Results for “Monte Carlo Code” — Page 26

DOE OSTI2020

Progress in modelling fast-ion D-alpha spectra and neutral particle analyzer fluxes using FIDASIM

FIDASIM is a code that models signals produced by charge-exchange reactions between neutrals and ions (both fast and thermal) in magnetically confined plasmas. With the ion distribution function as input, the code predicts the efflux to a neutral particle analyzer diagnostic and the photon radiance of Balmer-alpha light to a fast-ion D α diagnostic, in addition to many other related quantities. A new, parallelized version of the Monte Carlo code FIDASIM has been developed in Fortran90 that is substantially faster than the original interactive data language version. Modified algorithms include more accurate treatments of the time dependent collisional-radiative equations that describe neutral energy levels, of the cloud of ‘halo’ neutrals that surround the injected neutral beam, and of finite Larmor radius effects. Enhanced physics capabilities include modelling ‘passive’ signals from cold edge neutrals, the ability to treat general three-dimensional magnetic confinement configurations, and calculations of diagnostic-specific weight functions that enable tomographic reconstructions of the fast-ion distribution function. Neutral beam attenuation, beam emission, and fast-ion birth profiles are also modelled. Finally, the new algorithms have been successfully validated against experimental data and new features have been tested through benchmarks between two independently developed versions of the code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

NASA NTRS2000

An Improved Neutron Transport Algorithm for Space Radiation

A low-energy neutron transport algorithm for use in space radiation protection is developed. The algorithm is based upon a multigroup analysis of the straight-ahead Boltzmann equation by using a mean value theorem for integrals. This analysis is accomplished by solving a realistic but simplified neutron transport test problem. The test problem is analyzed by using numerical and analytical procedures to obtain an accurate solution within specified error bounds. Results from the test problem are then used for determining mean values associated with rescattering terms that are associated with a multigroup solution of the straight-ahead Boltzmann equation. The algorithm is then coupled to the Langley HZETRN code through the evaporation source term. Evaluation of the neutron fluence generated by the solar particle event of February 23, 1956, for a water and an aluminum-water shield-target configuration is then compared with LAHET and MCNPX Monte Carlo code calculations for the same shield-target configuration. The algorithm developed showed a great improvement in results over the unmodified HZETRN solution. In addition, a two-directional solution of the evaporation source showed even further improvement of the fluence near the front of the water target where diffusion from the front surface is important.

Heinbockel, John H.↗

DOE OSTI2012

Optimization of Monte Carlo transport simulations in stochastic media

This paper presents an accurate and efficient approach to optimize radiation transport simulations in a stochastic medium of high heterogeneity, like the Very High Temperature Gas-cooled Reactor (VHTR) configurations packed with TRISO fuel particles. Based on a fast nearest neighbor search algorithm, a modified fast Random Sequential Addition (RSA) method is first developed to speed up the generation of the stochastic media systems packed with both mono-sized and poly-sized spheres. A fast neutron tracking method is then developed to optimize the next sphere boundary search in the radiation transport procedure. In order to investigate their accuracy and efficiency, the developed sphere packing and neutron tracking methods are implemented into an in-house continuous energy Monte Carlo code to solve an eigenvalue problem in VHTR unit cells. Comparison with the MCNP benchmark calculations for the same problem indicates that the new methods show considerably higher computational efficiency. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Model Validation and Uncertainty Quantification on the KRUSTY Microreactor Design Using GRIFFIN Neutron Transport Code [Poster]

Argonne National Laboratory (ANL) and INL have developed a GRIFFIN steady state neutronics model for the multiphysics simulations of the Kilopower Reactor Using Sterling TechnologY (KRUSTY) microreactor in the Multiphysics Object Oriented Simulation Environment (MOOSE). The reliability of such deterministic neutronics models can be validated by comparing with computations from Monte Carlo codes (e.g. MCNP, SERPENT, OpenMC, Shift, etc). Furthermore, potential modeling/design improvements can be identified by incorporating uncertainty quantification (UQ), which can be performed by MOOSE’s Stochastic Tools Module (STM). KRUSTY is a prototype for a 5-kW thermal nuclear-powered space reactor. Its primary components consist of nuclear fuel, heat pipes, a control rod, a reflector, and the shielding. The fuel consists of 3 stacked U-7.65Mo cylinders with a hole in the center for the control rod. 8 liquid sodium heat pipes transfer fission energy from the solid fuel block to the Sterling power conversion system where the energy is extracted, and the cooled sodium flows back to the core via capillary action . The movable Boron Carbide control rod regulates the neutron population during startup or when a reactor temperature boost is needed . The beryllium oxide reflector is in 3 places in the reactor; it surrounds the core axially, it lies beneath the core on a platen, and it is present in the shim. The axial and lower reflectors rest on an adjustable stainless-steel platen that moves upward to cover the fuel and help the reactor reach criticality. Lastly, radial stainless steel surrounds the core offering protection from radiation exposure .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

