scholarly journals The Use of the MCNP Code for Radiation Damage Calculations

2021 ◽  
pp. 70-77
Author(s):  
Mohammad Hiwa ◽  
Keyword(s):  
Monte Carlo ◽  
Radiation Damage ◽  
Cross Section ◽  
Neutron Energy ◽  
Radiation Transport ◽  
Mcnp Code ◽  
Transport Calculation ◽  
Energy Irradiation ◽  
Basic Concepts ◽  
The Impact

This work gives a detailed analysis of the result of Monte Carlo physics practical using MCNP. This paper describes basic concepts of the Monte Carlo theory of radiation transport calculation and also discusses the variance and the history method as used in Monte Carlo Problem solving. Therefore, in this exercise the MCNP code has been used to solve and estimate the number of neutron flux. The paper investigated the impact of the primary radiation damage in iron by the neutron energy irradiation. The established measurement of radiation damage is the displacements per atom (dpa) in matter as a function of neutron energy. The simulations were carried out to calculate the dpa cross section.

THE IMPACT OF LO, NLO AND NNLO FOR THE HIGGS SEARCHING AT $\sqrt{s} = 7$ TeV OF LHC

2010 ◽  
Vol 25 (36) ◽  
pp. 3027-3031
Author(s):  
JIAN WANG ◽  
GUOMING CHEN ◽  
WEIMIN WU
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Higgs Mass ◽  
Monte Carlo Generator ◽  
Final States ◽  
Leading Order ◽  
Analysis Process ◽  
Signal Rate ◽  
Monte Carlo Studies ◽  
The Impact

Most of current Monte Carlo studies on the Higgs searching are based on LO, or NLO calculation. However, in recent years, the next-to-next-to-leading order (NNLO) corrections have been computed for some physics process, and found that the cross section increases the kinematics changes. As the results, the analysis results could be impacted by these high order QCD corrections. We use standard Monte Carlo generator for LO, as well as MC@NLO for NLO and ResBos for NNLO at 7 TeV of LHC to evaluate this impact for physics channel of the Higgs, mass at 165 GeV, to WW, then W decay to lepton and neutrino as the final states. We found the signal rate could be effected by ratio of 1:2.6:3.4 for LO, NLO and NNLO using the same standard H→WW→lνlν searching analysis process.6

Mathematical verification of the Monte Carlo maximum cross-section technique

2015 ◽  
Vol 21 (4) ◽  
Author(s):  
Victor S. Antyufeev
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Cross Section ◽  
Path Length ◽  
Radiation Transport ◽  
Free Path Length ◽  
Maximum Cross Section ◽  
The Monte Carlo Method ◽  
Section Technique ◽  
Probabilistic Proof

AbstractMaximum cross-section technique is used for solving problems of radiation transport by the Monte Carlo method to optimize the particles' free-path length modeling in inhomogeneous media. A probabilistic proof of a variation of this technique is proposed in the article.

scholarly journals On the W&Y interpretation of high-energy Drell-Yan measurements

2021 ◽  
Vol 2021 (2) ◽  
Author(s):  
Riccardo Torre ◽  
Lorenzo Ricci ◽  
Andrea Wulzer
Keyword(s):  
Monte Carlo ◽  
Differential Cross Section ◽  
Monte Carlo Simulations ◽  
Cross Section ◽  
Differential Cross ◽  
High Energy ◽  
New Physics ◽  
Leading Order ◽  
Theoretical Predictions ◽  
The Impact

Abstract High-energy neutral and charged Drell-Yan differential cross-section measurements are powerful probes of quark-lepton contact interactions that produce growing-with-energy effects. This paper provides theoretical predictions of the new physics effects at the Next-to-Leading order in QCD and including one-loop EW corrections at the single-logarithm accuracy. The predictions are obtained from SM Monte Carlo simulations through analytic reweighting. This eliminates the need of performing a scan on the new physics parameter space, enabling the global exploration of all the relevant interactions. Furthermore, our strategy produces consistently showered events to be employed for a direct comparison of the new physics predictions with the data, or to validate the unfolding procedure than underlies the cross-section measurements. Two particularly relevant interactions, associated with the W and Y parameters of EW precision tests, are selected for illustration. Projections are presented for the sensitivity of the LHC and of the HL-LHC measurements. The impact on the sensitivity of several sources of uncertainties is quantified.

scholarly journals Correlated energy uncertainties in reaction rate calculations

2020 ◽  
Vol 642 ◽  
pp. A41
Author(s):  
Richard Longland ◽  
Nicolas de Séréville
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Reaction Cross Section ◽  
Reaction Rate ◽  
Resonance Energy ◽  
Reaction Rates ◽  
Reaction Cross ◽  
Current Interest ◽  
Resonance Energies ◽  
The Impact

