Pebble Bed High Temperature Reactor Research Articles

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Influence of the Lagrangian Integral Time Scale Estimation in the Near Wall Region on Particle Deposition

In a high temperature pebble-bed reactor core where thousands of pebbles are amassed, the friction between the outer graphite layer of the fuel elements triggers the formation of carbonaceous dust. This dust is eventually conveyed by the cooling carrier phase and deposits in the primary circuit of the high temperature reactor. The numerical prediction of carbonaceous dust transport and deposition in turbulent flows is a key safety issue. Most particle tracking procedures make use of the Lagrangian integral time scale to reproduce the turbulent dispersion of the discrete phase. In the present Lagrangian particle tracking procedure, the effect of the Lagrangian integral time scale near the wall is thoroughly investigated. It is found that, in the linear sublayer, a value of the normalized wall normal component of the Lagrangian integral time scale lower that 4 delivers accurate particle deposition velocities. The value worked out here near the wall region is in accordance with Lagrangian integral time scales derived from recent direct numerical simulations.

A collective dynamics-based method for initial pebble packing in pebble flow simulations

In the simulation of pebble flow in Pebble-Bed High Temperature Reactors (PB-HTR's), high-fidelity methods, such as discrete element method (DEM), are usually employed to simulate the dynamic process of the pebble circulation. Such simulation normally takes extraordinarily long time to reach the dynamic equilibrium state, in which the pebble distribution is statistically steady. However, if an initial dense packing of pebbles can be provided, which is close to the realistic packing at the equilibrium state and can be easily implemented without much computational effort, then the high-fidelity pebble flow simulation can take much less time to reach the dynamic equilibrium state. In this paper, a collective dynamics-based method is developed to generate an initial pebble packing for the subsequent high-fidelity pebble flow simulations. In the new method, pebbles are packed by two processes: a sequential generation process allowing overlaps and an overlap elimination process based on a simplified normal contact force model. The latter provides an adaptive and efficient mechanism to eliminate the overlaps accounting for different overlap size and pebble size, thus can pack tens of thousands of pebbles within several minutes. Applications of the new method to packing pebbles in both cylindrical and annular core geometries are studied for two types of PB-HTR designs: HTR-10 and PBMR-400. The resulting packings show similar radial and axial packing fraction distributions compared to the dynamic equilibrium packing state produced by the DEM pebble flow simulation. Comparisons with other existing random packing methods, such as the gravitational deposition method, have shown that the developed method not only exhibits excellent computation efficiency, but also presents desirable potential in other applications as a general packing algorithm for packing mono-sized or poly-sized spheres in a large container.

Simplified models for pebble-bed HTR core burn-up calculations with Monteburns2.0 ©

This paper aims at comparing some simplified models to simulate irradiation cycles of Pu fuelled pebble-bed reactors with Monteburns2.0 © code. As a reference core, the PBMR-400 (proposed in the framework of the EU PUMA project, where this kind of core fuelled by a Pu and Pu–Np fuel has been studied) was taken into account. Pebble-bed High Temperature Reactor (HTR) cores consist of hundreds of thousands pebbles arranged stochastically in a cylindrical or annular space and each pebble is a single fuel element, and it is able to reach ultra-high burn-ups, i.e. up to 750 GWd/tHM (for Pu-based fuels). Additionally, pebble-bed cores are characterised by a continuous recirculation of pebbles from the top to the bottom of the core. Modelling accurately with current computer codes such an arrangement, in order to predict the behaviour of the core itself, is a very difficult task and any depletion code specifically devoted to pebble-bed burn-up calculation is not available at the moment. Because of limitations of the most common current MCNP-based depletion codes as well as huge calculation times, simplified models have to be implemented. After an analysis of the literature available on pebble-bed models for criticality and burn-up calculations, a preliminary assessment of the impact of different kind of simplified models for a Pu-Np fuelled Pebble-Bed Modular Reactor (PBMR), proposed in the framework of the EU PUMA project, is shown, particularly as far as burn-up prediction with Monteburns2.0 © code is concerned.

Automated Design and Optimization of Pebble-Bed Reactor Cores

This paper presents a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble bed cores. The method employs PEBBED, a reactor physics code specifically designed to solve for the asymptotic burnup state of pebble bed reactors in conjunction with a genetic algorithm to obtain a core with acceptable properties. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. A novel representation of the distribution of pebbles enables efficient coupling of the burnup and neutron diffusion solvers. Complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to manipulation using modern optimization techniques. The user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. For this study, the design of two pebble bed high-temperature reactor concepts subjected to demanding physical constraints demonstrated the technique’s efficacy.

