Monte Carlo Neutron Transport Code Research Articles

Coupled neutronics and thermal-hydraulics based on JMCT and FLUENT

A coupling interface between Monte Carlo neutron transport code JMCT2.2 and commercial CFD code FLUENT was developed by C++ language. The one-by-one mapping strategy that one cell in the MCNP model logically corresponded to the computational domain in the FLUENT model was used in this coupling method to realize simple partition of the physical model and refinement of partition of CFD mesh. The calculated results of a PWR single cell model and a PWR 33 fuel model were compared between this program and the MCNP-FLUENT program to validate its accuracy. The results demonstrate that this coupling method between JMCT code and FLUENT code is reasonable and it can accurately calculate the feedback parameters.

Physics-oriented optimization strategy for the energy lookup algorithm in continuous energy Monte Carlo neutron transport simulation

The continuous energy Monte Carlo method is a most high-fidelity and high-resolution method for neutron transport simulations in reactor physics with minimal approximations. However, one of the major disadvantages is that it is very time-consuming and computational-intensive for large scale whole-core simulations, especially for coupling with depletion analysis for realistic reactors. Some recent researches indicate that one of the principle performance bottlenecks for the problem lies in the energy lookup algorithm during the calculation of energy-dependent material cross sections. Therefore, two physics-oriented optimization strategies, based on making use of the physical characteristics of neutron transport behaviors, are developed to optimize the run-time performance of the algorithm for accelerating the energy lookup without any loss in precision and accuracy. The first optimization strategy is called Neighbored Material Cascade Grid (NMCG) which is a hybrid approach utilizing the key features of the cascade grid and double indexing method. The second optimization strategy is called Adaptive Optimal Logarithmic Grid (AOLG) which is a variation of the conventional logarithmic energy grid method utilizing the advantages of energy hash tables. The strategies are incorporated into a continuous energy Monte Carlo neutron transport code and tested on realistic whole-core reactor systems. The computational performance as measured by memory usage, elapsed runtime and overall speedup, associated with each of the optimization strategies are demonstrated in the whole-core Monte Carlo simulations. Depending on the complexity of the models, the number of nuclides in the material compositions and the utilization of different optimization strategies, overall speedup ratios of 1.2-1.7, relative to the conventional binary lookup algorithm, are routinely observed. Furthermore, the numerical results indicate that the run-time performance of the new physics-oriented optimization strategies performs a bit better than that of conventional optimization methods with existing approaches.

Criticality concerns of a group actinide co-crystallization separations approach to used nuclear fuel recycling

The co-crystallization of hexavalent actinides with uranyl nitrate hexahydrate has been established as a viable approach to achieving group actinide separation in used nuclear fuel recycle, while avoiding the difficulties associated with the solvent-extraction process altogether. A preliminary neutronics safety analysis has been performed to alleviate criticality concerns associated with this group hexavalent actinide co-crystallization approach. The basic uranyl nitrate hexahydrate material compositions used are representative of used fuel found at commercial pressurized water reactors all around the U.S. This material is subsequently doped with various percentages of either typical reactor grade plutonium or with the total transuranic vector, replacing the uranium atoms in the crystal lattice structure. Infinite and effective neutron multiplication factors were calculated using the state-of-the-art Monte Carlo neutron transport code MCNP in a variety of plausible material configurations, assuming an acceptable safety limit for criticality to be an effective neutron multiplication factor less than 0.95. The results indicate that the crystal structures form a very subcritical intrinsic geometry even with six water molecules being incorporated into the crystal structure. The results of this research suggest that significant opportunity exists for the safe transfer of the group actinide separation approach from a laboratory to an operational scale.

An improved target velocity sampling algorithm for free gas elastic scattering

We present an improved algorithm for sampling the target velocity when simulating elastic scattering in a Monte Carlo neutron transport code that correctly accounts for the energy dependence of the scattering cross section. The algorithm samples the relative velocity directly, thereby avoiding a potentially inefficient rejection step based on the ratio of cross sections. We have shown that this algorithm requires only one rejection step, whereas other methods of similar accuracy require two rejection steps. The method was verified against stochastic and deterministic reference results for upscattering percentages in 238U. Simulations of a light water reactor pin cell problem demonstrate that using this algorithm results in a 3% or less penalty in performance when compared with an approximate method that is used in most production Monte Carlo codes.

Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

This paper presents the validation of the continuous-energy Monte Carlo neutron-transport code MCS with the ENDF/B-VII.0 neutron cross-section library for the criticality safety analysis of PWR spent fuel pools and storage casks. The MCS code is developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the analysis of Pressurized Water Reactors (PWRs) with high fidelity and high performance. The validation is conducted with 279 selected critical benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The 279 validation cases are representative of PWR spent fuel pools and storage casks with 235U enrichment ranging from 2.35 wt% to 4.74 wt%, pin pitch ranging from 1.075 cm to 2.540 cm, moderator to fuel ratio (H/U) ranging from 0.4683 and 11.5398, Energy of the Average Lethargy causing Fission (EALF) ranging from 0.0109 eV to 1.0600 eV, without soluble boron and with soluble boron concentration ranging from 0.015 g/L to 5.030 g/L. The calculation of the effective neutron multiplication factor by MCS is validated by the comparison between experiment and calculation for the selected critical benchmarks. The Upper Safety Limit (USL) of the MCS code is established in accordance to the NUREG/CR-6698 guideline recommended by the NRC (US National Regulatory Commission). The full validation process and determination of USL based on the selected critical benchmarks was also repeated with the MCNP6.1 and SERPENT2.1.27 codes in order to compare the performances of MCS with other reactor analysis codes. This paper demonstrates the capability of the MCS code for the criticality safety analysis of PWR spent fuel pools and storage casks.

Analysis of simplified BWR full core with serpent-2/simulate-3 hybrid stochastic/deterministic code

This paper presents an evaluation of a Monte Carlo (MC) neutron transport code as a tool to generate nuclear cross section data for nuclear reactor core simulators. The goal is not to replace deterministic lattice calculations with MC simulations but to provide an audit tool. A boiling water reactor (BWR) assembly is modeled with a deterministic lattice physics code CASMO-5 and with a MC code SERPENT-2 for a simplified set of histories and branch cases. The work reported in this paper is a continuation of authors’ previous analysis of a BWR in the reference 13 and SERPENT-2 is used instead of SERPENT-1 to overcome excessive memory usage and execution time of the latter. CASMO-5 and SERPENT-2 are compared with those of a hybrid CASMO-5+SERPENT-2 scheme which uses the depletion solver of CASMO-5 and the transport solver of SERPENT-2. Their results are compared for an assembly model, in terms of k-inf, macroscopic cross sections, assembly discontinuity factors and diffusion coefficients. The differences are less than 1% except for diffusion coefficients of 5% difference. Subsequently, a nuclear cross section data library is generated for SIMULATE-3 core simulator. A full core depletion calculation is performed and k-eff, axial and radial power distributions are compared. The discrepancies of k-eff between CASMO-5/SIMULATE-3 and SERPENT-2/SIMULATE-3 are smaller than 250pcm, and the differences of axial relative power fraction (RPF) root mean square errors are smaller than 8%. The differences of radial RPF root mean square errors are smaller than 1.5%.

MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7

Abstract The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of 0.99994 ± 6.8E-6 (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

CFD/Monte-Carlo neutron transport coupling scheme, application to TRIGA reactor

A new computational model of the JSI TRIGA Mark II, coupling Monte Carlo neutron transport code TRIPOLI and fluid dynamics code CFX was built and verified with a set of new experimental data. A set of subroutines was developed to allow the communication between the Monte-Carlo transport code and CFD code. First, test of the coupling scheme is presented: for a given thermal power of the reactor, the coupled model numerically reproduced fuel temperature monitored during reactor operation and axial water temperature profile measured in the coolant channels. Then axial temperature profiles in the coolant channels were measured with a newly developed sensor during steady-state operation. Predictions of the coupled model are in expected agreement with experimental data recorded during reactor operations. Influence of the coupling has been investigated.