Context. Monte Carlo methods can be used to evaluate the uncertainty of a reaction rate that arises from many uncertain nuclear inputs. However, until now no attempt has been made to find the effect of correlated energy uncertainties in input resonance parameters. Aims. Our goal is to investigate the impact of correlated resonance energy uncertainties on reaction rates. Methods. Using a combination of numerical and Monte Carlo variation of resonance energies, the effect of correlations are investigated. Five reactions are considered: two fictional, illustrative cases and three reactions whose rates are of current interest. Results. The effect of correlations in resonance energies depends on the specific reaction cross section and temperatures considered. When several resonances contribute equally to a reaction rate, and when they are located on either side of the Gamow peak, correlations between their energies dilute their effect on reaction rate uncertainties. If they are both located above or below the maximum of the Gamow peak, however, correlations between their resonance energies can increase the reaction rate uncertainties. This effect can be hard to predict for complex reactions with wide and narrow resonances contributing to the reaction rate.

scholarly journals Uncertainty propagation based on correlated sampling technique for nuclear data applications

2020 ◽  
Vol 6 ◽  
pp. 8 ◽  
Author(s):  
Axel Laureau ◽  
Vincent Lamirand ◽  
Dimitri Rochman ◽  
Andreas Pautz
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Neutron Transport ◽  
Uncertainty Propagation ◽  
Computation Time ◽  
Sampling Technique ◽  
Nuclear Data ◽  
Correlated Sampling ◽  
Enriched Uranium ◽  
The Impact

A correlated sampling technique has been implemented to estimate the impact of cross section modifications on the neutron transport and in Monte Carlo simulations in one single calculation. This implementation has been coupled to a Total Monte Carlo approach which consists in propagating nuclear data uncertainties with random cross section files. The TMC-CS (Total Monte Carlo with Correlated Sampling) approach offers an interesting speed-up of the associated computation time. This methodology is detailed in this paper, together with two application cases to validate and illustrate the gain provided by this technique: the highly enriched uranium/iron metal core reflected by a stainless-steel reflector HMI-001 benchmark, and the PETALE experimental programme in the CROCUS zero-power light water reactor.

scholarly journals Using of Stopping and Range of Ions in Matter Code to Study of Radiation Damage in Materials

2020 ◽  
Vol 12 (4) ◽  
pp. 451-456
Author(s):  
Hiwa Mohammad Qadr ◽  
◽  
Ari Maghdid Hamad ◽  
Keyword(s):  
Radiation Damage ◽  
Proton Energy ◽  
Damage Parameter ◽  
Irradiation Damage ◽  
Ion Distribution ◽  
Atomic Displacements ◽  
Energy Irradiation ◽  
Definition Of ◽  
Displacement Per Atom ◽  
The Impact

The aim of this work to investigate the impact of the radiation damage in the materials by the proton energy irradiation. The damage parameter used in the evaluation is displacement per atom (DPA) in material as a function of proton energy. Stopping and Range of Ions in Matter (SRIM) code was used to calculate the total vacancy and the number of atomic displacements based on the Norgett-Robinson-Torrens model by difference energies for proton irradiation damage. The option of this code was calculated by using Ion Distribution and Quick Calculation of Damage (Kinchin-Pease) for Fe and Cu target and also Full damage cascade (F-C) was chosen for only Fe. The result is that, the prediction of the F-C model are higher than the K-P calculation. Comparisons has been made with an international standard definition of DPA.

scholarly journals EFFECT OF FISSION SOURCE SPECTRUM ON MONTE CARLO CALCULATION OF EX-CORE QUANTITIES

2021 ◽  
Vol 247 ◽  
pp. 02027
Author(s):  
Eva E. Davidson ◽  
Tara M. Pandya ◽  
Katherine E. Royston ◽  
Thomas M. Evans ◽  
Andrew T. Godfrey ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Neutron Source ◽  
Radiation Transport ◽  
Source Spectrum ◽  
Detector Response ◽  
Monte Carlo Code ◽  
Sampling Schemes ◽  
Transport Calculation ◽  
Water Reactors ◽  
Quantities Of Interest

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities to combine highfidelity in-core radiation transport with temperature feedback using MPACT and CTF with a follow-on fixed source transport calculation using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides a fission source to Shift for the follow-on radiation transport calculation. In past VERA releases, MPACT passed a spatially dependent source without the energy distribution to Shift. Shift then assumed a235U Watt spectrum to sample the neutron source energies. There were concerns that, in cases with burned or mixed oxide (MOX) fuel near the periphery of the core, the assumption of a235U Watt spectrum for the source neutron energies would not be accurate for studying ex-core quantities of interest, such as pressure vessel fluence or detector response. Therefore, two additional options were implemented in VERA for Shift to sample neutron source energies: (1) a nuclide-dependent Watt spectra for235U,238U,239Pu, and241Pu, and (2) to use the standard 51-energy group MPACT spectrum. Results show that the 51-group MPACT spectrum is not suitable for ex-core calculations because the groups have been fine-tuned for in-core calculations. Differences in relative detector response due to235U and nuclide-dependent Watt spectra sampling schemes were negligible; however, the use of nuclide-dependent Watt spectra for vessel fluence calculations was found to be important for fuel cycles with burned and fresh fuel.

Sign in / Sign up

Export Citation Format

Share Document