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Ultrasound in Medicine & Biology
April 2018
Numerical investigation of natural convection inside the containment with recovering passive containment cooling system using GASFLOW-MPI
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Yabing Li, Han Zhang, Jianjun Xiao, J.R. Travis, Thomas Jordan The condensation on containment shell drives the nearby gas downwards creating a natural convection in containments which plays a key role for long-term passive containment cooling. In late phase of accident where hydrogen and steam accumulate in containment, recovery of containment cooling can cause the loss of steam-inert status for the containment atmosphere leading to hydrogen risk in containment. This process is analyzed in this paper focusing on the natural convection driven by wall condensation, and its influence on hydrogen distribution. Firstly, both convective heat transfer model and condensation model are validated with two separate effect experiments. The analogy argument among monument, heat and mass transfer is adopted in GASFLOW-MPI to analysis monument, mass and energy transfer between structure surface and fluid. The simulation result shows good agreement with experiment data. Then a simplified containment model including two steam generator compartments and pressurizer compartment is built and analyzed with GASFLOW-MPI with a postulated accident condition. To avoid long time calculation, the initial condition is calculated with a methodology that is designed to estimate containment status during severe accident provided by EPRI. Two cases are simulated, one without steam injection, where the natural convection drives only by condensation. Other one considers the decay heat that is postulated as a constant steam injection to simulate the natural circulation in containments. Result shows that, during the containment cooling, a transient stratification will occur, leading to high concentration of steam in the dome while low concentration at bottom. This is because the condensed gas is driven downwards near containment shell, pushing steam-rich gas at bottom upwards. The stratification of steam results in a reverse stratification of hydrogen, with high concentration at bottom while low concentration in the dome. Combustibility cloud shows that there is still a stratification of combustibility in containment, though the hydrogen distribution is quite uniform at the end of computation. Therefore, the hydrogen risk should be concerned when implementing containment cooling, especially the local hydrogen concentrate happening at the bottom of containment.
April 2018
Effect of surface modification of silica nanoparticles by silane coupling agent on decontamination foam stability
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Jong Suk Sonn, Ju Yeon Lee, Seon Hui Jo, In-Ho Yoon, Chong-Hun Jung, Jong Choo Lim The effect of surface modification of silica nanoparticles by Dimethyldichlorosilane (DMDCS) on decontamination foam stability was investigated by the measurement of decaying foam volume with time using a Foamscan. The hydrophobicity of silica nanoparticles modified by DMDCS was characterized by active ratio via a floating test and contact angle analysis. Contact angle measurement has shown that silica nanoparticles surface become more hydrophobic as DMDCS concentration increases. Foam stability test in unmodified silica particles-surfactant mixtures revealed that silica nanoparticles-surfactant stabilized foams are much more stable than surfactant-stabilized foams and certain level of surfactant concentration is required for the synergy between silica nanoparticle and surfactant. In foam stability test with modified silica particle-surfactant mixtures, it was found that silica nanoparticles with the proper level of hydrophobicity shows the best performance in foam stability and this result was supported by optical and fluorescence microscope images.
April 2018
On a various soft computing algorithms for reconstruction of the neutron noise source in the nuclear reactor cores
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Seyed Abolfazl Hosseini, Iman Esmaili Paeen Afrakoti This paper presents a comparative study of various soft computing algorithms for reconstruction of neutron noise sources in the nuclear reactor cores. To this end, the computational code for reconstruction of neutron noise source is developed based on the Adaptive Neuro-Fuzzy Inference System (ANFIS), Decision Tree (DT), Radial Basis Function (RBF) and Support Vector Machine (SVM) algorithms. Neutron noise source reconstruction process using the developed computational code consists of three stages of training, testing and validation. The information of neutron noise sources and induced neutron noise distributions are used as output and input data of training stage, respectively. As input data, both the real and imaginary parts of numerical value of the neutron noise in the detector are used. In the present study, the neutron noise source of absorber of variable strength type is only considered. The neutron noise distributions in the detectors due to 2000 randomly generated neutron noise sources are calculated using the developed computational code based on Galerkin Finite Element Method (GFEM). As output data, the strength, frequency of occurrence and location (X and Y coordinates) of the considered neutron noise sources are used. The VVER-1000 reactor core is considered as the benchmarking problem for validation of performed simulation using developed computational code. All specifications of neutron noise source including strength, frequency and location of the neutron noise source are reconstructed with high accuracy. Finally, a sensitivity analysis of results to the number of active detectors in the reactor core is performed. A comparative study of the performance of different developed algorithms represents Decision Tree as the most appropriate one for reconstruction of the neutron noise source in the nuclear reactor cores.
