Code MCS Research Articles

Development of an Adjoint Flux Calculation Technique for Exact Perturbation Theory in Monte Carlo Code MCS

A calculation technique computing the adjoint flux of perturbed system is developed for the exact perturbation theory in Monte Carlo transport. By using a correlated sampling and iterated fission probability methods together, the adjoint flux of perturbed system is calculated during a forward Mote Carlo simulation of an unperturbed system. In the method, no additional particle simulation is required to compute the adjoint flux of perturbed system. The exact perturbation method is implemented in the Monte Carlo code MCS. The results of perturbation method for k-eigenvalue change are compared against differential operator sampling, adjoint-weighted perturbation, and direct perturbation methods for the Godiva benchmark and PLUS7 fuel assembly.

Modelling atomic relaxation and bremsstrahlung in the deterministic code STREAM

STREAM, developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST), is a deterministic neutron- and photon-transport code primarily designed for light water reactor (LWR) analysis. Initially, the photon module in STREAM did not account for fluorescence and bremsstrahlung photons. This article presents recent developments regarding the integration of atomic relaxation and bremsstrahlung models into the existing photon module, thus allowing for the transport of secondary photons. The photon flux and photon heating computed with the newly incorporated models is compared to results obtained with the Monte Carlo code MCS. The incorporation of secondary photons has substantially improved the accuracy of photon flux calculations, particularly in scenarios involving strong gamma emitters. However, it is essential to note that despite the consideration of secondary photon sources, there is no noticeable improvement in the photon heating for LWR problems when compared to the photon heating obtained with the previous version of STREAM.

Inverted fuel geometry implementation for a fast reactor: Potential improvements in neutronic and thermalhydraulic performance

This paper presents a detailed procedure for implementing the inverted fuel geometry to a fast reactor to improve its safety system and economy. The study is starting from the fuel unit cell, fuel assembly, reactor core, and burnup analysis. A proposed multivariable graph (v, ΔP, TFmax–Dco, VF) introduced at the fuel unit cell level provides comprehensive thermalhydraulic and neutronic parameters in a single graph, allowing for an efficient optimization process. The fuel unit cell study reveals that the inverted fuel design has a higher fuel volume fraction and lower core pressure drop than conventional pin-typed fuel. This is beneficial for the reactor economy and enhances the reliability of the safety system. With the inverted fuel design, the primary loop can save pumping power by up to 40 % and provides an excess driving force for natural circulation. The male–female axial grid structure separating the fuel assembly potentially eliminates coolant flow path restriction and fretting issues. The core is named the Inverted Core Fast Reactor (IC-FR), an LBE-cooled fast SMR designed to generate 60 MWth for at least 40 years of full-power operation without refueling and fuel shuffling. IC-FR is a transportable reactor and has load following capability that can be deployed for many applications, including marine and land-based applications, and stand alone or mixing power grid. The burnup study of IC-FR reveals that the balance of neutron leakage and fissile inventory yields a small reactivity swing (<1$) for 40 years. This study extensively utilizes Monte Carlo (MC) code MCS for neutronic calculation. Owing to the high computational expense of MC code, the approaches to optimize the MC usage are also presented.

Coupled neutronics/thermal-hydraulic analysis of ANTS-100e using MCS/RAST-F two-step code system

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

Group constants generation by Monte Carlo code MCS for LWR analysis

This paper presents the generation of few-group constants using the Monte Carlo (MC) code MCS for light-water reactor (LWR) analysis. The few-group constants are fed into the conventional diffusion nodal code to perform neutronic analysis. This study investigates three methods, B1, P1, and CASMO-4E (CM), to obtain the critical spectrum for correcting the cross-section for LWR analysis. The inflow transport correction based on the PN solution is integrated into the MCS in addition to the typical outflow transport correction in the MC codes. A numerical case study comparing the two-step code MCS/PARCS and MCS standalone results is required to determine the best combination in the transport/diffusion two-step to achieve a high-fidelity LWR analysis. Numerical tests with small and large LWR cores show that the CM method combined with inflow transport correction is the most accurate for two-step code MCS/diffusion in terms of the eigenvalues and power profiles.

Validation of MCS code for shielding calculation using SINBAD

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this paper, the radiation shielding capability of the MCS code is validated by simulating some selected SINBAD (Shielding Integral Benchmark Archive and Database) experiments. The whole validation was performed in two ways. Firstly, the functionality and computational rationality of the MCS code was verified by comparing the simulation results with those of MCNP code. Secondly, the validity and computational accuracy of the MCS code was confirmed by comparing the simulation results with the experimental results of SINBAD. The simulation results of the MCS code are highly consistent with the those of the MCNP code, and they are within the 2σ error bound of the experiment results. It shows that the calculation results of the MCS code are reliable when simulating the radiation shielding problems.

Analysis of several VERA benchmark problems with the photon transport capability of STREAM

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS – also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

Verification of a two-step code system MCS/RAST-F to fast reactor core analysis

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in keff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

Use of Monte Carlo code MCS for multigroup cross section generation for fast reactor analysis

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MWth metal-fueled SFR (MET-1000) and a 3600 MWth oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core keff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core keff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.

