Volume 88, Issue 3

In this study, radionuclide behavior in high-level radioactive waste (HLW) disposal repositories is complicated because of the spatial heterogeneity of porous media, coupled flow-transport mechanisms, and multiple chemical reaction processes. Discrepancies in the diffusion behavior of a non-sorbing tracer (HTO) and a reactive tracer ( 137 Cs) in porous media have long been recognized but are not yet fully understood, which hinders effective assessment of the capabilities of buffer materials. This paper was dedicated to exploring and explaining the discrepancies in the transport behavior of non-sorbing and reactive tracers through laboratory experiments and an investigation of contributing mechanisms. Our results showed that for a bentonite sample of the same thickness, 137 Cs has smaller apparent and less effective diffusion coefficients than those for HTO. These discrepancies can be attributed to the negative surface electric effects, atomic properties, and chemical reactions. In the case of bentonite samples with different thicknesses (0.5, 0.75, 2.0, 2.5 cm), the apparent and effective diffusion coefficients show an increasing trend with bentonite thickness. According to the experimental data and fitting results, the apparent and effective diffusion coefficients are highly related to bentonite thickness. Thus, scaling effects on transport parameters were proposed to explain the results, which were attributed to the nonuniform distribution of the pore space in the bentonite sample. The scale effect behavior of radionuclide was quantified through a regression analysis. The results can be used to improve buffer designs for radionuclides diffusion.

To design a compact heat sink, a simplified geometry, enhanced heat dissipation, and the minimum pressure drop should be taken into consideration. With this objective, an experimental investigation has been conducted with the corrugated plate-fin heat sink by varying the relative radius of corrugation and relative corrugation pitch in the range of 0.16–0.31, and 0.06–0.16, respectively, for the Reynolds number range of 6000–14,000. Experiments were conducted on a corrugated plate-fin heat sink using an open-loop experimental system comprising a test section of a rectangular channel measuring 2300 mm long, 180 mm wide, and 80 mm high. The corrugated fin creates higher disturbances caused by multiple separations and reattachments in the flow and thereby yielding a higher localized heat transfer coefficient and enhanced heat transfer from the system. The maximum fin performance is found to be 5.87 for the corrugated plate-fin heat sink corresponding to the relative radius of corrugation and relative corrugation pitch of 0.16 and 0.125, respectively.

The control drum is usually placed in the periphery of the active zone of the reactor core, and the control drum worth is generally small. This paper discusses different schemes to improve the control drum worth by the combination of neutron reflectors and absorbers, and the design optimization of the drums under the condition that external dimensions of the drums and positions of the drums relative to the core are fixed. The results show that BeO as the reflector, B 4 C and Gadolinium as the neutron absorber can obtain relatively higher drum worth. In addition, the control drum worth can also be appropriately improved by using a double-layer combination of moderator material and neutron absorber material, but it cannot be effectively improved by enlarging the central angle of the neutron absorber sector.

It is advised to use a new formula to calculate the (p, n) reaction cross section at 7.5 MeV. We propose six new parameters for the formula proposed by Broeders, C. and Konobeyev, A.Y. (2008. Systematics of (p,n) reaction crosssection. Radiochim. Acta 96: 387–397) to fit experimental data. It should be noted that our systematics is only applicable to isotopes with a ratio of 7.5 MeV to the reaction threshold above 1.3 (7.5/ E th  > 1.3). This is based on analytical calculations generated from the semi-empirical mass formula, the evaporation model, and the pre-equilibrium exciton model. We were able to find new parameters for the Broders et al. formula through this inquiry that guarantee a good fit with the revised experimental data (EXFOR2022) and provide a minimum value for the statistical parameters ∑ and χ 2 .

The research addressed the economic evaluation of nuclear energy, coal, oil and natural gas in various cycle configurations with different distillation technologies. A comparison between fossil and nuclear energy sources is discussed. A comprehensive review of all desalination plants using fossil and nuclear energy with cogeneration is also included and discussed. Computational model was used to evaluate nuclear and non-nuclear desalination plants. Four possibilities were discussed. In this first case, a nuclear gas cycle integrated with a MED desalination plant was found to have the lowest cost in the gas cycle. While the nuclear steam cycle integrated with RO has the lowest cost compared to oil and natural gas. In the third case, the combined nuclear cycle was discussed and it was found that the combined nuclear cycle associated with RO has the lowest cost. In the last case, the evaluation of a heat-only desalination plant was conducted among all sources, and it was found that the nuclear plant with RO and MED has the lowest cost. In addition, the completed nuclear desalination plant was subjected to five scenarios to calculate and estimate at which capacity the plant provides the best values. The results show that nuclear desalination with gas cycle is the most economical among the oil and natural gas options.

