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

Jang, Jaerim; Lemaire, Matthieu; Jeong, Sanggeol; Jeong, Eun; Jo, Yongmin

doi:10.1016/j.anucene.2017.12.054

Cited by 34 publications

(6 citation statements)

References 10 publications

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“…Multivariate design optimization of a reactor core is an iterative process that can require hundreds to thousands of simulations. Two radiation transport codes were used for the MicroURANUS core design: the deterministic Argonne Reactor Computation (ARC) code system 19‐22 and the Monte Carlo (MC) code MCS 23‐30 . Short execution time of the deterministic codes enables many scoping simulations to be performed at the heterogeneous unit‐cell level for fuel pin and lattice geometry optimization.…”

Section: Computer Codesmentioning

confidence: 99%

MicroURANUS : Core design for long‐cycle lead‐bismuth‐cooled fast reactor for marine applications

et al. 2021

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Summary A core design of MicroURANUS, a long‐cycle lead‐bismuth‐cooled fast nuclear reactor for marine applications, is presented. It aims to generate a power of 60MWth, which can be regulated during operation. MicroURANUS was designed to achieve a small burnup reactivity swing for 30 effective full‐power years of a lifetime without refueling. To attain these goals, a unit cell study with uranium oxide fuel was initiated to lay a solid foundation for core design, owing to size constraints. A reflector optimization was also performed to minimize neutron leakage. MicroURANUS adopts onion zoning constructed by two enrichment zones for flattening the power and lengthening the core lifetime. The coolant is driven by electromagnetic pumps to achieve inlet and outlet temperatures of 250°C and 350°C, respectively. MicroURANUS analyses were performed using the Argonne Reactor Computation suite and the Monte Carlo code MCS. Core performance features were analyzed for criticality, power profiles, fuel isotope mass inventory, reflector coefficients, reactivity feedback coefficients, and shutdown margin. Additional thermal‐hydraulic calculations were performed to confirm that the fuel and cladding temperatures were within the acceptable range. Furthermore, a load follow analysis using the quasi‐static reactivity balance method confirmed the feasibility of regulating power by adjusting the inlet coolant temperature.

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Section: Computer Codesmentioning

confidence: 99%

MicroURANUS : Core design for long‐cycle lead‐bismuth‐cooled fast reactor for marine applications

et al. 2021

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“…There are lot of research works dedicated to the problems of estimation radioactive emission and its influence onto humans and environment [2,3], calculation of subcriticality and prevention of the uncontrolled chain reaction [4], development additional radiation protection constructions etc. However, the thermal part of the complex concept of DSFSNF safety has not enough scientific attention.…”

Section: Fig 2 Resistance Thermometer and The Place Of Its Location In The System Of Thermal Monitoringmentioning

confidence: 99%

Scientific Basis of Thermal Safety Analysis of Dry Storage of Spent Nuclear Fuel on Zaporizhska NPP

Alyokhina¹,

Kostikov²,

Koriahina³

2020

Problems of Atomic Science and Technology

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Now only one Dry Storage Facility of Spent Nuclear Fuel (DSFSNF) is operated in Ukraine. It is the facility on Zaporizhska NPP. Many different thermal investigations were done for ventilated containers of DSFSNF. In this study the generalization of scientific approaches to the thermal safety assessment are carried out. The multi-stage approach to the definition of thermal state of containers' group, single container, spent fuel assemblies and fuel rods was developed. Detailed thermal profiles of spent fuel assemblies inside storage container were calculated. With usage of multi-stage approach the thermal simulations of the influence of outer factors onto thermal state of containers was carried out. Results of thermal investigations were generalized and factors, which are influence on thermal state of containers, are detected. The method of spent nuclear fuel thermal state prediction and suggestion for improving the system of thermal monitoring were proposed.

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“…In its current version, three kinds of calculations are allowed by MCS: neutron criticality runs for reactivity calculations, neutron fixed‐source runs for shielding problems, and photon fixed‐source runs for shielding problems. The validation of MCS neutron transport capability is conducted with ~300 benchmarks of the International Criticality Safety Benchmark Experimental Problem (ICBEP) database, the BEAVERS benchmark, and the VENUS‐2 and Hoogenboom benchmarks . MCS calculations are verified against reference transport codes for the prismatic very high temperature reactor (VHTR), the Advanced Power Reactor 1400 MW electricity (APR‐1400), the Jordan Research and Training Reactor (JRTR), and the boron‐free small modular pressurized water reactor (SMPWR) .…”

Section: Computer Codesmentioning

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

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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.

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

Cited by 34 publications

References 10 publications

MicroURANUS : Core design for long‐cycle lead‐bismuth‐cooled fast reactor for marine applications

MicroURANUS : Core design for long‐cycle lead‐bismuth‐cooled fast reactor for marine applications

Scientific Basis of Thermal Safety Analysis of Dry Storage of Spent Nuclear Fuel on Zaporizhska NPP

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

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