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.

Subplane collision probabilities method applied to control rod cusping in 2D/1D

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August 2018

An improved convergence rate for the prompt

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.

The sensitivity analysis for IHTS and SG due to the Large-scale Sodium-Water Reaction event in PGSFR

Publication date:

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.

Novel genetic algorithm for loading pattern optimization based on core physics heuristics

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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

A Next Generation Method for Light Water Reactor core analysis by using Global/Local Iteration Method with SP3

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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.

Continuous machine learning for abnormality identification to aid condition-based maintenance in nuclear power plant

Publication date:

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

Piecewise Diffusion Synthetic Acceleration scheme for neutron transport simulations in optically thick systems

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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 ${k}_{\mathit{eff}}$ 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).

A direct calculation method for subcritical multiplication factor in Reactor Monte Carlo code RMC

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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-

Application of advanced Rossi-alpha technique to reactivity measurements at Kyoto University Critical Assembly

Publication date:

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 $\text{n}(\text{t})={e}^{\mathit{At}}n(0)$. 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.

A new Krylov subspace method based on rational approximation to solve stiff burnup equation

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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.

Control of the reactor core power in PWR using optimized PID controller with the real-coded GA

Publication date:

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.

Efficacy analysis of hydrogen mitigation measures of CANDU containment under LOCA scenario

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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.

False alarm reducing in PCA method for sensor fault detection in a nuclear power plant

Publication date:

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

Wide pressure range condensation modeling on pure steam/steam-air mixture inside vertical tube

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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

Thermal hydraulic analysis of loss of flow accident in the JRR-3M research reactor under the flow blockage transient

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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 di mensionless number for t hermal and hydraulics ch aracterization (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 ‘ m

A new dimensionless thermal hydraulics parameter for the heat exchangers

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August 2018

Burnup characteristics analyses of graphite impurities in HTGR fuel element

Publication date:August 2018

**Source:**Annals of Nuclear Energy, Volume 118 Author(s): Jian Li, Ding She, Lei Shi In the fuel element of high-temperature gas cooled reactors (HTGRs), there exist various impurities in the graphite matrix, which are mainly originated from the raw materials, such as naturally occurring graphite, artificial graphite and resin. These impurities can reduce the excess reactivity of the reactor core, both in the initial criticality and during the overall operation history. In the practical physics design of HTGR, the equivalent boron content (EBC) method is usually used to describe the graphite impurities, where the impurity isotopes are simply converted to some amounts of boron by preserving the neutron absorption. Although it is found to be effective in criticality calculations, EBC may introduce errors in burnup analyses. This paper investigates the burnup characteristics of the graphite impurities. According to their behaviors during burnup, the impurities are classified into three categories, i.e. the high-burnable isotopes, low-burnable isotopes and unburnable isotopes. Moreover, a revised equivalent boron content (REBC) is proposed to describe the absorption rates of impurities, which can improve the computational accuracy in the burnup analyses of HTGR fuel.

Burnup characteristics analyses of graphite impurities in HTGR fuel element

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July 2018

Study on bubble and liquid velocities in an area-varying horizontal channel

Publication date:August 2018

**Source:**Annals of Nuclear Energy, Volume 118 Author(s): Thanh Tram Tran, Byoung Jae Kim, Hyun-Sik Park Two-fluid equations are widely used to simulate thermal-hydraulic phenomena in a nuclear reactor. Simulation accuracy depends on the modeling terms in the two-fluid equations. For a dispersed flow, the overall two-phase pressure drop by wall friction must be apportioned to each phase in proportion to the fraction of each phase (Kim et al., 2014). By applying this approach, the prediction of bubble phase velocity can be close to that of liquid for a fully developed flow in a horizontal pipe with a constant area. One may want to know what would happen in the area-varying channels. It is always true that the bubble density is much lower than the water density. Hence, the bubble would accelerate faster than the liquid in a nozzle in which the pressure decreases along the downstream; the bubbles would decelerate more quickly than the liquid in a diffuser in which the pressure increases along the downstream. The purpose of this study was to investigate those behaviors in an area-varying channel using the experimental data and MARS simulations. Experiments were made of turbulent bubbly flows in an area-varying horizontal channel. The velocities of two phases were measured with the help of the PIV technique. The experimental result showed that the two-phase velocities were no longer close to each other in the area-varying regions. The bubble was faster than the liquid in the nozzle region; in contrast, the bubble was slower than the liquid in the diffuser region. MARS code simulations were performed to assess the wall drag model. By replacing the original wall drag partition model in the MARS code with Kim’s one, the simulation results were consistent with experimental observations.

