Tehran Research Reactor Research Articles

Evaluation of U3O8–Al, UO2–Zr, and U3Si2–Al fuel plates in a hypothetical melting accident of Tehran research reactor (TRR)

Tehran research reactor (TRR) is a material test reactor (MTR) that uses U3O8–Al fuel plates. Fuel plates are usually used in research reactors due to their great thermal characteristics. The occurrence of various accidents in nuclear reactors has increased the efforts to develop accident tolerant fuels (ATF). For this reason, intermetallic compounds of uranium and silicon, have been introduced as candidates for ATF. In this study, the heat transfer and melting of fuel plates composed of U3O8–Al, UO2–Zr, and U3Si2–Al, in normal operation and hypothetical heat transfer impairment in the TRR have been studied. These compounds are chosen as candidates for probable replacement in the new design of the TRR. In this work, two-dimensional equations of heat transfer and enthalpy using the finite difference method have been solved and solid to liquid phase changes are discussed. According to the findings of this research, during a hypothetical heat transfer impairment, UO2–Zr fuel undergoes a phase change from solid to liquid after a longer period of time compared to the other two fuels. This delay is attributed to the relatively high melting point of UO2–Zr, which is notably higher than the melting points of the other two fuels. It has been shown that when the focus is on transferring heat from the fuel plate to the coolant, U3Si2–Al performs better than the other two types of fuel due to its higher thermal conductivity and relatively good melting behavior.

Gross alpha/beta and radionuclide activity concentrations in soil, plant and some fruits around the Tehran Research Reactor

Human activities usually have some contamination as effluents from chemical industries and radionuclides from nuclear reactors. For assessing the probable radioactive contamination in vicinity of Tehran Research Reactor, The gross alpha and beta radioactivity concentrations in soil, pine and cedar leaves and some selected fruits (fig, apple, berry and pomegranate) were investigated using an alpha/beta spectrometer during 2021–2022. Also, the concentrations of artificial and natural radionuclides in samples were investigated by the method of gamma spectroscopy. The gross alpha activity concentrations in soil, pine and cedar leaves and some selected fruits samples are from 0.05 to 0.35 Bq/gr and 0.07–0.31 Bq/gr and 0.04–0.18 Bq/gr, respectively. The gross beta activity concentrations in soil, pine and cedar leaves and some selected fruit samples are from 0.73 to 4.25 Bq/gr and 0.21–3.97 Bq/gr and 1.01–2.71 Bq/gr, respectively. Average activities concentration of natural radionuclide 232Th, 238U and 40K in soil, pine and cedar leaves and some selected fruits are 31.89–16.23–582.73 Bq/kg and 1.84–0.99–84.60 Bq/kg and 1.98–1.09–72.08 Bq/kg respectively. From artificial radionuclides, just 137Cs is recognized in soil sample and the range of 137Cs concentration in surface soils was observed to vary in the range 0.85–2.21 (Bq/kg). The result showed that the Tehran Research Reactor activities not have increased the environmental radioactivity and radiation level in the area.

Design of emergency solar energy system adjacent the nuclear power plant to prevent nuclear accidents and increase safety

Providing sufficient energy in emergency situations in nuclear power plants is of great importance due to their crucial role in improving power plant safety and eliminating disasters.Solar power is capable of providing a secure energy source which could be used for emergency power loads. The main goal of this study is application of solar energy for providing emergency loads of the Tehran research reactor. A 100 kW solar power plant connected to the grid and 400 kWh battery pack which is capable of providing 50 kw loads for 8 h is designed. As the first priority, the plant charges the battery pack and discharges excess energy to the power grid. Therefore, the proposed scheme, in addition to increasing safety, also has economic benefits. Design procedure of the power plant was carried out using PVsyst which is international well known software for designing solar power plants and results were investigated. Economic evaluation of the project is carried out using RETSCREEN software and different schemes are evaluated. Due to the critical nature of the power loads and their sensitivity, application of this scheme for providing emergency energy loads of the Tehran research reactor is suggested. Considering the expansion of nuclear safety and the protection of the environment and humanity, we request the respected editors and reviewers of the journal to have positive opinions about the publication of the article.

