Journal Sciences News
Ultrasound in Medicine & Biology
August 2018
Subplane collision probabilities method applied to control rod cusping in 2D/1D
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Aaron Graham, Benjamin Collins, Shane Stimpson, Thomas Downar The MPACT code is being jointly developed by the University of Michigan and Oak Ridge National Laboratory. It uses the 2D/1D method to solve neutron transport problems for reactors. The 2D/1D method decomposes the problem into a stack of 2D planes, and uses a high fidelity transport method to resolve all heterogeneity in each plane. These planes are then coupled axially using a lower order solver. Using this scheme, 3D solutions to the transport equation can be obtained at a much lower cost. One assumption made by the 2D/1D method is that the materials are axially homogeneous for each 2D plane. Violation of this assumption requires homogenization, which can significantly reduce the accuracy of the calculation. This paper presents two new subgrid methods to address this issue. The first method is polynomial decusping, a simple correction used to address control rods partially inserted into a 2D plane. The second is the subplane collision probabilities method, which is a more accurate, more robust subgrid method that can be applied to other axial heterogeneities. Each method was applied to a variety of problems. Results were compared to fine mesh solutions which had no axial heterogeneity and to Monte Carlo reference solutions generated using KENO-VI. It was shown that the polynomial decusping method was effective in many cases, but it had some limitations, with 3D pin power errors as high as 25% compared to KENO-VI. The subplane collision probabilities method performed much better, lowering the maximum pin power error to less than 5% in every calculation.
August 2018
An improved convergence rate for the prompt
August 2018
The sensitivity analysis for IHTS and SG due to the Large-scale Sodium-Water Reaction event in PGSFR
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Sang June Ahn, Gun Yeop Park, Kwi Lim Lee, Chiwoong Choi, Taekyeong Jeong, Jin Tae Kim, Seung Won Lee, Jae-Ho Jeong The Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). As a reactor coolant, it uses sodium to transfer the core heat energy to the turbine. Sodium has a chemical characteristics that it violently reacts with materials such as water and air. The Steam Generator (SG) has Intermediate Heat Transfer System (IHTS) sodium on the shell side and feed water on the tube side. When the Sodium-Water Reaction (SWR) event occurs by SG tube Double-Ended Guillotine Break (DEGB), high pressure waves and corrosive reaction products are produced which threaten the structural integrity of the IHTS and safety of the Primary heat Transfer System (PHTS). In the PGSFR, the SWR event is classified as Loss of Heat Sink (LOHS), which is one of the Design Basis Events (DBEs). The event is analyzed in terms of the structural integrity of the affected IHTS and SG. A series of behaviors such as (i) growth of hydrogen bubble by SWR and (ii) expansion/propagation of the high pressure waves, are calculated with a Sodium Water Advanced Analysis Method (SWAAM-II) code. For SWR event, a sensitivity analysis is performed on the location where the highest pressure occurs in the affected SG. Further, sensitivity analysis of design variables that can be affect the pressure seen by the main components is carried out. The parameters of the sensitivity analysis are (i) the burst pressure of the rupture disk, (ii) the gas pressure of the expansion tank and (iii) the distance between the rupture disk and the SG.
August 2018
Novel genetic algorithm for loading pattern optimization based on core physics heuristics
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): E. Israeli, E. Gilad A genetic algorithm based on novel genetic operators is implemented for the problem of nuclear fuel loading pattern optimization. This is achieved using rank selection or tournament selection and novel crossover operator and fitness function constructions, e.g., improved crossover and mutation operators by considering the chromosomes as permutations (which is a specific feature of the loading pattern problem) and the “stage fitness function” that separates the different objectives of the optimization. Another novel feature of the algorithm is the consideration of the geometric nature of the problem and the desired loading pattern solutions. A new geometric crossover is developed to utilize this geometric knowledge and its implementation exhibits good results. A comprehensive study is performed on the effect of different adaptive mutation strategies on the performances of the algorithm. The new algorithm is implemented and applied to two benchmark problems and used to study the effect of boundary conditions on the symmetry of the obtained best solutions.