Evaluation of radiation damage in belt-line region of VVER-1000 nuclear reactor pressure vessel

Regarding irradiation damage, the most sensitive region in the reactor pressure vessel (RPV) is the area adjacent to the reactor core; the so called belt-line region. Belt-lines and their related heat affected zones (HAZs) in the belt-line region are constantly under neutron and gamma irradiation and have a higher potential to various flaws. The microscopic defects produced in materials due to irradiation are referred to as radiation damage; the units of displacements per atom (DPA) are used routinely to characterize the extent of such damage. In this research, for a typical VVER-1000 (V446) RPV, the areas of the highest neutron radiation flux (belt-line) is first identified by using a Monte Carlo neutron transport code. This is used to calculate and validate the neutron flux and fluence at the belt-line. Employing the PTRAC output from MCNP and the SRIM Monte Carlo code, radiation damage in the RPV is simulated and assessed in terms of DPA. The integrated damage for a full effective year is evaluated to be order of 5E-04 displacements per cm thickness of RPV belt-line metal.

Analysis of BEAVRS two-cycle benchmark using RMC based on full core detailed model

With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive especially in criticality analysis of initial core and shielding calculations, due to its advantages of flexible geometry modeling and use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with different physics and feedback interactions and coupling. To perform the high fidelity multi-physics simulations of real reactors or benchmark calculations such as two-cycle BEAVRS benchmark based on measurement data of a practical nuclear power plant, several factors must be considered such as large scale detailed depletion, thermal-hydraulics feedback, on-the-fly nuclear cross section processing, criticality search and inter-cycle refueling. In this paper, the abilities mentioned above for multiple burnup cycles simulations in Hot Full Power condition of PWR full core have been developed in continuous-energy Reactor Monte Carlo neutron and photon transport code RMC, which is developed by Department of Engineering Physics at Tsinghua University, Beijing (Wang et al., 2015). RMC has the capacity for lifecycle simulations of nuclear reactor cores. The BEAVRS benchmark was selected as an example and RMC was applied to a full core, two cycle burnup calculation of BEAVRS. All of the parameters given in the BEAVRS benchmark have been calculated and compared with the measured values of BEAVRS benchmark. For other parameters such as pin power distributions, they are compared to the results of other codes. The results of RMC agree well with the measured values of BEAVRS benchmark and also agree well with those of other codes. This was the first time for a Monte Carlo code to perform the full core, two cycle calculation of BEAVRS. This work paves the way for Monte Carlo codes in life cycle simulations of nuclear reactor cores.

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of Their Contribution in Gamma Flux Calculations

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating [12]; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC) [3]. The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR — 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA — 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR — 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed.

On the calculation of angular neutron flux in MCNP

Modern Monte Carlo neutron transport codes offer many options for neutron flux and spectra calculations, however, they often lack the option to obtain the angular neutron flux in a region of the problem. The angular flux can also be obtained from deterministic programs, however, it includes biases due to discretization and other physical approximations. Therefore, a novel method for determining the angular neutron flux from the standard output of the MCNP is proposed in this paper. The method was also implemented as a set of Python libraries and tested in several examples. The results were then used to investigate the self-shielding effect in a realistic angular profile of the flux, i.e., the TRIGA research reactor.

Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies

The paper aims to develop an MCNP5-ORIGEN2 coupling scheme for burnup calculation. Specifically, the Monte Carlo neutron transport code (MCNP5) and the nuclides depletion and decay calculation code (ORIGEN2) are combined by data processing and linking files written in the PERL programming language. The validity and applicability of the developed coupling scheme are tested through predicting the neutronic and isotopic behavior of the “VVER-1000 LEU Assembly Computational Benchmark”. The MCNP5-ORIGEN2 coupling results showed a good agreement with the k-inf benchmark values within 600 pcm during the entire burnup history. In addition, the differences of isotopes concentration at the end of the burnup (40 MWd/kgHM) when compared with benchmark values were reasonable and generally within 6.5%. The developed coupling scheme also considered the shielding effect due to gadolinium isotopes and simulated well the depletion of isotopes as a function of the radial position in gadolinium bearing fuel rods.