Neutronic and thermal-hydraulic assessment of the TRR with new core designed based on tubular fuels

Abstract Herein, the feasibility study of the Tehran Research Reactor (TRR) with a new core designed based on tubular fuels from the neutronic, thermal-hydraulic, safety, and operational points of are investigated using MCNPX, WIMS, CITATION, Computational Fluid Dynamics (CFD), and RELAP codes. According to the results, the total neutron flux in the new core with tubular fuels is increased by more than 14.3 % compared with the current core of the TRR with plate-type fuels. Moreover, due to the higher fuel amount in the tubular compared with plate-type fuels (about 17 % in similar conditions), its effective multiplication factor is much higher than the TRR with plate-type fuels. Moreover, the results show that the maximum cladding temperature is sufficiently lower than 105 °C and the produced heat in the tubular fuel are removed without changing the current flow rate of the core. Furthermore, the maximum fuel temperature in tubular fuel is about 10 °C lower than the maximum fuel temperature in the current standard fuel element.

New insights to containment management in TRR during severe accident by spray and fan cooler measures

The corium behavior and its effects on containment behaviors along with mitigation measures are investigated for Tehran Research Reactor (TRR) by using the MELCOR and CONTAIN codes during a Severe Accident (SA). According to obtained results, the SAs analysis for the research reactor shows different behavior compared with power reactors. This behavior can be can be investigated in two phases; the initial phase: before the concrete ablation initiation up to 119 days and secondary phase: initiation concrete ablation from 119 to 125 days. During the initial phase, the containment pressure increases at a rate of 0.016 Pa/s and reaches 0.1138 MPa (the maximum pressure of the containment) during about 24 h. In this phase, the effects of the corium initial temperature, corium mass and relative humidity on the containment behavior are not significant. During the secondary phase, the average containment pressure increasing rate is about 12 times more than initial phase. Results show that containment management systems are need to maintain containment integrity. The suggested systems are to use a fan cooler (minimum power capacity 3 kW, coolant flow rate and temperature 0.05 Lit/s and 298 K) or spray (maximum spray droplet temperature 298 K with minimum spray flow 0.001 Lit/s) to mitigate SA.

Production assessment of 195mPt in Tehran research reactor

Production assessment of 195mPt in Tehran research reactor

Investigating the effect of data acquisition geometry and image reconstruction method on neutron computed tomography images at Tehran Research Reactor

تصویربرداری مقطع‌نگاری نوترونی یکی از کاربردهای مدرن رآکتورهای تحقیقاتی در سرتاسر جهان به‌شمار می‌رود. تلفیق قابلیت نمایش سه‌بعدی تصویربرداری مقطع‌نگاری با ویژگی‌های منحصر به فرد برهم‌کنش نوترون با مواد، می‌تواند اطلاعات بسیار ارزشمندی از ساختار داخلی مواد و تجهیزات را در اختیار محققین قرار ‌دهد. در این تحقیق، تأثیر هندسه داده‌برداری تجربی و روش بازسازی تصویر بر کیفیت تصاویر مقطع‌نگاری نوترونی سامانه تصویربرداری رآکتور تحقیقاتی تهران براساس شاخص وضوح و میزان تولید نویز مورد مطالعه قرار گرفته است. در مرحله داده‌برداری تجربی، دو نمونه مورد آزمون در دو هندسه با چرخش نیم‌صفحه‌ای با نموِ زاویه‌ای ‌5/0 و 1 درجه توسط باریکه نوترون پرتودهی شده‌اند. تصاویر نمونه با استفاده از الگوریتم‎های بازسازی تصاویر مقطع‌نگاری شامل روش‌های تحلیلی FBP و روش‌های تکرارشونده ART، SARTو SIRT به دست آمده‌اند. تصویر حجمی نمونه با روی هم قرار دادن پشته‌ای از مجموعه 1670 تایی از تصاویر مقطعی حاصل شده است. از زبان برنامه‌نویسی پایتون برای پردازش و بازسازی تصویر از افکنش استفاده شده است. برای حصول افکنش‌های با کیفیت مناسب برای بازسازی تصویر، ابتدا مشخصات سیستم داده‌برداری تجربی بهینه‌سازی شده‌اند. سپس، با استفاده از پیش‌پردازش، کیفیت تصاویر افکنش‎ها افزایش یافته است. براساس نتایج، اجزای نمونه مورد آزمون و عیوب موجود در آن در تمام تصاویر، از یک‌دیگر قابل تمایز می‌باشند. مقدار وضوح و آهنگ وضوح به نویز در هندسه با نمو زاویه‌ای کم‌تر و با استفاده از روش SIRT بهبود یافته است.

