Search

The aim of this paper is to assess the reliability and accuracy of the PSI standard method, used in many previous works, for the quantification of ND uncertainties in the SPERT-III RIA transient, by quantifying the discrepancy between the actual inserted reactivity and the original static reactivity worth and their associated uncertainties. The assessment has shown that the inherent S3K neutron source renormalization scheme, introduced before starting the transient, alters the original static reactivity worth of the transient CR and reduces the associated uncertainty due to the ND perturbation. In order to overcome these limitations, two additional methods have been developed based on CR adjustment. The comparative study performed between the three methods has showed clearly the high sensitivity of the obtained results to the selected approach and pointed out the importance of using the right procedure in order to simulate correctly the effect of ND uncertainties on the overall parameters in a RIA transient. This study has proven that the approach that allows matching the original static reactivity worth and starting the transient from criticality is the most reliable method since it conservatively preserves the effect of the ND uncertainties on the inserted reactivity during a RIA transient.

The risk of criticality accidents by fuel debris in the decommissioning of Fukushima Dai-ichi Nuclear Power Station is unlikely to be negligible. Therefore, an estimation of the consequences of suc...

This work reviewed the historical literature associated with the Dragon experiment and Water Boiler reactors operated at Los Alamos during the Manhattan Project. Frisch's invited talk given at the Fast Burst Reactor Conference held the University of New Mexico in Albuquerque, NM in 1969 is quoted. From the literature review, basic models for the Dragon experiment and for a Water Boiler type assembly (aqueous homogeneous reactor) were created that can be used for conducting multi-physics simulations for criticality excursion studies. This methodology utilizes the coupled neutronic-hydrodynamic method to perform a time-dependent dynamic simulation of a criticality excursion. MCNP was utilized to calculate important nuclear kinetic parameters that were incorporated into the models. Simulation results compared reasonably well with historic data.

Modification of the Monte Carlo depletion calculation code OpenMC was performed to enable the depletion calculation of the subcritical neutron multiplying system. With the modified code, it became ...

The work presented in this paper focuses on the first 10 years (2011–2020) of radiation test object (RTO) operations at the National Criticality Experiments Research Center. RTOs are subcritical co...

The work presented in this paper focuses on the first 10 years (2011–2020) of Godiva IV operations at the National Criticality Experiments Research Center (NCERC). Godiva IV is a fast burst critica...

The article presents the results of the calculation analysis of the criticality of the reactor pool compartment for VVER-1000 reactors at different stages of a severe accident. The criticality analysis of fuel-containing systems formed during the development of a severe accident is considered on the example of several discrete configurations that can be described by certain models. In reality, the course of a severe accident will be a continuous process of transition from the initial stage of the accident to its steady state at the end. However, depending on the fuel characteristics and conditions of development of a severe accident, any considered configuration of fuel assemblies and reactor pool materials may become final. The article presents a criticality analysis of the reactor pool compartment of the VVER-1000 reactor unit at the severe accident stages following the deformation of fuel rods and fuel assemblies at the initial stage of the accident due to water boil-off and temperature increase. On the basis of the models developed, the multiplication properties of the reactor pool compartment in the conditions of degradation of hexagonal tubes were investigated. Supercriticality (i.e. multiplication factor was greater than 1) was achieved in all two considered configurations of the water-uranium mixture. That is, under such conditions, the required level of subcriticality can be provided by a boron solution in water. The considered homogeneous single-layer and two-layer model of corium spreading on the support plate is safe in terms of criticality. However, the support plate has holes for the passage of the coolant to the fuel assembly. In the case when the melt or fuel debris enters the space under the support plate through these holes, they can form a uranium-water mixture with the remaining water, which at certain fuel and water ratios can become supercritical. In this case, the multiplication properties of the reactor pool compartment will depend on the ratio of impurities of the reactor pool and fuel assembly structural elements in the material under the support plate and/or the concentration of boric acid in the water. After damage of the reactor pool compartment floor, the melting of concrete begins. The considered configurations and material composition of the mixture of corium, concrete and water allow us to make a conclusion about the safety of this phase of a severe accident in terms of criticality.

The objective of this research is to the use of americium (AmO2) as a burnable absorber effectively instead of conventional gadolinium (Gd2O3) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO2 as a burnable absorber. From these studies, it is found that AmO2 can decrease the excess reactivity like Gd2O3 without changing the criticality life span and enrichment of 235U. A homogeneous mixture of the 0.20% AmO2+ 4.95% enriched UO2 fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO2+UO2 in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

A benchmark to verify the accuracy of neutron transport criticality solvers along the energy dimension was established. For the first time, the analytic solution of the flux amplitude was derived i...