MicroURANUS : Core design for long‐cycle lead‐bismuth‐cooled fast reactor for marine applications

A core design of MicroURANUS, a long-cycle lead-bismuth-cooled fast nuclear reactor for marine applications, is presented. It aims to generate a power of 60MWth, which can be regulated during operation. MicroURANUS was designed to achieve a small burnup reactivity swing for 30 effective full-power years of a lifetime without refueling. To attain these goals, a unit cell study with uranium oxide fuel was initiated to lay a solid foundation for core design, owing to size constraints. A reflector optimization was also performed to minimize neutron leakage. MicroURANUS adopts onion zoning constructed by two enrichment zones for flattening the power and lengthening the core lifetime. The coolant is driven by electromagnetic pumps to achieve inlet and outlet temperatures of 250°C and 350°C, respectively. MicroURANUS analyses were performed using the Argonne Reactor Computation suite and the Monte Carlo code MCS. Core performance features were analyzed for criticality, power profiles, fuel isotope mass inventory, reflector coefficients, reactivity feedback coefficients, and shutdown margin. Additional thermal-hydraulic calculations were performed to confirm that the fuel and cladding temperatures were within the acceptable range. Furthermore, a load follow analysis using the quasi-static reactivity balance method confirmed the feasibility of regulating power by adjusting the inlet coolant temperature.

UNCERTAINTY ANALYSIS OF SUB-EXERCISES IN UAM-SFR BENCHMARK WITH THE MCS CODE

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, sub-exercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core (MOX-3600) are tested by a Monte Carlo code MCS to perform uncertainty analysis. Classical perturbation theory and generalized perturbation theory are used to calculate sensitivity coefficients. Uncertainty is calculated by multiplying the sensitivity coefficients and relative covariance matrix from ENDF/B-VII.1 library.

MACROSCOPIC CROSS SECTIONS GENERATION BY MONTE CARLO CODE MCS FOR FAST REACTOR ANALYSIS

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.

Criticality benchmark of Monte Carlo code MCS for light water reactor fuel in transportation and storage packages

This paper presents a verification and validation (V&V) study of the Monte Carlo neutron transport code MCS for the criticality analysis of LWR fuel in transportation and storage casks/packages. A total of 173 critical experiments from the NUREG/CR-6361 report are analyzed with MCS and the ENDF/B-VII.1 cross section library. A preliminary verification is conducted by comparing the reaction rates. Subsequently, a criticality analysis of the 173 benchmarks is performed. The average value of the multiplication factors calculated by MCS for the 173 critical experiments is 1.00243 with a standard deviation of 291 pcm and extremum error values of −1451 pcm and +1661 pcm compared to the experimental results. The bias of the package (MCS + ENDF/B-VII.1 library) for the criticality analysis of LWR fuel transportation/storage casks is, therefore, estimated to be +243 pcm. This paper demonstrates the capability of the MCS code to perform the criticality analysis of LWR fuel transportation and storage packages.

Neutronic simulation of China experimental fast reactor start-up tests- part II: MCS code Monte Carlo calculation

The International Atomic Energy Agency and the China Institute of Atomic Energy proposed a coordinated research project (CRP) to establish a benchmark, based on the China Experimental Fast Reactor (CEFR) start-up tests which include fuel loading and criticality, measurements of control rod worth and reactivity coefficients, and foil activation measurements. As a participant in the CRP, the computational reactor physics and experiment laboratory of the Ulsan National Institute of Science and Technology has been studying and analyzing the neutronic performance of the CEFR. The verification and validation are performed for the simulations in the CEFR analysis. A verification process of the methodology was conducted to compare the predicted results with those obtained by a well-known MCNP6.1 Monte Carlo code. The results obtained by the MCS code were compared against the experimental data. The difference in various wire models at the critical state are evaluated in this study.

Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCS

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.

Validation of UNIST Monte Carlo code MCS for criticality safety calculations with burnup credit through MOX criticality benchmark problems

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.

Validation of nuclide depletion capabilities in Monte Carlo code MCS

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

RETRACTED: Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code

Abstract 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.

Extension of Monte Carlo code MCS to spent fuel cask shielding analysis

The shielding analysis of the spent fuel dry storage cask TN-32 is carried out using the continuous-energy Monte Carlo neutron- and photon-transport code MCS developed by the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The shielding analysis involves the transport simulation of the decay neutrons and gammas emitted by the spent fuel to obtain the neutron and gamma dose rate values outside of the cask. This deep-penetration problem is solved by means of weight window population control as variance reduction. The TN-32 cask calculations are repeated with the MCNP6 Monte Carlo code for verification purpose and a good agreement within three standard deviations is observed between MCS and MCNP6. The neutron dose rate on the cask surface and 2 m away from the cask can be calculated with MCS, whereas the photon dose rate calculation for positions outside of the cask will be further improved by employing more sophisticated variance reduction techniques.

Interpretation of two SINBAD photon-leakage benchmarks with nuclear library ENDF/B-VIII.0 and Monte Carlo code MCS

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article.