WE-B-207-00: CT Lung Cancer Screening Part 1

The US National Lung Screening Trial (NLST) was a multi-center randomized, controlled trial comparing a low-dose CT (LDCT) to posterior-anterior (PA) chest x-ray (CXR) in screening older, current and former heavy smokers for early detection of lung cancer. Recruitment was launched in September 2002 and ended in April 2004 when 53,454 participants had been randomized at 33 screening sites in equal proportions. Funded by the National Cancer Institute this trial demonstrated that LDCT screening reduced lung cancer mortality. The US Preventive Services Task Force (USPSTF) cited NLST findings and conclusions in its deliberations and analysis of lung cancer screening. Under the 2010 Patient Protection and Affordable Care Act, the USPSTF favorable recommendation regarding lung cancer CT screening assisted in obtaining third-party payers coverage for screening. The objective of this session is to provide an introduction to the NLST and the trial findings, in addition to a comprehensive review of the dosimetry investigations and assessments completed using individual NLST participant CT and CXR examinations. Session presentations will review and discuss the findings of two independent assessments, a CXR assessment and the findings of a CT investigation calculating individual organ dosimetry values. The CXR assessment reviewed a total of 73,733 chest x-ray exams that were performed on 92 chest imaging systems of which 66,157 participant examinations were used. The CT organ dosimetry investigation collected scan parameters from 23,773 CT examinations; a subset of the 75,133 CT examinations performed using 97 multi-detector CT scanners. Organ dose conversion coefficients were calculated using a Monte Carlo code. An experimentally-validated CT scanner simulation was coupled with 193 adult hybrid computational phantoms representing the height and weight of the current U.S. population. The dose to selected organs was calculated using the organ dose library and the abstracted scan parameters. This session will review the results and summarize the individualized doses to major organs and the mean effective dose and CTDIvol estimate for 66,157 PA chest and 23,773 CT examinations respectively, using size-dependent computational phantoms coupled with Monte Carlo calculations. Learning Objectives: Review and summarize relevant NLST findings and conclusions. Understand the scope and scale of the NLST specific to participant dosimetry. Provide a comprehensive review of NLST participant dosimetry assessments. Summarize the results of an investigation providing individualized organ dose estimates for NLST participant cohorts.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2015

WE-B-207-01: CT Lung Cancer Screening and the Medical Physicist: Background, Findings and Participant Dosimetry Summary of the National Lung Screening Trial (NLST)

The US National Lung Screening Trial (NLST) was a multi-center randomized, controlled trial comparing a low-dose CT (LDCT) to posterior-anterior (PA) chest x-ray (CXR) in screening older, current and former heavy smokers for early detection of lung cancer. Recruitment was launched in September 2002 and ended in April 2004 when 53,454 participants had been randomized at 33 screening sites in equal proportions. Funded by the National Cancer Institute this trial demonstrated that LDCT screening reduced lung cancer mortality. The US Preventive Services Task Force (USPSTF) cited NLST findings and conclusions in its deliberations and analysis of lung cancer screening. Under the 2010 Patient Protection and Affordable Care Act, the USPSTF favorable recommendation regarding lung cancer CT screening assisted in obtaining third-party payers coverage for screening. The objective of this session is to provide an introduction to the NLST and the trial findings, in addition to a comprehensive review of the dosimetry investigations and assessments completed using individual NLST participant CT and CXR examinations. Session presentations will review and discuss the findings of two independent assessments, a CXR assessment and the findings of a CT investigation calculating individual organ dosimetry values. The CXR assessment reviewed a total of 73,733 chest x-ray exams that were performed on 92 chest imaging systems of which 66,157 participant examinations were used. The CT organ dosimetry investigation collected scan parameters from 23,773 CT examinations; a subset of the 75,133 CT examinations performed using 97 multi-detector CT scanners. Organ dose conversion coefficients were calculated using a Monte Carlo code. An experimentally-validated CT scanner simulation was coupled with 193 adult hybrid computational phantoms representing the height and weight of the current U.S. population. The dose to selected organs was calculated using the organ dose library and the abstracted scan parameters. This session will review the results and summarize the individualized doses to major organs and the mean effective dose and CTDIvol estimate for 66,157 PA chest and 23,773 CT examinations respectively, using size-dependent computational phantoms coupled with Monte Carlo calculations. Learning Objectives: Review and summarize relevant NLST findings and conclusions. Understand the scope and scale of the NLST specific to participant dosimetry. Provide a comprehensive review of NLST participant dosimetry assessments. Summarize the results of an investigation providing individualized organ dose estimates for NLST participant cohorts.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2015