August 2018
A Next Generation Method for Light Water Reactor core analysis by using Global/Local Iteration Method with SP3
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Tzung-Yi Lin, Yen-Wan Hsueh Liu The Global/Local Iteration Method (GLIM) capability was implemented into the in-house developed nodal code, NuCoT, which was based on the Hybrid Nodal Green’s Function Method (HNGFM) with the diffusion/SP3 approximations. A procedure was proposed for obtaining the Even Parity Discontinuity Factor (EPDF) in SP3 method. The accuracy of NuCoT with GLIM was verified by using two benchmark problems, KAIST and C5G7, with reference solutions calculated by NEWT and MCNP respectively. The results of NuCoT show that the accuracy of the GLIM is comparable to the whole core pin-wise calculation. The SP3 calculation with EPDF is better than the diffusion result. For the KAIST problem using GLIM, the Root-Mean-Square (RMS) of pin power error is
August 2018
Continuous machine learning for abnormality identification to aid condition-based maintenance in nuclear power plant
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): R.M. Ayo-Imoru, A.C. Cilliers A condition-based maintenance (CBM) regime in a nuclear plant will result in eliminating unnecessary maintenance cost without jeopardizing the safety of the plant. The foundation of a good CBM regime is an accurate and timely fault detection. A method has been developed to identify transients and detect fault in a Nuclear power plant in transients. This is to aid condition-based maintenance in a nuclear power plant. This method was achieved by using the nuclear plant simulator as a dynamic reference. At steady state, a fault is easily detected but in transients, it is difficult. This gives rise to the introduction of a machine-learning tool like artificial neural networks (ANN) to train both the simulator and plant parameters. The neural network outputs of the plant and simulator are then compared and this results in a better identification of faults in transients.
August 2018
Piecewise Diffusion Synthetic Acceleration scheme for neutron transport simulations in optically thick systems
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Fran
August 2018
A direct calculation method for subcritical multiplication factor in Reactor Monte Carlo code RMC
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Xiaotong Shang, Ganglin Yu, Jing Song, Kan Wang The subcritical multiplication factor $k s$ with the existence of external source neutrons is important for the evaluation of the neutron multiplication M, especially in the accelerator-driven system (ADS) performance assessment. The effective multiplication factor $k eff$ calculated from the traditional source iteration method in Monte Carlo codes can’t fully describe the subcritical system, acquiring the spurious neutron flux distribution. A direct Monte Carlo method called modified source iteration method by external source is introduced to calculate the subcritical multiplication factor $k s$ directly, acquiring the real neutron flux at the same time. Compared with Monte Carlo fixed source calculation, the modified source iteration method has a great advantage in calculating the subcritical multiplication factor $k s$ and its statistical error directly, especially for systems with complex geometry and a variety of fissile materials where much post-processing work on the tally results from fixed source calculation is needed. Moreover, much calculation time can be saved through this method when obtaining the subcritical multiplication factor $k s$ with similar precision. This method has already been implemented in Reactor Monte Carlo code RMC including parallel calculation capability and can be effectively applied to the analysis of accelerator driven subcritical system (ADS).
August 2018
Application of advanced Rossi-alpha technique to reactivity measurements at Kyoto University Critical Assembly
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Chidong Kong, Jiwon Choe, Seongpil Yum, Jaerim Jang, Woonghee Lee, Hanjoo Kim, Wonkyeong Kim, Khang Hoang Nhat Nguyen, Tung Dong Cao Nguyen, Vutheam Dos, Deokjung Lee, Ho Cheol Shin, Masao Yamanaka, Cheol Ho Pyeon This study presents the first application of the advanced Rossi-alpha method (theoretically introduced by Kong et al., 2014) on the reactivity measurements in a research reactor: detector count signals at the Kyoto University Critical Assembly (KUCA) facility. The detector signals in the KUCA A-type core are analyzed by three subcriticality measurement methods: (1) Feynman-alpha (F-
August 2018
A new Krylov subspace method based on rational approximation to solve stiff burnup equation
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Xuezhong Li, Jiejin Cai A burnup equation can be solved with matrix exponential method and its solution can be written as $n ( t ) = e At n ( 0 )$. In burnup calculation, general Krylov Subspace Method can solute a matrix–vector efficiently in a subspace but fails to keep a high precision. To solve this problem, a new kind of Krylov Subspace Method, Generalized Minimal Residual Method (GMRES) is implemented, based on a rational approximation method. It shows its great advantage in computation speed, which is more than four times faster than the same kind of rational approximation solved in a whole space while its accuracy is also guaranteed. Some optimizations, such as shift-Invariant technique, precondition technique and restart technique, are also implemented on burnup calculation.
August 2018
Control of the reactor core power in PWR using optimized PID controller with the real-coded GA
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Seyed Mohammad Hossein Mousakazemi, Navid Ayoobian, Gholam Reza Ansarifar Load following is an importance topic in the Nuclear Power Plants (NPPs). One of the conventional and simplest ways is the use of Proportional-Integral-Derivative (PID) controller. The reactor power is simulated based on the point kinetic model. PID gains of a nonlinear time-varying system (a PWR NPP) are optimized and scheduled using real-coded genetic algorithm (GA). To this end, the objective function of the decision variables, include the overshoot, settling time and stabilization time (based on the Lyapunov approach) of the system is minimized. The presented control system track demand power level change within a wide range of time. The simulation results demonstrate good stability of this method and show high performance of the optimized PID gains to adapt any changes in the output power.
