Base de neutronique : coefficients de réactivité dans les réacteurs nucléaires

Reuss, Paul

doi:10.1051/anphys:200002003

Cited by 5 publications

(28 citation statements)

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“…These calculations are performed assuming a standard 17 Â 17 fuel assembly, with a fissile height of 2.15 m. The average quadratic discrepancy between the interpolations and the calculation is 2%. The model also optimizes the core enrichment in order to adapt the cycle length to the number of refuelling shutdowns required, and 8 Using the well-known approximation Bu n ð Þ ¼ 2n nþ1 Bu 1 ð Þ, where n is the refuelling strategy (1 for single batch, 2 for two batch), and Bu the burn-up [22]. J.…”

Section: Methodsmentioning

confidence: 99%

“…For Flexblue ® , the limited fissile height (2.15 m) is very favourable in terms of stability. In order to analyse the risk of xenon axial instability, a simplified conservative analytical model based on reference [11] has been used. The model estimates the maximum fissile height for which xenon oscillations are stable, for an axially uniform power profile (conservative hypothesis for stability), as a function of linear power (assuming a standard 17 Â 17 fuel assembly) and enrichment.…”

Section: Xenon Stability and Load Followingmentioning

confidence: 99%

“…In this kind of configuration called heterogeneous distribution, Gd 11 Cycle length, which means a final burn-up of 42 GWd/t UO2 . Fig.…”

Section: Fuel Optimisation For Full Refuelling Managementmentioning

confidence: 99%

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Flexblue^®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Ingremeau¹,

Cordiez²

2015

EPJ Nuclear Sci. Technol.

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Abstract. Flexblue ® is a 160 MWe, transportable and subsea-based nuclear power unit, operating up to 100 m depth, several kilometers away from the shore. If being underwater has significant safety advantages, especially using passive safety systems, it leads to two main challenges for core design. The first one is to control reactivity in operation without soluble boron because of its prohibitive drawbacks for a submerged reactor (system size, maintenance, effluents, and safety considerations). The second one is to achieve a long cycle in order to maximise the availability of the reactor, because Flexblue ® refuelling and maintenance will be performed in a shared support facility away from the production site. In this paper, these two topics are dealt with, from a neutronic point of view. Firstly, an overview of the main challenges of operating without soluble boron is proposed (cold shutdown, reactivity swing during cycle, load following, xenon stability). Secondly, an economic optimisation of the Flexblue ® core size and cycle length is performed, using the QUABOX/CUBBOX code. Thirdly, the fuel enrichment and poisoning using gadolinium oxide are optimized for full core or half core refuelling, with the DRAGON code. For the specific case of the full core refuelling, an innovative heterogeneous configuration of gadolinium is used. This specific configuration is computed using a properly adapted state-of-the-art calculation scheme within the above-mentioned lattice code. The results in this specific configuration allow a reactivity curve very close to the core leakage one during the whole cycle.

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Section: Methodsmentioning

confidence: 99%

Section: Xenon Stability and Load Followingmentioning

confidence: 99%

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Flexblue^®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Ingremeau¹,

Cordiez²

2015

EPJ Nuclear Sci. Technol.

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“…Let H denote the Boltzmann operator, r the reactivity, f and f + respectively the flux and adjoint flux of a system, DH the perturbation operator, P the production operator, and Dr the reactivity variation due to DH. Application of first order perturbation theory on the system (fuel) for reactivity [11] gives:…”

Section: Equivalent Plutonium Contentmentioning

confidence: 99%

Artificial neural network surrogate development of equivalence models for nuclear data uncertainty propagation in scenario studies

Krivtchik

Blaise

Coquelet-Pascal

2017

EPJ Nuclear Sci. Technol.