Study on bubble and liquid velocities in an area-varying horizontal channel

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July 2018

Molten Salt Reactors and Thorium Energy, edited by Thomas J. Dolan

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): J.L. Kloosterman

Molten Salt Reactors and Thorium Energy, edited by Thomas J. Dolan

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July 2018

Numerical analysis on the dynamic behaviors of a graphite-moderated molten salt reactor based on MOREL2.0 code

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Kun Zhuang, Liangzhi Cao Molten salt reactors (MSRs) are selected as one of six generation-IV nuclear power systems due to its remarkable advantages in terms of inherent safety, nuclear non-proliferation and economy. Liquid fuel makes the neutronics and thermal-hydraulics characteristics of MSRs different from those of conventional solid fuel reactors. In this study, transients perturbed by control rod ejection, local perturbation and over feeding U235 were analyzed for a graphite-moderated MSR based on a 3D coupled neutronics/thermal hydraulics code MOREL2.0. The effective delayed neutron fraction decreased with the rise of fuel mass flow. The core power changed more fiercely and even occurred supercritical phenomenon for larger fuel mass flow when the same reactivity was introduced. Local blockage introduced negative reactivity, and MSR maintained a safety state. While local overcooling resulted in positive reactivity introduction, and core power increased first then tended to be stable due to the negative temperature feedback. Furthermore, the response of core power at the transient of over feeding fissile isotopes was simulated. Those numerical results provide valuable information for the research and design of this new generation reactor.

Numerical analysis on the dynamic behaviors of a graphite-moderated molten salt reactor based on MOREL2.0 code

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July 2018

Application of three-dimensional looped network analysis method to the core of prismatic very high temperature gas-cooled reactor

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Jeong-Hun Lee, Hyoung Kyu Cho, Goon-Cherl Park A hydrogen production system coupled to very high temperature gas-cooled nuclear reactor (VHTR) is considered to be one of the most promising ways to achieve massive hydrogen production. The core of the prismatic VHTR consists of hexagonal graphite fuel blocks and reflector blocks. There exist gaps between graphite blocks; the vertical gap is referred to as a bypass gap and the horizontal gap is referred to as a cross gap. The coolant flows through these gaps as well as the coolant channels, thereby forming a complicated flow field in the core. A looped network analysis method has been introduced for fast and simple flow distribution analysis. It has great strengths in analyzing a complex network in a short computational time; however, the looped network analysis has been used generally to analyze two-dimensional flow networks in practical applications. Because the flow network of the core of prismatic VHTR is three-dimensional, a new methodology to apply the looped network analysis to a three-dimensional flow network was proposed in this study. This paper introduces the geometrical configuration of the reactor core of the prismatic VHTR, the looped network analysis method and its extension to the three-dimensional flow network, applied constitutive relations for closure and finally, the validation results against available multi-block bypass flow experimental data.

Application of three-dimensional looped network analysis method to the core of prismatic very high temperature gas-cooled reactor

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July 2018

Analysis of a small-scale reactor core with PARCS/Serpent

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): F. Fejt, J. Frybort Cross section homogenization is a challenging task that has been implemented mainly for full-core calculations of nuclear power plants. Small-scale reactors started to be a main point of interest for this kind of analysis during last few years. This paper presents a suitable homogenization strategy for a small core of VR-1 reactor (28.6

Analysis of a small-scale reactor core with PARCS/Serpent

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July 2018

Application of an algebraic turbulent heat flux model to a backward facing step flow at low Prandtl number

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Andrea De Santis, Afaque Shams Turbulent heat transfer represents a considerably challenging phenomenon from the modelling point of view. In the RANS framework, the classical Reynolds analogy provides a simple and robust approach which is widely employed for the closure of the turbulent heat flux term in a broad range of applications. At the same time, there is an ever growing interest in the development and assessment of advanced models which would allow, at least to some extent, for the relaxation of the simplifying assumptions underlying the Reynolds analogy. In this respect, the use of algebraic closures for the turbulent heat flux has been proposed in the literature by different authors as a viable approach. One of these algebraic closures has been extended for its application to low Prandtl number fluids in various flow regimes, by means of calibration and assessment of the model against some basic test cases, in what is known as the AHFM-NRG+ model. In the present work the AHFM-NRG+ is applied for the first time to a relatively complex configuration, i.e. a backward facing step in both forced and mixed convection regimes with a low Prandtl working fluid, and assessed against reference DNS data. The obtained results suggest that the AHFM-NRG+ is able to provide more accurate predictions for the thermal field within the domain and for the heat transfer at the wall in comparison to the Reynolds analogy assumption. These encouraging results indicate that the AHFM-NRG+ can be considered as a promising model to improve the accuracy in the simulation of the turbulent heat transfer in industrial applications involving low Prandtl fluids.

Application of an algebraic turbulent heat flux model to a backward facing step flow at low Prandtl number

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July 2018

Solution of the stochastic point kinetics equations using the implicit Euler-Maruyama method

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Daniel Suesc

Solution of the stochastic point kinetics equations using the implicit Euler-Maruyama method

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July 2018

Confirmation of Wilks’ method applied to TRACE model of boiling water reactor spray cooling experiment

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Travis Mui, Tomasz Kozlowski Wilks’ formula has been frequently used to quantify the minimum amount of computational work required to meaningfully assess a model’s uncertainty, due to its nonparametric statistical nature that does not require knowledge of the distribution of the qualifying parameters of interest. Additionally, this method allows for any number of input uncertain parameters in the simulation model. This is favorable due to considerable computational expense of typical nuclear safety simulations, providing a quantifiable number of code executions that can statistically verify a desired level of safety. However, there are various existing definitions and uses of Wilks’ theorem in such scenarios, which the present study aims to investigate and quantify for a real thermal-hydraulics experiment used for reactor safety licensing. In this work, the U.S. NRC TRACE thermal-hydraulics code was chosen to simulate the separate-effect spray cooling tests performed by ASEA-ATOM for licensing of BWR SVEA-64 fuel. The computational model was evaluated by performing forward uncertainty quantification (UQ) using Dakota as the analysis tool and code driver on 31 identified sensitive parameters. Using this validated model, the TRACE model was sampled 1000

Confirmation of Wilks’ method applied to TRACE model of boiling water reactor spray cooling experiment

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July 2018

Simulation of VVER-1000 startup physics tests using the MCU Monte Carlo code

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Artem Bikeev, Mikhail Kalugin, Anna Shcherenko, Denis Shkarovsky The main objective of this work is to make the quality test and verification of a full-scale computer model of the VVER-1000 reactor with the detailed description of the geometry and material composition. Using the computer model and MCU Monte Carlo code, we simulated the experiments conducted at the stage of physical start-up of the third power unit of Rostov NPP (Russia) in December 2014. The following neutron-physical characteristics were calculated: effective neutron multiplication factor, critical boric acid concentration, efficiency of single control rod cluster, integral and differential efficiency of control rod group, scram system reactivity worth, and reactivity coefficients. The calculated neutron-physical characteristics were compared with the measured ones. The comparison shows agreement with the measured data over the range of the experimental error and statistical uncertainty of the calculation. To reduce the impact of statistical uncertainty on the calculated differential control rod reactivity worth and reactivity coefficients we applied a special approach for the estimation of these parameters.

Simulation of VVER-1000 startup physics tests using the MCU Monte Carlo code

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July 2018

Improvement in the simulation of detector readings using a high fidelity local flux reconstruction-based method

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Mohamed Dahmani, Brian Phelps, Tahar Mohamed Sissaoui Even with current computing capabilities, detailed full core three-dimensional (3-D) transport calculations are still not practical. However, if we are satisfied with knowing only the average values of spatial flux distributions, the 3-D diffusion solution will constitute the final solution. On the other hand, in reactor design and safety analysis, direct information about the local flux distribution for the heterogeneous assemblies is required to assess the design and determine the safety margins. For this reason, after having solved the full-reactor-core problem, we have to look into the possibilities of recovering in a second step the information on local properties of single heterogeneous assemblies. In particular, the detector readings at detector locations are derived using these global homogenized parameters by applying appropriate numerical methods such as advanced interpolations. In this paper, we propose a method based on flux reconstruction to calculate the simulated detector readings in three-dimensions with high fidelity. Data from detector readings are very important in ensuring optimal reactor operations as well as in detecting any deviations from normal operations. Thus, calculating the detector readings with high fidelity will allow improvements to operating and safety margins. To validate this method, comparisons between detector reading simulation results and measurements from an operating CANDU reactor will be conducted and results will be presented.

Improvement in the simulation of detector readings using a high fidelity local flux reconstruction-based method

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July 2018

Validation progress and exploratory analyses of three-dimensional simulation code for BWR in-vessel core degradation

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Tsuyoshi Okawa Based on the information obtained from the severe accident at the Fukushima Daiichi Nuclear Power Plant Units 1–3 in 2011, Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) started development of a detailed three-dimensional simulation code for in-vessel core degradation. The simulation code aims at applying to BWR core degradation assessment with three-dimensional geometry. The present paper describes the progress of validation and the assessment of BWR in-vessel core degradation for future application by parametric surveys. In-vessel core degradation phenomena were validated mainly using the CORA-18 experiment at Karlsruhe Institute of Technology to estimate axial- and lateral-directional motion of molten corium. Additionally, metallic melt relocation around a BWR core support plate and molten corium breakup in a lower head were validated by the XR-2 experiment at Sandia National Laboratory, and the FARO L-19 and the KROTOS K-37 experiments at Joint Research Centre facilities of the European Commission, respectively. The code results match severe accident relevant phenomena qualitatively well. Furthermore, as a validation process, the exploratory analyses using 1/4 Sector-Core Geometry in a hypothetical severe accident condition have been carried out for a typical middle-power-level BWR with the following variable parameters as initial core conditions: a water level, a decay heat level, a radial power shape and an oxidation of fuel cladding. These efforts showed that the code could be used for application to the actual core degradation evaluation with three-dimensional geometry in the near future.