August 2018
Efficacy analysis of hydrogen mitigation measures of CANDU containment under LOCA scenario
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Wonjun Choi, Seon Oh Yu, Sung Joong Kim Mitigation of hydrogen risk has been a critical objective to assure ultimate safety of the nuclear power plant (NPP). Since the CANDU containment showed potential vulnerability over hydrogen explosion during severe accident, efficacy of hydrogen mitigation measures equipped in the current CANDU NPP need to be investigated thoroughly. Herein we report the effect of mitigation measures such as spray, igniter, passive autocatalytic recombiner (PAR), local air cooler (LAC), and filtered containment venting system (FCVS) on the hydrogen behavior using MELCOR 2.1 code. Wolsong Unit 1 was selected for target NPP and Loss of Coolant Accident (LOCA) was assumed as a plausible severe accident scenario. The analysis showed that the hydrogen concentration in the lower region of the containment was higher in the unmitigated base case. The application of FCVS with other safety measures successfully prevented the containment failure. Expected safety function of the FCVS was to filter out the fission products but substantial amount of oxygen was also vented out. This leads to the oxygen starvation in the containment and prevention of the containment failure in the long run.
August 2018
False alarm reducing in PCA method for sensor fault detection in a nuclear power plant
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Wei Li, Minjun Peng, Qingzhong Wang Principal component analysis (PCA) is applied for fault detection of sensors in a nuclear power plant (NPP) in this paper. In order to reduce the false alarms of T2 and Q statistics during fault detection, two different methods are further proposed in this paper. One is statistics-based method which generates second confidence limits for T2 and Q statistics, and then false alarms are reduced based on the second confidence limit during test. The other is iteration-based method which reduces false alarms during modeling. Measurements beyond the first confidence limit of T2 or Q statistics are successively removed from the training data through iteration process. Finally, sensor measurements from a real NPP are acquired to train and test the proposed methods. On one hand, simulation results show that the proposed PCA model is capable of detecting the faulty sensors no matter with small or major failures. On the other hand, simulations also indicate that the PCA model combined with statistics-based and iteration-based methods simultaneously makes more contribution to the timeliness and effectiveness of sensor fault detection compared with the PCA model only with statistics-based or iteration-based method.
August 2018
Wide pressure range condensation modeling on pure steam/steam-air mixture inside vertical tube
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Young-Jong Chung, Hyungjun Kim, Ho-Sik Kim, Soo-Hyung Kim, Kyoo-Hwan Bae The elimination of active systems and their replacement with passive systems are emphasized to improve a system reliability in advanced nuclear power plants such as SMART. For accident situations, a condensate heat exchanger having vertical tubes is an ultimate heat sink to remove residual heat generated in the core. To adequately analyze the system behaviors, a realistic condensation model with pure steam or a steam-air mixture in the condensate heat exchanger is important for the various thermal hydraulic conditions. Improved correlations are proposed to analyze the thermal hydraulic behaviors for wide pressure conditions of up to 6
August 2018
Thermal hydraulic analysis of loss of flow accident in the JRR-3M research reactor under the flow blockage transient
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Yuchuan Guo, Guanbo Wang, Dazhi Qian, Heng Yu, Bo Hu, Simao Guo, Xiangmiao Mi The transient thermal-hydraulic behavior of loss of flow accident (LOFA) in the JRR-3M 20MW pool-type research reactor under flow blockage is investigated using RELAP5/MOD3.4 code. The core is divided into 7 hot channels, 1 average channel and 1 bypass channel to take into account the interaction between the blocked channel and adjacent channels. Blockage ratios considered in the work includes 0%, 40%, 50%, 60%, 70%, 80% and 100%. MDNBR and fuel central temperature are investigated to estimate the integrity of the fuel plate. Compared with the accident consequence of flow blockage, the results indicate that it is more likely for LOFA under flow blockage to occur departure from nucleate boiling (DNB) in the same blockage ratio. Even when the flow channel is totally blocked, the fuel temperature is still within the safety limit, which is set as 400
August 2018
A new dimensionless thermal hydraulics parameter for the heat exchangers
Publication date: August 2018
Source:Annals of Nuclear Energy, Volume 118 Author(s): Muhammad Ilyas, Fatih Aydogan Thermal and hydraulic considerations are crucial for the design of an efficient heat exchanger (HX) which allow boiling. Steam generator is one of the biggest and most expensive components of most nuclear power plants. Therefore, design perfections of heat exchangers can lead to improved design of steam generators. Two-phase heat transfer coefficient (HTC) strongly depends on the prevailing flow regime for a given surface pattern as well. The enhancement in HTC is associated with an increase in frictional pressure drop. A single parameter for characterization of heat transfer surfaces based on thermal and hydraulic consideration doesn’t exist. A new dimensionless number for thermal and hydraulics characterization (DITCH) of the heat exchanger surface has been derived for evaluation of heat transfer surfaces. It captures the hydraulic behavior of coolant due to phase change and wall friction. Its value characterizes different flow regimes of two phase flow. Analysis of a selected data reveals that DITCH increases with quality ‘x’ and the mass flow rate ‘