Nuclear desalination has been identified as an alternative option with much lower carbon dioxide emissions to provide fresh water by driving high capacity desalination plants. This work considers a theoretical analysis of using nuclear desalination to provide fresh water in three selected Saudi Arabian cities. It presents a theoretical model that integrates the characteristics of nuclear reactor, power cycle and desalination blocks. The power block includes three steam turbines and five feed water heaters. It is coupled via low pressure turbine to a multiple effect desalination unit integrated with a thermal vapor compressor encompassing eight effects, seven feed preheaters and a down condenser. The output includes work generated as function of fuel mass and reactor type, enrichment percentage, power and water production with different nuclear reactor type. Desalination performance indicators such as the fresh water production rate, gain output ratio (GOR), specific energy consumption (SEC) and specific cooling water mass flow rate have been evaluated and analyzed as function of sea water temperature for three specific Saudi cities. It was found that these indicators reflect better performance along a year for Jizan city than for Jubail and Tabuk. The case of Jizan city gives over the whole year more uniform values of water production rates, gain output ratio, specific energy consumption and cooling water mass flow rates.

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) with reference to the Advanced Power Reactor 1400 MW (APR1400) for transient and design basis accidents (DBAs) simulation. An experiment for a 4-inch cold leg top-slot break was performed at ATLAS to resolve a safety issue that the loop seal reformation (LSR) of APR1400 could lead to the increase of peak cladding temperature. In addition, the experimental data has been utilized to validate system codes within a framework of domestic standard problem (DSP) program organized by KAERI in collaboration with Korea Institute of Nuclear Safety (KINS). In this study, the experiment has been analyzed by thermal-hydraulic system analysis code, MARS-KS 1.5 and a comparison with experimental and calculation results has been performed. Since the top-slot break is not a typical break geometry for safety analyses, this study aims at examining the applicability of MARS-KS to the top-slot break accident where the LSR occurs repeatedly. The results revealed that overall physical behavior during the accident was predicted by the code, appropriately. MARS-KS showed the excursion of the peak cladding temperature because of the LSR as in experiment. It has been confirmed that the core integrity was maintained because the temperature excursion by the LSR was not large enough to alter the acceptance criteria. In addition, it is presented the results of the sensitivity analysis of parameter that affect the figure of merits.

A project was initiated to assess the sustainability of Indonesia’s planned nuclear energy system using the IAEA INPRO Methodology to develop an awareness of sustainability issues to support nuclear energy development strategic planning and decision making. Accordingly, for a nuclear energy system to be sustainable, fulfilment of current needs should not compromise the ability of future generations to meet theirs. The sustainability of the planned nuclear energy system was assessed based on the basic principles, user requirements and criteria in the areas of economics, infrastructure, waste management, proliferation resistance, physical protection, environment, and safety. The assessments covers the sustainability of a large reactor completed in 2014 to support the Pre-Feasibility Study in Bangka Belitung, and of a small and medium-sized reactor (SMR) currently ongoing to support the study in West Kalimantan. The results indicate that evidence of achieving sustainability exists in many respects, but there remain gaps to be addressed in due time on aspects such as economics, investment climate, availability of technology, and global development on non-renewable resources. A preliminary study on fuel cycle arrangements is being performed in parallel, focussing on collecting data including on fuel cycle services abroad. Collecting data has been the main challenge for a newcomer country that has yet to decide on the exact reactor technology. The paper is intended to evaluate the performance of Indonesia’s nuclear energy program using the INPRO Methodology to address the sustainability of its planned NES, i.e., reactors and fuel cycle facilities, and to evaluate the effectiveness of the efforts to achieve the purpose. Awareness of sustainability issues plays a key role in nuclear energy policy, especially for determining the energy mix by 2040, to achieve the Net Zero Emission policy target by 2060 or earlier.

