Reactor Assembly Research Articles

ON-THE-FLY INTERPOLATION OF CONTINUOUS TEMPERATURE-DEPENDENT THERMAL NEUTRON SCATTERING DATA IN RMC CODE

Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future.

Physics-informed reinforcement learning optimization of nuclear assembly design

Optimization of nuclear fuel assemblies if performed effectively, will lead to fuel efficiency improvement, cost reduction, and safety assurance. However, assembly optimization involves solving high-dimensional and computationally expensive combinatorial problems. As such, fuel designers’ expert judgement has commonly prevailed over the use of stochastic optimization (SO) algorithms such as genetic algorithms and simulated annealing. To improve the state-of-art, we explore a class of artificial intelligence (AI) algorithms, namely, reinforcement learning (RL) in this work. We propose a physics-informed AI optimization methodology by establishing a connection through reward shaping between RL and the tactics fuel designers follow in practice by moving fuel rods in the assembly to meet specific constraints and objectives. The methodology utilizes RL algorithms, deep Q learning and proximal policy optimization, and compares their performance to SO algorithms. The methodology is applied on two boiling water reactor assemblies of low-dimensional (∼2×106 combinations) and high-dimensional (∼1031 combinations) natures. The results demonstrate that RL is more effective than SO in solving high dimensional problems, i.e., 10 × 10 assembly, through embedding expert knowledge in form of game rules and effectively exploring the search space. For a given computational resources and timeframe relevant to fuel designers, RL algorithms outperformed SO through finding more feasible patterns, 4–5 times more than SO, and through increasing search speed, as indicated by the RL outstanding computational efficiency. The results of this work clearly demonstrate RL effectiveness as another decision support tool for nuclear fuel assembly optimization.

Depletion capabilities in the OpenMC Monte Carlo particle transport code

A depletion solver has been implemented in OpenMC and is described herein. The depletion solver is implemented in Python and interfaces with OpenMC’s transport solver through a C++ application programming interface, which enables an in-memory transport-depletion coupling. Multiple integration methods for advancing in time have been implemented and exhibit tradeoffs in cost, accuracy, and memory use. For all time integration methods, evaluation of the matrix exponential is performed by using the incomplete partial fraction form of the Chebyshev rational approximation method.Simulations of a pressurized water reactor (PWR) pincell and a sodium-cooled fast reactor (SFR) assembly were carried out with OpenMC and Serpent. For both problems, the use of a high-fidelity depletion chain results in predictions of keff that agree within 20–30 pcm between OpenMC and Serpent. Predicted actinide concentrations were found to agree to a fraction of a percent, and most fission product concentrations were found to agree within 1%. The few cases where larger differences were observed can be attributed either to differences in how the energy dependence of fission product yields is handled or deficiencies in the nuclear data used. OpenMC simulations of the PWR and SFR problems using a simplified 228-nuclide depletion chain demonstrate that it achieves accuracy close to that of the full, high-fidelity depletion chain with respect to the studied responses.

Study on algorithms for solving depletion sparse matrix based on DESCAR module

Depletion solver with Chebyshev rational approximation method (DESCAR) was developed by Nuclear Power Institute of China (NPIC) to solve depletion equation. In this paper, the algorithms for solving depletion sparse matrix are studied, including symbolic Lower-Triangle and Upper-Triangle (LU) decomposition using elimination directed acyclic graphs (EDAGS), symbolic LU decomposition using symmetric reduction (FSNZ), symbolic LU decomposition using partial path-symmetric reduction (FPNZ), symbolic decomposition using elimination tree (ETREE), the direct sparse solver using pardiso routine in Intel math kernel library (PDMKL). An optimized direct factorization algorithm (ODFZ) is proposed to improve the speed of solving sparse matrix. The above methods are developed and tested in DESCAR module using a 17 × 17 pressurized water reactor (PWR) assembly containing Gd2O3-UO2 fuel rods. Results show that ODFZ method has highest calculation efficiency, the calculation speed is about 3.5 times that of PDMKL method.

