Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment

Zheng, Meiyin; Tian, Wenxi; Wei, Hongyang; Zhang, Dalin; Wu, Yingwei; Qiu, Suizheng; Su, G.H.

doi:10.1016/j.anucene.2013.08.020

Cited by 33 publications

(14 citation statements)

References 6 publications

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“…A “modified predictor corrector” approach is used in MCORE to increase the calculation precision and adjust the time steps for the two codes, which is a combination of the traditional “predictor approach” and “middle time approach.” The validity of MCORE was verified by comparing the calculation results with benchmark data and published results for fast reactors. The MCORE results agreed well with the benchmark data …”

Section: Calculation Tools and Modelssupporting

confidence: 77%

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

et al. 2019

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Summary Lead‐based fast reactors (LFRs) have unique advantages in the development of a SMR, which has attracted a lot of attention in recent years. In this paper, an optimized design for a lead‐bismuth small modular reactor was studied on the basis of the design of SUPERSTAR. This paper aims to propose an improved LFR core scheme to enhance the neutronic performance as well as the thermal‐hydraulic safety of the reference reactor. Advanced nitride fuel is adopted in which the plutonium is used as the driven fuel, while thorium is used as the fertile fuel. Subchannel analysis was performed in the assembly design using an in‐house subchannel code, SUBAS, and an 11 × 11 scheme with a pitch‐to‐diameter (P/D) ratio of 1.4 was chosen. Using the modified assembly, the core was redesigned using the coupled code MCORE. The active core was divided into four zones with different enrichment of 239Pu to extend the core lifetime and flatten the power distribution. The main kinetic parameters and reactivity coefficients were obtained. Neutronic performance at different operation times was also studied. The maximum radial power peak factor was 1.28, while the maximum total power peak factor was 1.737. During the whole lifetime, the reactivity swing was 0.926$, which was below the limit of 1$. The subchannel study of the core flow distribution showed that a flow distributor is needed to further improve the flow distribution capability. The peaking cladding temperature was 508.7°C, and the maximum fuel center temperature was 723.4°C, both of which do not exceed the limit temperature. Compared with features of SUPERSTAR, the peaking cladding temperature was well improved and the lifetime extended.

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Section: Calculation Tools and Modelssupporting

confidence: 77%

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

et al. 2019

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“…According to preliminary thermal analysis, the fuel temperature is assumed 2,000 K, temperature of the heat pipes and matrix is 1,200 K, and temperature of the control drums and reflector is 900 K for neutronics analysis. A code MCORE (Zheng et al, 2014) coupling MCNP and depletion code ORIGEN is used to analyze the core lifetime. Only 1/6 of the reactor core is modeled due to the symmetry.…”

Section: Neutronics Calculation Calculation Methods and Modelmentioning

confidence: 99%

Reactor Core Design and Analysis for a Micronuclear Power Source

et al. 2018

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Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

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“…MCORE, which couples MCNP and ORIGEN, is used to perform neutronics and depletion calculations in the present paper (Zheng et al, 2014c). MCORE has been verified by VVER-1000 LEU assembly benchmark and OECD/NEA fast reactor benchmark.…”

Section: Assembly Simulationmentioning

confidence: 99%

Performance of radial fuel shuffling sodium cooled Breed and Burn reactor core

Zheng

Tian

Zhang

et al. 2016

Annals of Nuclear Energy

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a b s t r a c tIn this paper, performance of radial fuel shuffling of sodium cooled Breed and Burn reactor core is investigated. Neutronics and depletion calculations are carried out by MCORE based on the ENDF/B-VII data library. Thermal-hydraulic analysis is performed based on a self-developed parallel channels model steady-state code SAST. First of all, the assembly parameters are optimized from neutronics and thermal-hydraulics. The results show that assembly with 127 fuel rods and P/D of 1.12 is the best design. Secondly, the core critical features and radial power peaking factors of the inward and outward fuel shuffling strategy under different fuel shuffling periods are carried out. The results show that k eff of the beginning of equilibrium cycle (BOEC) and end of equilibrium cycle (EOEC) of both the inward and outward fuel shuffling strategy parabolically vary with the fuel shuffling period. Power peaking factor of the inward fuel shuffling strategy decreases with the increase of fuel shuffling period exponentially, while that of the outward fuel shuffling strategy increases with the fuel shuffling period exponentially. The outward fuel shuffling strategy with shuffling period of 500 days performs better in core critical features and radial power peaking factors. Reactivity coefficients of the optimized core are calculated. Finally, preliminary thermal-hydraulic analysis of the optimized core is performed. The results show that maximum cladding interface temperature and maximum fuel temperature are all within the acceptable limits.

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Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment

Cited by 33 publications

References 6 publications

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

Reactor Core Design and Analysis for a Micronuclear Power Source

Performance of radial fuel shuffling sodium cooled Breed and Burn reactor core

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