Measurement of leakage neutron spectra from graphite cylinders irradiated with D-T neutrons for validation of evaluated nuclear data

A benchmark experiment for validation of graphite data evaluated from nuclear data libraries was conducted for 14MeV neutrons irradiated on graphite cylinder samples. The experiments were performed using the benchmark experimental facility at the China Institute of Atomic Energy (CIAE). The leakage neutron spectra from the surface of graphite (Φ13cm×20cm) at 60° and 120° and graphite (Φ13cm×2cm) at 60° were measured by the time-of-flight (TOF) method. The obtained results were compared with the measurements made by the Monte Carlo neutron transport code MCNP-4C with the ENDF/B-VII.1, CENDL-3.1 and JENDL-4.0 libraries. The results obtained from a 20cm-thick sample revealed that the calculation results with CENDL-3.1 and JENDL-4.0 libraries showed good agreements with the experiments conducted in the whole energy region. However, a large discrepancy of approximately 40% was observed below the 3MeV energy region with the ENDF/B-VII.1 library. For the 2cm-thick sample, the calculated results obtained from the abovementioned three libraries could not reproduce the experimental data in the energy range of 5–7MeV. The graphite data in CENDL-3.1 were verified for the first time and were proved to be reliable.

Benchmarking of 232Th evaluation by a 14.8 MeV neutron leakage spectra experiment with slab samples

An integral benchmark experiment for neutron leakage spectra from ThO2 samples has been performed with a 14.8MeV D-T neutron source at China Institute of Atomic Energy (CIAE) to validate the evaluated nuclear data related to the design of Thorium Molten Salt Reactor (TMSR). The leakage spectra from ThO2 samples were measured at 60° and 120° emission angles with the Time of Flight (TOF) method. The measured spectra are well reproduced as a whole by the Monte Carlo neutron transport code simulations with the ENDF/B-VII.1, JENDL-4.0 and CENDL-3.1 evaluated nuclear data, while some discrepancies in the major reaction channels including (n,el), (n,inl), (n,2n) and (n,3n), still exist. Detailed information on the neutron product distribution with the evaluated data from the above three libraries is presented to explain the discrepancies. From the comparisons, the data of the above three libraries need to be improved as required in different energy ranges.

Neutronic analysis of JET external neutron monitor response

The power output of fusion devices is measured in terms of the neutron yield which relates directly to the fusion yield. JET made a transition from Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) during 2010–11. Absolutely calibrated measurement of the neutron yield by JET neutron monitors was ensured by direct measurements using a calibrated 252Cf neutron source (NS) deployed by the in-vessel remote handling system (RHS) inside the JET vacuum vessel.Neutronic calculations were required in order to understand the neutron transport from the source in the vacuum vessel to the fission chamber detectors mounted outside the vessel on the transformer limbs of the tokamak. We developed a simplified computational model of JET and the JET RHS in Monte Carlo neutron transport code MCNP and analyzed the paths and structures through which neutrons reach the detectors and the effect of the JET RHS on the neutron monitor response. In addition we performed several sensitivity studies of the effect of substantial massive structures blocking the ports on the external neutron monitor response. As the simplified model provided a qualitative picture of the process only, some calculations were repeated using a more detailed full 3D model of the JET tokamak.

Full 3-D core calculations with refueling for the OPAL Research Reactor using Monte Carlo Code Serpent 2

Monte Carlo (MC) neutron transport codes have been extensively used for more than three decades to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. The irruption of affordable low-cost high performance computer resources in the last decade allows to consider some initially unexpected applications, such as full core burnup calculations or cell level modeling for few group parameters calculations. In this work the concern of the potential use of Monte Carlo codes to perform full 3-D calculations including burnup for a state of art Research Reactor is analyzed, regarding aspects related to accuracy, performance and resources requirements. For such purpose Serpent 2 v.1.24 Code, developed by VTT Technical Research Centre of Finland is used for full core burnup calculations of the 20MWth OPAL Research Reactor. This code is the second version of a brand-new Monte Carlo code designed to perform burn dependent cell-level and full 3-D core calculations using optimized schemes to diminish the computational effort. In past works the first version of Serpent Code was tested as a cell-level-code to model the MTR-type fuel assemblies from OPAL Research Reactor, obtaining fairly good results. Further works were developed for full 3-D models, where several parameters such as critical configurations, in-core thermal neutron flux profiles and effective delayed neutron parameters were obtained and compared to experimental data and other codes results, showing a very good performance. In the present work, a full 3-D model is developed using specifications and high quality experimental data from IAEA Technical Report Series. This model is used to perform full-core 3-D calculations including burnup and refueling for the first six operating cycles without the aim of an external calculation code. To perform such task, an ad hoc code to manipulate Serpent 2 restart files was developed in order to model the overall full core burnup problem without the help of any other calculation code. The results were compared with the experimental data available showing a very good agreement. Finally several aspects of the computational issues related with this modeling such performance, scalability and resources requirements are discussed, showing that the use of full core 3-D MC models including burnup for small cores represents nowadays a feasible alternative for specific calculations.