A nodal dynamic model for control system design and simulation of an MHTGR core

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high-temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power-distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium–plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 ( n, γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX.

PBMR radionuclide source term analysis validation based on AVR operating experience

The determination of radionuclide source terms is vital for any reactor design and licensing safety evaluation. This paper provides an overview of the PBMR analysis tools with specific focus on the modelling of mobile and deposited radionuclide source terms within the pressure boundary of a typical pebble-bed high temperature reactor (HTR). The main focus is on the Dust and Activity Migration and Distribution (DAMD) software code system that models the activation, migration and time-dependent distribution of dust and atomic particles in an HTR such as the AVR and PBMR. Since DAMD provides a time-dependent systems integrated model of HTR designs, most of the obvious physical phenomena relevant to source terms are at play. These include the neutron flux, activation cross-sections, radioactive decay, dust production rates, dust impurity levels, dust filter capabilities, dust particle size distributions, thermal–hydraulic parameters influencing the migration and distribution of particles throughout the main power system and subsystems, and helium coolant leakage and make-up rates. At this stage the DAMD calibration and validation is mainly based on the operational data, experiments and measurements made during 21 years of operating life of the AVR. The comparisons of the DAMD results with various AVR measurements provide confidence in the use of DAMD for the PBMR design and safety evaluations. In addition, sensitivity analyses are performed with DAMD to determine the bounding system parameters that drive the migration and distribution of radionuclides. The use of DAMD to evaluate design configurations, e.g. the effect of the introduction and placement of filters on the radionuclide distribution, is also shown. In conclusion, the importance of a systems modelling approach for radionuclide transport and distribution within the pressure boundary of a typical HTR system, is demonstrated. Since the DAMD code system is calibrated and validated against the AVR measurements it can be concluded that the radionuclide source term phenomena in the AVR, resulting in the measured AVR contamination levels, is taken into account in the design and safety evaluation of the PBMR.

Evaluation of experiments in the AVR with the DALTON–THERMIX coupled code system

The DALTON–THERMIX coupled code system for pebble-bed high temperature reactors was further validated by simulation of two dynamic experiments in the AVR. The validation included a rapid mass flow reduction experiment and a depressurized loss of forced cooling (DLOFC) experiment. For the mass flow reduction experiment, the rotational speed of the blowers dropped rapidly from 100% to 50% in 62 s. In the DLOFC experiment, the long term behavior (100 h) of the reactor was calculated and compared with experiments. Besides the transient calculation, each experiment included a coupled steady state calculation to determine the initial condition for the transient. The calculated results of the new code system agree very well with the experimental ones, from which it is concluded that the code system is very well suited for the design and optimization of future (V)HTR cores.

In-core fuel management optimization of pebble-bed reactors

A reduction of the power peak in the core of High Temperature pebble-bed reactors is attractive to decrease the maximum fuel temperature and to increase fuel performance. A calculation procedure was developed, which combines fuel depletion, neutronics and thermal–hydraulics to investigate the impact of a certain (re)loading scheme for the pebble-bed type HTR. The procedure has been applied to a model of the Pebble Bed Modular Reactor (400 MW) design. This paper shows that an important reduction in axial power peaking can be achieved by adopting a multi-pass recycling scheme for the pebbles. By dividing the core into several radial fuel zones in combination with multi-pass recycling the power profile can be flattened in the radial direction. For a core with two fuel zones the impact on the k eff and maximum power density as a function of the zone size has been investigated. A heuristic method has been used to find the optimal pebble loading pattern for several (re)loading schemes. Using this method a reduction of the maximum power density from 10.0 to 8.2 MW/m 3 has been achieved for a core with three radial fuel zones. The effects of the improved power profiles on the fuel temperature during normal operation and a Depressurized Loss Of Coolant (DLOFC) accident have been analyzed. It was found that the optimized power profile results in a reduction of the maximum fuel temperature of 80 °C and 300 °C for normal operation and DLOFC conditions, respectively.

