Hasil untuk "Nuclear engineering. Atomic power"

Marco A.B. Zanoni, Zhe Liang

ChengYi Hou, RuiXuan Zhao, HuiJun Zhang et al.

FeCr alloys are promising for cladding due to their thermal stability and radiation resistance, but their atomic-scale mechanical behaviors under varying temperatures is not yet well understood. Traditional empirical potentials are unreliable at high temperatures due to oversimplified assumptions. The deep potential (DP) model offers a more accurate and efficient alternative for predicting high-temperature alloy behavior. Here, we develop a deep potential model for FeCr alloys using a dataset obtained from density-functional theory (DFT) and the DP-GEN active learning framework. Molecular dynamics(MD) simulations based on the DP model show that a typical Fe3Cr alloy has a tensile strength of 15 GPa at 1200 K with a 25 % reduction in stress. This difference is attributed to the pinning effect of Cr atoms on dislocation slip and the strengthening induced by short-range ordering in Fe3Cr bonds. Compared to the MEAM potential, the DP model predicts a fracture strain of 32 % for FeCr alloys, which is in agreement with ductile characteristics observed in experiments. These results elucidate the microscopic mechanical behavior and failure mechanisms of FeCr alloys, paving the way for the development of high-performance FeCr alloys for high-temperature applications.

Wentao Wang, Congxin Yang, Yanlei Guo et al.

The lead-cooled fast reactor (LFR) is a representative design of Generation IV nuclear reactors. To investigate the flow field characteristics of LFR coolant within an axial-flow pump, computational fluid dynamics (CFD) simulations were conducted for both lead-bismuth eutectic (LBE) and water under five distinct flow rates. Results indicate that LBE consistently exhibits significantly higher head and efficiency than water at all operating points. While the theoretical head difference between LBE and water is minimal, hydraulic losses within the impeller and guide vanes are markedly lower for LBE. The variation trends of skin friction coefficients are similar for both media, yet LBE demonstrates substantially lower coefficients on blade surfaces. The vorticity and turbulent kinetic energy (TKE) in the impeller and guide vanes exhibit a strong positive correlation. In the blade wake region, areas of high vorticity show significant overlap with zones of elevated TKE. The distribution characteristics of both TKE and vorticity remain consistent across the two media, with water demonstrating slightly higher values than LBE.

Evgeniy Adamov, Boris Gabaraev, Nikolay Gorin et al.

Ugyu Jeong, Hyeongjin Byeon, Jaeyeong Park

M. Criado, E. Torres, M.J. de Hita

Ion-exchange resins (IERs) are used for the decontamination of effluents from chemical impurities and radioisotopes in the primary and secondary cooling circuits of pressured water reactors. Cementation in OPC matrices (CA, cement and fly ash) has been the preferred alternative for the conditioning of spent IERs (SIERs). Due to the progressive supply shortage of ash in the international market, the technical feasibility of other alternative cementitious formulations (CAS, cement, ash and slag, and CS, cement and slag) to confine this waste has been assessed. In this study, aspects such as the maximum waste loading, physic-chemical characteristics of the wasteforms, or their radionuclide retention capacity were evaluated. The maximum SIERs surrogates admitted for each formulation was 7.5 % wt./wt. resin/binder (12 % volume). For all wasteforms, boron released from SIERs surrogates seemed to interact with calcium compounds of the raw materials, retarding the reaction kinetics and setting, especially in the CS sample, and delaying the formation of the gel and the portlandite. No formation of boron compound but substitution of silicon tetrahedral by boron tetrahedral in the gel structure. CAS wasteforms sample with immobilised resin, independently of the SIERs surrogates content, exhibited an improved performance compared to the CA formulation currently in use.

Alistair F. Holdsworth, Zizhen Feng, Ruth Edge et al.

When studying hazardous materials such as spent nuclear fuel (SNF), the minimisation of sample volumes is essential, together with the use of chemically-similar surrogates where possible. For example, the bulk behaviour of urania (UO₂) can be mimicked by appropriately-engineered thin films of sufficient thickness, and inactive materials such as ceria (CeO₂) can be used to study the effects within radioactive systems used to fuel nuclear fission. However, thin film properties are sensitive to the preparative method, many of which require the use of highly toxic precursors and specialised apparatus (e.g., chemical vapour deposition). To address this, we present the development of a flexible, tuneable, scalable method for the preparation of thin-film CeO₂ SIMFUEL models with a thickness of ≈5 μm. The effects of γ irradiation (up to 100 kGy) and dopants including trivalent lanthanides (Ln³⁺) and simulant ε-particles on the structure and long-term leaching of these systems under SNF storage conditions were explored, alongside the context of this within further work. It was found that the sensitivity of CeO₂ films to reduction upon irradiation, particularly in the presence of simulant ε-particles, resulted in increased leaching of Ce (as Ce^III), while trivalent lanthanides (Nd³⁺ and Eu³⁺) had a minimal effect on Ce leaching.

Hongsheng Chen, Hongxing Xiao, Chongsheng Long et al.

The actual swelling of AgInCd absorber might exceed the predicted swelling value after years of service in pressurized water reactors, and the chemical and microstructural changes of AgInCd absorber induced by transmutation reactions are the main reason for the swelling acceleration of AgInCd absorber. In the present study, a model for calculating the irradiation swelling of AgInCd absorber in nuclear control rods is developed according to chemical and microstructural changes of AgInCd absorber. In this model, the chemical compositions of AgInCd absorber as a function of the thermal neutron fluence are firstly calculated, and then the volume of AgInCd absorber after irradiation is obtained on the basis of the crystallographic parameters of phases in the AgInCd absorber, and the irradiation swelling of AgInCd absorber is finally calculated. The crystallographic parameters can be obtained by preparing the simulated AgInCd alloys and fitting the experimental data. The model calculating results of irradiation swelling are in good agreement with the actual swelling data in literature. More importantly, the present model can well explain the EPRI results of the acceleration in the diametral swelling rate above 6–8 × 1020 n/cm2 and the decrease in the diametral swelling rate above about 2 × 1021 n/cm2.

T.P. Kaloni, A. Prudil, D.E. Spearot et al.

F.G. Di Lemma, J.F. Jue, T.L. Trowbridge et al.

In this study we compared the microstructure evolution of U-Mo fuel foils produced with and without heat treatment at low burn-up via advanced post-irradiation examination. The aim of this study is to observe after irradiation the ways in which the fabrication processes have influenced fuel behavior at early-stage irradiation, as for very low burn up microstructural studies are lacking. In this work it was observed that the larger grain size detected in the heat-treated samples before irradiation led to decreased grain refinement after irradiation. Grain refinement was associated with the presence of small nano-size bubbles and precipitates. This phenomenon is hypothesized to influence early fuel swelling during reactor irradiation. Grain refinement was also observed to increase in regions where γ-U decomposition was present. Thus, it was enhanced in the samples fabricated without heat treatment. The heat treatment also increased the thickness of the U-Mo/Zr interface, namely of the UZr2 layer. However, the influence of this layer on fuel performance needs further investigation. On one side, it may contribute to better mechanical bonding; on the other, it may influence swelling and blistering in the interaction layer as porosity increases when this layer is increased. This was observed especially in the presence of increased area containing low Mo concentration, and thus containing a higher fraction of the α-U phase, which is highly susceptible to irradiation induced swelling. Strong evidence of reverse transformation under irradiation (α-U + γ′-U2Mo → bcc γ-U) was observed in these samples. While the precipitates (carbides and oxide) seem to be unaffected by the irradiation at these low burnups. However, further analyses are necessary at higher burn-up to assess the exact impact different heat treatments have on fuel performance.

Soo Min Lee, Chansoo Choi, Bangho Shin et al.

Recently, a new monkey computational phantom, called Visible Monkey, was developed for non-ionizing radiation studies in animal research. In this study, we extended its applications to ionizing radiation studies by implementing the voxel model of the Visible Monkey into three general-purpose Monte Carlo (MC) codes: MCNP6, PHITS, and Geant4. The implementation work for MCNP and PHITS was conducted using the LATTICE, UNIVERSE, and FILL cards. The G4VNestedParameterisation class was used for Geant4. Then, organ dose coefficients (DCs) for idealized photon beams in the antero-posterior direction were calculated using the three codes and compared, showing excellent agreement (differences <3%). Additionally, organ DCs in other directions (postero-anterior, left-lateral, and right-lateral) were calculated and compared with those of the newborn and 1-year-old reference phantoms. Significant differences were observed (e.g., the stomach DC of the monkey was 5-fold greater than that of the 1-year-old phantom at 0.03 MeV) while the differences tended to decrease with increasing energy (mostly <20% at 10 MeV). The results of this study allows conducting MC simulations using the Visible Monkey to estimate organ-level doses, which should be valuable to support/improve monkey experiments involving ionizing radiation exposures.

Byungkuk Lee, Seungchul Yang, Dongyong Kwak et al.

The existing wet reconversion processes for the recovery of scraps generated in manufacturing of nuclear fuel are complex and require several unit operation steps. In this study, it is attempted to simplify the recovery process of high-quality fuel-grade UO2 powder. A novel wet reconversion process for uranyl nitrate hexahydrate solution is suggested by using a newly developed pulsed fluidized bed reactor, and the resultant chemical characteristics are evaluated for the intermediate ammonium uranate hydrate product and subsequently converted UO2 powder, as well as the compliance with nuclear fuel specifications and advantages over existing wet processes. The UO2 powder obtained by the suggested process improved fuel pellet properties compared to those derived from the existing wet conversion processes. Powder performance tests revealed that the produced UO2 powder satisfies all specifications required for fuel pellets, including the sintered density, increase in re-sintered density, and grain size. Therefore, the processes described herein can aid realizing a simplified manufacturing process for nuclear-grade UO2 powders that can be used for nuclear power generation.

Rokhmat Arifianto, Robi Sulaiman, Slamet Slamet et al.

Badan Riset dan Inovasi Nasional memiliki fasilitas riset yang digunakan untuk penelitian terhadap bahan nuklir yang memiliki potensi bahaya radiasi yang dapat membahayakan pekerja. Untuk mengurangi potensi bahaya radiasi, dilakukan pemantauan radiasi secara rutin oleh pekerja. Untuk mengoptimalkan pemantauan radiasi tersebut, dilakukan pengembangan 3 (tiga) buah monitor radiasi dengan menggunakan 1 (satu) detektor radiasi Sintilasi (S) dan 2 (dua) buah detektor GM (GM dan GMT) yang nantinya dapat terpasang dan memantau radiasi secara real-time. Pengujian terhadap 3 (tiga) monitor radiasi yang dikembangkan dilakukan dalam penelitian ini untuk mengetahui keakuratan pengukuran masing-masing monitor radiasi. Pengujian yang dilakukan adalah pengujian regresi linear untuk mendapatkan model konversi dari satuan cacah per detik ke satuan mikrosievert per jam. Selain itu, dilakukan pengujian ANOVA untuk melihat apakah ada perbedaan signifikan antara pengukuran laju dosis dari ketiga monitor radiasi dibandingkan dengan surveymeter yang telah terkalibrasi. Pengujian Tukey HSD dilakukan untuk menguji masing-masing monitor radiasi hasil pengembangan dan dibandingkan dengan surveymeter yang terkalibrasi. Hasil pengujian regresi linear antara surveymeter (GS) dengan ketiga monitor radiasi didapatkan koefisien determinasi diatas 0,95. Pengujian ANOVA yang dilakukan didapatkan bahwa terdapat perbedaan rata-rata hasil pengukuran laju dosis radiasi yang signifikan dari masing masing monitor radiasi. Hasil pengujian Tukey HSD menunjukkan bahwa hanya salah satu monitor radiasi yang memiliki rata-rata nilai pengukuran laju dosis radiasi yang tidak berbeda secara signifikan dengan rata-rata nilai pengukuran laju dosis radiasi dari surveymeter terkalibrasi. Oleh karena itu monitor radiasi GMT yang telah dikembangkan merupakan monitor radiasi yang telah layak untuk digunakan dalam pemantauan radiasi pada fasilitas riset bahan nuklir. Kata kunci: Detektor radiasi, regresi linear, ANOVA

Dae Hee Hwang, Yonghee Kim

For the future energy-mix policy for carbon neutrality, demand for the capability of load-follow operation has emerged in nuclear power plants in order to accommodate the intermittency of renewable energy. The short-term decay heat analysis is also required to evaluate the decay heat level varied by the power level change during the load-follow operation, which is a very important parameter in terms of short-term decay heat removal during a grace time. In this study, the short-term decay heat level for 10 days after the shutdown was evaluated for both seasonal and daily load-follow cases. Additionally, the nuclide-wise contribution to the accumulated decay heat for 10 days was analyzed for further understanding of the short-term decay heat behavior. The result showed that in the seasonal case, the decay heat level was mainly determined by the power level right before the shutdown and the amount of each nuclide was varied with the power variation due to the long variation interval of 90 days. Whereas, in the daily case, the decay heat level was strongly impacted by the average power level during operation and meaningful mass variations for those nuclides were not observed due to the short variation interval of 0.5 days.

