• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1240-1249
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• 2017

An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly $2{\times}$. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly $3-4{\times}$, with a corresponding 15-20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of $2{\times}$. In total, the improvements yield roughly a $7-8{\times}$ speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.27-36
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• 2020

Natural convection and thermal stratification phenomena are found in large water pools that are being used as heat sinks for decay heat removal from the reactor core using passive heat removal systems. In this study, the two-phase (water and air) natural convection and thermal stratification phenomena with conjugate heat transfer in the rectangular enclosure were investigated numerically using ANSYS Fluent 17.2 code. The transient numerical simulations of these phenomena in the full-scale computational domain of the experimental facility were performed. Generation of water vapour bubbles around the heater rod and evaporation phenomena were included in this numerical investigation. The results of numerical simulations are in good agreement with experimental measurements. This shows that the natural convection is formed in region above the heater rod and the water is thermally stratified in the region below the heater rod. The heat from higher region and from the heater rod is transferred to the lower region via conduction. The thermal stratification disappears and the water becomes well mixed, only after the water temperature reaches the saturation temperature and boiling starts. The developed modelling approach and obtained results provide guidelines for numerical investigations of thermal-hydraulic processes in the water pools for passive residual heat removal systems or spent nuclear fuel pools considering the concreate walls of the pool and main room above the pool.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2816-2829
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• 2021

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

• KSII Transactions on Internet and Information Systems (TIIS)
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• v.13 no.4
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• pp.1961-1974
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• 2019

In this letter, a harmonic-mean-based dual-antenna selection scheme at relay node is proposed in two-way relaying networks (TWRNs). With well-designed distributed orthogonal concatenated Alamouti space-time block code (STBC), a dual-antenna selection problem based on the instantaneous achievable sum-rate criterion is formulated. We propose a low-complexity selection algorithm based on the harmonic-mean criterion with linearly complexity $O(N_R)$ rather than the directly exhaustive search with complexity $O(N^2_R)$. From the analysis of network outage performance, we show that the asymptotic diversity gain function of the proposed scheme achieves as $1/{\rho}{^{N_R-1}}$, which demonstrates one degree loss of diversity order compared with the full diversity. This slight performance gap is mainly caused by sacrificing some dual-antenna selection freedom to reduce the algorithm complexity. In addition, our proposed scheme can obtain an extra coding gain because of the combination of the well-designed orthogonal concatenated Alamouti STBC and the corresponding dual-antenna selection algorithm. Compared with the common-used selection algorithms in the state of the art, the proposed scheme can achieve the best performance, which is validated by numerical simulations.

• Structural Engineering and Mechanics
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• v.71 no.6
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• pp.699-712
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• 2019

The reinforced concrete lining of hydraulic pressure tunnels tends to crack under high inner water pressure (IWP), which results in the inner water exosmosis along cracks and involves typical hydro-mechanical interaction. This study aims at the development, validation and application of an indirect-coupled method to simulate the lining cracking process. Based on the concrete damage plasticity (CDP) model, the utility routine GETVRM and the user subroutine USDFLD in the finite element code ABAQUS is employed to calculate and adjust the secondary hydraulic conductivity according to the material damage and the plastic volume strain. The friction-contact method (FCM) is introduced to track the lining-rock interface behavior. Compared with the traditional node-shared method (NSM) model, the FCM model is more feasible to simulate the lining cracking process. The number of cracks and the reinforcement stress can be significantly reduced, which matches well with the observed results in engineering practices. Moreover, the damage evolution of reinforced concrete lining can be effectively slowed down. This numerical method provides an insight into the cracking process of reinforced concrete lining in hydraulic pressure tunnels.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1869-1877
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• 2021

This paper presents the neutron dose rate analysis of the new CONSTOR® RBMK-1500/M2 storage cask intended for the spent nuclear fuel storage at Ignalina Nuclear Power Plant in Lithuania. These casks are designed to be stored in a new "closed" type interim storage facility, with the capacity to store up to 202 CONSTOR® RBMK-1500/M2 casks. In 2016 y, the "hot trials" of this new facility were conducted and 10 CONSTOR® RBMK-1500/M2 casks loaded with the spent nuclear fuel were transported to the dedicated storage places in this facility. During "hot trials", the dose rate measurements of the CONSTOR® RBMK-1500/M2 casks were performed as the dose rate is one of the critical parameter to control and it must be below design (and safety) criteria. Therefore, having the actual data of the spent nuclear fuel characteristics, the neutron dose rate modeling of the CONSTOR® RBMK-1500/M2 cask loaded with this particular fuel was also performed. Neutron dose rate modeling was performed using MCNP 5 computer code with very detailed geometrical representation of the cask and the fuel. The obtained modeling results were compared with the measurement results and it was revealed, that modeling results are generally in good agreement with the measurements.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2386-2394
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• 2022

