Application of genetic algorithm for optimization of control rods positioning in a fast breeder reactor core

Abstract The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.

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Optimization of Primary Loop Pump Power of a Loop-type Liquid Metal Fast Breeder Reactor with Annular Fuel Using Genetic Algorithm

Advances in Intelligent Systems and Computing - Proceedings of Fourth International Conference on Soft Computing for Problem Solving ◽

10.1007/978-81-322-2217-0_29 ◽

2014 ◽

pp. 339-352

Author(s):

Shubhajit Karmakar ◽

Swarnendu Sen ◽

Sanjib Kumar Acharyya

Keyword(s):

Genetic Algorithm ◽

Pump Power ◽

Liquid Metal ◽

Fast Breeder Reactor ◽

Primary Loop

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Reactor Physics Parameters of Alternate Fueled Fast Breeder Reactor Core Designs

Nuclear Technology ◽

10.13182/nt79-a32142 ◽

1979 ◽

Vol 42 (2) ◽

pp. 123-132

Author(s):

D. R. Haffner ◽

R. W. Hardie

Keyword(s):

Reactor Core ◽

Fast Breeder Reactor ◽

Reactor Physics

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Possibility of Different Fuel Cycles Usage in GT-MHR

10th International Conference on Nuclear Engineering, Volume 4 ◽

10.1115/icone10-22404 ◽

2002 ◽

Cited By ~ 1

Author(s):

N. Kodochigov ◽

Yu. Sukharev ◽

E. Marova ◽

N. Ponomarev-Stepnoy ◽

E. Glushkov ◽

...

Keyword(s):

Fuel Cycle ◽

Block Design ◽

Average Power ◽

Reactor Core ◽

Fuel Burnup ◽

Fuel Load ◽

Fuel Cycles ◽

Partial Cycle ◽

Control Rods ◽

Neutronic Characteristics

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

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Development of Safety-Enhanced Fast Reactor by Using Minor Actinide Bearing Internal Blanket

Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors ◽

10.1115/icone2020-16746 ◽

2020 ◽

Author(s):

Sho Fuchita ◽

Satoshi Takeda ◽

Koji Fujimura ◽

Toshikazu Takeda ◽

Kazuhiro Fujimata

Keyword(s):

Fast Reactor ◽

Inner Core ◽

Reactor Core ◽

Minor Actinide ◽

Core Damage ◽

Burnable Absorber ◽

Sensitivity Studies ◽

Core Height ◽

Gas Expansion ◽

Control Rods

Abstract For a 750MWe sodium-cooled fast reactor core using MOX fuel, safety-enhancement measures have been studied to reduce the risk of core damage under unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents. As passive measures the followings are considered: 1) adoption of the axial heterogeneous core configuration with sodium plenum and Gas Expansion Modules (GEMs) to lower sodium void reactivity for ULOF, and 2) addition of minor actinides (MAs) as burnable absorber and fertile nuclides to the internal blanket in the inner core to reduce burnup reactivity for UTOP. In this study, configurations of the safety-enhanced core were optimized based on sensitivity studies as follows. Firstly, effects of 1) above on the sodium void reactivity were evaluated by changing the inner core height, B-10 content of the upper shield, GEMs, and standby position of the backup control rods, which are the dominant factors of core behavior in the event of ULOF. Secondly, the effects of 2) above on the burnup reactivity were evaluated by changing the MA content in the internal blanket and the burnup period, which are the dominant factors of UTOP. Finally, by utilizing sensitivity analysis results, the safety-enhanced core which satisfies the provisional design goals has been developed. This core has negative sodium void reactivity and burnup reactivity less than 1 $.

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