WE-B-207-02: CT Lung Cancer Screening and the Medical Physicist: A Dosimetry Summary of CT Participants in the National Lung Cancer Screening Trial (NLST)

The US National Lung Screening Trial (NLST) was a multi-center randomized, controlled trial comparing a low-dose CT (LDCT) to posterior-anterior (PA) chest x-ray (CXR) in screening older, current and former heavy smokers for early detection of lung cancer. Recruitment was launched in September 2002 and ended in April 2004 when 53,454 participants had been randomized at 33 screening sites in equal proportions. Funded by the National Cancer Institute this trial demonstrated that LDCT screening reduced lung cancer mortality. The US Preventive Services Task Force (USPSTF) cited NLST findings and conclusions in its deliberations and analysis of lung cancer screening. Under the 2010 Patient Protection and Affordable Care Act, the USPSTF favorable recommendation regarding lung cancer CT screening assisted in obtaining third-party payers coverage for screening. The objective of this session is to provide an introduction to the NLST and the trial findings, in addition to a comprehensive review of the dosimetry investigations and assessments completed using individual NLST participant CT and CXR examinations. Session presentations will review and discuss the findings of two independent assessments, a CXR assessment and the findings of a CT investigation calculating individual organ dosimetry values. The CXR assessment reviewed a total of 73,733 chest x-ray exams that were performed on 92 chest imaging systems of which 66,157 participant examinations were used. The CT organ dosimetry investigation collected scan parameters from 23,773 CT examinations; a subset of the 75,133 CT examinations performed using 97 multi-detector CT scanners. Organ dose conversion coefficients were calculated using a Monte Carlo code. An experimentally-validated CT scanner simulation was coupled with 193 adult hybrid computational phantoms representing the height and weight of the current U.S. population. The dose to selected organs was calculated using the organ dose library and the abstracted scan parameters. This session will review the results and summarize the individualized doses to major organs and the mean effective dose and CTDIvol estimate for 66,157 PA chest and 23,773 CT examinations respectively, using size-dependent computational phantoms coupled with Monte Carlo calculations. Learning Objectives: Review and summarize relevant NLST findings and conclusions. Understand the scope and scale of the NLST specific to participant dosimetry. Provide a comprehensive review of NLST participant dosimetry assessments. Summarize the results of an investigation providing individualized organ dose estimates for NLST participant cohorts.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2017

Electron Acceleration at Pulsar Wind Termination Shocks

We study the acceleration of electrons and positrons at an electromagnetically modified, ultrarelativistic shock in the context of pulsar wind nebulae. We simulate the outflow produced by an obliquely rotating pulsar in proximity of its termination shock with a two-fluid code that uses a magnetic shear wave to mimic the properties of the wind. We integrate electron trajectories in the test-particle limit in the resulting background electromagnetic fields to analyze the injection mechanism. We find that the shock-precursor structure energizes and reflects a sizable fraction of particles, which becomes available for further acceleration. We investigate the subsequent first-order Fermi process sustained by small-scale magnetic fluctuations with a Monte Carlo code. We find that the acceleration proceeds in two distinct regimes: when the gyroradius r{sub g} exceeds the wavelength of the shear λ, the process is remarkably similar to first-order Fermi acceleration at relativistic, parallel shocks. This regime corresponds to a low-density wind that allows the propagation of superluminal waves. When r{sub g}<λ, which corresponds to the scenario of driven reconnection, the spectrum is softer.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2014

First measurements of D{sub α} spectrum produced by anisotropic fast ions in the gas dynamic trap

Angled injection of eight deuterium beams in gas dynamic trap (GDT) plasmas builds up the population of fast ions with the distribution function, which conserves a high degree of initial anisotropy in space, energy, and pitch angle. Unlike the Maxwellian distribution case, the fast ion plasma component in GDT cannot be exhaustively characterized by the temperature and density. The instrumentation complex to study of fast ions is comprised of motional Stark effect diagnostic, analyzers of charge exchange atoms, and others. The set of numerical codes using for equilibrium modeling is also an important tool of analysis. In the recent campaign of summer 2014, we recorded first signals from the new fast ion D-alpha diagnostic on GDT. This paper presents the diagnostic description and results of pilot measurements. The diagnostic has four lines of sight, distributed across the radius of an axially symmetric plasma column in GDT. In the present setup, a line-integrated optical signal is measured in each channel. In the transverse direction, the spatial resolution is 18 mm. Collected light comes to the grating spectrometer with the low-noise detector based on a charge-coupled device matrix. In the regime of four spectra stacked vertically on the sensor, the effective spectral resolution of measurements is approximately 0.015 nm. Exposure timing is provided by the fast optical ferroelectric crystal shutter, allowing frames of duration down to 70 μs. This number represents the time resolution of measurements. A large dynamic range of the camera permits for a measurement of relatively small light signals produced by fast ions on top of the bright background emission from the bulk plasma. The fast ion emission has a non-Gaussian spectrum featuring the characteristic width of approximately 4 nm, which can be separated from relatively narrow Gaussian lines of D-alpha and H-alpha coming from the plasma periphery, and diagnostic beam emission. The signal to noise ratio varies from approximately ten for the central channel to approximately five for the outermost channel. We used the special set of Monte Carlo codes to fit the measured spectra. The shape of model fit shows a good agreement with the experimental fast ion D-alpha spectrum.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

SU-E-T-627: Precision Modelling of the Leaf-Bank Rotation in Elekta’s Agility MLC: Is It Necessary?

Purpose: To demonstrate the method used to determine the leaf bank rotation angle (LBROT) as a parameter for modeling the Elekta Agility multi-leaf collimator (MLC) for Monte Carlo simulations and to evaluate the clinical impact of LBROT. Methods: A detailed model of an Elekta Infinity linac including an Agility MLC was built using the EGSnrc/BEAMnrc Monte Carlo code. The Agility 160-leaf MLC is modelled using the MLCE component module which allows for leaf bank rotation using the parameter LBROT. A precise value of LBROT is obtained by comparing measured and simulated profiles of a specific field, which has leaves arranged in a repeated pattern such that one leaf is opened and the adjacent one is closed. Profile measurements from an Agility linac are taken with gafchromic film, and an ion chamber is used to set the absolute dose. The measurements are compared to Monte Carlo (MC) simulations and the LBROT is adjusted until a match is found. The clinical impact of LBROT is evaluated by observing how an MC dose calculation changes with LBROT. A clinical Stereotactic Body Radiation Treatment (SBRT) plan is calculated using BEAMnrc/DOSXYZnrc simulations with different input values for LBROT. Results: Using the method outlined above, the LBROT is determined to be 9±1 mrad. Differences as high as 4% are observed in a clinical SBRT plan between the extreme case (LBROT not modeled) and the nominal case. Conclusion: In small-field radiation therapy treatment planning, it is important to properly account for LBROT as an input parameter for MC dose calculations with the Agility MLC. More work is ongoing to elucidate the observed differences by determining the contributions from transmission dose, change in field size, and source occlusion, which are all dependent on LBROT. This work was supported by OCAIRO (Ontario Consortium of Adaptive Interventions in Radiation Oncology), funded by the Ontario Research Fund.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

SU-E-T-169: Evaluation of Oncentra TPS for Nasopharynx Brachy Using Patient Specific Voxel Phantom and EGSnrc

Purpose: Nasopharnx carcinoma (NPC) treatment is being carried out using Ir-192 HDR seeds in Mehdieh Hospital in Hamadan, Iran. The Oncentra™ TPS is based on optimized TG-43 formalism which disregards heterogeneity in the treatment area. Due to abundant heterogeneity in head and neck, comparison of the Oncentra™ TPS dose evaluation and an accurate dose calculation method in NPC brachytherapy is the objective of this study. Methods: CT DICOMs of a patient with NPC obtained from Mehdieh Hospital used to create 3D voxel phantom with CTCREATE utility of EGSnrc code package. The voxel phantom together with Ir-192 HDR brachytherapy source were the input to DOSXYZnrc to calculate the 3D dose distribution. The sources were incorporate with type 6 source in DOSXYZnrc and their dwell times were taken into account in final dose calculations. Results: The direct comparison between isodoses as well as DVHs for the GTV, PTV and CTV obtained by Oncentra™ and EGSnrc Monte Carlo code are made. EGSnrc results are obtained using 5×10{sup 9} histories to reduce the statistical error below 1% in GTV and 5% in 5% dose areas. The standard ICRP700 cross section library is employed in DOSXYZnrc dose calculation. Conclusion: A direct relationship between increased dose differences and increased material density (hence heterogeneity) is observed when isodoses contours of the TPS and DOSXYZnrc are compared. Regarding the point dose calculations, the differences range from 1.2% in PTV to 5.6% for cavity region and 7.8% for bone regions. While Oncentra™ TPS overestimates the dose in cavities, it tends to underestimate dose depositions within bones.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2012

CPR methodology with new steady-state criterion and more accurate statistical treatment of channel bow

An overview is given of existing CPR design criteria and the methods used in BWR reload analysis to evaluate the impact of channel bow on CPR margins. Potential weaknesses in today's methodologies are discussed. Westinghouse in collaboration with KKL and Axpo - operator and owner of the Leibstadt NPP - has developed an optimized CPR methodology based on a new criterion to protect against dryout during normal operation and with a more rigorous treatment of channel bow. The new steady-state criterion is expressed in terms of an upper limit of 0.01 for the dryout failure probability per year. This is considered a meaningful and appropriate criterion that can be directly related to the probabilistic criteria set-up for the analyses of Anticipated Operation Occurrences (AOOs) and accidents. In the Monte Carlo approach a statistical modeling of channel bow and an accurate evaluation of CPR response functions allow the associated CPR penalties to be included directly in the plant SLMCPR and OLMCPR in a best-estimate manner. In this way, the treatment of channel bow is equivalent to all other uncertainties affecting CPR. Emphasis is put on quantifying the statistical distribution of channel bow throughout the core using measurement data. The optimized CPR methodology has been implemented in the Westinghouse Monte Carlo code, McSLAP. The methodology improves the quality of dryout safety assessments by supplying more valuable information and better control of conservatisms in establishing operational limits for CPR. The methodology is demonstrated with application examples from the introduction at KKL. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Serpent and MCNP Calculations of the Energy Deposition in the Transformational Challenge Reactor

This paper focuses on the calculation of the energy deposition in the Transformational Challenge Reactor by two major Monte Carlo codes: Serpent and MCNP. The first software computation relies on Kinetic Energy Released per unit Mass (KERMA) factors while the second one relies on Q-values. The results from these two independent computation methodologies are in very good agreement; however, Serpent runs much faster than MCNP (for the same computational model) and allows for a detailed energy deposition distribution from a 1-mm-side square mesh with a relative statistical error between 0.5% and 1%. This detailed energy deposition is suitable for multiphysics analyses aimed at design optimizations. In order to calculate the energy deposition, Serpent needs enhanced ACE files (distributed by the software developers). Unlike other Monte Carlo software that uses inputs based on Python or Java languages, the Serpent input syntax is very similar to that of MCNP; a Python script can convert a MCNP input to a Serpent input in seconds. For simulations not requiring the calculation of the energy deposition, Serpent can also read nuclear data from MCNP ACE files, which eventually improves the comparison of the results of the two codes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