April 2018
Numerical analysis of the 2D C5G7 MOX benchmark using PL equations and a nodal collocation method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): M.T. Capilla, C.F. Talavera, D. Ginestar, G. Verd
April 2018
VHTR core analysis with McCARD and DeCART codes for high temperature engineering test reactor benchmark
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Jinsu Park, Tae Young Han, Deokjung Lee, Hyun Chul Lee This paper presents a Very High Temperature Gas-Cooled Reactor (VHTR) core analysis of the McCARD and DeCART-2D with High Temperature Engineering Test Reactor (HTTR) benchmark problem. The numerical results of the HTTR benchmark problem are demonstrated, the capability for VHTR core analysis of McCARD is validated by using the experiment data, and that of DeCART-2D is verified by comparing the numerical results with those of McCARD. The detail HTTR benchmark problem and specification of core geometry and material are also described. Also, a method of approximation for the complex geometry of the VHTR core is described for modelling in computational code. In terms of the multiplication factor and temperature coefficient, the numerical results of McCARD are easily compared with the HTTR experiment data. Because the DeCART-2D can solve 2-dimensional (2D) geometry, the simulation of 2D HTTR fuel rod, block, and core problem is performed with DeCART-2D and McCARD. The numerical results between them are around 200–400
April 2018
The influence of the air fraction in steam on the growth of the columnar oxide and the adjacent
April 2018
A two-dimensional experimental investigation on the sloshing behavior in a water pool
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Songbai Cheng, Shuo Li, Kejia Li, Nan Zhang, Ting Zhang Studies on the pool sloshing behavior are important for the improved evaluation of energetic potential of a large whole-core-scale molten fuel pool that might be formed during a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). Motivated to understand the characteristics of this behavior, in this study a series of simulated experiments was conducted by injecting nitrogen gas into a Two-Dimensional (2D) rectangular water pool through a nozzle positioned at the center of pool bottom. To achieve a comprehensive understanding, experimental parameters, including nitrogen gas pressure (
April 2018
A non-equilibrium multiphase model for pressure escalation in fuel coolant interaction during explosion
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Mingjun Zhong, Yuan Zhou, Tao Huang, Yanhua Yang A non-equilibrium model is used to simulate the multiphase interactions of molten fuel and coolant at explosion stage. In the current model, fluids are liquid coolant, vapor, molten drop and molten fragment. According to the non-equilibrium concept, an energy balance is built at vapor film surrounding the melt. Heat transferred from the melt to the vapor film interface directly participates to evaporate the liquid coolant, without heating the bulk liquid. Mass transfer between molten drop and molten fragment are dominated by thermal and hydrodynamic fragmentation mechanisms. The conservation equations are solved by the modified MCBA-SIMPLE method, with a semi-implicit iteration. Shock tube tests in two-phase medium and two simulations of the KROTOS experiments are used to validate the model. Simulations of KROTOS42 are performed for sensitivity study on fragmentation model, melt diameter, void fraction and melt temperature. Results reveal that the non-equilibrium multiphase model can give reasonable predictions on pressure generation and escalation in fuel coolant interaction. Fragmentation model analysis reveals that thermal fragmentation at trigger stage has unobvious effect on hydrodynamic fragmentation and hydrodynamic fragmentation is the major factor on pressure escalation. With the escalation of pressure, hydrodynamic fragmentation based on stripping mechanism results in a weaker pressurization against Rayleigh–Taylor mechanism. It is suggested that Rayleigh Taylor instability will be more preponderant at higher relative velocity in vapor explosion. Besides, the sensitivity study indicates that the steam generation and quenching process during the premixing have significant effect on explosion. Further, the model is preliminarily applied to a reactor simulation and integrity of the cavity wall is evaluated.
April 2018
Reactor power monitoring using Cherenkov radiation transmitted through a small-bore metallic tube
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): G. Bentoumi, B. Benson, P.K. Chan, M. Gaudet, G. Li, L. Li, P. Samuleev, B. Sur The use of a small-bore highly reflective metallic tube has been demonstrated at a pool-type nuclear reactor for transmitting Cherenkov radiation emitted from a quartz cylinder placed near the reactor core. The study revealed the promising prospect of using a metallic tube to remotely make localized measurements in order to independently monitor reactor power using Cherenkov light generated in the water pool or inside an attached radiator. This study confirmed qualitatively and quantitatively that commercially available stainless steel tubes with highly reflective inner surfaces can adequately transport Cherenkov light, produced in the proximity of a SLOWPOKE-2 reactor core operated at 20
April 2018
Neutronic/thermal-hydraulic coupling analysis of natural circulation IPWR under ocean conditions
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Genglei Xia, Bing Wang, Xue Du, Chenyang Wang In this paper, a thermal-hydraulic analysis code under non-inertial system was developed based on a RELAP5/MOD3 code by establishing dynamic simulation models of typical ocean conditions. The simulation results under periodic force field were in good agreement with the reference data, proving the accuracy of the program. In order to analyze the reactor core flow distribution and power distribution under different ocean conditions, the thermal-hydraulic code was coupled with a two-group three-dimensional neutron kinetics code. Based on the modified RELAP5 code, the effects of heeling, heaving and rolling conditions on neutronic/thermal-hydraulic coupling characteristics of natural circulation integrated pressurized water reactor (IPWR) were studied. The results indicate that the uneven distribution of coolant flow increases with the increasing inclination angle, but the reactor power distribution is subject to little changes. This trend leads to an uneven distribution of coolant temperature at the core outlet. The flow or power fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. Furthermore, the impact of the heaving amplitude is more significant than the heaving period, whether on the flow or power.
April 2018
Comparative neutronic study of homogeneous and heterogeneous thorium fuel based core design in a lead-cooled fast reactor
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Luis-Carlos Ju
April 2018
Effect and treatment of angular dependency of multi-group total cross section and anisotropic scattering in fine-mesh transport calculation
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Fan Xia, Tiejun Zu, Hongchun Wu The whole core heterogeneous high-fidelity transport calculation is widely researched recently, where the spatial meshes reach sub-pin level to provide accurate results. In this context, two issues, namely angular dependency of total cross section and anisotropic scattering, which greatly affect the precision of multi-group transport calculation are assessed in this paper. Two pin cell problems respectively fueled with UO2 and MOX are calculated. The numerical results show that neglecting angular dependency of total cross section leads to over-estimation of flux in resonance groups in fuel pellet and under-estimation of eigenvalue. Using angular-flux-weighted total cross sections can effectively reduce the error of flux in resonance groups. For UO2 fuel pin cell, angular dependency of total cross section is more important than anisotropic scattering. While for MOX fuel pin cell, anisotropic scattering is more important. And then SPH method and a newly developed method called auxiliary term method are introduced and tested on UO2 and MOX pin cell and 3
April 2018
Numerical investigation on the bubble separation in a gas-liquid separator applied in TMSR
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Junlian Yin, Yalan Qian, Tingting Zhang, Dezhong Wang The fission gas removal system plays a critical role in the development of the Thorium Molten Salt reactors. An axial gas-liquid separator is adopted in the gas removal system. To predict the bubble trajectories in swirling flow is essential for designing such gas-liquid separator, since the separation efficiency is closely related to the bubble trajectory. In this paper, we proposed a numerical method to predict the bubble motion. This method is a modified Lagrangian approach in that the velocity of the continuous phase is obtained by approximating the velocity profiles from CFD. Combing the known velocity distribution with explicit mathematical expression and the force model for a single bubble, a mathematical model to calculate the bubble motion is well posed. Calculations with various bubble sizes and Reynolds numbers were carried out. By comparing the simulation results with the experimental data, we concluded that the numerical results agree well with the experimental data. The maximum error of the separation length is less than 10%, which is accurate enough for the determination of the dimension of the separator.