In this paper, smooth and notched bar tensile tests of austenitic stainless steel 304 are performed, covering four different multi-axial stress states and six different strain rate conditions, to investigate the effect of the stress triaxiality and strain rate on fracture strain. Test data show that the measured true fracture strain tends to decrease with increasing stress triaxiality and strain rate. The test data are then quantified using the Johnson-Cook (J-C) fracture strain model incorporating combined effects of the stress triaxiality and strain rate. The determined J-C model can predict true fracture strain overall conservatively with the difference less than 20%. The conservatism in the strain-based acceptance criteria in ASME B&PV Code, Section III, Appendix FF is also discussed.

• Computers and Concrete
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• v.31 no.1
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• pp.71-84
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• 2023

In this paper, the effects of the thickness of cubic samples on the tensile strength of concrete blocks were studied using experimental tests in the laboratory and numerical simulation by the particle flow code in three dimensions (PFC3D). Firstly, the physical concrete blocks with dimensions of 150 mm×190 mm (width×height) were prepared. Then, three specimens for each of seven different samples with various thicknesses were built in the laboratory. Simultaneously with the experimental tests, their numerical simulations were performed with PFC3D models. The widths, heights, and thicknesses of the numerical models were the same as those of the experimental samples. These samples were tested with a new tensile testing apparatus. The loading rate was kept at 1 kg/sec during the testing operation. Based on these analyses, it is concluded that when the thickness was less than 5 cm, the tensile strength decreased by increasing the sample thickness. On the other hand, the tensile strength was nearly constant when the sample thickness was raised to more than 5 cm (which can be regarded as a threshold limit for the specimens' thickness). The numerical outputs were similar to the experimental results, demonstrating the validity of the present analyses.

• Structural Engineering and Mechanics
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• v.86 no.1
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• pp.139-156
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• 2023

The reinforced concrete lining in the hydraulic pressure tunnel tends to crack during the water-filling process. The lining will be detached from the surrounding rock due to the inner water exosmosis along concrete cracks. From the previous research achievements, the cohesive element is widely adopted to simulate the concrete crack but rarely adopted to simulate the lining-rock interface. In this study, the zero-thickness cohesive element with hydro-mechanical coupling property is not only employed to simulate the traditional concrete crack, but also innovatively introduced to simulate the lining-rock interface. Combined with the indirect-coupled method, the hydro-mechanical coupling algorithm of the reinforced concrete lining in hydraulic pressure tunnels is proposed and implemented in the finite element code ABAQUS. The calculated results reveal the cracking mechanism of the reinforced concrete lining, and match well with the observed engineering phenomenon.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2604-2612
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• 2023

Spent nuclear fuel and long-lived radioactive waste must be carefully handled before disposing them off to a geological repository. After the pre-storage period in water pools, spent nuclear fuel is stored in casks, which are widely used for interim storage. Interim storage in casks is very important part in the whole cycle of nuclear energy generation. This paper presents the results of the numerical study that was performed to evaluate the thermal behavior of a metal-concrete CONSTOR RBMK-1500 cask loaded with spent nuclear fuel and placed in an open type interim storage facility which is under fire conditions (steady-state, fire, post-fire). The modelling was performed using the ANSYS Fluent code. Also, a local sensitivity analysis of thermal parameters on temperature variation was performed. The analysis demonstrated that the maximum increase in the fuel load temperatures is about 10 ℃ and 8 ℃ for 30 min 800 ℃ and 60 min 600 ℃ fires respectively. Therefore, during the fire and the post-fire periods, the fuel load temperatures did not exceed the 300 ℃ limiting temperature set for an RBMK SNF cladding for long-term storage. This ensures that fire accident does not cause overheating of fuel rods in a cask.