Validation progress and exploratory analyses of three-dimensional simulation code for BWR in-vessel core degradation

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July 2018

Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Yuan Gao, Christopher R. Hughes, Christopher R. Greulich, James E. Tulenko, Andreas Enqvist, James E. Baciak This study presents a detailed comparison of the dose rate distributions of a dry TN-32 fuel cask with two geometry models and two cross-sectional datasets. The accuracies of radiation dose rate estimation and computational efficiencies of each geometry model with two cross-sectional data-sets are compared. The use of automated variance reduction techniques can significantly improve the computational efficiency of such a realistic, deep penetration problem that involves radiation transport from different volumetric sources, thereby eliciting only a small statistical error. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is a computational sequence within the SCALE 6.2 code package based on consistent adjoint driven importance sampling (CADIS), a type of automated variance reduction technique. Homogenous and full fuel assembly models are built herein, and two nuclear cross-section libraries (V7-200N47G and continuous energy) are applied in this work. Based on the detailed comparisons, we found that neutron dose rate estimation is more dependent on geometry modeling than on cross-section data. For neutron-induced gamma rays, the dose rate distribution depends on both the spatial self-shielding effect and the cross-section library. The primary gamma rays respectively contribute to the total dose rate by

Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

Publication date:

July 2018

Comparative analysis on the influence of the MAAP4 phenomenological model parameters for the severe accident source term for different plant designs and accident scenarios

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Seong Woo Kang, Man-Sung Yim Considerable efforts have been taken to estimate the radioactive source term by using integrated computer models, but uncertainties exist in using these models, partly due to the fact that the results cannot be validated using real-world large-scale experiments. The purpose of this research is to identify the MAAP4 phenomenological model parameters with large influences in predicting environmental releases. The identified parameters were compared for different accident scenarios and plant configurations. Two-step screening of the influential model parameters through the Latin hypercube-rank correlation and the one-at-time method was used for the comparative analysis. From the integral analyses, it was found that the source term error propagation was much larger in the containment and in the secondary system. However, error propagation was not necessarily larger inside the containment than in the secondary system. There are common phenomenological parameters with significant influences on the source term simulations: primary depressurization, molten core-concrete interaction, and cladding oxidation models. More focused research regarding the effect of aforementioned identified phenomenological models on the source term calculations would be necessary to better predict the amount of environmental radionuclides release for the severe accidents. However, the detailed uncertainty and sensitivity analysis involving phenomenological model parameters may have to be done in a plant-specific way. Some phenomenological model parameters that showed significant impact on the global source term uncertainty in one plant type did not necessarily have the same impact on another plant. A sensitivity study involving a typical PWR may not have same implications in other designs, even if a same accident scenario with similar accident phenomena is analyzed. Such sensitivity study would be much more beneficial for standardized plant designs.

Comparative analysis on the influence of the MAAP4 phenomenological model parameters for the severe accident source term for different plant designs and accident scenarios

Publication date:

July 2018

Investigations of sample reactivity worth measurement in a fast neutron reactor with the inverse kinetics method

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Shumiao Wang, Haojun Zhou, Zhongxiong Bai, Xiaoqiang Fan, Yanpeng Yin A new method for the measurement of sample reactivity worth in a fast neutron reactor named the inverse kinetics method is proposed in the paper. The sample reactivity worth could be obtained by measuring the reactivity step change in the process of sample fetching and placing in the delayed critical reactor. Compared with the traditional period method, the advantage is that the accuracy of reactor reactivity control will not exert any influence on the uncertainty of reactivity worth measurement. The inverse kinetics method has been used to measure the reactivity worth of Ô20

Investigations of sample reactivity worth measurement in a fast neutron reactor with the inverse kinetics method

Publication date:

July 2018

Adaptive expansion order for diffusion Variational Nodal Method

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Boning Liang, Hongchun Wu, Yunzhao Li The Variational Nodal Method (VNM) has been employed as the diffusion module in our PWR core analysis code Bamboo-Core within our PWR fuel management code system NECP-Bamboo. It expands the nodal volumetric flux and surface partial currents into the sums of orthogonal basis functions without using the transverse integration technique. To reduce the extra computing cost by the uniform expansion order setting, an adaptive expansion order technique has been developed in this paper. After estimating the net currents between each pair of neighboring nodes by using the Coarse-Mesh Finite-Difference (CMFD) technique, it estimates the required expansion orders in each node analytically. This technique increases the complexity of the code, but reduces the computational efforts both in computing time and memory storage by a factor of about 5 and 4, respectively. In addition, the CMFD acceleration is also employed to further improve the performance of the code. It is demonstrated by the numerical results that the CMFD acceleration technique can provide a speedup ratio of about 17.