April 2018
Research on the induced radioactivity of HTR-PM’s reactor pressure vessel: A comparative study between FLUKA, KORIGEN and QAD-CGA
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Wenqian Li, Sheng Fang, Hong Li The reactor pressure vessel (RPV), which is made of steel containing impurities, will be irradiated by a strong neutron field during the long-term operation of the reactor. Induced radioactivity due to neutron irradiation, which may affect the health of staffs during the maintenance and decommissioning, was studied in the case of the High Temperature Reactor – Pebble-bed Module (HTR-PM). The activation products and their corresponding specific activities were calculated by KORIGEN, which is based on deterministic method, and by the Monte Carlo code FLUKA. A comparison of these two programs shows good agreement, and the differences of specific activities are within one order of magnitude for 60% of the nuclides. Furthermore, the remanent dose rates at shutdown time and after 30
April 2018
Adjoint neutron flux calculations with Tripoli-4®: Verification and comparison to deterministic codes
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Nicholas Terranova, Guillaume Truchet, Igor Zmijarevic, Andrea Zoia The possibility of computing adjoint-weighted scores by Monte Carlo methods is a subject of active research. In this respect, a major breakthrough has been achieved thanks to the rediscovery of the so-called Iterated Fission Probability (IFP) method, which basically maps the calculation of the adjoint neutron flux into that of the neutron importance function. Based on IFP, we have recently developed the calculation of effective kinetics parameters and sensitivity coefficients to integral reactor responses in the Monte Carlo production code Tripoli-4®. In view of the next release of the code, we have added a new routine allowing for the calculation of the adjoint angular flux (and more generally adjoint-weighted sources) in eigenvalue problems, which can be useful for code-code comparisons with respect to deterministic solvers. In this work we analyse the behaviour of the adjoint angular flux as a function of space, energy and angle for a few benchmark configurations, ranging from mono-kinetic transport in one-dimensional systems to continuous-energy transport in fuel assemblies. The Monte Carlo adjoint flux profiles are contrasted to reference curves, where available, and to simulation results obtained from ERANOS and APOLLO2 deterministic codes.
April 2018
Analysis of nuclear fire safety by dynamic complex algorithm of fuzzy theory and system dynamics
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Tae Ho Woo The cognitive architecture is investigated for the disaster management in the nuclear accident site. The fire-based nuclear facility catastrophe is a very significant considerable factor in the aspect of the nuclear safety. There are simulation results as the graphs by each time step composing of the simulations of 3 factors of Fire Caution, Fire Control, and Personal Factor. Each graph shows the instability of the designed event. There is the simulation result for Instability of Fire Protection System. The highest value is 0.67584 in 56th year in which the system is very unstable in fire protections. In the graph, the later time part of the operations has higher values comparatively. This could be explained by the system aging of the plant. In the system dynamics (SD) method combined with fuzzy set algorithm, the tractable modeling is liable to be adapt to any kinds of situations in the NPPs. Needless to say, the complex algorithm could be usable in the cases of normal as well as emergency circumstances.
April 2018
Experimental study on 1/28 scaled NGNP HTGR reactor building test facility response to depressurization event
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Se Ro Yang, Ethan Kappes, Thien Nguyen, Rodolfo Vaghetto, Yassin Hassan The analysis and understanding of air ingress events are an important aspect of the design of high-temperature gas-cooled reactor (HTGR) accident scenarios. These include depressurized loss of forced cooling (D-LOFC) events that allow for the possibility of air ingress into the reactor pressure vessel as a result of a break in the helium pressure boundary, which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. To characterize air ingress into the vented low-pressure containment of the next generation nuclear plant HTGR during hypothetical moderate-sized D-LOFC break accidents experimentally, a 1/28 scaled simplified reactor building model was established. A non-dimensional similarity approach was employed for scaling of the experimental facility. Three experiments were designed and conducted to study the dynamic response to the accident scenario. The experimental results suggest an increment in the flow area of the check valve between the reactor cavity (CV1) and steam generator cavity (CV3). Furthermore, qualitative analysis was conducted on the experimental data.