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Abstract. Scenario studies simulate the whole fuel cycle over a period of time, from extraction of natural resources to geological storage. Through the comparison of different reactor fleet evolutions and fuel management options, they constitute a decision-making support. Consequently uncertainty propagation studies, which are necessary to assess the robustness of the studies, are strategic. Among numerous types of physical model in scenario computation that generate uncertainty, the equivalence models, built for calculating fresh fuel enrichment (for instance plutonium content in PWR MOX) so as to be representative of nominal fuel behavior, are very important. The equivalence condition is generally formulated in terms of end-of-cycle mean core reactivity. As this results from a physical computation, it is therefore associated with an uncertainty. A state-of-the-art of equivalence models is exposed and discussed. It is shown that the existing equivalent models implemented in scenario codes, such as COSI6, are not suited to uncertainty propagation computation, for the following reasons: (i) existing analytical models neglect irradiation, which has a strong impact on the result and its uncertainty; (ii) current black-box models are not suited to cross-section perturbations management; and (iii) models based on transport and depletion codes are too time-consuming for stochastic uncertainty propagation. A new type of equivalence model based on Artificial Neural Networks (ANN) has been developed, constructed with data calculated with neutron transport and depletion codes. The model inputs are the fresh fuel isotopy, the irradiation parameters (burnup, core fractionation, etc.), cross-sections perturbations and the equivalence criterion (for instance the core target reactivity in pcm at the end of the irradiation cycle). The model output is the fresh fuel content such that target reactivity is reached at the end of the irradiation cycle. Those models are built and then tested on databases calculated with APOLLO2 (for thermal spectra) and ERANOS (for fast spectra) reference deterministic transport codes. A short preliminary uncertainty propagation and ranking study is then performed for each equivalence models.

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“…All of this section is based on the graduate-level course of Alain Hébert at the École Polytechnique de Montréal. An exhaustive presentation of this thema can be found in reference, 2 which inspired this section, but also in Reuss and Massimo's documents, 3,4 this last one being more focused on high-temperature reactors.…”

Section: Neutronics Equationsmentioning

confidence: 99%

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

Descotes¹

2011

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In Very High Temperature Reactors (VHTRs), the long mean-free-path and large migration area of neutrons leads to spectral influences between fuel and reflector zones over long distances. This presents significant challenges to the validity of the classic two-step approach of cross section preparation wherein infinite lattice transport calculations are performed on relatively small physical domains (e.g. single assembly) in order to compute homogenized few-group cross sections for whole core analysis.Effects of the inner and outer reflectors challenge the classical two-step approach, while burnable poison locations affect neighboring assemblies as well. Use of transuranics-only (TRU) Deep Burn fuel in a prismatic VHTR (DB-VHTR) presents the additional challenge of producing vastly different neutron spectra between fresh and burned fuel.Modeling the VHTR also requires the ability to treat in the code the hexagonal geometry, the presence of TRISO particles, whose distribution in the fuel compacts is stochastic, and large lattice domains. Currently, DRAGON can treat these features: the hexagonal geometry is in place, the large domains may be treated by the method of characteristics (MOC) or collision probability (CP), and the doubleheterogeneity treatment can be used for TRISO particles.Therefore, in order to complement the analysis and design activities for Deep Burn in prismatic High Temperature Reactors (HTRs), this study was performed in order to provide some initial guidance on the necessary procedures to model the DB-VHTR with sufficient accuracy. The purpose of this study is to provide evaluation of the tradeoffs in accuracy between domain size in the lattice calculation and number of energy groups used in the whole core calculation. The lattice calculations were performed using the DRAGON-4 code and the whole core calculations using the Idaho National Laboratory code INSTANT, both of which are described in the following sections. The accuracy of the whole-core calculations was judged by comparison of Eigenvalues and ¿ssion rate shapes to results from MCNP calculations. The number of groups used in the lattice calculations was fixed at 295 and two different domain sizes were used, a single block and a supercell. Primarily of interest in this study were the effects of the reflector on the spectrum and, in turn, the adequacy of the cross section sets.Also included in this report are results from a preliminary study aimed at determining the effect of the presence of reflector on the burnup characteristics in a peripheral block. Depletion was performed in a supercell in order to observe the differences across a block of interest.The results showed that using the Method of Characteristics (MOC) in DRAGON, both a single block and a supercell leads to good predictions of the Eigenvalue and of the relative fission rates of each fuel block in the core if enough groups are kept in INSTANT (e.g., 26 groups). The supercell path produced the best results compared to the equivalent calculations using the single-block path....

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Base de neutronique : coefficients de réactivité dans les réacteurs nucléaires

Cited by 5 publications

References 0 publications

Flexblue^®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Flexblue^®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Artificial neural network surrogate development of equivalence models for nuclear data uncertainty propagation in scenario studies

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

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Base de neutronique : coefficients de réactivité dans les réacteurs nucléaires

Cited by 5 publications

References 0 publications

Flexblue®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Flexblue®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Artificial neural network surrogate development of equivalence models for nuclear data uncertainty propagation in scenario studies

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

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Product

Resources

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Flexblue^®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling

Flexblue^®core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling