• Nuclear Engineering and Technology
• /
• v.50 no.5
• /
• pp.780-787
• /
• 2018

This article presents a security module based on a field programmable gate array (FPGA) to mitigate man-in-the-middle cyber attacks. Nowadays, the FPGA is considered to be the state of the art in nuclear power plants I&C systems due to its flexibility, reconfigurability, and maintainability of the FPGA technology; it also provides acceptable solutions for embedded computing applications that require cybersecurity. The proposed FPGA-based security module is developed to mitigate information-gathering attacks, which can be made by gaining physical access to the network, e.g., a man-in-the-middle attack, using a cryptographic process to ensure data confidentiality and integrity and prevent injecting malware or malicious data into the critical digital assets of a nuclear power plant data communication system. A model-based system engineering approach is applied. System requirements analysis and enhanced function flow block diagrams are created and simulated using CORE9 to compare the performance of the current and developed systems. Hardware description language code for encryption and serial communication is developed using Vivado Design Suite 2017.2 as a programming tool to run the system synthesis and implementation for performance simulation and design verification. Simple windows are developed using Java for physical testing and communication between a personal computer and the FPGA.

• Nuclear Engineering and Technology
• /
• v.54 no.10
• /
• pp.3888-3896
• /
• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Nuclear Engineering and Technology
• /
• v.49 no.1
• /
• pp.216-223
• /
• 2017

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and $10^{th}$ value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2377-2386
• /
• 2021

To model the atmospheric dispersion of radionuclides released from nuclear accident is very important for nuclear emergency. But the uncertainty of model parameters, such as source term and meteorological data, may significantly affect the prediction accuracy. Data assimilation (DA) is usually used to improve the model prediction with the measurements. The paper proposed a parameter bias transformation method combined with Lagrangian puff model to perform DA. The method uses the transformation of coordinates to approximate the effect of parameters bias. The uncertainty of four model parameters is considered in the paper: release rate, wind speed, wind direction and plume height. And particle swarm optimization is used for searching the optimal parameters. Twin experiment and Kincaid experiment are used to evaluate the performance of the proposed method. The results show that the proposed method can effectively increase the reliability of model prediction and estimate the parameters. It has the advantage of clear concept and simple calculation. It will be useful for improving the result of atmospheric dispersion model at the early stage of nuclear emergency.

• Journal of Nuclear Fuel Cycle and Waste Technology
• /
• v.1 no.1
• /
• pp.37-47
• /
• 2013

This paper describes basic plans for the development of a radioactive waste disposal facility with the introduction of Nuclear Power Plants (NPPs) for Kenya. The specific objective of this study was to estimate the total projected waste volumes of low- and intermediate-level radioactive waste (LILW) expected to be generated from the Kenyan nuclear power programme. The facility is expected to accommodate LILW to be generated from operation and decommissioning of nuclear power plants for a period of 50 years. An on-site storage capacity of 700 $m^3$ at nuclear power plant sites and a final disposal repository facility of more than 7,000 $m^3$ capacity were derived by considering Korean nuclear power programme radioactive waste generation data, including Kori, Hanbit, and APR 1400 nuclear reactor data. The repository program is best suited to be introduced roughly 10 years after reactor operation. This study is important as an initial implementation of a national LILW disposal program for Kenya and other newcomer countries interested in nuclear power technology.

• Nuclear Engineering and Technology
• /
• v.37 no.2
• /
• pp.177-184
• /
• 2005

The Multivariate State Estimation Technique (MSET) is being used in Nuclear Power Plants for sensor and equipment condition monitoring. This paper presents the use of regularization methods for optimizing MSET's predictive performance. The techniques are applied to a simulated data set and a data set obtained from a nuclear power plant currently implementing empirical, on-line, equipment condition monitoring techniques. The results show that regularization greatly enhances the predictive performance. Additionally, the selection of prototype vectors is investigated and a local modeling method is presented that can be applied when computational speed is desired.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1878-1884
• /
• 2023

To lay a solid basis for prosperity and competitiveness, countries should achieve balance in the three fundamental aspects: energy availability, energy affordability and ecological balance. Nuclear energy has attracted international interest as one of the most crucial environmental quality strategies. The objective of this study is to analyze the non-linear link between nuclear energy and environmental quality in the top-10 nuclear energy consumer countries (USA, China, Russia, France, Canada, Spain, Sweden, South Korea, Ukraine, and Germany). Earlier research employed panel data methodologies to examine the linkage between nuclear energy and the environment, despite the fact that many nations did not independently demonstrate such a correlation. On the alternative, this study uses a novel approach known as 'Quantile-on-Quantile,' which allows for the analysis of time-series dependence in each country by giving universal yet country-specific insights into the relationship between the variables. Estimates show that the consumption of nuclear energy improves environmental quality by lowering ecological footprint in the majority of the nations studied at certain quantiles of data. Moreover, the data demonstrate that the degree of asymmetries between our variables changes by nation, emphasizing the importance of policymakers exercising caution when adopting nuclear energy and environmental quality regulations.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1775-1782
• /
• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3599-3616
• /
• 2023

After a nuclear power plant (NPP) accident, it would be helpful to predict the movement of the radioactive plume emitted from the NPP as accurately as possible to protect the nearby population. Radioactive plumes are mainly affected by wind direction and speed. Since it is difficult to identify the wind direction and speed immediately after the accident, a good understanding of the historical wind data could save many lives and ensure smoother evacuation procedures. In this study, wind data for the past 10 years are analyzed for the five NPPs in the Republic of Korea (ROK). The analyzed data include wind direction and wind speed from 2012 to 2021. In particular, the characteristics of the wind field blowing from the NPPs to the nearest densely populated regions are examined. Finally, suggestions to improve evacuation plans are made.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1593-1603
• /
• 2023

In deterministic and Monte Carlo transport codes, b-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available b-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code - named Time Dependent OpenMC or OpenMC(TD)- time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.