April 2018
Acceleration of 3D pin-by-pin calculations based on the heterogeneous variational nodal method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Tengfei Zhang, Hongchun Wu, Liangzhi Cao, Yunzhao Li The pin-by-pin method with homogenization performed at the pin cell level is of considerable interest in thermal reactor physics calculations. However, significant computational costs are required in pin-by-pin calculations due to the fine spatial meshes. To resolve this issue, we present in this paper the three-dimensional (3D) heterogeneous variational nodal method (VNM), along with dedicated acceleration methods. The combined methods can be utilized to perform pin-by-pin diffusion calculations more efficiently. 3D response matrix (RM) equations are formulated, allowing for incorporation of multiple pin cells into one coarse node without altering the original pin cell configuration. Within the nodes, finite elements in the x-y plane and orthogonal polynomials in the axial direction are employed to describe the piecewise constant heterogeneous geometry. On the nodal interfaces, orthogonal polynomials in the x-y interfaces and piecewise constants in the axial interfaces are adopted to approximate neutron current distributions. The resulting RM equations are solved by the standard Red-Black Gauss-Seidel (RBGS) iteration. Matrix reordering (MR) acceleration and parallelization tailored to the RM formation are incorporated. The coarse nodes acceleration (CNA) is investigated by combining homogenized pin cell nodes into larger heterogeneous nodes. A series of meshing schemes are examined with a small modular reactor core problem. Results show that the implementation of MR and parallelization effectively reduces the RM formation time. Besides, with sufficient radial interface expansion order, CNA is able to reproduce the results obtained with fine node calculations. Furthermore, it is demonstrated that judicious choice of coarse nodes substantially accelerate the RBGS iteration. The combined acceleration schemes achieve favorable accuracy-efficiency trade-off.
April 2018
Feasibility study for detection of reactor state changes during severe accidents via external gamma radiation measurements
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): J
April 2018
Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part II: DYN3D calculations
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Reuven Rachamin, S
April 2018
Reactivity determination using the hybrid transport point kinetics and the area method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Paolo Picca, Roberto Furfaro This paper presents an application of the hybrid transport point kinetic (HTPK) technique to the reactivity determination in subcritical reactor configurations. The mathematical model of the HTPK, initially proposed by Picca et al. (2011) to simulate the time-dependent neutron transport, is here extended to incorporate delayed emissions. The classical area method (Sj
April 2018
Thermal-hydraulic analysis of a 7-pin sodium fast reactor fuel bundle with a new pattern of helical wire wrap spacer
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Seong Dae Park, Han Seo, Yeong Shin Jeong, Dong Wook Jerng, In Cheol Bang Thermal-hydraulic analysis was performed with a 7-pin fuel bundle by using CFD. The ordinary wire-wrapped spacers are wound on the fuel pins in the same direction. Counter flow is predicted to occur in all sub-channels due to this pattern. This flow was confirmed in this analysis work. A new type of arrangement for wire-wrapped spacers, called the U-pattern, is presented to provide favorable flow for coolant mixing. In this pattern, 7 pins are designated in a group, the center pin has no wrapping spacer, and the pins surrounding the center pin have alternate winding directions of the spacer. Superior mixing effect and the uniform flow were confirmed in this pattern. This pattern has about 30
April 2018
Validation of STRCS code for calculation of fission-product transport in reactor coolant system during severe accidents
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Mehdi Khaki Arani, Mohammad Hossein Esteki, Navid Ayoobian Transport and deposition of fission products within the primary coolant circuit during severe accident condition have significant effect on the amount of released radionuclides in the containment. In this study, a new code, STRCS, compatible with RELAP5/SCDAP code, has been developed that calculates the fission products’ transport through the piping system. The homogeneous and heterogeneous nucleation, aerosol agglomeration, aerosol and vapor deposition and aerosol resuspension are considered in this code. The developed code was validated through international standard problems STROM SR11 and PHEBUS FPT-1 experiments. The STRCS predictions are in good agreement with the ISPs published results, although the obtained results deviate from the PHEBUS FPT-1 experimental data.
April 2018
Numerical study on temperature fluctuation of upper plenum in FBR with a more realistic model
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Qiong Cao, Shiyao Liu, Daogang Lu, Hongyuan Li, Mu Chang The thermal fatigue caused by the temperature fluctuation is more serious in FBR than that in PWR due to thermal characteristic. Limited by the ability of the computer, the simplified jet models were used to study the temperature fluctuation in published literature, which can not fully reflect the real characteristic of temperature fluctuation of upper plenum in FBR. In present study, the temperature fluctuation of upper plenum with a more realistic model (i.e. three heads of assemblies and the part of central column) in FBR was simulated with LES method. By comparison, the velocity and temperature fields are more complex in a more realistic model than that of the simplified jet model. The characteristic of temperature fluctuation was obtained around the central column. In a more realistic model, the fluid produces intense mixing near interface of the adjacent different temperature regions, which lead to severe temperature fluctuation. The distributions of temperature and temperature fluctuation intensity are asymmetric. The temperature field is divided into three regions with high temperature range
April 2018
Quantification of control rod worth uncertainties propagated from nuclear data via a hybrid high-order perturbation and efficient sampling method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Chen Hao, Fu Li, Wenqi Hu, Yunfei Zhang, Qiang Zhao Based on the high order perturbation theory and few-group neutron diffusion equation, the formula to evaluate control rod worth is derived for various core state, which is equivalent to establish a function between CRW and eigenvalue, few-group homogenized cross sections, flux and adjoint flux. Then, an effective hybrid high order perturbation and efficient sampling (HOPES) method to quantify the uncertainty of control rod worth propagated from uncertainties in input parameters is proposed. To verify the validity of the HOPES method, a three dimensional mini core model with typical AP1000 fuel assemblies in a 3
April 2018
SN transport method for neutronic noise calculation in nuclear reactor systems: Comparative study between transport theory and diffusion theory
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Mona Bahrami, Naser Vosoughi In this paper, the neutron noise based on transport theory and diffusion noise theory using Green’s function technique is calculated. As the neutron noise is used for core diagnostic, surveillance and monitoring, calculation of neutron noise precisely can play an important role in monitoring and safety. We compare the accuracy of Green’s function based on transport and diffusion theory in order to survey the differences between these theories. In this study some deviation between results obtained two theories are observed, and the impact of dimensions, cross sections and frequency on the results investigated.