Determination of integral turbulence model parameters as applied to the calculation of flows in fuel assemblies of fast reactors in porous-body approximation

This work aimed to correct the integral turbulence model developed earlier for assemblies of smooth rods. Two variants of the fuel assembly design were considered. In the first variant, the fuel rods were spaced using spacer grids. The presence of a spacer grid does not require a change in the form of the system of equations but leads to a change in the form of the resistance tensor and the generation of turbulence in the spacer grid region. In the second variant, a wire-wrapped fuel bundle was analyzed. The presence of a wire-wrapped fuel bundle requires an additional term in the equation for the conservation of momentum and change in the form of the resistance tensor. The simulations were obtained by CFD code ANSYS CFX and aimed at the determination of parameters involved in an integral model of turbulence being developed for modeling nuclear-reactor cores and heat exchangers in anisotropic porous-body approximation.

Study of neutronic characteristics for the VVER reactor assembly with five various burnable poisons using MCNPX code

The present work investigates the neutronic characteristics of VVER reactor assembly with various burnable poisons. The MCNPX code is used to design VVER-1000 LEU Assembly model. At first gadolinium oxide (Gd2O3) as integral burnable absorber is used, the results are compared with the results of the reference benchmark. Different fuels (uranium oxide, thorium oxide and thorium plutonium oxide) and five various burnable poisons are used in the study. Gadolinium Oxide (Gd2O3) and Erbium Oxide (Er2O3) are used together in the same assembly as IBAs that are mixed homogenously with 30 uranium fuel rods and Dysprosium Titanite (Dy2O3–TiO2) is coated the fuel rod with 0.001 cm thickness. Also, Iridium Oxide (IrO2) used in guide tube as BPR and finally Minor Actinides are used as both IBAs or BPRs. A number of parameters are determined at BOL such as thermal neutron flux, axial power distribution and pin power distribution. Also, four physical parameters are studied, Keff, βeff, MTC and FTC. The burnup calculation is performed for VVER-1200. The comparison of the multiplication factor variation with depletion (for 730 days) illustrated that Iridium Oxide in Th–F assembly has a faster rate of depletion than other Burnable absorbers. Using Iridium Oxide shortens the core life time due to the transformation process of Ir-191 to Ir-192. Moreover, the concentration of actinides for the different assemblies at EOC were discussed. In addition, the effect of using BPs in different fuel assemblies on the concentration of 135Xe and 149Sm production are also performed.

РАСЧЕТНОЕ МОДЕЛИРОВАНИЕ ДОЖИГАНИЯ ТВС ЭНЕРГОБЛОКОВ № 1 И № 2 ЛЕНИНГРАДСКОЙ АЭС ПОСЛЕ ВЫВОДА ИЗ ЭКСПЛУАТАЦИИ

The RBMK-1000 reactor operates in a continuous refueling mode. The reactor core comprises FAs across the burn-up spectrum, from fresh fuel assemblies to the most burnt-up ones. Following the reactor shutdown for decommissioning, the majority of irradiated fuel assemblies (IFA) have the potential for further use (the so-called afterburning) at operating NPP units. Most of the irradiated fuel assemblies have a burn-up fraction, which is far from the specified threshold value. The neutron multiplication factor in the cell exceeds the core average value and, therefore, the IFA loading makes it possible to increase the reactivity margin to the desired value. Reuse of spent fuel assemblies in reactors at other power units offers a number of advantages. In economic terms, afterburning is able to reduce the use of fresh fuel assemblies (FFA). In terms of radioactive waste handling, afterburning is able to reduce the number of irradiated fuel assemblies, which require long-time storage in a spent nuclear fuel repository thus reducing the IFA and radioactive waste disposal loads. JSC NIKIET has proposed the use of irradiated fuel assemblies from units 1 and 2 of Leningrad NPP shut down after decommissioning in reactors at units 3 and 4 of the same NPP.

Analysis of the drop dynamics of control rod assembly in a LBE-cooled research reactor

The lead-based fast reactor is very attractive as one of the Generation-IV nuclear energy systems due to high-power density, heavy liquid metal coolants with a high boiling point as well as harmless interaction with air or water. Nuclear heat is generated from very high neutron flux that is regulated and controlled by the drop of the Control Rods (CRs) for safe operation. The resisting effect of the Lead Bismuth Eutectic (LBE) due to high density on the driving force of the CR Assembly (CRA) during the fall is crucial, which some delays may lead to unavoidable accidents. The mathematical approach has been applied to investigate the CRA dynamics of lead-based research reactor. The non-linear differential equations governing the motion of the CRA as well as the LBE flow have been ascertained to capture the pertinent parameters. Using the Fourth-Order Runge-Kutta iteration method, a computer program was developed with MATLAB and it can be used to calculate the drop time of the CRA and other important parameters described and presented in the study. Compared with the results of the experiment in the available literature, the program developed was proved to be applicable and reliable. The results showed that among the LBE flow characteristics, the density is the most influential parameter on the drop time and tends to have a positive relationship with the drop time. The total CRA mass contributes to the driving force of the CRA to reduce drop time when increased. Finally, the result of the theoretical model will serve as a basis to analyse the auxiliary mechanism and thus, help make informed choice on LBE flow parameters during modification to improve safety.

A multiscale heat transfer model for nuclear reactor assemblies

We develop a new multiscale approach to investigate the heat transfer in a nuclear assembly. Fuel rods are represented as mono-dimensional (1D) inclusions into a three-dimensional (3D) bulk domain representing the coolant. More precisely, the interaction between the fuel rod grid and the coolant is modeled by means of projection operators that couple the 1D and 3D problem taking into account the actual dimension of the rods together with the cladding layers. Due to the dimensional reduction, the inclusions do not need to conform with the bulk computational grid. As a result, the computational challenges of the numerical discretization, due to the high geometrical complexity of the problem, are significantly reduced, as confirmed by the numerical simulations presented in this work.

Comparative evaluation of the biochemical methane potential of waste activated sludge acetic acid and cellulose substrates under mesophilic and thermophilic anaerobic digestion

Knowing the biochemical methane potential of a substrate to biogas production is a preliminary to possible selection of the pre-anaerobic digestion technology and identifying the optimum reaction time. Following standard method to reactor assembly, feed preparation and monitoring of the specific methane yield, the two temperature systems are compared among substrates. As a result, the rate of biochemical methane yield from the waste activated sludge anaerobic digestion in the thermophilic case is much faster than the mesophilic. The specific methane yield and volatile solid removal of the waste activated sludge batch anaerobic digestion is 74% of the theoretical and 81% for the thermophilic while it is 57% theoretical and 76% for the mesophilic, respectively. However, the best methane yield is recorded for the acetic acid followed by the cellulose and sludge substrates, signifying the importance of sludge pretreatment to enhance degradability in anaerobic digestion. Also, the optimum reaction time for thermophilic and mesophilic systems is different.

A catalytic composite membrane reactor system for hydrogen production from ammonia using steam as a sweep gas

For catalytic reactions involving H2 extraction, the membrane reactor is an attractive option for enhancing the equilibrium and kinetics while eliminating excessive purification steps. In this study, a steam carrier adopted composite membrane reactor system is developed to produce pure H2 (>99.99%) from ammonia with high H2 productivity (>0.35 mol-H2 gcat−1 h−1) and ammonia conversion (>99%) at a significantly reduced operating temperature (<723 K). Coupling of a custom developed palladium/tantalum composite metallic membrane and ruthenium on lanthanum-doped alumina catalysts allowed stable operation of the membrane system with significant mass transfer enhancement. Various reactor assemblies involving as-fabricated membranes and catalysts are experimentally compared to suggest the optimal configuration and operating conditions for future applications. Steam is adopted as a sweep gas, presenting efficient H2 recovery (>91%) while replacing conventionally utilized noble carrier gases that require additional gas separation processes. The steam carrier presents similar membrane reactor performance to that of noble gases, and the water reservoir used for steam generation acts as an ammonia buffer via scrubbing effects. Finally, electricity generation is demonstrated using a commercial fuel cell along with process simulation, substantiating potential of the proposed membrane system in practical applications for H2 production from ammonia and on-site power generation.

МОДЕЛИРОВАНИЕ ПРОЦЕССОВ ТЕПЛО- И МАССООБМЕНА В ТВС БЫСТРЫХ РЕАКТОРОВ В РАМКАХ ПОКАНАЛЬНОГО МЕТОДА РАСЧЕТА. ОБОБЩЕННЫЕ ХАРАКТЕРИСТИКИ ОБМЕНА ДЛЯ ОДНОФАЗНЫХ ПОТОКОВ ЖИДКИХ МЕТАЛЛОВ

The results of experimental and computational-theoretical studies of the transverse interchannel molecular-turbulent and convective exchange of mass, momentum and energy are presented in the framework of the channel-by-channel model of thermohydraulic calculation of fuel assemblies of fuel elements of the fast reactors core with liquid metal coolants for nominal, non-nominal and accident conditions. A complete system of inter-channel exchange coefficients is obtained for closing the system of macro-transport equations in a wide range of parameters, taking into account the fuel assemblies deformation. It was shown that transverse convective exchange in fuel assemblies with spacing by wire wrapped is caused by a deficit of static pressure in the region behind winding at the surface of the rod, and the high intensity of molecular-turbulent exchange in tight bundles is explained by the appearance of secondary flows in cells. The heat exchange between cells due to the thermal conductivity of the fuel rods is determined by the thermal modeling parameter of the fuel rods for the first harmonic. The effect of wire wrapped on the flow can be modeled by periodically changing hydraulic resistance to the transverse coolant flow. The integral model of convective inter-channel exchange in fuel assemblies with distance wire wrap is substantiated. Taking into account the centrifugal effect and calculating the interchannel exchange coefficients taking into account the local distribution of parameters allows us to clarify the thermohydraulic characteristics in the shaped bundles. For the characteristic parameters of fuel assemblies of fast reactors (Pe ≥ 50; S/d ≥ 1,1; h/d ≤ 30; ε1 &lt; 1) the predominant is convective inter-channel exchange due to the spacing of fuel rods by wire wrapped. For highly heat-conducting fuel compositions, a significant contribution to the inter-channel heat exchange can be made by the exchange due to the thermal conductivity of the fuel elements. At a low flow rate (Pe &lt;50), the contribution of molecular-turbulent exchange and heat exchange increases due to the thermal conductivity of the fuel rods.

Investigating the Influence of Heavy Oil Recovery by In Situ Combustion during Air Injection as EOR Technique

Heavy oil is one of the most useful energy resources specially in the times of crises when other resources are not present in profusion. However, Occurrence of heavy oil in unconsolidated sands is one the most challenging factor to recover the heavy oil. Therefore, in this study the main focus is derived towards the extraction of heavy oil with optimistic procedure called air injection. For the research, a reactor assembly was developed for the experimental work on air (21% oxygen) injection into heavy oil (12.59 °API) reservoir. Total 13 kinetics runs were conducted on unconsolidated cores by varying the parameters involved system pressure, flow rate (air flux), oxidation temperature (heat input), and rock formation (sand matrix). It was found that the process is very dependent on operating conditions employed, as oxygen consumption rate was very dependent on air flux. Increase of air flux from 15.19 to 22.78 m3/m2-hr resulted in slightly increasing rates of oxygen consumption over the temperature range under investigation. The temperature difference also shows great effect on the high temperature oxidation. The pressure and porous media also have great impact on the combustion behavior. The influence of individual parameter was obtained from analysis of the inlet oxygen and composition of flue gases from the combustion cell. Indeed, the oxygen conversion was too less to evaluate the kinetic data at temperature less than 250 °C while for oxidation reactions, the oxygen statistics analyzed from temperature above than 350 °C. The experimental results reveal that the average maximum peak temperature was 440 °C, and the oxidation reaction process at high temperature was very effective in terms of produced carbon oxides with an average percentage of 9.5% CO2, 5.5% CO in flue gases. Oil displacement was observed from the analysis of flue gases, consequently; incremental oil recovery was achieved between 56%-80% under high temperature oxidation (HTO) conditions.

Investigating the Influence of Heavy Oil Recovery by In Situ Combustion during Air Injection as EOR Technique

Heavy oil is one of the most useful energy resource special in the times of crises when other resources are not present in profusion. However, Occurrence of heavy oil in unconsolidated sands is one the most challenging factor to recover the heavy oil. Therefore, in this study the main focus is derived towards the extraction of heavy oil with optimistic procedure called air injection. For the research, a reactor assembly was developed for the experimental work on air (21% oxygen) injection into heavy oil (12.59 °API) reservoir. Total 13 kinetics runs were conducted on unconsolidated cores by varying the parameters involved system pressure, flow rate (air flux), oxidation temperature (heat input), and rock formation (sand matrix). It was found that the process is very dependent on operating conditions employed, as oxygen consumption rate was very dependent on air flux. Increase of air flux from 15.19 to 22.78 m 3 /m 2 -hr resulted in slightly increasing rates of oxygen consumption over the temperature range under investigation. The temperature difference also shows great effect on the high temperature oxidation. The pressure and porous media also has great impact on the combustion behavior. The influence of individual parameter was obtained from analysis of the inlet oxygen and composition of flue gases from the combustion cell. Indeed, the oxygen conversion was too less to evaluate the kinetic data at temperature less than 250 °C while for oxidation reactions, the oxygen statistics analyzed from temperature above than 350 °C. The experimental results reveal that the average maximum peak temperature was 440 °C, and the oxidation reaction process at high temperature was very effective in terms of produced carbon oxides with an average percentage of 9.5% CO 2 , 5.5% CO in flue gases. Oil displacement was observed from the analysis of flue gases, consequently; incremental oil recovery was achieved between 56%-80% under high temperature oxidation (HTO) conditions.