An uncertainty quantification method relevant to material test reactors

Within material test reactor calculations, energy dependent flux and reaction rate uncertainties are typically not quantified when performing as-run analyses to determine the neutron field experienced by the experiment. When high fidelity Monte-Carlo codes are used in such analyses, straight forward methods to calculate output uncertainties are not available, instead expert opinion is used to postulate computational uncertainties. New methods to propagate uncertainties through these high fidelity simulations are available when sufficient computational power is available. A tool is developed here for sampling any part of an MCNP input from random distributions to determine output uncertainties based on those inputs. Another tool is developed to sample nuclear data cross-section in ACE format using multi-group nuclear data covariances. The Total Monte-Carlo Method and Gesellschaft für Anlagen-und Reaktorsicherheit method (GRS) are implemented and compared to one another as well as MCNP sensitivity and uncertainty calculations. The methods were applied to the Godiva critical sphere k-eigenvalue, the UAM pin-cell benchmark energy dependent flux and reaction rates, and the Advanced Test Reactor energy dependent flux within an experimental location. Furthermore, the two methods agree well, with GRS allowing for an order of magnitude speedup for reaction rate uncertainty calculations and several orders of magnitude for eigenvalue uncertainty calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4 calculations. In this paper, the neutron-photonic calculation scheme used for the neutron-gamma flux evaluation in the OSIRIS core periphery will be first detailed. Then, a comparison of calculated and measured values of neutron and gamma fluxes will be introduced. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Simulation of SPND Beta-Current in AP1000 core

The Self-Powered Neutron Detectors (SPNDs) are widely used in Generation-III commercial PWR due to its small size and external-power free character. It can measure the in-core neutron flux density immediately to obtain the power profile in the core. To better understand the actual state of the simulated core, a SPND beta-current simulation module has been established in the NECP-Bamboo code. Firstly, neutron simulation is carried out in the lattice code Bamboo-Lattice together with nuclide depletion simulation to obtain the depleted composition of the entire lattice including the SPND, and the few-group parameters including density and cross section of emitter material. Secondly, the depleted lattice can then be piped into a Monte-Carlo code for neutron-photon-electron simulation to obtain the electron escape efficiency from the emitter of the SPND to the corresponding collector. Thirdly, the electron escape efficiency can be parameterized into the function of the lattice state parameters similarly to the common few-group constants. Finally, the beta-current was calculated by the core code Bamboo-Core simultaneously with the simulation of the core operation process. The beta-current of the Vanadium SPND in the AP1000 core was simulated and analyzed in this paper. Comparison between simulation and measurement of the beta-current distribution in the specific core state is shown in this paper to verify the effectiveness of the method and the code. The radial normalized current distribution is in good agreement with the measurement. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Monte Carlo Simulations of the Water Draining Experiment of Giacint Critical Assembly

The MCNP6 computer program has been successfully extended to simulate reactor dynamics problems with moving parts of the geometries. Different from the dynamic method developed in other Monte Carlo codes, a movement scheme has been developed to account for the geometrical parts motion during the particle random walk. The MCNP6 computer program has been used to simulate two transient experiments of the Giacint critical assembly. The MCNP6 calculated total neutron flux was compared with that from the Serpent simulation. An excellent agreement was obtained between the results of the two Monte Caro computer programs. The MCNP6 calculated total neutron flux was also compared with the two measured transients. The MCNP6 results predicted a faster transient than the experimental data. The MCNP6 transient simulation was improved with an adjusted geometrical model which shifts the fuel rods slightly to match the measured reactivity worth due to the drained water. (c) 2021 Elsevier Ltd. All rights reserved.

MCNP, Moving Geometries, Monte Carlo Reactor Trans↗

DOE OSTI2017

Verification of High-Fidelity Neutronics Code PROTEUS for C5G7 Benchmark Problems

Under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the 3D whole core transport code PROTEUS has been developed at Argonne National Laboratory to perform the detailed 3D transport calculation for various reactor types including irregular or complex geometric reactors. PROTEUS includes two distinct transport solvers: SN and MOC. Among them, the MOC solver based on the extruded geometry, thus named MOCEX, can produce the faithful 3-D transport solutions by employing the rigorous and robust 2D/3D MOC formulation. This formulation overcomes the drawbacks of a full 3D MOC method requiring enormous memory and computational resources and resolves the instability issues of the 2D/1D method as well. The solution of MOCEX converges to that of a Monte Carlo code as the discretization of angle, space and energy is refined, unlike the 2D/1D method. As a verification and validation (V and V) effort of PROTEUS-MOCEX, this paper presents verification test results using the well-known 3D C5G7 benchmark problems by solving and comparing eigenvalues and pin powers with MCNP solutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