April 2018
An Adjoint Proper Orthogonal Decomposition method for a neutronics reduced order model
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Stefano Lorenzi This paper deals with the use of Reduced Order Methods for neutronics modelling. This approach is used whether both accuracy and computational efficiency are required. A very popular category of these methods is based on projection approaches which use spatial basis and test functions for the development of the reduced order model. The selection of the spatial basis and test functions used in the projection phase is a crucial issue since it has an impact on the accuracy and the computational cost. In this work, different methods for the creation of the spatial basis and the test functions are analysed. In particular, an Adjoint Proper Orthogonal Decomposition (APOD) method is proposed combining the properties of the Proper Orthogonal Decomposition and the use of the adjoint flux as test function in the neutronics framework. The different methods are applied to create a spatial neutronics model for the ALFRED reactor. The simulation results show that the APOD method gives better results compared to the other methods (Modal Method and standard Proper Orthogonal Decomposition) increasing the accuracy of the reduced order model or minimizing the computational cost.
April 2018
Nuclide depletion capabilities in the Shift Monte Carlo code
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Gregory G. Davidson, Tara M. Pandya, Seth R. Johnson, Thomas M. Evans, Aarno E. Isotalo, Cole A. Gentry, William A. Wieselquist A new depletion capability has been developed in the Exnihilo radiation transport code suite. This capability enables massively parallel domain-decomposed coupling between the Shift continuous-energy Monte Carlo solver and the nuclide depletion solvers in ORIGEN to perform high-performance Monte Carlo depletion calculations. This paper describes this new depletion capability and discusses its various features, including a multi-level parallel decomposition, high-order transport-depletion coupling, and energy-integrated power renormalization. Several test problems are presented to validate the new capability against other Monte Carlo depletion codes, and the parallel performance of the new capability is analyzed.
April 2018
Fully ceramic microencapsulated fuel in prismatic high temperature gas-cooled reactors: Analysis of reactor performance and safety characteristics
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Cihang Lu, Briana D. Hiscox, Kurt A. Terrani, Nicholas R. Brown Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tristructural isotropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar. The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design. The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238U in the fuel. In one case considered, this will increase the absorption of neutrons in 238U, and the resultant impact of 239Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to
April 2018
Development and assessment of a parallel computing implementation of the Coarse Mesh Radiation Transport (COMET) method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Kyle Remley, Farzad Rahnema The reactor physics (neutronics) method of the Coarse Mesh Radiation Transport (COMET) code has been used to solve whole core reactor eigenvalue and power distribution problems. COMET solutions are computed to Monte Carlo accuracy on a single processor with several orders of magnitude faster computational speed. However, to extend the method to include on-the-fly depletion and incident flux response expansion function calculations via Monte Carlo an implementation for a parallel execution of deterministic COMET calculations has been developed. COMET involves inner and outer iterations; inner iterations contain local (i.e., response data) calculations that can be carried out independently, making the algorithm amenable to parallelization. Taking advantage of this fact, a distributed memory algorithm featuring domain decomposition was developed. To allow for efficient parallel implementation of a distributed algorithm, changes to response data access and sweep order are made, along with considerations for communications between processors. These changes make the approach generalizable to many different problem types. A software implementation called COMET-MPI was developed and implemented for several benchmark problems. Analysis of the computational performance of COMET-MPI resulted in an estimated parallel fraction of 0.98 for the code, implying a high level of parallelism. In addition, wall clock times on the order of minutes are achieved when the code is used to solve whole core benchmark problems, showing vastly improved computational efficiency using the distributed memory algorithm.
April 2018
Efficient simultaneous solution of multi-physics multi-scale nonlinear coupled system in HTR reactor based on nonlinear elimination method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Han Zhang, Jiong Guo, Fu Li, Yunlin Xu, T.J. Downar Pebble bed HTR reactor core is a globally coupled system involving neutronics and thermal-hydraulic field. At the same time, the locally coupled relations exist in fuel sphere temperature, local neutron flux and pebble bed temperature. What’s more, several types of fuel spheres with different burnup and fission power must be treated separately. Due to the essential difference between local and global coupling properties, the method, named nonlinear elimination (NlEm), is developed in this paper to solve the multi-physics/multi-scale system of HTR in a tightly coupled, two-level nonlinear form. In the NlEm framework, the local coupled variables and the global coupled variables are treated separately at two different levels to enhance the computational efficiency. For comparison, the Jacobian-free Newton Krylov (JFNK) method and the Picard method are utilized to solve the multi-physics/multi-scale system respectively. The numerical results reveal that the computational efficiency of NlEm is 6.7
April 2018
Numerical simulation of drainage performance in a drain device of steam generator
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Yong Mei, Shengjie Gong, Chi Wang, Hanyang Gu, Bingbin Ying, Yingxi Song As the last stage of a moisture separator, the drain device is an important part of the secondary separator which can affect the separation efficiency. The emphasis of this paper is placed on studying the flow structure and predicting the free liquid surface height in the drain device. Four different turbulence models including k -
April 2018
Magnetic variation and power density of gravity driven liquid metal magnetohydrodynamic generators
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Drew Ryan, Corey Loescher, Ian Hamilton, Robert Bean, Adam Dix Magnetohydrodynamics (MHD) is the study of electrically conducting fluids flowing through applied magnetic fields. MHD can be applied in power generation to produce electricity with no moving mechanical parts. By not using mechanical parts, MHD generators may potentially produce electricity with low capital costs. This form of electrical generation can be paired with advanced nuclear reactors to make the reactors more economically competitive. This paper studies a vertical gravity driven MHD generator system, specifically maximizing the electrical power density output. For an MHD system, there will be a specific magnetic field that, when applied, will produce a maximum power density for the system.
