability type-II in the high power region. Then the parametric study is carried out to understand the relation between the parameters of system and two types of instability. The parametric study reveals that lengthening the riser has an unstable effect on system stability. Thus, lengthening the riser causes a reduction in the stability region in the both low power and high power levels. Also it can be observed that by increasing the form loss coefficient at the inlet of heated section or in the downcomer section, the stability region expands, however by increasing the form loss coefficient at the outlet of heated section or in the upper horizontal section, the stability region decreases consequently

[en] Utilization of irradiation facilities in Tehran research reactor (TRR) requires proper computational tools to deliver accurate and precise results. In this paper validity of different schemes are checked against experimental measurements. A reference core with a neutron flux trap in the middle is chosen as the reference model for this purpose

[en] An anodic electrophoretic deposition (EPD) technique is used to create a uniform TiO₂ thin film coating on boiling thin steel plates (1.1 mm by 90 mm). All of the effective parameters except time of the EPD method are kept constant. To investigate the effect of gamma irradiation on the critical heat flux (CHF), the test specimens were irradiated in a gamma cell to different doses ranging from 100 to 300 kGy, and then SEM and BET analysis were performed. For each coated specimen, the contact angle and capillary length were measured. The specimens were then tested in a boiling pool for CHF and boiling heat transfer coefficient. It was observed that irradiation significantly decreases the maximum pore diameter while it increases the porosity, pore surface area and pore volume. These surface modifications due to gamma irradiation increased the CHF of the nano-coated surfaces compared to that of the unirradiated surfaces. The heat transfer coefficient (HTC) of the nano-coated surfaces irradiated at 300 kGy increased from 83 to 160 kW/(m² K) at 885 kW/m² wall heat flux by 100%. The CHF of the irradiated (300 kGy) and unirradiated surfaces are 2035 kW/m² and 1583 kW/m², respectively, an increase of nearly 31%

[en] Determination of peaking factor, normally requires wire-scanning or gold activation techniques. These techniques are rather awkward and slow. A more versatile technique based on miniature neutron detector (MND) is developed and used for Tehran research reactor. (orig.)

[en] Highlights: • A Fuel Burnup Measurement System was designed and developed in TRR using a non-distractive gamma-ray spectroscopy method. • The irradiated fuel burnup was measured in TRR through absolute activity determination of ¹³⁷Cs indicator. • The operating histories of the irradiated fuel assemblies of TRR were obtained carefully. • Results from reactor physics calculations were verified using non-distractive gamma-ray spectroscopy. In this work, a nondestructive gamma scanning technique has been applied to determine the irradiated fuel burnup of Tehran Research Reactor (TRR). Thereby, a system was designed and installed at the rim of the TRR pool, which includes a High-Purity Germanium (HPGe) detector and associated electronics-devices, a proper shield, a collimator, and an elevator to radioactive fuel handling for any longitudinal and transverse fuel movement. In the system, it also was possible to measure fuel burnup for fuels with short cooling times at the minimum distance between the fuel and the detector. Five Standard Fuel Elements (SFEs) have been studied with U3O8Al fuel in burnup range of 14%–60% FIMA and cooling time range of 60 days–550 days. Then, by analyzing the gamma-ray emitted from 137Cs isotope as a fuel burnup indicator, an axial profile of fuel burnup was measured in the active fuel length of 61.5 cm and an active width of 6 cm. Finally, verification of results from reactor physics calculations was conducted based on fuel burnup measurements using a nondestructive gamma scanning technique that represents a well enough agreement between calculations and measurements.

[en] Highlights: • The transient heat transfer equation of an annular fuel is solved numerically using the orthogonal collocation method (OCM). • The OCM results were validated by comparison with FVM, FDM and analytical method. • The computational efficiency of the schemes were compared in terms of CPU time and accuracy. • The OCM with fewer nodes is much faster than the FVM and FDM with comparable accuracy. - Abstract: In this paper, radial temperature distribution in an annular fuel rod is calculated using the orthogonal collocation method (OCM) and the finite volume method (FVM). Using the non-dimensional transient heat transfer equation and defining the relative radius of the nodal points inside each solid region, the annular fuel rod is transformed into the interval [0,1]. By doing this, one can use the Legendre polynomial roots in the interval [0,1] as the collocation points. In order to solve the system of equations, the boundary conditions appropriate to different discretization schemes were applied. To validate the proposed scheme, results were compared with those reported in the literatures, finite different method (FDM) and analytical solution. It was found that although FVM and FDM are simpler to implement, the OCM with fewer nodes is much faster than the FVM for comparable accuracy.

[en] Highlights: •We assessed the possibility of fuel test at TRR core from thermal-hydraulic point of view. •We applied standard codes and developed complementary computer programs for the study. •Safety related thermal-hydraulic parameters were studied during the irradiation experiment. -- Abstract: Following domestic fabrication of nuclear fuels in Iran, it is necessary to investigate fuel material behavior, fission gas release, fuel swelling, cladding material behavior and fuel integrity of domestic fuels at different burnup in a research reactor during irradiation. Currently, Tehran research reactor is the sole operating research reactor which can be used for fuel irradiation experiments in the country. In this regard, standard codes as well as developed complementary computer programs are applied to verify thermal-hydraulic performance of irradiating a domestic rod-type fuel assembly of natural UO₂ pellets in Tehran research reactor core, which itself contains 20% enriched plate-type U₃O₈–Al fuels. Maximum temperatures of fuel, clad and coolant, onset of nucleate boiling, onset of flow instability and departure from nucleate boiling during irradiation experiment are investigated by subchannel analysis as indicators to verify the reactor core safe operation during the experiment. The results give the confidence that during this irradiation experiment, thermal-hydraulic steady state safety criteria of the mixed-core are satisfied and the fuel irradiation experiment does not induce any significant operational change.