Investigating the Influence of Heavy Oil Recovery by In Situ Combustion during Air Injection as EOR Technique

Heavy oil is one of the most useful energy resource special in the times of crises when other resources are not present in profusion. However, Occurrence of heavy oil in unconsolidated sands is one the most challenging factor to recover the heavy oil. Therefore, in this study the main focus is derived towards the extraction of heavy oil with optimistic procedure called air injection. For the research, a reactor assembly was developed for the experimental work on air (21% oxygen) injection into heavy oil (12.59 °API) reservoir. Total 13 kinetics runs were conducted on unconsolidated cores by varying the parameters involved system pressure, flow rate (air flux), oxidation temperature (heat input), and rock formation (sand matrix). It was found that the process is very dependent on operating conditions employed, as oxygen consumption rate was very dependent on air flux. Increase of air flux from 15.19 to 22.78 m 3 /m 2 -hr resulted in slightly increasing rates of oxygen consumption over the temperature range under investigation. The temperature difference also shows great effect on the high temperature oxidation. The pressure and porous media also has great impact on the combustion behavior. The influence of individual parameter was obtained from analysis of the inlet oxygen and composition of flue gases from the combustion cell. Indeed, the oxygen conversion was too less to evaluate the kinetic data at temperature less than 250 °C while for oxidation reactions, the oxygen statistics analyzed from temperature above than 350 °C. The experimental results reveal that the average maximum peak temperature was 440 °C, and the oxidation reaction process at high temperature was very effective in terms of produced carbon oxides with an average percentage of 9.5% CO 2 , 5.5% CO in flue gases. Oil displacement was observed from the analysis of flue gases, consequently; incremental oil recovery was achieved between 56%-80% under high temperature oxidation (HTO) conditions.

Automated Glycan Assembly in a Variable-Bed Flow Reactor Provides Insights into Oligosaccharide-Resin Interactions.

A pressure-based variable-bed flow reactor built for peptide synthesis and capable of real-time monitoring of resin swelling was adapted for automated glycan assembly. In the context of the solid-phase synthesis of several oligosaccharides, the coupling efficiencies, resin growth patterns, and saccharide solvation during the synthesis were determined. The presented work provides the first estimation of on-resin oligosaccharide solvation and an alternative technique to UV–vis monitoring.

Decay heat analysis of system-integrated modular advanced reactor (SMART) using SCALE 6

In this study, the decay energy productions rate of the SMART reactor core and assemblies were analyzed by using SCALE/Triton depletion sequences. The fuel elements and assemblies were modeled based on technical and operational design data. Time and problem dependent cross sections were generated for the reactor and used in the fuel depletion and decay analysis. The heat generation calculations during operation showed an energy production of 21.6 MW at the beginning of the cycle, the contribution of decay heat to the total reactor power (P/Po) was 6.5% in the BOC and drops to 6.3% toward EOC. Post-shutdown calculations of the decay energy produced per assembly resulted in a decreased from 0.356, 0.369, 0.363 MW at shutdown to 0.261, 0.271 and 0.266 kW after five years, for assembly type A, B, and C respectively.Decay heat is generated mainly from short-lived fission products; the fast decay of these radionuclides results in an extreme drop in the decay heat generation hours after shutdown. In the first day after reactor shutdown, the ratio of decay heat to total reactor power (P/Po) drops from 6.3% to only 0.53%. System ability to remove this amount of heat from the reactor core during this period is crucial to the reactor safety. Loss of coolant and failure to remove decay heat is a substantial threat to fuel integrity and may lead to overheating and core meltdown.The reactor core decay heat was calculated for 30 years cooling period; results show that core energy decreases exponentially with time, from 20.8 MW to only 6.8 kW after 30 years. The decay heat as a function of time was calculated and compared to the total reactor power. The energy production of the top 30 heat generating radionuclides and their contribution to reactor decay heat were also estimated.