April 2018
CAD modeling study on FLUKA and OpenMC for accelerator driven system simulation
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Jin-Yang Li, Long Gu, Hu-Shan Xu, Nadezda Korepanova, Rui Yu, Yan-Lei Zhu, Chang-Ping Qin Accelerator driven system (ADS) is an advanced nuclear system with high inherent safety feature, which can transmute the nuclear waste including minor actinides (MA) and long-lived fission products (LLFPs) into short-lived fission products. In order to accomplish the neutronic simulation in ADS, a conventional way can be divided into two steps, namely the physical process of proton beam bombarding spallation target (FLUKA) and the neutron transport process in subcritical reactor coupling with a spallation neutron source (OpenMC). However, there are some difficulties in the modeling process for ADS, such as the inconsistent modeling process between FLUKA and OpenMC codes, and the error-prone process of constructing complex geometries. Therefore, it is imperative to build a code system that can convert the CAD modeling files into the input cards of Monte Carlo codes. In this context, a code system named CAD-PSFO (FreeCAD based parsing script for FLUKA and OpenMC) has been developed to solve the modeling conversion problem in ADS, in which basic geometry and boolean logic classes have been established with a mapping relationship to FLUKA and OpenMC codes, and ray-casting technology based iteration method has been proposed to solve the problem of higher-order surfaces in complex geometries. Finally, CAD-PSFO has been verified and validated by comparing with a reference model, two benchmarks, and two simulation tests.
April 2018
Zero-dimensional transient model of large-scale cooling ponds using well-mixed approach
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Ahmed Ramadan, Reaz Hasan, Roger Penlington Nowadays, nuclear power plants around the world produce vast amounts of spent fuel. After discharge, it requires adequate cooling to prevent radioactive materials being released into the environment. One of the systems available to provide such cooling is the spent fuel cooling pond. The recent incident at Fukushima, Japan shows that these cooling ponds are associated with safety concerns and scientific studies are required to analyse their thermal performance. However, the modelling of spent fuel cooling ponds can be very challenging. Due to their large size and the complex phenomena of heat and mass transfer involved in such systems. In the present study, we have developed a zero-dimensional (Z-D) model based on the well-mixed approach for a large-scale cooling pond. This model requires low computational time compared with other methods such as computational fluid dynamics (CFD) but gives reasonable results are key performance data. This Z-D model takes into account the heat transfer processes taking place within the water body and the volume of humid air above its surface as well as the ventilation system. The methodology of the Z-D model was validated against data collected from existing cooling ponds. A number of studies are conducted considering normal operating conditions as well as in a loss of cooling scenario. Moreover, a discussion of the implications of the assumption to neglect heat loss from the water surface in the context of large-scale ponds is also presented. Also, a sensitivity study is performed to examine the effect of weather conditions on pond performance.
April 2018
Modal analysis of the helical tube in a small nuclear reactor’s steam generator using a finite element method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Xin Guo, Jun Cai, Yandong Wang A finite element method was used to simulate the vibration characteristics of helical tubes in a SMART reactor. It was found that vibrational characteristics of larger coil diameter helical tubes are different from smaller coil diameter tube (142
April 2018
Sensitivity and uncertainty analysis for the PWR online power-distribution monitoring with NECP-ONION system
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Zhuo Li, Liangzhi Cao, Hongchun Wu, Wei Shen, Yong Liu, Chenghui Wan In this paper, the sensitivity and uncertainty analysis model has been proposed and implemented in the NECP-ONION system, which is capable of performing the 3D online power-distribution monitoring for PWR by applying the harmonics expansion method. A new method called the analytical method is used to quantify the relative sensitivity coefficients of the monitored power distributions with respect to the detector signals; then the “sandwich” rule is used to quantify the uncertainties of the monitored power distributions with the sensitivity-analysis results based on the uncertainty-propagation method. The BEAVRS benchmark is used to verify the sensitivity and uncertainty analysis capability in NECP-ONION. The results of the direct numerical perturbation method are used as reference to verify the sensitivity coefficients calculated with the analytical method. The results of the statistical sampling method and the analytical method are compared to each other for the uncertainty verification. Then, the sensitivity and uncertainty analyses with various expansion orders of harmonics and core burnup levels are carried out. Numerical results show that, the expansion order is the trade-off of the monitoring accuracy and uncertainty, while the sensitivity and uncertainty analysis is useful for the determination of the expansion order.
April 2018
Fuel cycle analysis of molten salt reactors based on coupled neutronics and thermal-hydraulics calculations
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Shengcheng Zhou, Won Sik Yang, Tongkyu Park, Hongchun Wu A new fuel cycle analysis code named FAMOS was developed for molten salt reactor (MSR) applications by extending the DIF3D code to model the transport of delayed neutron precursors and by coupling the extended DIF3D code with an in-house multi-channel thermal-hydraulics solver and an in-house nuclide depletion solver for the simulation of fuel depletion, reprocessing and feeding. FAMOS can also calculate the effective delayed neutron fraction accounting for the motion of fuel salt, the neutron generation time and the reactivity feedback coefficients. In order to verify the FAMOS code, the molten salt fast reactor (MSFR) proposed in the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System) project was analyzed using FAMOS and the calculation results were compared with the reported results. The numerical results indicate that FAMOS is capable of simulating the fuel cycle process of MSRs with good accuracy. FAMOS was applied to the fuel cycle analysis of the molten salt breeder reactor (MSBR) as well. The effects of the temperature distribution and the drift of delayed neutron precursors on the neutronics characteristics and fuel cycle performances of MSBR were evaluated using three different calculation models. The drift of delayed neutron precursors and the temperature distribution shift the axial power distribution upward and downward, respectively, and the latter effect is stronger than the former one. However, the impacts of the temperature distribution and the drift of delayed neutron precursors on the depletion and buildup behaviors of actinides and fission products in MSRs are negligible. It is also possible to achieve a 233U conversion ratio larger than 1.0 without protactinium removal and thus to operate the MSBR by only feeding 232Th, except for the initial start-up stage.
April 2018
Visual observations of flow patterns in downward air-water two-phase flows in a vertical narrow rectangular channel
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Tae Hyeon Kim, Vikrant Siddharudh Chalgeri, Woosung Yoon, Byong Jo Yun, Ji Hwan Jeong Flow patterns in downward air–water two-phase flows in a narrow rectangular channel were visually observed. A test section was constructed using transparent acrylic plate with a test channel 2.35
April 2018
Continuous-energy perturbation methods in the MORET 5 code
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): A. Jinaphanh, N. Leclaire The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. A perturbation procedure using first order Taylor expansion series in the continuous-energy calculation for isotopic concentrations and density perturbations have been recently implemented. It relies on the calculation of an adjoint weighted sensitivity coefficient with respect to isotopic concentration. Thus, the MORET code now offers the users the possibility to use either the correlated sampling or taylor series as the perturbation method. The present paper provides a description of continuous-energy perturbation methods available in the MORET code focusing on the Taylor expansion series, which have been implemented in the last release of the code. Then, some considerations about the algorithm and implementation are presented. Finally the verification is presented using several ICSBEP and OECD/NEA benchmarks and some configurations of the French proprietary experimental program MIRTE.
April 2018
The effect of eccentric loading in spent fuel pool criticality safety analyses
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Walid A. Metwally, Abdulrahman S. Al Awad Eccentric fuel loading is one of the studies performed in a spent nuclear fuel criticality safety analysis. The study normally considers fuel assemblies placed at one side or corner in the rack cell. The purpose of this work is to study the effect of random placement of fuel assemblies on reactivity. MCNP was used to model the high density spent fuel pool and the depleted PLUS7 fuel. A total of 1400 cases with randomly displaced fuel assemblies in the rack cells, with and without neutron absorbers, were prepared. The results showed that none of the k-effective values from the random displacement cases exceed that of the base case where the fuel assemblies were centered in the rack cell. This suggests that eccentric fuel loading studies may not be needed in all criticality safety analyses with randomly displaced fuel assemblies.
April 2018
Studies on the subcooled boiling in a fuel assembly with 5 by 5 rods using an improved wall boiling model
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Tenglong Cong, Rui Zhang, Lijuan Chen, Xiang Zhang, Ting Yu An improved wall boiling model with consideration of thin film heat transfer was employed with the Eulerian two-fluid model to investigate the three-dimensional flow boiling characteristics in a fuel assembly with 5 by 5 rod bundle and a vaned grid. Models were validated by using the experimental data for subcooled boiling, post-dryout heat transfer and critical heat flux. The calculated results agreed well with experimental data. Thermohydraulics in the rod bundle were obtained, including temperature, velocity, phasic volume fraction and pressure, based on which, the effects of mixing vane on the localized thermohydraulics were studied. Mixing vane can significantly reduce the cross section averaged vapor fraction and increase the heat transfer capacity due to the swirl effects on two-phase flow; however, it will increase the localized vapor fraction on the heated surface, which may result in the anticipation of boiling crisis, i.e., reach the critical heat flux. Besides, influences of heat flux profile on the flow and heat transfer characteristics were also obtained by comparing two cases with uniform and cosine heat flux distribution along the axial direction. Finally, the impacts of thin film heat transfer on the wall heat partition were investigated, proving that thin film heat transfer played an essential role in the modeling of wall boiling when the vapor fraction at heated surface exceeding 0.25.
April 2018
Detailed neutronic analysis of a MOX-fueled metal-cooled reactor
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Zeeshan Jamil, Bin Wu, Shengpeng Yu, Qi Yang, Muhammad Salman Khan, Muhammad Younas Ali, Liqin Hu With the prominent features of different U-Pu fuel compositions, the MOX-fueled fast spectrum metal-cooled reactors have been of increasing interest for recent years worldwide. Present study invokes a detailed investigation on the calculations for neutronic analysis and it conducts a comparative study for neutron physics parameters of a metal-cooled fast spectrum reactor – the BFS-62-3A. This work is an amalgamation of four tasks. In the first part, a detailed comparative analysis was performed to perform a code-to-code verification using the available results of three Monte Carlo codes including SuperMC, MCNP, and Serpent, and a deterministic one, the DYN3D-MG with the employment of continuous neutron energy cross-sections. The experimental results of BFS-62-3A benchmark were used to assess the potentiality of the aforementioned reactor physics codes. For most of the integral parameters, the SuperMC was found to be on the leading edge. In the second part, the effects of data libraries including ENDF/B-7.1, ENDF/B-7.0, ENDF/B-6.6, JEFF3.2, and HENDL3.0 on the simulations performed using SuperMC code for evaluating k-eff and radial fission rates were all assessed. The third part incorporates the investigation of the fission rates’ deviations in stainless steel radial reflector by changing the density of the reflector. The decrease of density by 5% was found to be in good agreement with the benchmark. In the last part, that has a special importance to the concept pertaining to safety-enhanced Sodium-cooled Fast Reactor (SFR) core, the reactivity of the critical assembly was studied by calculating the sodium void reactivity effect. The simulation results of SuperMC code agreed well with the available experimental and simulation results. The present study has enabled SuperMC code to pass another big milestone on dealing with complex and advanced nuclear systems.