A study of 239Pu production rate in a water cooled natural uranium blanket mock-up of a fusion–fission hybrid reactor

The 239Pu production rate is important data in neutronics design for a natural uranium blanket of a fusion–fission hybrid reactor, and the accuracy and reliability should be validated by integral experiments. The distribution of 239Pu production rates in a subcritical natural uranium blanket mock-up was obtained for the first time with a D-T neutron generator by using an activation technique. Natural uranium foils were placed in different spatial locations of the mock-up, the counts of 277.6 keV γ-rays emitted from 239Np generated by 238U capture reaction were measured by an HPGe γ spectrometer, and the self-absorption of natural uranium foils was corrected. The experiment was analyzed using the Super Monte Carlo neutron transport code SuperMC2.0 with recent nuclear data of 238U from the ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0u2, JEFF-3.2 and CENDL-3.1 libraries. Calculation results with the JEFF-3.2 library agree with the experimental ones best, and they agree within the experimental uncertainty in general with the average ratios of calculation results to experimental results (C/E) in the range of 0.93 to 1.01.

Validation of the Serpent 2-DYNSUB code sequence using the Special Power Excursion Reactor Test III (SPERT III)

The Special Power Excursion Reactor Test III (SPERT III) is studied using the Serpent 2-DYNSUB code sequence in order to validate it for modeling reactivity insertion accidents (RIA) in PWRs. The SPERT III E-core was a thermal research reactor constructed to analyze reactor dynamics. Its configuration resembles a commercial PWR on terms of fuel type, choice of moderator, coolant flow and system pressure. The initial conditions of the rod ejection accident experiments (REA) performed cover cold startup, hot startup, hot standby and operating power scenarios. Eight of these experiments were analyzed in detail. Firstly, multi-dimensional nodal diffusion cross section tables were created for the three-dimensional reactor simulator DYNSUB employing the Monte Carlo neutron transport code Serpent 2. In a second step, DYNSUB stationary simulations were compared to Monte Carlo reference three-dimensional full scale solutions obtained with Serpent 2 (cold startup conditions) and Serpent 2/SUBCHANFLOW (operating power conditions) with a good agreement being observed. The latter tool is an internal coupling of Serpent 2 and the sub-channel thermal-hydraulics code SUBCHANFLOW. Finally, DYNSUB was utilized to study the eight selected transient experiments. Results were found to match measurements well. As the selected experiments cover much of the possible transient (delayed super-critical, prompt super-critical and super-prompt critical excursion) and initial conditions (cold and hot as well as zero, little and full power reactor states) one expects in commercial PWRs, the obtained results give confidence that the Serpent 2-DYNSUB tool chain is suitable to model REAs and other RIAs in PWRs.

Development of One-way-coupling Methodology between Severe Accident Integral Code MELCOR and Monte Carlo Neutron Transport Code SERPENT

A SARCAM (Severe Accident Re-Criticality Assessment Methods) code is under development. It allows data exchange between Severe Accident integral computer code MELCOR and Monte Carlo neutron transport code SERPENT. The basic purpose of this framework is a systematic assessment of the neutron multiplication properties of the molten corium, debris and core remnants, during and after a severe accident in a light water reactor. The paper describes the current status of the methodology development for the lower plenum calculations and provides simulations (test) results for the BWR reactor based on the Fukushima Unit 3 (1F3) design.