MC<sup>2</sup>-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

Lee, Chang-Ho; Yang, Won Sik

doi:10.1080/00295639.2017.1320893

Cited by 96 publications

(62 citation statements)

References 15 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Burnup region-dependent 33-group cross section sets were generated from the ENDF/B-VII.0 data library using the ETOE-2/MC2-3/TWODANT sequence of codes of the ARC package. [26] [27] Specifically, the MC2-3 code first calculates 0-dimensional (0-D) neutron spectra and condenses point-wise cross sections prepared by ETOE-2 into an ultrafine group structure (~2000 groups). For non-multiplying regions such as reflectors and shield, leaking spectra from the fueled region is used for the condensation.…”

Section: Neutronics Modelmentioning

confidence: 99%

3D in-core fuel management optimization for breed-and-burn reactors

Hou

Qvist

Kellogg

et al. 2016

Progress in Nuclear Energy

View full text Add to dashboard Cite

Section: Neutronics Modelmentioning

confidence: 99%

3D in-core fuel management optimization for breed-and-burn reactors

Hou

Qvist

Kellogg

et al. 2016

Progress in Nuclear Energy

View full text Add to dashboard Cite

“…An infinite-homogeneous composition of three isotopes is considered with a four-group energy structure. Cross sections are obtained using the fast reactor cross section generator, MC 2 -3 [5]. Microscopic cross sections and isotope densities are provided in Section A for the convenience of replicating results.…”

Section: Resultsmentioning

confidence: 99%

An Analytic Benchmark for the Solution to the Isotopic Fission Spectrum Mixture Problem

Dawn

2020

View full text Add to dashboard Cite

In neutron transport calculations, it is common to specify material compositions in terms of constituent isotopes. Material compositions may be described by isotope number densities and associated microscopic cross sections. For general reaction cross sections, the macroscopic cross sections of a composition are simply the summation of the sum of the products of isotope number densities and microscopic cross sections. A notable exception is the mixture of the neutron fission spectrum in fissile material. To demonstrate proper and improper mixture of the neutron fission spectrum, a zero-dimensional test problem is developed. Using the test problem, it is demonstrated that the improper mixing of the fission spectrum results in an eigenvalue error of approximately 65 pcm. Though this may seem small, any error in such a simple calculation is unacceptable. In similar infinite-homogeneous test problems, eigenvalue errors of more than 300 pcm have been observed. It is also shown that the error due to fission spectrum mixing can be exaggerated for computer program verification purposes. It is concluded that the proper mixture of the neutron fission spectrum is essential for accurate neutron transport simulations. The effect of improper neutron fission spectrum mixture is demonstrated and quantified and this test problem may be used for verification purposes in the future.

show abstract

“…A suite code package for fast reactor analysis, called Argonne Reactor Computation (ARC) 20 and developed by ANL, was used in this work. The package consists of three main modules: (a) Multigroup Cross-section generation Code (MC 2 -3) which prepares problem-dependent ultrafine group cross sections, 21 (b) TWODANT which generates ultrafine group flux from the Boltzmann transport equation solution using the discrete ordinate method, 22 and (c) REBUS-3 which performs nodal diffusion and depletion calculations for fast reactor fuel analyses. 23 In the first step of a fast reactor analysis, TWODANT generates the ultrafine group regionwise flux spectra using the ultrafine group cross sections (XS) from MC 2 -3.…”

Section: Fast Reactor Analysis Code System Arcmentioning

confidence: 99%

“…Therefore, the SMLFR core can achieve a small reactivity swing (around 1000 pcm) and a long lifetime. The design and the analysis of this core are performed with the ARC fast reactor analysis code suite MC 2 ‐3/TWODANT/REBUS‐3 developed by Argonne National Laboratory (ANL) and the inhouse MCS Monte Carlo (MC) code developed at Ulsan National Institute of Science and Technology (UNIST) . The ARC code system is used for optimization analysis due to its calculation speed (ARC calculations are very fast and not time‐consuming), and the Monte Carlo code MCS is used to verify the results of the ARC code system for the final core candidate.…”

Section: Introductionmentioning

confidence: 99%

Core design of long‐cycle small modular lead‐cooled fast reactor

et al. 2018

View full text Add to dashboard Cite

Summary A core design of small modular liquid‐metal fast reactor (SMLFR) cooled by lead‐bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off‐grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low‐enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S‐CO2) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the core with low‐enriched uranium (LEU) performs the function of core ignition. The center region plays the role of a breeding blanket to increase the core lifetime for long cycle operation. The core working fluid inlet and outlet temperatures are 300°C and 422°C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi‐static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal‐hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the conceptual core is favorable in terms of self‐controllability, which is the first step towards inherent safety.

show abstract

MC²-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

Cited by 96 publications

References 15 publications

3D in-core fuel management optimization for breed-and-burn reactors

3D in-core fuel management optimization for breed-and-burn reactors

An Analytic Benchmark for the Solution to the Isotopic Fission Spectrum Mixture Problem

Core design of long‐cycle small modular lead‐cooled fast reactor

Contact Info

Product

Resources

About

MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

Cited by 96 publications

References 15 publications

3D in-core fuel management optimization for breed-and-burn reactors

3D in-core fuel management optimization for breed-and-burn reactors

An Analytic Benchmark for the Solution to the Isotopic Fission Spectrum Mixture Problem

Core design of long‐cycle small modular lead‐cooled fast reactor

Contact Info

Product

Resources

About

MC²-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis