MOSAP (modular severe accident analysis program) is an integrated severe accident analysis code for nuclear power plants independently developed by Xi’an Jiaotong University. In order to verify the rationality and accuracy of MOSAP, it was used to model and calculate the nuclear accident process of Three Mile Island Unit 2 (TMI-2). In the model established by the MOSAP, the two cold pipe sections and two main pumps in each loop of TMI-2 were simplified into one cold pipe section and one main pump. The important emergency core cooling system was simulated in the primary loop system, and the pressurizer and its spray and heater devices were also simulated. The core area was divided into 16 axial segments and 4 radial rings. The lower head was divided into 5 sections in the radial direction and 5 layers in the thickness direction. The steady-state calculation results of the MOSAP are in good agreement with the TMI-2 system parameters, indicating that the models established by MOSAP can simulate the operation of the power plant well. This model was used for accident transient calculations and the calculation results of key parameters in the accident process such as the pressure of the primary loop system, the water level of the pressurizer, the water level of the core, the fuel temperature, the amount of hydrogen produced, and the mass of molten material in the lower plenum were obtained and compared with the measured accident values and the calculated values of the internationally recognized severe accident analysis program ASTEC (V2.2). The results show that the calculated values of the pressure of the primary loop system, the water level of the pressurizer, the water level of the core, and fuel temperature in the MOSAP are in good agreement with the measured accident values and the calculated values of ASTEC. The hydrogen production value calculated by the MOSAP is about 541 kg and the total mass of debris in the lower plenum is about 18 t, both of which are closer to the measured values of TMI-2 than those calculated by the ASTEC. ASTEC can’t be used to calculate the significant migration of molten material towards the lower plenum, which is due to the solidification of the molten material at the core lower plate caused by the certain water level held in the lower part of the core. Overall, the deviation between the calculated key parameters of MOSAP and the measured values of TMI-2 is within 20%. MOSAP can simulate the main processes and phenomena of severe accidents such as core overheating and cladding oxidation, core material melting, migration and relocation, core reflooding, and lower head response.
YU Ting1, ZHANG Yinyin2, ZHANG Ruizhi3, JIN Wenlei2, LUO Yingting2, ZHU Shengfeng3, HE Hui1, YE Guoan1, GONG Helin4,
Spent fuel reprocessing is an important nuclear energy process which aimed at recovering resources and managing radioactive materials to control potential hazards. In this field, Purex technology is widely used for its high efficiency, scalability, and wide applicability. Purex technology, a liquid-liquid extraction technique to separate and purify uranium and plutonium from nuclear fuel, plays a key role in spent fuel reprocessing, enabling reprocessing and recycling of nuclear fuel, reducing the release of radioactive nuclear waste, and improving the efficiency of nuclear energy resources. Meanwhile, as an emerging technology, machine learning has attached wide attention and has been applied in the field of Purex, such as the selection of ligands and ionic liquids, the prediction of ligand properties, and so on. In this paper, machine learning is combined with distribution ratio prediction, which is defined as the distribution ratio of ionic liquids in different phases, which can reflect the extraction rate of ions, and plays an important role in Purex computer simulation, so the distribution ratio prediction model can help researchers to choose the optimal experimental conditions, optimize the process, and reduce the experimental cost and time. Since the traditional mathematical model of uranium distribution ratio leads to at least 15% prediction error, in this paper, three classical machine learning models (namely, random forest, support vector regression and K-nearest neighbor) were constructed to predict the distribution ratios of uranium, plutonium, and HNO3 in the 30%TBP/kerosene-HNO3 system. These models were trained based on different datasets, and their hyper-parameters were optimized using algorithms such as grid search, Bayesian optimization, and K-fold cross-validation. The results show that random forest achieves the best results in distribution ratio prediction. The average absolute relative error (AARE) of uranium distribution ratio prediction reaches 7.73%, which is about 7% higher than that of the traditional model. In addition, plutonium and HNO3 distribution ratios are also predicted to verify the generalizability of the machine learning model, and the highest of 11.6% and 13.7% are achieved. The machine learning model prediction results show that the machine learning method proposed in this paper achieves better performance than the traditional distribution ratio mathematical model, effectively improves the accuracy of uranium distribution ratio prediction, and performs well in plutonium and HNO3 distribution ratio prediction.
Hanan Akhdar, Ahmed S. Haidyrah, Mohammad W. Marashdeh
et al.
This pioneering work presents a novel approach for manufacturing B-doped ZrO2, creating in-situ ZrB2O5 within the ZrO2 matrix. A cost-effective hydrothermal synthesis using zircon mineral yields a material with exceptional radiation shielding properties. Incorporating B into the ZrO2 lattice combines ZrO2's gamma attenuation with B's neutron shielding against ionizing radiation. XRD, SEM, and EDX characterization revealed structural properties, phase compositions, and elemental distributions. Monte Carlo simulation and XCOM theoretical modeling investigated B atoms' role in shielding parameters across 0.0332–2.506 MeV γ-ray energies. Results show increased B concentration slightly decreases shielding properties while enhancing mechanical properties including crystallite size, lattice distortions, and dislocation density. B concentration increases from 2.1 to 5.4 wt% decrease LAC values to 65.959–58.613 cm-1 (0.0332 MeV), 0.464–0.461 cm-1 (0.511 MeV), and 0.208–0.270 cm-1 (2.506 MeV). B concentration up to 5.4 wt% enhances mechanical properties without significantly affecting radiation shielding, especially at intermediate energies.
BackgroundHL-2M tokamak is a new-generation magnetic confinement fusion plasma device in China, which has realized the high parameter operation mode with 1 MA plasma current.PurposeThis study aims to investigate the turbulent transport associated with the internal transport barrier (ITB) using gyrokinetic calculations.MethodsNumerical simulation was performed based on the gyrokinetic theory. The turbulent transport relevant to ITB was studied using the gyrokinetic toroidal code (GTC) combined with the equilibrium of HL-2M tokamak. A filtering zonal flow was considered in analyzing the influence of zonal flow on turbulent saturation level. A time evolution analysis of turbulent poloidal spectrum was conducted to investigate the effect of different wavelength modes on turbulent transport.ResultsThe results show that the turbulent transport at ITB saturates twice in succession, and the calculated average ion heat transport diffusivity is approximately twice that of the first saturation level. Moreover, the short-wave mode kθρi~2.15 dominates the first turbulent transport saturation, whereas the long-wave mode kθρi~0.49 dominates the second turbulent transport saturation. Specifically, an "M" shape distribution of the radial heat transport diffusivities is obtained at the transport barrier position during the turbulent saturation period. Finally, the minimum radial heat transport diffusivity during the turbulent saturation period occurs near the ITB where a maximum plasma temperature and density gradient occurs.ConclusionsThe turbulent transport at ITB may be dominated by two types of microinstabilities at different stages of turbulence development. Turbulent energy of the system is inversely cascaded from the modes with short wavelengths to those with long-wavelengths. The results are in good agreement with the theoretical prediction of the ITB.
90Sr(T1/2=28.9 a) is a highly toxic nuclide with high heat release and strong exothermic properties, and the content of 90Sr in radioactive waste must be accurately measured before disposal. The separation of 90Sr from radioactive waste is the key to accurate measurement. By changing the synthesis reaction path and optimizing the synthesis reaction parameters, this work solves the problems of hydrogenation hazard and purification difficulties in the synthesis reaction of di-tert-butyldicycloh-exyl-18-crown-6(DtBuCH18C6), and achieves the synthesis of high purity products. Using DtBuCH18C6 as the extraction agent, the solvent extraction method was used to study the effects of diluent type, extractant concentration, nitric acid concentration and organic to aqueous phase ratio on the strontium extraction performance, so as to determine the optimal extraction conditions for the selective separation of Sr2+, and establish an effective and rapid separation method for Sr2+, which laid a foundation for the future extraction of 90Sr from the high-level waste liquid.
Nuclear engineering. Atomic power, Chemical technology
LIN Wuhui1, , ZHANG Yibang2, 3, DU Jinqiu3, LUO Zhu4, CAO Shaofei5, TUO Fei6
A large amount of artificial radionuclides have been released into the ocean, contributing to serious nuclear pollution in marine environment, and arising public concerns and worry around the world. The Fukushima-derived artificial radionuclides can also be used as tracers to reveal the migration, transformation processes, and fate of artificial radionuclides in the ocean. The most polluted port within less than 1 km from the Fukushima Daiichi Nuclear Power Plant (FDNPP) was focused on in this study. The most polluted port near the FDNPP serves as windows to reflect progresses and effectiveness of decommissioning of the FDNPP, which is inaccessible for public and many other counties around the world. Historical activities of 134,137Cs in seawater, marine sediment, and marine fish were reconstructed from April 2011 to October 2023 on the basis of over 1 000 reports from Ministry of Economy, Trade and Industry of Japan, Nuclear Regulation Authority of Japan, and Tokyo Electric Power Company. The patterns of the three-stage evolution of 134,137Cs in seawater, the four-stage evolution of 134,137Cs in sediments, and the three-stage evolution of 134,137Cs in marine fish were proposed to quantify the activity levels and effective half-lives (EHL) of 134,137Cs at different stages. The evolutions of historical 134,137Cs in seawater, sediment, and marine fish were closely related to multiple countermeasures of decommissioning at the FDNPP, including the relocation of the drainage channels during June 2014 to April 2015, seabed covering of port in April 2015, removal of highly contaminated retained water in December 2015, filling of tunnels and towers in December 2015, and completed construction of sea-side impermeable walls in February 2016. The longest EHL of 134,137Cs in marine sediment indicates the memory effect of marine sediment and its persistent and dominated contribution to 134,137Cs in marine fish. Additionally, a highly consistent activity ratio of 134Cs to 137Cs (about 1.0) was simultaneously calculated in seawater, sediment, and marine fish, indicating the transferring of the Fukushima-derived 134,137Cs in multiple matrices in the marine environment. The temporal variation of concentration factor of 137Cs in marine fish was also constructed to reveal the dynamic processes of the enrichment and uptake of 137Cs in marine fish from seawater. The relatively high value of concentration factor of 137Cs in marine fish was observed during the initial period of nuclear accident followed by a decline in concentration factor of 137Cs to about 100 L/kg. This study would provide scientific evaluations for the effectiveness of the decommissioning of the FDNPP and the consequences of Fukushima contaminated water discharged into the ocean.
ZHENG Wei1, HE Xuedong1, YIN Huaqiang1, DU Bin1, LI Haoxiang1, MA Tao1, PU Yang2, WANG Shangjun2
The helium coolant in the primary circuit of the high-temperature gas-cooled reactor (HTGR) contains trace of impurities such as CO, H2, H2O, and CH4, which have an adverse effect on the structural materials at elevated temperature. Mainly, the corrosion behaviors include oxidation, decarburization, and carburization, depending on the impurity composition and corrosion temperature. Inconel 617 is the reference candidate material for steam generators of HTGR, which may be corroded by trace impurities in helium at high temperature. In order to explore the corrosion mechanism of the superalloy in impure helium and establish a prediction model of decarbonization phenomenon, the corrosion experiments of Inconel 617 were carried out at 980℃ in the impure helium. The gas phase data and corrosion behaviors of the alloy were analyzed by gas chromatograph (GC), field emission scanning electron microscopy (FESEM) with energy-dispersive X-ray spectroscopy (EDS) system, and X-ray diffraction (XRD). The mechanism of decarbonization is elucidated by chemical thermodynamics, which indicates that the driving force of carbon transfer is the carbon potential difference between alloy and environment, and more specifically, the carbon activity difference. Then, the prediction model of the decarbonization reaction was established. The critical temperature (TA) at which the corrosion behavior occurs can be obtained by thermodynamic calculation, and it is a function of the partial pressure of carbon monoxide. This model is in good agreement with the experimental data in this study and previous works with different contents of carbon monoxide. The results show that even if the impurity level is very low, it can still induce corrosion behavior. On this basis, the effects of pre-oxidation and corrosion temperature on the decarbonization reaction of alloy were investigated. When the temperature is reduced, there is no more obvious decarbonization phenomenon of the alloy, which indicates that it is an effective way to avoid decarbonization. However, after the pre-oxidation in the air at high temperature, Inconel 617 still has carbon loss, which may be due to the imperfect oxide layer formed in the air. In order to improve the compactness of the alloy oxide layer, surface modification such as coating may be more effective. For the impurity content, this study shows that Inconel 617 has strong decarburization behavior in the impure helium with very low impurity content. Therefore, the low level of impurity is not the goal of coolant purification in HTGR, and the more reasonable impurity scheme should be selected according to model prediction and experimental analysis.
The year 2023 marks the 35th anniversary of the establishment of the Beijing Tandem Accelerator Nuclear Physics National Laboratory. Accelerators and nuclear reactors are the two main tools for studying nuclear science. In 1988, the Tandem Laboratory was officially founded, serving as a significant research hub for nuclear physics in our country. It has consistently played a leading role in nuclear science innovation, achieving 140 000 h of stable operation. The laboratory has undertaken research in nuclear physics research, nuclear data measurement, nuclear physics applications, and interdisciplinary studies. This has resulted in a series of internationally recognized basic and technological achievements that meet national major demands, fostering a group of outstanding talents. It has provided solid support for the continuous development of nuclear physics research and nuclear technology strategy in our country. This article provides a comprehensive overview of the 35 years of development of the Tandem Laboratory.
Shunsuke Uchida, Masanori Naitoh, Hidetoshi Okada
et al.
In a severe accident of a BWR plant, zirconium alloy of fuel cladding reacts with steam becoming a heat source for core melting and becoming a hydrogen source for hydrogen explosion. There are a lot of publications giving empirical formulas for energy and hydrogen production due to the zirconium alloy-steam reaction. Most of the formulas are based on a parabolic law, which means the reaction rate and hydrogen production rate decrease monotonously with the reaction time due to the protective oxide layers developed on the alloy surface. However, when the oxide layers are defective for some reason, their effectiveness against the reaction is weakened and the reaction rate will increase. In order to confirm the stability of the oxide layers, an integral-type experiment on the zirconium alloy-steam reaction was set up, hydrogen production rate due to the reaction was measured as a function of exposure time and surface temperature, and then, the oxide layers on the specimen were examined after the exposure. It was confirmed that (1) the analytical results based on Cathcart's formula well explained the experimental results, and (2) the hydrogen production rate did not decrease simply with the exposure time but the decrease in effective oxide layer density resulted in loss of protectiveness and the reaction rate increased.
Medical physics. Medical radiology. Nuclear medicine, Nuclear engineering. Atomic power
ZOU Hang;LIANG Liang;ZHANG Qian;SONG Peitao;ZHAO Qiang
Time-dependent, whole-core calculation with high-fidelity pin-resolved details serves an important role in multi-physics reactor applications. However, the limitation of the current CPU has become a sore point. In order to maintain the high accuracy and reduce the computational burden, it is necessary to develop the advanced technology to speed-up the transient calculation. Recently, heterogeneous computing has been increasingly widespread used with the high-performance computing (HPC) systems, which is widely equipped in HPC clusters. Compared to CPU, current GPU has much higher rate of FLOPs/s (floating-point operations per second), larger memory bandwidth, and lower energy consumption per FLOP. These features leverage the parallel compute engine in GPUs to solve many complex computational problems in a more efficient way. Method of Characteristics (MOC) has been widely used in GPU-based whole-core transport calculation for its feature of natural massive parallelization. Based on the above points, a GPU-based 2D MOC transient fixed source problem (TFSP) solver is implemented on lattice physics code ALPHA (advanced lattice physics code based on heterogeneous architecture). In this paper, the full implicit method (FIM) were adopted to solve the TFSP. The Jacobi transport sweep algorithm was introduced in MOC TFSP solver. The transport sweep over energy groups and polar angles were performed for each ray segment during the ray tracing. To accelerate the convergence of the TFSP, the pin-based coarse mesh finite difference (CMFD) was implemented in the TFSP solver. In order to parallelize the CMFD TFSP solver on GPU, the red-black ordering algorithm was introduced in the CMFD TFSP solver. The results of two problem are presented including the TWIGL benchmark problem and the 2D MINI-CORE benchmark problem to verify the capability of the GPU-based TFSP solver. Numerical results demonstrate that the TFSP solver built in ALPHA has the desired accuracy. The percent difference of core power history between ALPHA and the reference are less than 0.5%. Compared with serial CPU-based solver, the speedup of the GPU-based solver is between 2.0x to 6.0x and the speedup is in inverse proportion to the time ratio of the CMFD calculation.
The concentration of the radionuclide 137Cs on sediment in watershed in Palembang has been analyzed. This study aims to determine the influence of sampling location and the water quality indicators of water pH, sediment pH, conductivity, turbidity, and sediment type on the concentration of 137Cs and to determine the distribution pattern of 137Cs in sediments. Sampling was conducted at seven stations spaced approximately 5 km apart, placed from the western end to the eastern end of the Musi river segment located within Palembang City.Sediment samples were prepared and their 137Cs contents were measured with gamma spectrometry. The results showed that their 137Cs concentrations ranged from below MDC (minimum detectable concentration) to 1.51 Bq/kg. This was within the 1×103 Bq/kg limit set by the quality standard. The varied and very low concentrations of 137Cs are thought to have originated from global fallout. The location point of sampling affects the concentration of radionuclide 137Cs while the characteristics of water quality are do not. The 137Cs concentration spread pattern in Musi sediment is influenced by tidal currents and river morphology.
The use of computed tomography (CT) has become a common practice in medical diagnosis in Indonesia. Its number, however, is not matched by the availability of dedicated-performance-check phantoms. This paper aims to describe the design, construction, and evaluation of an in-house phantom for CT performance check that accommodates both radiation dose measurement and image quality performance checks. The phantom is designed as laser-cut polymethyl methacrylate (PMMA) slabs glued together to form a standard cylindrical shape, with spaces to place dose measurement and image quality modules. In this paper, measurement results on both aspects are discussed and compared with standard phantoms and other works. For dose measurement, the constructed phantom exhibited the greatest absolute discrepancy against the reference standard phantom of 8.89 %. Measurement of the CT number linearity and modulation transfer function (MTF) yielded, at most, 7.51 % and 5.07 % discrepancies against Catphan 604, respectively. Meanwhile, although found to be more linear in the phantom-based contrast linearity test, the use of the in-house phantom for clinical image contrast threshold determination requires further study. For noise power spectrum (NPS) measurement, accurate results were obtained within a limited range of spatial frequency.
The EU-DEMO First Wall (FW) will be a relatively thin structure. In order not to damage this layer, heat loads distributed onto the wall should be carefully controlled. In the case of transient events, as for example plasma disruptions, the steady-state heat load limit (∼1-2MW/m2) can be largely exceeded for a timespan sufficiently long to cause damages. Therefore, when the control system detects an upcoming disruption, Shattered Pellet Injection (SPI) or Massive Gas Injection (MGI) mitigation techniques can be employed to inject impurities and switch off the plasma safely. In the present work, the Monte-Carlo ray-tracing code CHERAB is used to compute the radiative heat load distribution on the EU-DEMO Plasma Facing Components (PFCs) due to a mitigated plasma disruption. By applying ad-hoc techniques to improve the quality of the Monte Carlo calculation, we obtain a peak radiative load of ∼490MW/m2 on the PFCs, which is ∼25% lower than previous estimates.
The panorama of research reactors in the world is at a turning point, with many old ones being shutdown, a very few new ones under construction and many newcomer countries interested to get access to one or to build one domestic research reactor or zero-power reactor. In this evolving context, several actions have been set up to answer this international collaboration need: the IAEA has launched the ICERR initiative, the OECD/NEA is proposing the P2M joint project proposal. In France, the Jules Horowitz Reactor (JHR), under construction at CEA Cadarache, within an International Consortium, will be one of the few tools available for the industry and research in the next decades. The paper presents some update of its construction, its experimental capacities and the European support through FP7 and H2020 tools. This paper provides also some insights of international tools (ICERR, P2M) and about the International Group on Research Reactors (IGORR) and how they complement or interact with the JHR.
Bourg Stéphane, Geist Andreas, Adnet Jean-Marc
et al.
Processes such as PUREX allow the recovery and reuse of the uranium and the plutonium of GEN II/GEN III reactors and are being adapted for the recycling of the uranium and the plutonium of GEN IV MOX fuels. However, it does not fix the sensitive issue of the long-term management of the high active nuclear waste (HAW). Indeed, only the recovery and the transmutation of the minor actinides can reduce this burden down to a few hundreds of years. In this context, and in the continuity of the FP7 EURATOM SACSESS project, GENIORS focuses on the reprocessing of MOX fuel containing minor actinides, taking into account safety issues under normal and mal-operation. By implementing a three-step approach (reinforcement of the scientific knowledge => process development and testing => system studies, safety and integration), GENIORS will provide more science-based strategies for nuclear fuel management in the EU.
Adlin López Díaz, Juan Miguel Martín, Amalia Pérez
et al.
In 2013, the European Association of Nuclear Medicine Dosimetry Committee recommends a “Standard Operational Procedures for Hyperthyroidism Pre-Therapeutic Dosimetry” based on the assessment of the individual 131I uptake and kinetics. To estimate the 3D dose delivery deviations from prescribed dose during patient specific application of this SOP, a computer Matlab application was developed and verified. It was design to execute: radiopharmaceutical curve fitting, cumulated activity calculations, functional thyroid mass estimation, obtain the therapeutic planning activity to warranty the prescribed dose and produce the 3D planning dose map and related dosimetry parameters. 6 patients with 150-400Gy prescribed dose data planning (average 241,67Gy) were analysed using the developed application. The developed system was verify successfully using a test image phantom and 6 known pharmacokinetics data. The tridimensional thyroid volume cumulated activity and dose distributions were heterogeneous. 3D dose distribution showed standard deviations between 18.01-27.08 % of prescribed dose. The differences between maximum and minimum dose value per voxel/MBq were 74-129%. According to the result, between 50,2 % and 71,4 % of patient’s thyroid will be treat with a dose of DP±20 % of planned dose, the rest will be overdose or sub dose. Conclusions: the 3D treatment planning dose distribution were completely no-homogenous, the significant difference observed should be study in the future more deeply in order to optimized the hyperthyroidism iodine treatment.
Nuclear engineering. Atomic power, Medical physics. Medical radiology. Nuclear medicine
When the equipment which is related to safety or important to power production is installed in nuclear power plant units (NPPs), verification of equipment Electromagnetic Susceptibility (EMS) must be performed. The low-frequency radiated magnetic field susceptibility (RS101) test is one of the EMS tests specified in U.S NRC (Nuclear Regulatory Commission) Regulatory Guide (RG) 1.180 revision 1. The RS101 test verifies the ability of equipment installed in close proximity to sources of large radiated magnetic fields to withstand them. However, RG 1.180 revision 1 allows for an exemption of the low-frequency radiated magnetic susceptibility (RS101) test if the safety–related equipment will not be installed in areas with strong sources of magnetic fields. There is no specific exemption criterion in RG 1.180 revision 1. EPRI TR-102323 revision 4 specifically provides a guide that the low-frequency radiated magnetic field susceptibility (RS101) test can be conservatively exempted for equipment installed at least 1 m away from the sources of large magnetic fields (>300 A/m). But there is no exemption criterion for equipment installed within 1 m of the sources of smaller magnetic fields (<300 A/m). Since some types of equipment radiating magnetic flux are often installed near safety related equipment in an electrical equipment room (EER) and main control room (MCR), the RS101 test exemption criterion needs to be reasonably defined for the cases of installation within 1 m. There is also insufficient data regarding the strength of magnetic fields that can be used in NPPs. In order to ensure confidence in the RS101 test exemption criterion, we measured the strength of low-frequency radiated magnetic fields by distance. This study is expected to provide an insight into the RS101 test exemption criterion that meets the RG 1.180 revision 1. It also provides a margin analysis that can be used to mitigate the influence of low-frequency radiated magnetic field sources in NPPs. Keywords: Electromagnetic compatibility (EMC), Electromagnetic susceptibility (EMS), Low-frequency radiated magnetic field susceptibility, Radiated susceptibility test method, RS 101 test exemption criterion