A New Approach to Generalization of Experimental Data on Heat Transfer to Fluids in Supercritical Region

Abstract The paper contains the results of analysis of heat transfer regimes in the case of forced turbulent flow of water and modeling fluids in channels of various configurations at supercritical pressures. Two new complex criteria were proposed for describing heat transfer in the pseudo-phase transition region. A system of equations suitable for the engineering calculation of heat transfer in fuel assemblies of nuclear supercritical water reactors is presented. A comparison of the calculations of heat transfer coefficient (HTC) with experimental data for the regimes of normal, deteriorated and mixed heat transfer is made.

The effect of soluble boron and gadolinium distribution on neutronic parameters of small modular reactor assembly

Early commercial PWR first cores started up without using burnable absorbers. This results in short first core cycle lengths, positive moderator temperature coefficients and high relative power peaking. Burnable absorbers are introduced to address these first core limitations. In the present work, the first part studies the effect of boron (as burnable poisons) concentration on a number of the small modular reactor assembly neutronic parameters. Among these parameters are the infinite multiplication factor, the depletion of U-235, the production of fissile plutonium isotopes (Pu-239+Pu-241), the total neutron flux, prompt neutron lifetime and effective fraction of delayed neutrons versus burnup. The effect of combination between soluble boron in moderator and gadolinium oxide mixed with uranium oxide is also investigated. For accomplishing this, MCNPX is used for modeling three small modular reactor assemblies. The first one is boron free model of enrichment 5% U-235. The other two assemblies use the same enrichment with soluble boron content 1000 and 2000 ppm respectively. The final conclusions of this part showed that the addition of soluble boron with moderator has a significant effect on the variation of K-inf versus burnup at the beginning and the end of cycle. Furthermore, the addition of soluble boron has an effect on the depletion of U-235. It is found that the weight percent of U-235 at 60 GWd/ton for the assemblies containing soluble boron is higher than that of the boron free assembly. For the mass of produced total fissile plutonium at the end of life, it is found that it increases for the models containing soluble boron. For fresh fuel, the effective fraction of delayed neutrons is directly proportional to the concentration of boron because of the higher content of U-235.The second part aims to study the effect of gadolinium enrichments and distributions on the reactivity of the assembly. Firstly, a comparison is carried out between non poisoned and poisoned assembly. Secondly, reactivity is studied for different models of SMR assemblies containing different numbers of gadolinium pins with (4% Gd2O3). Thirdly, reactivity and radial power distribution are analyzed for five different assemblies containing gadolinium rods of 12% Gd2O3.Using Gd2O3 for controlling excess reactivity and pin power peaking factor of the SMR fuel assembly has been also investigated.It was found that the five arrangements loaded with 88% UO2 and 12% Gd2O3 achieve a flat reactivity and radial power distribution. Therefore, these fashions could be used in the SMR core to guarantee the safe operation of the reactor. The results showed that a satisfactory value of the peaking factor is obtained at the beginning and the end of life for each assembly.

Using covariance weighted euclidean distance to assess the dissimilarity between integral experiments

Integral experiments especially criticality experiments help a lot in designing either new nuclear reactor or criticality assembly. The calculation uncertainty of the integral parameter which is introduced in by the nuclear data uncertainty is larger than the experimental uncertainty for most high-enriched uranium metal experiments, therefore the integral experiment is still very useful. There are lots of integral experiments have been done and documented. It should be considered carefully that which integral experiments should be used in applications. For instance, if the aim of the application is to validate the criticality design of a new reactor, integral experiments which are similar to the new reactor should be used. There are several similarity measures which have been used to assess the similarity between integral experiments, such as E similarity measure, G similarity measure and C similarity measure. But, there is no standard rule to choose which similarity measure should be used to assess the similarity between integral experiments in specific application. Another shortage of these similarity measures is that the thresholds of these similarity measures which should be set to judge whether the integral experiments are similar to each other or not have no clear physical meaning. In this paper, we will analyze the existing similarity measures which have been used to assess the similarity between integral experiments, and test some other similarity or dissimilarity measures which have been used in other research fields. After testing the Tanimato similarity measure and Euclidean distance, we find that the covariance weighted Euclidean distance is well suit to assess the dissimilarity between integral experiments, and the physical meaning of its threshold is clear. We recommend using covariance weighted Euclidean distance to assess the dissimilarity between integral experiments.