The spent nuclear fuel discharged from power reactors is a very important problem facing the future of using power reactors in electricity production. This paper focuses on the thermal-hydraulic behaviour of the VVER spent fuel in the vault dry storage system under forced convection mode, which is experimentally and numerically investigated. For this purpose, a test rig is designed and constrained to simulate the cooling loop vault system that contains four spent fuel assemblies discharged from the VVER reactor, which are represented by four electric heaters. A numerical simulation is performed by the ANSYS-CFX fluid dynamics code. The effects of decay heat generation and inlet air velocity are investigated as an operating condition. Also, the effect of the type of the Vault System tube material is being studied. The results show that the increase in the inlet air velocity improves the coolability of the fuel, while the increase in decay heat leads to a decrease in the coolability of the fuel. The used velocity range is (0.1 < V < 0.5 m/s) for inlet coolant air and heater power (20 < P < 100 W). Three tube materials (aluminum, copper, and stainless steel) were evaluated for mechanical properties, including thermal conductivity, to assess the feasibility of their use as tubes in the spent fuel storage.

A correlation for the volume porosity of irradiated UO 2 fuel as a sum of the individual contributions of the pore and swelling porosity in terms of the local density and fractional matrix swelling is developed. By using the existing low temperature low burnup model of swelling, evolution of the matrix swelling and porosity terms are calculated for UO 2 fuel in the low burnup but in the high burnup two various models are applied, the one with considering grain recrystallization and another without it but with the method of Xe depletion measurement. The purpose of the paper is comparison of fuel swelling behavior between two models at high burnups. The bulk swelling and porosity evolution in both methods are also validated by experimental data. Finally, one conclusion from this comparison is obtained, as the method which considering grain recrystallization has more rational behavior in the fuel swelling and porosity.

Mobile Robots (MR) are currently used across a variety of different sectors and have military, nuclear and industrial applications among others. In unmanned systems, teleoperation sensors, navigation instruments, control systems and radiation sensors can be fixed on the MR to perform required tasks such as radiological scanning, identifying, and surveying the contaminated environment that has been exposed to radiation. In this work, an estimation of the mobile robot location and the optimum path for time-delay compensation for MR teleoperation are investigated. As the MR teleoperation has a stochastic nature, the kinematics equations are modeled using stochastic differential equations (SDEs). Afterwards, these SDEs are solved using Numerical algorithms such as Euler–Maruyama algorithm which is used to approximate SDEs solution with the aid of MATLAB. Additionally, the results are discussed and depicted in tables and figures. Finally, the simulated results for the solution are performed and are found to highly agree with the ideal path of the simulated MR. This result is of great importance to be used in the case of nuclear emergency response and mitigation.

In heavy ion fusion using the inertial confinement fusion (ICF) approach, firstly, the deposited energy of heavy ions in the target and, secondly, the charged products resulting from fusion reactions in the plasma of the fuel capsule are key and necessary points. In this paper, we used the ICF method for the core of a spherical fusion reactor simulation filled with multi-layer fuel capsules with foam using symmetrical irradiation from 32 different directions by two heavy ion beams of Cs and Pb with radiation energies of 8 and 10 GeV, respectively. Then we simulated the process of penetration and deposited energy of the beams inside the core of this reactor using GEANT4 code. The results of our simulations show that if the atomic number of radiation beams increases, the amount of beam stopping power increases, which is in agreement with existing theories. Also, by changing parameters such as the type and energy amount of the radiation beam, thickness, and the type of material selected in the layers of the desired fuel capsules, the amount of the penetration depth, the produced secondary particles, the stopping power per unit volume of fuel capsule and the reactor core will change. Eventually, these variations will cause a change in deposited energy gain inside the core of a spherical fusion reactor. The obtained maximum deposited energy due to the two selective Pb + and Cs + beams with 8 and 10 GeV energies in this study is related to DT fuel compared to the two neutron free-fuels of D 3 He and P 11 B. It can be seen that energy gain increases significantly with changing beam energy from 8 to 10 GeV, but for both selected energy, the enhancement of DT energy gain compared to D 3 He and P 11 B is not so significant.