Evaluation of the produced 177Lu-specific activity via direct and indirect methods using Tehran Research Reactor

در پژوهش حاضر به بررسی امکان تولید رادیونوکلید لوتسیوم-177، Lu177، در رآکتور تحقیقاتی تهران، TRR، پرداخته شده است. Lu177 با داشتن ویژگی‏های واپاشی هسته‏ای مناسب و هم‏چنین رفتار شیمیایی مطلوب، یک رادیونوکلید ایده‏آل جهت تهیه رادیوداروهای درمانی است. برخی از رادیوداروهای حاوی Lu177 در حال حاضر در کشور در درمان انواع سرطان استفاده می‏شود و بسیاری از آن­ها تحت آزمایشات بالینی هستند. رادیونوکلید Lu177 را می‏توان یا مستقیماً از طریق واکنش Lu177Lu(n,γ)176 یا به طور غیرمستقیم طی واکنش تولید کرد. بازده پرتودهی Lu177 در هر دو مسیر تولید به صورت تجربی در 14 روز تابش در شار نوترونی حرارتی 1-s 2-cm 1013×5، تعیین شده و با محاسبات نظری مربوطه مقایسه شد. اثر شار نوترون حرارتی بالاتر و هم‏چنین کاهش درصد غنی­سازی Yb176 بر تغییرات اکتیویته / اکتیویته ویژه حاصل، با استفاده از نرم­افزار متلب به صورت تئوری مورد بررسی قرار گرفت. مشخص شد در هر دو روش تولید، شار نوترون بالاتر منجر به تولید اکتیویته / اکتیویته ویژه بالاتر می‏شود. با این‏حال، هیچ تأثیری بر سرعت افت اکتیویته ویژه طی دو هفته پس از پایان پرتودهی ندارد. به‏علاوه استفاده از ایتربیم-176 با غنای پایین‏تر در روش غیرمستقیم، منجر به کاهش اکتیویته ویژه سریع­تر طی دو هفته پس از پایان پرتودهی می ­شود.

The design and construction of a collimator holder to equip beam tube D of the Tehran research reactor

The continued safe and efficient operation of a research reactor depends on the behavior of the structural materials making up its main components. Many of these materials are difficult to replace in view of hard-to-reach locations after installation or even storage after years of use due to radiation from high exposure to neutrons inside the research reactor structure. The neutron collimator is a major part of the beam-related devices used in a research reactor. This device is installed in the beam tube via a holder. The collimator holder used as a structural component of beam tube D in the Tehran Research Reactor (TRR) has been made of cast iron. This component has been installed inside the beam tube and subjected to irradiation for 30 years. As a result, it has become severely damaged and must be transferred to radioactive waste storage. A necessary measure to recommission beam tube D in TRR is selecting a replacement with higher quality. Suitable candidates for this purpose would be materials that retain the neutron beam quality in addition to being adequate for long lifetime management. In this case, lower induced activity, shorter half-life of produced radioisotopes, and shorter cooling time are desirable. The present work used ORIGEN2 and MCNPX2.6 computational codes to compare various materials, such as barite concrete (type BA), Portland iron concrete, and stainless steel (316), in terms of the induced activity for various reactor operation periods, radiation level for various cooling periods, and neutron and photon beam parameters at the beam tube nozzles. The results indicated that replacing cast iron with Barite concrete (type BA) retains the quality of the neutron beam. In addition, 11.2, 2.81, and 3.63 times more activity was induced in stainless steel, Portland iron concrete, and cast iron, respectively, compared to type-BA barite concrete after 30 years under identical conditions. Since barite concrete is economical and simpler to manufacture, it was a suitable replacement for cast iron as the material for the collimator holder in beam tube D in TRR. In addition, Barite concrete holder was constructed based on the calculations made in this research. The results showed a good agreement between dosimetry experimental data and calculation results with a 3 % error.

