scholarly journals Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology

2013 ◽  
Vol 2013 ◽  
pp. 1-9 ◽  
Author(s):  
Mario Matijević ◽  
Dubravko Pevec ◽  
Krešimir Trontl
Keyword(s):  
Cross Sections ◽  
Neutron Transport ◽  
National Laboratory ◽  
General Purpose ◽  
Monte Carlo Code ◽  
Oak Ridge ◽  
Benchmark Data ◽  
Selection Of ◽  
Good Agreement ◽  
Reactor Dosimetry

Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

scholarly journals Collisions of Nucleons with Atoms: Calculated Cross Sections and Monte Carlo Simulation

2021 ◽  
Vol 9 ◽  
Author(s):  
Francesc Salvat ◽  
José Manuel Quesada
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Nuclear Reactions ◽  
Impact Ionization ◽  
Neutron Transport ◽  
Emission Spectra ◽  
Monte Carlo Code ◽  
Dose Distributions ◽  
Elastic Collisions ◽  
Proton Impact

After a summary description of the theory of elastic collisions of nucleons with atoms, we present the calculation of a generic database of differential and integrated cross sections for the simulation of multiple elastic collisions of protons and neutrons with kinetic energies larger than 100 keV. The relativistic plane-wave Born approximation, with binding and Coulomb-deflection corrections, has been used to calculate a database of proton-impact ionization of K-shell and L-, M-, and N-subshells of neutral atoms These databases cover the whole energy range of interest for all the elements in the periodic system, from hydrogen to einsteinium (Z = 1–99); they are provided as part of the penh distribution package. The Monte Carlo code system penh for the simulation of coupled electron-photon-proton transport is extended to account for the effect of the transport of neutrons (released in proton-induced nuclear reactions) in calculations of dose distributions from proton beams. A simplified description of neutron transport, in which neutron-induced nuclear reactions are described as a fractionally absorbing process, is shown to give simulated depth-dose distributions in good agreement with those generated by the Geant4 code. The proton-impact ionization database, combined with the description of atomic relaxation data and electron transport in penelope, allows the simulation of proton-induced x-ray emission spectra from targets with complex geometries.

Experimental Program for Investigating the Influence of Cladding Defects on Burst Pressure

2006 ◽  
Author(s):  
Shengjun (Sean) Yin ◽  
B. Richard Bass ◽  
Wallace J. McAfee ◽  
Paul T. Williams
Keyword(s):  
Finite Element ◽  
National Laboratory ◽  
Maximum Load ◽  
Plastic Instability ◽  
Experimental Program ◽  
Finite Element Models ◽  
Ductile Tearing ◽  
Oak Ridge ◽  
Clad Layer ◽  
Good Agreement

An experimental program was conducted by the Heavy-Section Steel Technology Program at the Oak Ridge National Laboratory (ORNL) to evaluate the structural significance of defects found in the unbacked cladding of the Davis-Besse vessel head. ORNL conducted total 13 clad burst tests with unflawed/flawed specimens. Failure pressure data from those tests indicated a high degree of repeatability for the tests performed in the clad burst program. Unflawed clad burst specimens failed around the full perimeter of the disk from plastic instability; an analytical model for plastic collapse was shown to adequately predict those results. The flawed specimens tested in the program failed by ductile tearing of the notch defect through the clad layer. Analytical interpretations that utilized 3-D finite element models of the clad burst specimens were performed for all tests. Fractographic studies were performed on failed defects in the flawed burst specimens to verify the ductile mode of failure. Comparisons of computed results from 3-D finite element models with measured gage displacement data (i.e., center-point deflection and CMOD) indicated reasonably good agreement up to the region of instability. For tests instrumented with the CMOD gage, good agreement between calculated and measured CMOD data up to the onset of instability implies that ductile tearing initiated near the maximum load and (with a small increase in load) rapidly progressed through the clad layer to produce failure of the specimen.

scholarly journals Processing and application of nuclear data for low temperature criticality assessment

2020 ◽  
Vol 239 ◽  
pp. 14006
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Tara Hanlon ◽  
Richard Hiles ◽  
Malcolm Lingard ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Low Temperature ◽  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Monte Carlo Code ◽  
Temperature Range ◽  
Nuclear Data Library ◽  
Criticality Assessment ◽  
Good Agreement

Until recently, criticality safety assessment codes had a minimum temperature at which calculations can be performed. Where criticality assessment has been required for lower temperatures, indirect methods, including reasoned argument or extrapolation, have been required to assess reactivity changes associated with these temperatures. The ANSWERS Software Service MONK® version 10B Monte Carlo criticality code, is capable of performing criticality calculations at any temperature, within the temperature limits of the underlying nuclear data in the BINGO continuous energy library. The temperature range of the nuclear data has been extended below the traditional lower limit of 293.6 K to 193 K in a prototype BINGO library, primarily based on JEFF-3.1.2 data. The temperature range of the thermal bound scattering data of the key moderator materials was extended by reprocessing the NJOY LEAPR inputs used to produce bound data for JEFF-3.1.2 and ENDF/B-VIII.0. To give confidence in the low temperature nuclear data, a series of MONK and MCBEND calculations have been performed and results compared against external data sources. MCBEND is a Monte Carlo code for shielding and dosimetry and shares commonalities to its sister code MONK including the BINGO nuclear data library. Good agreement has been achieved between calculated and experimental cross sections for ice, k-effective results for low temperature criticality benchmarks and calculated and experimentally determined eigenvalues for thermal neutron diffusion in ice. To quantify the differences between ice and water bound scattering data a number of MONK criticality calculations were performed for nuclear fuel transport flask configurations. The results obtained demonstrate good agreement with extrapolation methods. There is a discernible difference in the use of ice and water data.

Numerical and Experimental Investigation of the Flow in the SNS Jet Flow Target

2015 ◽  
Author(s):  
Charlotte Barbier ◽  
Mark Wendel ◽  
David Felde ◽  
Michael C. Daugherty
Keyword(s):  
Numerical Simulations ◽  
Radiation Transport ◽  
Fluid Dynamic ◽  
National Laboratory ◽  
Jet Flow ◽  
Numerical Approach ◽  
General Purpose ◽  
Temperature Fields ◽  
Beam Power ◽  
Oak Ridge

Computational Fluid Dynamic (CFD) numerical simulations were performed for the flow inside the Spallation Neutron Source jet-flow target vessel at Oak Ridge National Laboratory. Different flow rates and beam conditions were tested to cover all the functioning range of the target, but for brevity, only the nominal case with a mass flow rate of 185 kg/s and a beam power of 1.54MW is presented here. The heat deposition rate from the proton beam was computed using the general-purpose Monte Carlo radiation transport code MCNPX and the commercial CFD code ANSYS-CFX was used to determine the flow velocity in the mercury and the temperature fields in both the mercury and the stainless steel vessel. Boundary conditions, turbulence model and mesh effects are presented in depth. To validate the numerical approach, Particle Imagery Velocimetry (PIV) measurements on a water-loop setup with an acrylic jet-flow target mock-up were performed and compared to the numerical simulations. It was found that a sustained wall jet was developed across the whole length of the vulnerable surface, confirming the good design of the jet-flow target. Overall, good agreements were observed between the experiments and the simulations: the velocity contours and the shape of the recirculation zone near the side baffle are qualitatively similar. However, some differences were also observed that underlines the shortcomings of both the numerical simulations and the experimental measurements.

scholarly journals Burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006

2021 ◽  
Vol 8 (4) ◽  
pp. 10-19
Author(s):  
Tiep Nguyen Huu ◽  
Dung Nguyen Thi ◽  
Phu Tran Viet ◽  
Thanh Tran Vinh ◽  
Ha Pham Nhu Viet
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Wide Temperature Range ◽  
Collision Probability ◽  
Shielding Effect ◽  
Probability Method ◽  
Monte Carlo Code ◽  
Benchmark Data ◽  
Fuel Assemblies ◽  
Burnup Calculation

The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (lowenriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006. The new depletion capability of MCNP6 was applied in the burnup calculation of the VVER-1000 LEU benchmark assembly. The OTF (on-the-fly) methodology of MCNP6, which involves high precision fitting of Doppler broadened cross sections over a wide temperature range, was utilized to handle temperature variation for heavy isotopes. The collision probability method based PIJ module of SRAC2006 was also used in this burnup calculation. The reactivity of the fuel assembly, the isotopic concentrations and the shielding effect due to the presenceof the gadolinium isotopes were determined with burnup using MCNP6 and SRAC2006 incomparison with the available published benchmark data. This study is therefore expected to reveal the capabilities of MCNP6 and SRAC2006 in burnup calculation of VVER-1000 fuel assemblies.

scholarly journals DEVELOPMENT OF MPACT FOR FULL-CORE SIMULATIONS OF MAGNOX GAS-COOLED NUCLEAR REACTORS

2021 ◽  
Vol 247 ◽  
pp. 06041
Author(s):  
Brian J. Ade ◽  
Nicholas P. Luciano ◽  
Andrew J. Conant ◽  
Cole A. Gentry ◽  
Shane G. Stimpson ◽  
...  
Keyword(s):  
Neutron Transport ◽  
National Laboratory ◽  
Test Problems ◽  
High Fidelity ◽  
Validation Data ◽  
Oak Ridge ◽  
Section Data ◽  
Axial Dimension ◽  
And Performance ◽  
Full Core