Criticality control of random media such as fuel debris is one of the most important safety issues in postaccident management. A 1/fβ spectrum randomizing model is expected to simulate such random ...

In the early days of nuclear energy in Brazil, a reactor named Argonaut, designed at Argonne National Laboratory, reached criticality at the Institute of Nuclear Engineering (IEN). The presence of ...

The deployment of molten salt reactors requires validation of the computational tools used to support the licensing process. The Molten Salt Reactor Experiment (MSRE), built and operated in the 196...

The change in the temperature of the nuclear reactor components (fuel, moderator, coolant, and structural materials) is considered to be a significant source of reactivity variation. This change must be taken in account during criticality calculations for safety analysis of the reactor. Hence, the exact representation of temperature in the calculations is very important. In this paper, two PWR assemblies are simulated, solid 16 ⨯ 16 and annular 12 ⨯ 12 fuel assemblies. The infinite multiplication factor and its temperature dependent parameters are calculated for both fuel assemblies. Adjusted temperature dependent libraries are created using makxsf code to exactly represent the different temperature values used in the calculations. It is shown that the results obtained using adjusted cross section libraries are more reliable. The two fuel assembly types follow the same behavior despite the differences in their geometrical configuration. The introduction of annular fuel has a very small effect on the investigated neutronic parameters because the moderator to fuel ratio is preserved.

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

To proceed with the decommissioning of the Fukushima Daiichi Nuclear Power Station, analyses of unexpected fuel debris criticality accidents are needed. Supercritical transient analyses have been c...

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (inter-library comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of 16O, 9Be, 235U, 238U, and S(α,β) of 1H in H2O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

A dynamical system under extreme physical disorder has the tendency to evolve toward the equilibrium state characterized by an inverse power law power spectrum. In this paper, a practical, implemen...

In order to support the development and deployment of uranium fuels with enrichment beyond 5%, additional criticality safety methodologies are needed to prevent the possibility of criticality accid...

This work presents an assessment of the applicability of existing benchmark critical experiments to the criticality safety code validation for a large-capacity high-assay low-enriched uranium (HALE...

The evaluation of fissile mass inside radioactive waste drums is essential for radioactive waste management, nuclear safety, and criticality issues. However, passive and active neutron measurements can be strongly impacted by the uncertainty on the neutron source position within the drum and by matrix attenuation effects. Therefore, an imaging panel proposed by Proportional Technologies Inc., composed of seven Boron-coated straw (BCS) detectors has been tested to localize neutron interactions, in view to reduce uncertainties associated with plutonium or uranium position inside radioactive waste drums. A numerical model of the imaging panel has been developed and validated from a comparison with experimental profiles obtained with a 252Cf source. A passive measurement system equipped with 12 such imaging panels has been designed by numerical simulation, in view to provide information on neutron source location in a 118-L radioactive waste drum filled with organic, metallic, or mixed organic–metallic matrices. Additionally, an experimental setup dedicated to active measurements with a D-T neutron generator has been implemented to test the imaging panel. Prompt fission neutron signals have been recorded, which is induced by thermal interrogating neutrons in fissile material samples. This article presents 2-D images indicating the position of fissile materials. Consequently, BCS imaging panels open interesting prospects to reduce the uncertainty associated with plutonium or uranium localization both in passive and active neutron measurements.

New versions of ENDF/B and JEFF data libraries have been released during the past two years with significant updates in the neutron reaction sublibrary and the thermal neutron scattering sublibrary. In order to get a more comprehensive impression of the criticality quality of these two latest neutron data libraries, and to provide reference for the selection of the evaluated nuclear data libraries for the science and engineering applications of the Reactor Monte Carlo code RMC, the criticality benchmarking of the two latest neutron data libraries has been performed. RMC was employed as the computational tools, whose processing capability for the continuous representation ENDF/B-Ⅷ.0 thermal neutron scattering laws was developed. The RMC criticality validation suite consisting of 116 benchmarks was established for the benchmarking work. The latest ACE format data libraries of the neutron reaction and the thermal neutron scattering laws for ENDF/B-Ⅷ.0, ENDF/B-Ⅶ.1, and JEFF-3.3 were downloaded from the corresponding official sites. The ENDF/B-Ⅶ.0 data library was also employed to provide code-to-code validation for RMC. All the calculations for the four different data libraries were performed by using a parallel version of RMC, and all the calculated standard deviations are lower than 30pcm. Comprehensive analyses including the C/E values with uncertainties, the δ k / σ values, and the metrics of χ 2 and | Δ | > , were conducted and presented. The calculated keff eigenvalues based on the four data libraries generally agree well with the benchmark evaluations for most cases. Among the 116 criticality benchmarks, the numbers of the calculated keff eigenvalues which agree with the benchmark evaluations within 3σ interval (with a confidence level of 99.6%) are 107, 109, 112, and 113 for ENDF/B-Ⅶ.0, ENDF/B-Ⅶ.1, ENDF/B-Ⅷ.0 and JEFF-3.3, respectively. The present results indicate that the ENDF/B-Ⅷ.0 neutron data library has a better performance on average.

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.

To improve the efficiency of MC criticality calculation, an acceleration method of fission source convergence which gives an improved initial fission source is proposed. In this method, the MC global homogenization is carried out to obtain the macroscopic cross section of each material mesh, and then the nonlinear iterative solution of the SP3 equations is used to determine the fission source distribution. The calculated fission source is very close to the real fission source, which describes its space and energy distribution. This method is an automatic computation process and is tested by the C5G7 benchmark, the results show that this acceleration method is helpful to reduce the inactive cycles and overall running time.

A series of critical experiments were conducted at the National Criticality Experiments Research Center (NCERC) in Nevada to evaluate the operational performance of a compact reactor that eventuall...

This work presents two different methods for quantifying and propagating the uncertainty associated with fuel composition at end of life for cask criticality calculations. The first approach, the computational approach uses parametric uncertainty including those associated with nuclear data, fuel geometry, material composition, and plant operation to perform forward depletion on Monte-Carlo sampled inputs. These uncertainties are based on experimental and prior experience in criticality safety. The second approach, the data-driven approach relies on using radiochemcial assay data to derive code bias information. The code bias data is used to perturb the isotopic inventory in the data-driven approach. For both approaches, the uncertainty in keff for the cask is propagated by performing forward criticality calculations on sampled inputs using the distributions obtained from each approach. It is found that the data driven approach yielded a higher uncertainty than the computational approach by about 500 pcm. An exploration is also done to see if considering correlation between isotopes at end of life affects keff uncertainty, and the results demonstrate an effect of about 100 pcm.

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores. Keywords: Monte Carlo, Core simulation, Zero power physics test, Validation, MCS

We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO2 and (Th+Pu)O2 mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O2, achieved about 24% higher fuel cycle length than the UO2 case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found. Keywords: Pebble-bed, HTR-10, Burnup, Thorium based fuel

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations. Keywords: MCS/CTF, MCS/TH1D, BEAVRS cycle 2, Quarter-core depletion

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (Keff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%–13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap. Keywords: Research reactors, Neutron flux, OpenMC, Flux trap, HEU, LEU, IAEA

Removal of fuel debris is regarded as one of the most important operations in the decommissioning of the Fukushima Daiichi nuclear power station (1F-NPS) to decrease long-term risk. To begin the op...

In the Monte Carlo method for criticality calculation, the convergence-in-distribution check of the sample mean of tallies can be approached in terms of the influence range of autocorrelation compa...

Neutronic analysis of a conceptual heat pipe-cooled micronuclear reactor with 70 % enriched uranium nitride fuel is carried out by modeling the core and peripheral control drum movement to estimate the power distribution. The core configuration results in non-uniformities and hotspots. For the heat removal, empirical formulae have been used in the case of sodium, lithium, and potassium working fluids. The neutronic simulation was carried out by the OpenMC code. It has been found that the radial flux peaking as high as ~20 % can occur at various stages of the drum movement. The novelty of this research is the investigation of the effect of variable enrichment on the overall system multiplication, which can form the basis for optimal fuel distribution. It has been found that non-uniform fuel distribution can mitigate peaking factors, and thus reduce the hotspots. This analysis is useful for the design optimization of compact micro nuclear reactors for underwater, portable and space propulsion systems.

Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed. Keywords: Research reactor, First criticality, Jordan research and training reactor, JRTR, Inverse multiplication method

This paper aims at extending the work performed by means of the A N Boundary Element – Response Matrix method in a previous paper, devoted to criticality problems, to non-homogeneous problems ( i.e. the problems that involve a fixed external source, as in case of the accelerator driven systems). As in the preceding paper, the interest is given to the 3D, xyz multigroup systems in which linearly anisotropic scattering is allowed. Standard interface conditions have been adopted and no external intervention in order to improve the numerical results, such as some kind of discontinuity factors, has been introduced. The latter choice is motivated by the need of establishing a clean test of the performances of the A N method in itself. Results of 2D and 3D multigroup fixed source problems obtained with the method described in this paper are compared with those obtained by well assessed reference codes such as the MCNP Monte Carlo code and the PARTISN discrete ordinates code. Even in the absence of discontinuity factors and despite of the intrinsically approximate character of the A N method, the accuracy of the numerical results is more than satisfactory. Finally the computational time is in most cases much shorter than that required by the chosen reference code.