Nalbandyan Ashkhen, Cammi Antonio, Lorenzi Stefano et al.

In this paper, CFD assessment of the DYNASTY natural circulation loop, adopting a RANS turbulence modeling approach, is performed using the OpenFOAM open source toolbox. The DYNASTY facility is designed to investigate the stability and dynamics of heat-generating fluids, in particular molten salts, in a natural or forced circulation regime and as such, it is one-of-a-kind, large scale facility for studying the natural circulation in presence of distributed heating. In this work, a CFD model of the facility is set up and validated by comparing the model results to experimental data obtained during the initial testing campaign of the facility, with water as working fluid. In particular, the equilibrium state of the system is investigated in terms of the mass flow dynamic behaviour and the temperature difference across the cooler section of the loop. It is shown that the CFD simulations adopting the k − ω SST turbulence model best reflect the experimental results. The CFD results are also in agreement with a simplified 1D modeling as well as an analytical solution.

Artavazd M. Sujyan, Viktor I. Deev, Vladimir S. Kharitonov

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

Dmitry M. Dronov, Aleksandr V. Gontovoy, Yelena N. Sarkisyan et al.

Power facilities use large amounts of water to cool steam in the steam turbine condensers, and lubricating oils, gas and air in turbine sets. The key requirement for the quality of cooling water is to ensure normal vacuum in condensers. Cooling water must not form mineral and biological deposits and corrosion products in the system. Deposits of mineral salts in the condenser tube system, as well as in auxiliary cooling systems, lead to deterioration in heat exchange and a major decrease in the cost effectiveness of the power equipment operation, and require the heat-exchange equipment to be periodically cleaned. The source water used for cooling is normally taken from nearby water bodies (large rivers or lakes). Circulating water supply systems are used most commonly: these systems use repeatedly the same water inventory for cooling, and require only small amounts of water added to make up for evaporation losses. Coolers, in this case, are cooling towers, spray pools and evaporation ponds. The water chemistry should ensure the operation of equipment without any damage to its components or the loss of efficiency caused by the corrosion of the internal surfaces as well as without scale and sludge formation. It is exactly when using circulating water supply that a stabilizing treatment program is the most practicable way to ensure a cost-effective and environmentally friendly mode of operation. To inhibit scaling processes on the heat-exchange surfaces of the turbine condenser tubes at the Novovoronezh NPP’s unit 5, the cooling water was treated with the NALCO 1392 inhibitor. The results of the NALCO 1392 inhibitor pilot tests in the circulating water supply system (with a cooling pool) are presented.

M. T. Saidin, A. A. Rahman, H. H. Harun et al.

This study aims to compare entrance surface dose (ESD) values measured with nanoDot Al2O3:C optically-stimulated luminescence dosimeter (OSLD) and guidance level set under the second national dose survey which utilized old-version LiF:Mg,Ti thermoluminescence dosimeter (TLD). In this study, we conducted a dosimetric assessment for posteroanterior chest X-ray (PA-CXR) examinations performed at various community clinics in Perak, Malaysia. These clinics were selected as they were excluded from the first and second national dose survey conducted in Malaysia in 1993-1995 and 2005-2009, respectively. The ESD is obtained by mounting the OSLD on the surface of polymethyl methacrylate (PMMA) slabs. The PMMA slabs were then exposed to X-ray based on the current practice of respective clinics. The results show that the 3rd quartile of ESDs ranged from 0.180 mGy to 0.229 mGy which is less than the recommended guidance level of the second national dose survey by 77 %. ESD measured using OSLD was found to be lower than the guidance values recommended from the second national dose survey. The finding showed a good competency of the radiographer to optimize radiological practice specifically in routine X-ray examination.

Nak Hyun Kim, Jong Bum Kim, Sung Kyun Kim

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.