The MPACT code, jointly developed by Oak Ridge National Laboratory and University of Michigan, is designed to perform high-fidelity light water reactor (LWR) analysis using wholecore pin-resolved neutron transport calculations on modern parallel-computing hardware. MPACT uses the subgroup method for resonance self-shielding, while the primary neutron transport solver uses a 2D/1D method that is based on the method of characteristics (MoC) for the x-y planes coupled with a 1D diffusion or transport solver in the axial dimension. Additional geometry capabilities are currently being developed in MPACT to support hexagonal-pitched lattices, as well as interstitial geometry (i.e., control rods at the corner of four adjacent pin cells). In this research, the MPACT method is tested on gas-cooled reactors by applying MPACT to full-core MAGNOX reactor test problems. MAGNOX test problems were chosen due to the availability of high-quality reactor design and validation data (available through an ongoing collaboration with the National Nuclear Laboratory in the United Kingdom) and the existence of a relatively complex axial power shape that is expected to challenge the MPACT method. MPACT’s convergence for partial- and full-core problems will be tested and verified. MPACT will be compared with high-fidelity continuous-energy Monte Carlo simulations to verify core reactivity, power distributions, and performance of the available cross section data libraries and energy group structures.

A Time-Dependent Solver for Coupled Neutron-Transport Thermal-Mechanics Calculations and Simulation of a Godiva Prompt-Critical Burst

2014 ◽  
Author(s):  
Carlo Fiorina ◽  
Manuele Aufiero ◽  
Sandro Pelloni ◽  
Konstantin Mikityuk
Keyword(s):  
Neutron Transport ◽  
National Laboratory ◽  
Time Dependent ◽  
Discrete Ordinates ◽  
Los Alamos National Laboratory ◽  
Preliminary Validation ◽  
Tightly Coupled ◽  
Enriched Uranium ◽  
Paul Scherrer ◽  
Good Agreement

The present paper describes a first step taken at the Paul Scherrer Institut in the development of a new multi-physics platform for reactor analysis. Such platform is based on the finite-volume software OpenFOAM and aims at a tightly coupled description of neutron transport, thermal mechanics and fluid dynamics. For this purpose, a steady-state 3-D discrete ordinates/thermal-mechanics solver was first developed in collaboration with the Politecnico di Milano. The present work briefly discusses such solver and its preliminary validation, which will be described in detail in parallel publications. It then focuses on its extension to time-dependent simulations. The solver is first tested by simulating different step-wise reactivity insertions in a critical configuration constituted by an infinite slab of highly enriched uranium. Subsequently, a super-prompt-critical power burst in the Godiva reactor has been simulated. Godiva was a spherical assembly of highly enriched uranium built and operated at the Los Alamos National Laboratory (US) during the Fifties. A prompt-critical transient in such system configures as a quick power excursion (up to ∼10 GW), which causes a temperature rise, and a subsequent reactivity reduction via expansion of the sphere. The overall transient lasts for few fractions of a millisecond. Results obtained with the newly developed model have been compared to experimental results, showing a relatively good agreement.

scholarly journals New search for mirror neutron regeneration

2019 ◽  
Vol 219 ◽  
pp. 07002
Author(s):  
L.J. Broussard ◽  
K.M. Bailey ◽  
W.B. Bailey ◽  
J.L. Barrow ◽  
K. Berry ◽  
...  
Keyword(s):  
Magnetic Field ◽  
Small Angle ◽  
Small Angle Neutron Scattering ◽  
National Laboratory ◽  
General Purpose ◽  
High Flux ◽  
Neutron Lifetime ◽  
Neutron State ◽  
Oak Ridge ◽  
Cold Neutrons

The possibility of relatively fast neutron oscillations into a mirror neutron state is not excluded experimentally when a mirror magnetic field is considered. Direct searches for the disappearance of neutrons into mirror neutrons in a controlled magnetic field have previously been performed using ultracold neutrons, with some anomalous results reported. We describe a technique using cold neutrons to perform a disappearance and regeneration search, which would allow us to unambiguously identify a possible oscillation signal. An experiment using the existing General Purpose-Small Angle Neutron Scattering instrument at the High Flux Isotope Reactor at Oak Ridge National Laboratory will have the sensitivity to fully explore the parameter space of prior ultracold neutron searches and confirm or refute previous claims of observation. This instrument can also conclusively test the validity of recently suggested oscillation-based explanations for the neutron lifetime anomaly.

A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification

2018 ◽  
Author(s):  
Wankui Yang ◽  
Baoxin Yuan ◽  
Songbao Zhang ◽  
Haibing Guo ◽  
Yaoguang Liu ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Research Reactor ◽  
Neutron Transport ◽  
Reactor Core ◽  
Deep Penetration ◽  
Monte Carlo Code ◽  
Deterministic Theory ◽  
Continuous Energy ◽  
Transport Calculation ◽  
Good Agreement

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing keff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But keff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the keff difference between these two methods remains. Even though the keff results in this benchmark are less than keff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

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