Criticality search is one of the important aspects to perform accurate lifecycle analysis of nuclear reactor in Monte Carlo code, in which the geometry criticality search is particularly important for some kinds of reactor systems, such as BWR and so on, to use control rods to adjust the criticality and power in lifecycle. To solve the Monte Carlo geometry criticality search problem correctly and efficiently, the Universe Based Geometry Criticality Search (UBGCS) capability along with the Non-Converged Sub-Iteration (NCSI) acceleration method is newly developed and implemented in reactor physics Monte Carlo code RMC. The UBGCS capability relies on the multi-layer universe geometry description structure in RMC and could perform eigenvalue criticality search not only to single or several surfaces, but also complex cells and universes combination. By using the NCSI method, the RMC’s UBGCS capability could achieve accurate and efficient Monte Carlo criticality search for complex problems. Based on the newly developed UBGCS capability and other advanced techniques in RMC, the lifecycle analysis of a nuclear heating small modular reactor concept NHR200-II is taken out to have a better understanding of the reactor physics characteristics of this reactor. NHR200-II is a new 200 MWt commercial size heating reactor using control rods as power regulating system developed by INET, Tsinghua University according to the new energy and environment requirements in China. In this article, the first detailed Monte Carlo model of NHR200-II has been build including different types of complex fuel assemblies, cruciform control rods, full core structures and so on. The cold zero power and hot zero power state of the NHR200-II reactor are analyzed based on the detailed model. Also using the Monte Carlo reactor physics/thermal hydraulics and UBGCS capabilities in RMC, the full core pin-by-pin burnup calculation of NHR200-II are performed and analyzed.

A global variance reduction (GVR) method based on the SPN method is proposed. First, the global multi-group cross-sections are obtained by Monte Carlo (MC) global homogenization. Then, the SP3 equation is solved to obtain the global flux distribution. Finally, the global weight windows are approximated by the global flux distribution, and the GVR simulation is performed. This GVR method is implemented as an automatic process in the RMC code. The SP3-coupled GVR method was tested on a modified version of C5G7 benchmark with a thickened water shield. The results show that the SP3-coupled GVR method can improve the efficiency of MC criticality calculation.

The high-temperature reactor pebble-bed module (HTR-PM) is a modular high-temperature gas-cooled reactor demonstration power plant. Its first criticality experiment is scheduled for the latter half of 2021. Before performing the first criticality experiment, a prediction calculation was performed using PANGU code. This paper presents the calculation details for predicting the HTR-PM first criticality using PANGU, including the input model and parameters, numerical results, and uncertainty analysis. The accuracy of the PANGU code was demonstrated by comparing it with the high-fidelity Monte Carlo solution, using the same input configurations. It should be noted that keff can be significantly affected by uncertainties in nuclear data and certain input parameters, making the criticality calculation challenge. Finally, the PANGU is used to predict the critical loading height of the HTR-PM first criticality under design conditions, which will be evaluated in the upcoming experiment later this year.

The present paper focuses on the modeling of the CROCUS reactor with the French MORET 5 continuous energy code and compares keff, anti-reactivity effects (variation of keff normalized to keff due to the insertion of an absorber), sensitivity coefficients and kinetics parameters with those provided by other reference codes. The aim is mainly to extend the experimental validation/verification database of the MORET 5 code to use reactor physics applications. A suite of about 1400 benchmarks is already available for criticality but only a few are dedicated to reactor physics applications. The CROCUS reactor configurations therefore constitute an opportunity to test and improve the implementation of kinetics parameters and sensitivity coefficients in the code and make a verification of these physical quantities through the comparison with other codes such as MCNP. Moreover, it offers the opportunity to quantify the effect of nuclear data processing through the use of two different data processing codes.

The duplex fuel pellet includes UO2 components as seed and ThO2 blanket, originally designed to increase the breeding rate of fissile materials in light water breeder reactors. However, the seed portion alone concentrates most of the energy released from the fission reaction, which can make the seed temperature too high. The duplex fuel potential for the high-temperature reactor (HTR) using TRISO fuel has never been researched. MCNPX simulation was conducted to compare the neutronics performance of normal TRISO fuel with that of TRISO duplex fuel. Two types of HTR fuel rod, the cylinder and the annular rod, were simulated and compared to see whether the duplex neutron characteristics would be better in the cylinder or annular fuel compact design. The TRISO duplex rods produced higher burnup than normal rods up to 5% higher for cylinder and 6% for annular rods. The TRISO duplex rod is also capable of maintaining longer criticality, producing smaller kinf changes and reducing the potential for plutonium build-up. The proposed TRISO duplex conceptual fuel design has an interesting neutronic properties, which offers possibilities for future studies.

Nuclear energy has been one of the sustainable energy sources, but after the Fukushima Daiichi nuclear accident, large-scale light water reactors are losing price competitiveness due to the...

In this research, the Coarse Mesh Finite Difference (CMFD) acceleration technique in both adjoint and forward mode is applied to the MOC kernel, to develop a new code called PerMOC. A new approach in the adjoint MOC kernel which is called thermal-up-scattering-like iteration scheme reduces the total number of adjoint iteration sweeps. In additions, more speed-up is achieved using a parallel transport sweeping in MOC kernel. Validation test for PerMOC C-Sharp code is carried out for four distinct criticality benchmarks in which C5G7 is one of those. In the case of C5G7 benchmark, the best-calculated multiplication factor has 51 pcm differences with the reference case. Furthermore, the newly proposed adjoint solver reduces about 30% the number of adjoint transport sweeps in comparison with the normal adjoint solver. The present article describes the PerMOC methodologies and illustrates its ability to model large problems with acceptable accuracy.

With the advancements in technology (both experimental and computational) the determination of the“true” experimental phonon spectrum became more accessible. In this work a methodology for producing thermal scattering libraries from the experimental data (namely the DFT + oClimax method) for lucite (C5O2H8)n is discussed. Double differential scattering cross section (DDSCS) experiments were performed at the Spallation Neutron Source of Oak Ridge National Laboratory (SNS ORNL). New scattering kernel evaluations, based on the phonon spectrum for (C5O2H8)n,were created using oClimax and NJOY2016 codes. In order to compare and asses the performance of the newly created library, the experimental setup was simulated using MCNP6.1. Compared to the current ENDF/B-VIII.0, the resulting RPI (C5O2H8)n library improved the calculation of both double differential scattering and total scattering cross sections. A set of criticality benchmarks containing (C5O2H8)n from HEU-MET-THERM resulted in an overall improved calculation of K eff . The DFT + oClimax method is shown to be the most comprehensive method for analysis of moderator materials, due to the fact that it can be verified against all data measured at VISION, ARCS and SEQUOIA neutron spectrometers at SNS ORNL, and experimental total scattering cross section measurements. This method also provides a new technique for calculating any phonon spectrum-related quantities such as scattering law kernel, specific heat capacity, thermal conductivity, etc. for any solid state material.

An optimization algorithm has been developed for the first time for application to International Criticality Safety Benchmark Evaluation Project (ICSBEP) subcritical neutron multiplication inference benchmark experiments. The optimization algorithm is a genetic algorithm for nuclear data evaluation adjustments, specifically applied to subcritical benchmark measurements. The algorithm has been tested and yields improvement in (C-E)/E values of subcritical benchmark observables of interest. In this work, the genetic algorithm is applied to improvement of fission neutron multiplicity distribution parameters using several subcritical neutron multiplication inference benchmarks; specifically a series of reflected 4.5 kg α -phase spherical plutonium benchmarks. The algorithm results suggest changing the mean ( ν ¯ ) and standard deviation ( σ ) of the number of neutrons emitted by 240 Pu in spontaneous fission from 2.1510 to 2.1460 and from 1.1510 to 1.1395, respectively. In addition, the standard deviation of the number of neutrons emitted by 239 Pu in induced fission should remain unchanged at 1.1400. These changes are all within 1 standard deviation.

Critical Experiment Device (cold) was established for research on the physics characteristics of Molten Salt Reactor (MSR) in 1970s, at Shanghai Institute of Applied Physics, China. Neutron flux distribution was measured for different core structures and the conversion ratio of the core containing thorium fuel was determined on the basis of the preliminary criticality experiments. The purpose of this work is to compare the calculated flux distribution and thorium-uranium conversion ratio with the measured results in this device. Calculations were accomplished by using MCNP5 code with nuclear data from ENDF/B-VII.0, ENDF/B-VII.1, JEFF 3.2, JENDL 4.0, CENDL 3.1 and ROSFOND 2010 libraries. The results show that the calculated results are in good agreements with the experimental data. Findings in this work have a great significance to the design and benchmark verification of a cold MSR. However, more hot experiments are needed to study the nuclear properties of the liquid fuel MSR.