Uncertainty analysis of a typical research reactor performance for radioisotope production

The accurate neutron flux in the different regions of irradiation positions is very important parameter in any research reactor performance with highly effects on uncertainty of produced activities. The produced activities in a research reactor core is highly dependent on 3 main groups of activity measurement, reactor and sample. The significance of these parameters is comprehensively investigated and determined numerically for a Tehran Research Reactor (TRR) as a typical research reactor. This is done using precise MCNPX 2.6.0 simulations, several experiments during reactor operation, and measurements in laboratory. The sample’s location uncertainty is one of the main parameters in the neutron flux uncertainty. Further, shadowing of samples and control rods, samples density, self-shielding, and also power measurement mechanism could be noticeable and give rise to distinctive uncertainty in the final results. It is determined that all the considered parameters could lead up to 17.8% uncertainty in the final results at the TRR which is remarkable.

Feasibility study to increase the reactor power at natural convection mode in Tehran Research Reactor (TRR) through a hybrid thermal-hydraulic simulation and analysis using the RELAP5 code and Computational Fluid Dynamic (CFD) modeling by ANSYS-FLUENT

Some irradiating applications can be used economically at a low power range and energy consumption. Moreover, removing the decay heat by the natural convection plays a very crucial role in Design Basis Accidents (DBAs) such as Loss of Flow Accident (LOFA) or Station Black Out (SBO). In such cases, the reactor can be operated by natural convection instead of forced convection. Such an operation needs careful thermal-hydraulic analyses to confirm nuclear safety. In this paper, Tehran Research Reactor (TRR) is chosen as the case study. A careful hybrid methodology is taken into account for safety analyses and visualizing the fluid streamlines and temperature profiles assuming both short-term and long-term cooling terms and conditions. Firstly, the RELAP5 code is used for the relevant simulation and safety analysis. Moreover, it is also used to trace any perturbation or oscillation and for the hot spot responses due to the flow regime changes and the reactor dynamic. Then the ANSYS-FLUENT is used for further Computational Fluid Dynamic (CFD) simulation of natural convection flow in detail. A modified 3-D porous media is modeled to simulate and observe the overall behavior of the reactor core in the reactor pool for long-term operations. The Boussinesq approximation is assumed for the buoyancy force. The pressure drop along the porous media is replaced with an accurate model of the Standard Fuel Element (SFE). Moreover, the k- ω turbulence model is also included in the Navier-Stokes equations to fix approximations. The relevant correlations have been also chosen for the heat transfer from the reactor pool to ambient air by convection, radiation, and surface evaporation. Finally, an accurate 2-D model of the hot-channel is also used for the accurate single-channel analysis regarding short-term operating conditions. Presented velocity vectors and temperature profiles depict the domain and behavior of the natural convection in both fuel channels and the reactor pool as well. Concerning both short-term and long-term issues, results are quite promising to increase the maximum operating limit of the natural convection mode from 100 kW to 200 kW. Moreover, regarding short-term operations or emergency conditions, this limit can be also increased to 400 kW due to the safety limit of 378 K for the maximum permissible fuel cladding temperature to protect the fuel from long-term corrosions. Keeping this limit also prohibits any perturbation or oscillation as well.