Adaptive expansion order for diffusion Variational Nodal Method

Publication date:

July 2018

A deterministic method for PWR loading pattern optimization

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Fariz B. Abdul Rahman, John C. Lee, Fausto Franceschini A new deterministic approach to core loading optimization is presented. Selection of fuel enrichment and burnable poison (BP) arrangement in a pressurized water reactor (PWR) core are optimally determined subject to power peaking factor (PF) constraints. The deterministic approach is based on the method of Lagrange multipliers and the direct adjoining approach for the inequality power PF constraint. The optimality conditions are derived using calculus of variations resulting in the presence of jump conditions and a linear control Hamiltonian. The optimal control problems are addressed by simultaneously minimizing the Hamiltonian using the gradient method and performing Newton's method in different regions of the core to produce viable solutions for the controls. Without active control of the power profile in PWRs during depletion, our methodology is able to satisfy the power PF constraint at all times by proper loading selection at the beginning of cycle (BOC). This is achieved by promoting the proper burnup path forward in time during the calculation of the forward optimality conditions. This information is utilized in the adjoint burnup calculation, which acts to propagate the optimal control path back to the BOC during the calculation of the adjoint optimality conditions. The methodology has been developed into a new code DMCO (Deterministic Multi-Control Optimization) which can produce two-dimensional two-group depletion results in approximately seven minutes per optimization iteration on a personal computer. No application of heuristics is required and the code is capable of finding solutions from initial control distributions. A preliminary case study is presented for the Westinghouse AP600 first cycle fuel loading design which achieved an extended fuel cycle while satisfying power PF constraints.

A deterministic method for PWR loading pattern optimization

Publication date:

July 2018

Neutronic analysis of the LVR-15 research reactor using the PARCS code

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Antonio Dambrosio, Marek Ru

Neutronic analysis of the LVR-15 research reactor using the PARCS code

Publication date:

July 2018

Comparison of HELIOS-2.1 and SCALE-6.1 codes on pin-cell model

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Sabina Maharramova, William Beere, Knut Eitrheim, Ole Reistad, Tord Walderhaug Deterministic HELIOS-2.1 and SCALE-6.1 codes are compared using pin-cell models for light water reactor (LWR) and heavy water reactor (HWR) cases. The main objective of this study is to identify the origins of any discrepancies between compared codes. The infinite multiplication factor kinf , flux distribution, absorption, fission, production reaction rates, and burn-up dependent concentrations of major fuel isotopes, are investigated herein and compared. Comparison of kinf has shown that the codes are in good agreement for both the LWR and HWR cases. The codes showed differences in the isotope number density of up to 6% in the case of prominent isotopes, and for 235U and 239Pu at 60

Comparison of HELIOS-2.1 and SCALE-6.1 codes on pin-cell model

Publication date:

July 2018

Fast and accurate GPU PWR depletion calculation

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): A. Heimlich, F.C. Silva, A.S. Martinez Fast and accurate simulations of isotopic inventory are of fundamental importance to the conceptual design, refueling, and disposal of fuel from nuclear power plants (NPP). The determination of the fuel’s isotopic composition requires a high computational effort arising from the complexity of solving the large system of coupled ordinary differential equations (ODE’s). The system of ODE’s is related to the physical mesh, fuel and coolant temperatures, time history of power, previous isotopic concentrations, and radioactive decay chains under analysis. This study surveys two methods used to simulate fuel burn-up on pressurized water reactors (PWR) implemented in Graphics Processor Unit (GPU). The accuracy of methods was also studied, by comparing the inventory simulation of one cycle burn-up looking for the benchmark obtained from Chebyshev Rational Approximation Method (CRAM), and shows the Square-Root-Mean-Error (SRME) less than 0.0001%. A performance comparison of sequential version and GPU methods exhibit a speed improvement exceeding one hundred times.

Fast and accurate GPU PWR depletion calculation

Publication date:

July 2018

On the use of CALPHAD-based enthalpy-temperature relations in suboxidized corium plane front solidification modelling

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): V. Tiwari, R. Le Tellier The knowledge of in-vessel corium behaviour and the associated risk of vessel failure are matters of prime interest within the framework of Severe Accident studies in a Light Water Reactor. As corium behaviour is dependent on associated thermochemical and thermohydraulic phenomena, its modelling within integral codes requires coupling between lumped parameter thermal models and thermochemical models. Such integral thermal models consist of mass and energy conservation equations that require inputs related to thermochemical properties of the materials, which are closely related to the state variables. In particular, the closure of energy conservation equations requires enthalpy-temperature relations. In the framework of multicomponent systems, the dependence of such relations to chemical composition is of importance and should be treated adequately to obtain a more accurate description of the phases depicted by the model. An approach to do so is to keep a general formulation of energy conservation equations in terms of specific enthalpies instead of substituting simplified enthalpy-temperature relations on a case-by-case basis in order to obtain models with an explicit temperature formulation. The enthalpy-temperature relation is then considered as an “Equation-of-State” (EOS) that can be written as: H : T , ( w j ) j

PIV study of velocity distribution and turbulence statistics in a rod bundle

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Xing Li, Zhengpeng Mi, Sichao Tan, Ruiqi Wang, Xiaoyu Wang In reactor fuel assembly in PWR, spacer grids with mixing vanes are used in rod bundles to enhance flow mixing and heat transfer. Thus, the spacer grids have significant influence on the economy and safety of reactors. This paper focuses on visualization research on the single phase flow in rod bundle with spacer grids by Particle Image Velocimetry (PIV), and a measurement of multi-subchannels in the both transverse and longitudinal directions was conducted at the Reynolds number of 10,400. Matching Index of Refraction (MIR) technique was implemented to accurately measure the internal flow field in the rod bundle. The flow statistics of rod bundle were obtained, such as mean velocity, turbulence intensity and Reynolds stress. The velocity distribution and turbulence statistics in the longitudinal flow field are presented downstream the spacer grid. The streamwise development of symmetrical vortex array in the transverse flow field is also evaluated. A comparison of Reynolds stresses in the longitudinal and the transverse flow field is performed to reveal the momentum transfer process in three-dimensional flow field. Experimental results benefit the development and evaluation of spacer grid, and also provide the whole-field validation for numerical simulations.

view: 123
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On the use of CALPHAD-based enthalpy-temperature relations in suboxidized corium plane front solidification modelling

Publication date:

July 2018

Dependence assessment in human reliability analysis using an evidential network approach extended by belief rules and uncertainty measures

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Xinyang Deng, Wen Jiang Because of the potential relevance among human errors, dependence assessment for human actions plays a very important role in human reliability analysis. Several typical methods have been developed for that task. However, in previous studies various uncertainties in analyst’s judgment and expert’s knowledge for dependence assessment is not fully taken into consideration, especially the epistemic uncertainty in expert’s knowledge is often ignored. In this paper, a belief function theory is employed to simultaneously model the probabilistic uncertainty and epistemic uncertainty within analyst’s judgment and expert’s knowledge. Mainly, a novel evidential network approach extended by belief rules and uncertainty measures is proposed, then based on that a new framework for dependence assessment is presented and its effectiveness is validated through an illustrative case study. This work, on one hand, gives an extended evidential network model on the basis of belief rules and uncertainty measures to implement dimension reduction and uncertainty reasoning; On the other hand, it presents a novel and effective framework for dependence assessment in human reliability analysis.

Dependence assessment in human reliability analysis using an evidential network approach extended by belief rules and uncertainty measures

Publication date:

July 2018

Numerical investigation of the core outlet temperature fluctuation for the lead-based reactor

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Lizhi Wang, Guowei Wu, Jin Wang, Ming Jin, Yong Song The temperature fluctuations induced by incomplete mixing of coolants with different temperature may cause thermal fatigue at the components of the lead-based reactor core outlet. Thus the accurate analysis of the phenomenon is very crucial for reactor safety operation. In this paper, the temperature fluctuations of the lead-based reactor core outlet were simulated by using large eddy simulation (LES) method in the simplified core outlet models. In order to analyze the temperature fluctuation sensitivity for the fuel assembly design parameters, such as the fuel assembly size and the gap between two adjacent fuel assemblies, five geometry models were constructed with different fuel assembly design parameters. The time histories of temperature fluctuations at different monitoring points on the center of three fuel assemblies were obtained. Then the amplitudes and the power spectrum density (PSD) of temperature fluctuations were analyzed, in order to compare temperature fluctuations of different geometry models at the same locations of core outlet. Finally the distribution characteristics of core outlet temperature fluctuations were obtained in axial directions, and the temperature fluctuation sensitivity with fuel assembly parameters was also analyzed based on the amplitudes, PSD and the normalized root-mean square temperature analysis. It is found that the temperature fluctuation intensity is enhanced with the increase of the gap size between adjacent fuel assemblies and the opposite edge width of each fuel assembly. The analysis results could provide important references for optimized design and engineering guidance of lead-based reactor.

Numerical investigation of the core outlet temperature fluctuation for the lead-based reactor

Publication date:

July 2018

Application of continuous adjoint method to steady-state two-phase flow simulations

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Guojun Hu, Tomasz Kozlowski Verification, validation and uncertainty quantification (VVUQ) have become a common practice in thermal-hydraulics analysis. An important step in the uncertainty analysis is the sensitivity analysis of various uncertainty input parameters. The common approach for computing the sensitivities, e.g. variance-based and regression-based methods, requires solving the governing equation multiple times, which is expensive in terms of computational effort. An alternative approach for computing the sensitivities is the adjoint method. The cost of solving an adjoint equation is comparable to the cost of solving the governing equation. Once the adjoint solution is obtained, the sensitivities of various parameters can be obtained with little effort. However, successful adjoint sensitivity analysis of the two-phase flow is rare. In this work, an adjoint sensitivity analysis framework is developed for the two-phase two-fluid model based on a new upwind numerical solver. The adjoint sensitivity analysis framework is tested with a steady-state boiling pipe problem. Results show that the adjoint sensitivity analysis framework is working as expected. The sensitivities obtained with the adjoint method are verified by the sensitivities obtained with a forward method.