April 2018
An angular biasing method using arbitrary convex polyhedra for Monte Carlo radiation transport calculations
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Joel A. Kulesza, Clell J. Solomon, Brian C. Kiedrowski This paper presents a new method for performing angular biasing in Monte Carlo radiation transport codes using arbitrary convex polyhedra to define regions of interest toward which to project particles (DXTRAN regions). The method is derived and is implemented using axis-aligned right parallelepipeds (AARPPs) and arbitrary convex polyhedra. Attention is paid to possible numerical complications and areas for future refinement. A series of test problems are executed with void, purely absorbing, purely scattering, and 50% absorbing/50% scattering materials. For all test problems tally results using AARPP and polyhedral DXTRAN regions agree with analog and/or spherical DXTRAN results within statistical uncertainties. In cases with significant scattering the figure of merit (FOM) using AARPP or polyhedral DXTRAN regions is lower than with spherical regions despite the ability to closely fit the tally region. This is because spherical DXTRAN processing is computationally less expensive than AARPP or polyhedral DXTRAN processing. Thus, it is recommended that the speed of spherical regions be considered versus the ability to closely fit the tally region with an AARPP or arbitrary polyhedral region. It is also recommended that short calculations be made prior to final calculations to compare the FOM for the various DXTRAN geometries because of the influence of the scattering behavior.
April 2018
Incorporation TACOM and SPAR-H into the operating procedure of nuclear power plants
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Xiaosong Zhao, Rui Zhang, Shumeng Zhao, You Wu, Zhen He The digital control system has improved the equipment reliability and the productivity of nuclear power plants (NPPs) greatly. However, the human error in main control room of digital control systems has become a major influence for the safety and economic efficiency. Based on the task complexity and human reliable analysis, this study analyzes the procedural tasks in the main control room. Standardized Plant Analysis Risk Human Reliability Analysis (SPAR-H) method lacks cognitive activity analysis and the results are not accurate due to the subjective assessment of performance shaping factors (PSFs). The TACOM (Task Complexity) quantification method is then introduced to SPAR-H method to gain precise results. Finally, the steam generator tube ruptures is analyzed using the improved method and comparison is made between the result and empirical data. The result shows that the improved method can effectively identify task complexity factor to achieve a more accurate prediction of the human error probability, and the improvement of the prediction is significant to the safe operation of nuclear power plant. Moreover, the improved method can distinguish the tasks that with similar difficulties which is very important for the task selection in time-urgent situation.
April 2018
A critical heat flux model for saturated flow boiling on the downward curved heated surface
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Qiang Hu, Xiao Yan, Shanfang Huang, Junchong Yu The critical heat flux (CHF) distribution on the outer surface of the lower head is a crucial parameter to assess the coolability limits of the in-vessel retention strategy through external reactor vessel cooling. Several CHF correlations concerning the orientation angle of heated wall have been proposed while the theoretical analysis is relatively insufficient. Based on the liquid microlayer dryout mechanism, a theoretical CHF model for the saturated flow boiling on the downward curved heated surface has been proposed in this work. With a thorough analysis of the vapor blanket behavior in the near-wall region, the effects of the mass velocity and orientation angle of the heated wall on the CHF are considered in present model. The well agreement between the predicted CHF and the experimental data suggests the validity and applicability of present CHF model to assess the CHF limits on the outer surface of the lower head under in-vessel retention.
April 2018
Validation and evaluation of the ADVANTG hybrid code on the ICSBEP labyrinth benchmark experiment
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Domen Kotnik, Alja
April 2018
Neutron noise simulation in the nuclear reactor core based on the average current nodal expansion method
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): A. Abed, N. Poursalehi In this work, the zeroth order of average current nodal expansion method (ACNEM) is developed for the neutron noise simulation of nuclear reactors core with two dimensional rectangular fuel assemblies. At first, the static calculation is performed for the forward treatment of diffusion equation. Then the forward neutron noise is earned by solving the diffusion equation in the frequency domain using the zeroth order of ACNEM. For the neutron noise calculation in the domain of reactor core, the noise source is considered as an “absorber with variable strength” type i.e. the absorption cross section can be changed in the selected material. In order to evaluate the accuracy of exploited scheme, the neutron noise simulation is performed for two well-known test cases including 2-D LRA BWR and 2-D BIBLIS PWR. For benchmarking purpose, the adjoint noise calculation is done for comparing results with the forward approach using a conventional relation in an elected non-zero frequency. Also the contrast of results is illustrated between the neutron noise in the zero frequency and the corresponding earned static fluxes. Totally, numerical results of problems validate the accuracy of the neutron noise simulation using the proposed method.
April 2018
Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Jaerim Jang, Wonkyeong Kim, Sanggeol Jeong, Eun Jeong, Jinsu Park, Matthieu Lemaire, Hyunsuk Lee, Yongmin Jo, Peng Zhang, Deokjung Lee This paper presents the validation of the continuous-energy Monte Carlo neutron-transport code MCS with the ENDF/B-VII.0 neutron cross-section library for the criticality safety analysis of PWR spent fuel pools and storage casks. The MCS code is developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the analysis of Pressurized Water Reactors (PWRs) with high fidelity and high performance. The validation is conducted with 279 selected critical benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The 279 validation cases are representative of PWR spent fuel pools and storage casks with 235U enrichment ranging from 2.35

Analysis of wall temperature jump of China Generation IV SFR Steam Generator
Publication date: April 2018
Source:Annals of Nuclear Energy, Volume 114 Author(s): Xianping Zhong, Jiyang Yu, Shengyu Yan, Muhammad Saeed, Yaodong Li This paper presents a thermal-hydraulic model of the sodium-heated once-through steam generator and its application to analyze the wall temperature jump of China Generation IV SFR Steam Generator (CSSG). Firstly, this model is verified with experimental data and another design code, which proves the reliability of this model. Secondly, the heat transfer characteristics of CSSG is obtained by using this model. Thirdly, the location and magnitude of maximum water wall temperature jump are chosen as figures of merit for the sensitivity analysis for CSSG. Based on numerical calculations and conservative consideration, it is concluded that the sodium inlet temperature and the water inlet mass flow rate must be strictly controlled in the operation of CSSG, and that the maximum water wall temperature jump is set to around 30
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