The design and optimization of a radiation shield for the Tehran Research Reactor's neutron diffraction facility

In this paper, a novel optimal radiation shield for the Tehran research reactor's neutron powder diffraction facility was designed to reduce undesirable neutron and gamma dose rates. A layer-by-layer methodology was considered based on the design approach to achieve a dose rate of less than 30 μSv/hr. MCNPX was used for this purpose as a Monte Carlo radiation transport code. A combination of neutron and gamma absorbers and moderators was employed to create this shield. The results indicated that the shield's six distinct layers met the design objective, totaling 65 cm in thickness. This newly designed shield comprises 15 cm paraffin, 15 cm Portland iron concrete, another 15 cm paraffin, 5 cm solid boric acid, another 10 cm Portland iron concrete, and 10 cm lead. The neutron and photon dose rates are 9.35 and 12.81 μSv/hr, respectively, which is one-sixteenth and one-third of the current facility shield. Validation of the simulation was accomplished by comparing the results for the current shield to experimental data.

Study of the 199Au nanoparticles production parameters via irradiation of platinum target by using thermal neutrons

In this study, the production parameters of 199Au nanoparticles (199AuNPs) have been investigated by a two-part study. The first part is about investigating the indirect method of producing non-carrier-added (NCA) 199Au radionuclide. MCNPX-2.6, TALYS-1.9, and ALICE/ASH-0.1 codes were applied as the theoretical approach to simulate the core of Tehran research reactor (TRR) for determining the activity of 199Au, specifying the production yield of 199Au, and calculating the excitation function of P198t(n,γ)P199t→A199u reaction. As the corresponding experimental approach, two 11 mg and 15.5 mg samples of enriched 198Pt metal powder were irradiated by thermal neutrons for 21 h and 10 min. The liquid-liquid extraction (LLX) technique has been used with a different solvent for each sample. LLX using ethyl acetate and LLX using Di-(2-Ethylhexyl) phosphoric acid (HDEHP) were applied for the 15.5 mg and the 11 mg samples respectively. The chemical yield of 199Au was calculated more than %99 for the 15.5 mg sample, and more than %80 for the 11 mg sample. The second part is about synthesizing 199AuNPs in an average size of 50 nm by using Turkevich method.

Evaluation of fast fission factor in a typical pool type research reactor

One of the important factors of a nuclear reactor core is the fast fission factor. This paper calculates this parameter based on space and energy-dependent method using the PTRAC card of MCNPX code. Tehran research reactor (TRR) is taken as a case study, and the parameter analyses are performed on the reactor core. Fast fission factor in TRR is evaluated regarding temperature effect, control rod positions, and fuel assembly positions. Using the PTRAC card, helpful information on fast fission factors is achieved throughout the reactor core. One MCNPX runs to return a data file about neutron interaction that can be analyzed many times in different manners to reveal this useful information. The method is simple and can be applied to any nuclear reactor core. The results obtained by this method can help nuclear reactor designers and nuclear reactor fuel managers to have a precise evaluation of the parameter. The method proposed in this paper for fast fission factor calculation is compared with the results previously published in the literature.

Collimated neutron beam design for TRR thermal column

Tehran Research Reactor (TRR) is the main neutron source in Iran which can be used for different applications of neutrons such as neutron radiography and neutron therapy. TRR has a thermal column which can provide high intensity flux of thermal neutrons for users. The aim of this study is to design a neutron collimator for TRR thermal column to produce parallel neutron beam with suitable intensity of thermal neutrons. To achieve this goal, Monte Carlo code of MCNX has been used to evaluate different configurations, geometries and materials of neutron collimator. The results show that the final selected configuration can provide a uniform thermal neutron beam with a flux of 1.21E+13 (cm-2·s-1) which is suitable for many different neutron applications.

Effect of certain metrological parameters on the extension of urgent protective action zone in Tehran research reactor accidents

In a core melting accident of a reactor, the release of radioactive material into the environment is inevitable. The amount of effective dose received by people around these facilities and consequently the urgent protective action zone (UPZ) strongly depend on the climatic conditions at the time of the accident and afterwards. In this study, the effect of wind speed, air stability class, temperature, and rainfall on the total effective dose equivalent (TEDE) at different time/place intervals of a hypothetical partial core melt accident around a research reactor was investigated using simulation method. The results showed that the stability class E, warm temperature, and light rain could lead to the highest dose of individuals. In addition, very low wind speeds were recommended for the facilities located in cities with high density of high-rise buildings and/or tropical areas in this regard.