Application of continuous adjoint method to steady-state two-phase flow simulations

Publication date:

July 2018

Benchmark problems in aerosol evolution: Comparison of some exact and DSMC results

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Isaac Saldivar, Fernando De La Torre Aguilar, Matthew Boraas, Sudarshan K. Loyalka Aerosols are generated in many normal or accident situations associated with the nuclear enterprise. For a good understanding and modeling of the nuclear source term, for example, good experimental data and computational programs relating to aerosol evolution are needed. In the past several years there has been an effort to explore use of the Direct Simulation Monte Carlo (DSMC) approach for such estimations to improve fidelity of computations to the actual physics and chemistry of the accidents. An integral part of these efforts has been verification and validation of the DSMC technique against other available results wherever possible. This paper explores verification of DSMC against one existing and two new benchmark problems covering condensation, coagulation, deposition and two-component aerosols. The simulations compare well with the exact results, providing further confidence in the use of DSMC.

Benchmark problems in aerosol evolution: Comparison of some exact and DSMC results

Publication date:

July 2018

Simulation analysis of an open natural circulation for the passive residual heat removal in IPWR

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Nan Jiang, Minjun Peng, Tenglong Cong In this paper, an innovative passive safety system (PSS) for the integrated pressurized water reactor (IPWR) IP200 was designed, which aims to compact the equipment layout of PSS and apply in floating nuclear power plants. In this design, the reactor vessel (RV) is connected directly to the containment to form an open loop, and a natural circulation can be established in the loop for the residual heat removal. In order to evaluate the mitigation of PSS conservatively, an extreme scenario of SBLOCA (small break LOCA) along with station blackout (SBO) is simulated by Relap5 code. The temperature distribution and flow characteristics during the transient and the long-term cooling are both calculated to explain the pivotal thermal–hydraulic phenomena in the PSS. Furthermore, characteristics under different thermal boundaries and structural parameters are compared to discuss the coupling effects between circulation capacity and heat exhaust. The results show that the accumulation of non-condensable gas in containment with a proper pressure is conducive to restrain the instability of two-phase natural circulation. Nevertheless, a single-phase circulation will be formed eventually to reduce the heat transfer efficiency if the initial pressure is too high. Besides, increasing the heat transfer area of ultimate heat sink helps to enhance the cooling performance of natural circulation during the start transient, but it will increase the size of heat exchanger and reduce the averaged heat transfer coefficient of convection. This study determines the key thermal parameters of PSS and can provide a reference for practical engineering design.

Simulation analysis of an open natural circulation for the passive residual heat removal in IPWR

Publication date:

July 2018

Towards the accurate numerical prediction of thermal hydraulic phenomena in corium pools

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Afaque Shams The knowledge of heat transfer in corium pools is one of the important issues for corium retention, as it defines the safety margin for vessel integrity. In this regard, in the 90’s the BALI experimental program was performed at the CEA, France. The principal idea was to create a database regarding the heat transfer distribution at corium pool boundaries for in-vessel and ex-vessel configurations at high internal Rayleigh number (1015 to 1017). One of the tasks within the ongoing IVMR project, part of the HORIZON 2020 program, is to assess the up-to-date CFD turbulence models over a wide range of Rayleigh number for the homogenous pool tests of the BALI experiments. In the present study, the assessment of three different turbulence models is performed for two BALI test cases. These turbulence models include a linear k-

Towards the accurate numerical prediction of thermal hydraulic phenomena in corium pools

Publication date:

July 2018

Hougaard process stochastic model to predict wall thickness in Flow Accelerated Corrosion

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Mahendra Prasad, V. Gopika, Arunkumar Sridharan, Smrutiranjan Parida, Avinash J. Gaikwad Nuclear Power Plants (NPPs) operate at very high temperature and high fluid mass flow rate which are favourable for Flow Accelerated Corrosion (FAC) resulting in wall thickness reduction in pipes, bends and other geometries. The prediction of progressive reduction in pipe wall thickness is required for safety of operating NPP. In this paper, Hougaard process stochastic model is proposed for prediction of wall thickness reduction in pipes between two consecutive in-service-inspections. The probability distribution function (PDF) for the Hougaard process is computationally unstable near the origin. Hence, its saddle point approximation was used along with method of maximum likelihood (MLE) to derive the mathematical expressions for the three parameters with generally given time interval between in-service-inspection and corresponding changes in wall thickness. The gamma process model fit and linear fit to data have also been carried out. The predictions of change in pipe wall thickness from probabilistic model are validated by using (a) Experimental data for FAC for 58° bend pipe and (b) NPP feeder pipe data on FAC. The results compare well with the experimental and field data used in analysis.

Hougaard process stochastic model to predict wall thickness in Flow Accelerated Corrosion

Publication date:

July 2018

Lattice physics evaluation of 35-element mixed oxide thorium-based fuels for use in pressure tube heavy water reactors

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Ashlea V. Colton, Blair P. Bromley A series of 2-D lattice physics calculations with depletion were carried with WIMS-AECL Version 3.1 out as part of exploratory scoping studies to evaluate various thorium-based fuel bundle concepts for potential application in pressure tube heavy water reactors (PT-HWRs). Fuel bundles concepts investigated consisted of a cluster of 35 fuel elements arranged in two rings (14

Lattice physics evaluation of 35-element mixed oxide thorium-based fuels for use in pressure tube heavy water reactors

Publication date:

July 2018

Voxelization-based high-efficiency mesh generation method for parallel CFD code GASFLOW-MPI

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Fujiang Yu, Han Zhang, Yabing Li, Jianjun Xiao, Andreas Class, Thomas Jordan Hydrogen safety analysis is an important issue in the field of the nuclear severe accident. Fast, robust and accurate simulation of the complex hydrogen behavior in the nuclear containment is the first but critical step for hydrogen risk mitigation. GASFLOW-MPI is a widely used CFD numerical tool which provides the fast and reliable prediction, because of its parallel computational capability and well-validated models. However, due to its manual input-card based mesh generation, the whole meshing process is slow and lacks user friendliness. Therefore, a fast automatic mesh generation module is of importance and profit for practical industrial applications. Nowadays, Computer Aid Design (CAD) has become the formal standard for project design and preview. In this work, a mesh generation module is developed for GASFLOW-MPI, which directly uses the CAD file as the input for automatic mesh generation. The module exploits a voxelization-based method, which seeks to generate a Cartesian mesh by tracing rays directed into the geometry. The complexity of the mesh generation algorithm is also analyzed. Three models, including a toy problem, a steam generator compartment model, and a complex full scale reactor containment model are used to validate the new developed automated mesh generation module. The results show that meshes are generated at fast construction speed and well match the original CAD models.

Voxelization-based high-efficiency mesh generation method for parallel CFD code GASFLOW-MPI

Publication date:

July 2018

Thermal-hydraulic code for rewetting analysis in a PWR experimental loop

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Sabrina P. Alves, Amir Z. Mesquita, Hugo C. Rezende, Daniel A.P. Palma The safety of nuclear power plants is determined by their protection against the possible outcomes of postulated accidents. One of the most important accidents is the loss of coolant in the core (Loss-of-Coolant Accident – LOCA). A process of fundamental importance in the event of a LOCA in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Centre (CDTN) has been developing programmes since the 1970s to make Brazil independent in the field of reactor safety analysis. To that end, a Rewetting Test Facility (ITR in Portuguese) was designed, assembled and commissioned. This facility aims at investigating the phenomena involved in the thermal hydraulic reflood phase of a LOCA in a PWR nuclear reactor. The objective of this work is the analysis of physical and mathematical models that govern the rewetting phenomenon. Thus, a simulation code was developed for the Rewetting Test Facility (ITR), which represents a PWR core cooling channels. The thermohydraulic code was named Rewet. The results obtained with Rewet code were compared with the experimental results of the ITR and with the results of the Hydroflut code, the program used until then. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the rewetting front for two typical tests using the two-code calculation and experimental results. In all cases, there was a better adjustment of the Rewet results in relation to those of the Hydroflut. The result simulated by the Rewet code for the rewetting time also came closer to the experimental results more than that obtained with the Hydroflut code.

Thermal-hydraulic code for rewetting analysis in a PWR experimental loop

Publication date:

July 2018

Optimal neutron population growth in accelerated Monte Carlo criticality calculations

Publication date:July 2018

**Source:**Annals of Nuclear Energy, Volume 117 Author(s): Ignas Mickus, Jan Dufek We present a source convergence acceleration method for Monte Carlo criticality calculations. The method gradually increases the neutron population size over the successive inactive as well as active criticality cycles. This helps to iterate the fission source faster at the beginning of the simulation where the source may contain large errors coming from the initial cycle; and, as the neutron population size grows over the cycles, the bias in the source gets reduced. Unlike previously suggested acceleration methods that aim at optimisation of the neutron population size, the new method does not have any significant computing overhead, and moreover it can be easily implemented into existing Monte Carlo criticality codes. The effectiveness of the method is demonstrated on a number of PWR full-core criticality calculations using a modified SERPENT 2 code.

Optimal neutron population growth in accelerated Monte Carlo criticality calculations

Publication date:

PIV study of velocity distribution and turbulence statistics in a rod bundle

Publication date:

view: 123

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