Gamma dose rate determination of TRR irradiated fuel assemblies

Determination of gamma radiation dose-rate emitted from the irradiated nuclear fuel and prediction of the resulting dose rates are essential issues in the context of the design and construction of transport and storage casks for the spent nuclear fuels. This paper aims to evaluate the dose rate of gamma radiation emitted from the irradiated nuclear fuel in the Tehran Research Reactor (TRR) through different computational and measurement methods. First, four irradiated fuel assemblies with different operating histories have been taken into account, and the dose rate of emitted gamma radiation has been measured using Amber dosimeters at different distances from the fuel assembly. Then, an axial profile of gamma radiation dose-rate has been measured by placing 10 Amber dosimeters at the distance of 6 cm from each other along a 61.5 cm active length of the fuel assembly with a burnup of 56.45% of FIMA. The performed dosimetry experiment was simulated using the computational ORIGEN2.1 and MCNPX2.7 codes to verify the performance of the computational code in predicting the gamma radiation dose rate; the obtained results have been then compared with measurements. For calculating the dose rate, first, a spectrum and flux density of the gamma radiation emitted from the nuclear fuels with different irradiation histories at 18 energy groups has been accurately predicted by ORIGENT2.1, followed by calculating four fuel assemblies-induced dose by MCNPX2.7. Findings illustrate a very good agreement between the results from calculations and those of measurements.

Investigation of safety aspects during steady state operation of Tehran research reactor fuel test loop

In order to enhance making use of potential capabilities of Tehran research reactor (TRR) as a material test reactor (MTR), a 10-bar fuel test loop is designed to provide required infrastructure to perform irradiation tests on domestic fuels in Tehran research reactor. Tehran research reactor with appropriate neutron flux in irradiation positions in the core can provide required neutron fluence for irradiation tests on nuclear fuels. TRR fuel test loop with a wide range of adjustable coolant flow rate, pressure and temperature, within the range applicable in research reactors, has capability to simulate thermal-hydraulic conditions of various research reactors to perform irradiation tests on research reactors’ fuels under the same condition as the reactor core of concern. This paper presents an introductory to the systems and safety features of the TRR fuel test loop as well as the results of analyses performed to investigate neutronic and thermal-hydraulic safety parameters of TRR core during fuel test loop operation and the results of steady state analysis of the loop.

A multi-moderator neutron spectrometer for use in BNCT studies of the Tehran research reactor

The Tehran Research Reactor is the only appropriate and available neutron source in Iran for clinical boron neutron capture therapy (BNCT). One of the requirements for BNCT is to carefully evaluate and measure the therapeutic neutron beam (epithermal neutrons) as well as the fast and thermal neutron components for successful treatment. In this research, a multi-moderator neutron spectrometer (MMNS) with LiI(Eu) scintillator as neutron counter was proposed for these measurements in the range of 10−11 eV to 15 MeV. The results confirmed promising precision of the designed MMNS for the epithermal spectrum; however, the angular dependency of the therapeutic beam due to any probable change in the beam-shaping assembly should be considered.

Numerical and experimental study of control rods effects on neutronic power in TRR

Power of nuclear reactors are measured through different methods. Neutron Detectors are usually used for online power monitoring of nuclear reactors. The measured power by neutronic detectors is affected by the positions of control rods (CRs) in the core. The main contribution of this research is to investigate the effect of the CRs position on power monitoring in Tehran Research Reactor (TRR) from experimental and simulation aspects. The MTR_PC package is used to evaluate the effect of partial insertion of the CRs on the power calibration at the operating core of the TRR. The maximum impact of CRs position on the NDs response is about 25% during a cycle. Results of the thermal power calibration experiment are used to investigate the CRs effect on power calibration. In the first experiment, the results of two thermal power calibrations, with the same monitoring power 2 MW at the different configurations of CRs, are compared. Experimental investigation illustrates 9.5% changes in the real thermal power of TRR, while simulation predicts about 11%. In the second experiment, the power of TRR is increased to 5 MW and the thermal power calibration has been repeated in the new position of CRs and its result is compared with 5 MW. Both of the experiment and modeling results predict a similar effect of CRs on power calibration.