Analysis of Pebble-Fueled Zone Modeling Influence on High-Temperature Reactor Core Calculations

The pebble bed gas-cooled reactor is one of the most promising concepts among the Generation III+ and Generation IV reactors. Currently, the pebble bed modular reactor (PBMR) design, both U and Pu and minor actinide fueled, is being developed. Modeling the arrangement of coated particles (CPs) inside a spherical region like a pebble seems to be an important issue in the frame of calculations. To use the (relatively) old Monte Carlo codes without any correction, some approximations are often introduced. Recent Monte Carlo codes like MCNP5 and some new original subroutines that we have developed allow the possibility of obtaining more detailed and more physically correct geometrical descriptions of this kind of system. Some studies on modeling pebbles and pebble bed cores have already been carried out by other researchers, but these works are substantially limited to AVR-type UO2-fueled pebbles. However, the impact of approximated models on fuel mass, reactivity, and reactor life prediction has not yet been investigated for new PBMR-type pebbles.At the same time, an assessment of introducing a stochastic CP arrangement is not so widespread. Analyzing two PBMR pebbles, one Pu- and the other U-fueled, this paper focuses on quantifying errors due to the different approximations generally used to describe the CP lattice inside a high-temperature reactor pebble bed core, as far as mass of fuel, reactivity, and burnup simulation are concerned. This aim was reached also through a new feature implemented in the MCNP5 code, i.e., capability to treat (pseudo) stochastic geometries. Later, we compared the initial mass of fuel, keff, and isotopic evolution versus burnup of some approximated pebble models with the reference model, built by means of this new MCNP5 feature.

DEM simulations and experiments of pebble flow with monosized spheres

Pebble flow in the core of a pebble bed high-temperature reactor (HTR) is a typical granular flow with a huge number of monosized spherical pebbles, where the kinematics of the pebbles needs to be predicted more accurately. In this study the discharge of monosized glass beads from a semi-cylindrical silo has been investigated using the Discrete Element Method (DEM), in which the particle–particle interaction model is based on contact mechanics with measured friction coefficients for precisely predicting the flow behaviour. Particle flow experiments were carried out with a transparent semi-cylindrical silo. Two cases were considered: one is for tracking the trajectories of identified particles with 11,500 glass beads, and the other is for the ‘double-zone’ flow for estimating the mixing evolution with 10,460 glass beads. Comparisons between the experimental and DEM results show good agreement, which indicates the applicability of DEM for predicting the pebble flow in a HTR.

Development and testing of analytical models for the pebble bed type HTRs

The pebble bed type gas cooled high temperature reactor (HTR) appears to be a good candidate for the next generation nuclear reactor technology. These reactors have unique characteristics in terms of the randomness in geometry, and require special techniques to analyze their systems. This study includes activities concerning the testing of computational tools and the qualification of models. Indeed, it is essential that the validated analytical tools be available to the research community. From this viewpoint codes like MCNP, ORIGEN and RELAP5, which have been used in nuclear industry for many years, are selected to identify and develop new capabilities needed to support HTR analysis. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP. The coupled MCNP-ORIGEN code is used to estimate the burnup and the refuelling scheme. Results obtained from Monte Carlo analysis are interfaced with RELAP5 to analyze the thermal hydraulics and safety characteristics of the reactor. New models and methodologies are developed for several past and present experimental and prototypical facilities that were based on HTR pebble bed concepts. The calculated results are compared with available experimental data and theoretical evaluations showing very good agreement. The ultimate goal of the validation of the computer codes for pebble bed HTR applications is to acquire and reinforce the capability of these general purpose computer codes for performing HTR core design and optimization studies.

Analyses of a restrained growth graphite irradiation creep experiment

Graphite is used as a moderator, reflector and structural component in pebble bed and prism High Temperature Reactors (HTRs). It is fortunate to reactor designers that irradiated graphite shows remarkably high creep behaviour under the influence of fast neutron irradiation at temperatures far below those required for significant creep strains to be generated in unirradiated graphite. This creep behaviour is important in the design of nuclear graphite reactor cores because the self-induced shrinkage stresses generated in typical core components during irradiation can be relieved. However, there are no reliable data on high fluence irradiation creep and the mechanistic understanding of the irradiation creep is insufficiently developed to reliably extrapolate to the high fluences expected of graphite in future HTR designs. The understanding of irradiation creep is further complicated because it has been experimentally observed that irradiation creep strain in graphite modifies other properties in particular the coefficient of thermal expansion. In addition modified changes in Young's modulus in crept specimens have been reported and it has also been postulated that irradiation creep may also modify dimensional change. The assessment of irradiation creep in graphite components is based on empirical laws derived from data obtained from small samples irradiated in a materials test reactor. However, due to the complicated irradiation rigs required and the amount of dimensional and property measurements needed to be taken, constant stress irradiation creep experiments are difficult and very expensive to carry out successfully. However, restrained creep experiments are simple to implement, less expensive and can be easily included as part of other, more conventional irradiation graphite experimental programmes. However, in the past, the disadvantage of these experiments has been that the results have been difficult to interpret using the then available analytical methods. In this paper the restrained creep experiment is revisited and analysed numerically and the possible benefit of using a restrained creep experiment in future graphite irradiation experiments is investigated. It is shown that a numerical simulation of the restrained creep experiment behaviour would be an essential tool to ensure that the stress within the specimen remains within defined limits so that specimen failure can be avoided.

Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten

The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on 18 October 2005 after 10 reactor cycles totaling 249 efpd and a maximum burn-up of 11.07% FIMA. The objective of the HFR-EU1bis test was to irradiate five HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, better use of fuel) and thus reduced waste production. The central temperature of all pebbles was kept as closely as possible at 1250 °C and held constant during the entire irradiation, with the exception of HFR downtime and power transients. This is the expected maximum central fuel temperature of a pebble bed VHTR with a coolant outlet temperature of 1000 °C. HFR-EU1bis should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: • increased central fuel temperature of 1250 °C compared to 1000–1200 °C in earlier irradiation tests; • irradiation to a burn-up close to 16% FIMA, which is double the license limit of the HTR-Modul; due to a neutronics data processing error, the experiment was prematurely terminated at 11.07% FIMA maximum so that this objective was not fully achieved; • confirmation of low coated particle failure fractions due to temperature, burn-up and neutron fluence. This paper provides the irradiation history of the experiment including data on fission gas release. Post-irradiation examinations at NRG Petten and JRC-ITU Karlsruhe included the verification of the received neutron fluences, burn-up and spectrum. They will be followed shortly by safety-relevant heating tests at JRC-ITU to verify fission product retention by out-of-pile heating tests beyond 1600 °C.

Modeling the multiaxial strength of H-451 nuclear grade graphite

The core of a prismatic High Temperature Reactor (HTR) is constructed from an array of nuclear graphite components. Similarly, the core of a Pebble Bed HTR is confined by large graphite blocks which define the (annular) core geometry. In both HTR designs the large graphite components act as neutron moderator and reflector as well as providing mechanical support to the active core. During reactor operation the graphite components of the core are subjected to complex stress states. Consequently, core designers need a suitable theory of failure. Both deterministic (e.g., maximum principal stress theory) and probabilistic (e.g., Weibull failure theory) have been considered. To test candidate failure theories a multiaxial testing program was conducted at Oak Ridge National Laboratory on H-451 graphite. Large specimens (∼27 cm length) were subjected to combined axial stress (tension and compression) and internal pressure. A total of 59 specimens were tested at 9 stress ratios in the first and fourth stress quadrants. Here, we report the basis and performance of a microstructurally based graphite fracture model and multiaxial strength data for grade H-451 graphite, along with the application of the model to predict the failure envelope for H-451 graphite in the first and fourth multiaxial quadrants.

Analytical Calculation of the Average Dancoff Factor for Prismatic High-Temperature Reactors

In the present studies we performed the analytical calculation of the average Dancoff factor for prismatic high-temperature reactors; in this type of core, the fuel elements consist of small fuel grains (TRISO particles) randomly dispersed in a moderator (graphite) matrix and confined to a cylindrical volume (fuel pin). By definition, the Dancoff factor is the probability that a neutron leaving a fuel kernel hits uncollided another fuel kernel in the same fuel pin, which represents the intrapin contribution, or in another pin, which represents the interpin contribution. Similar studies have already been performed for pebble bed high-temperature reactors, where spheres (fuel pebbles) play the role of the cylinders; consequently, we retained the physical model describing an infinite lattice of unit cells, each containing a pair of concentric spheres, where the inner sphere is filled with a mixture of fuel grains and moderator and the outer one is filled with pure moderator, and we derived the mathematical model for the case of concentric cylinders. The physical model is grounded on the chord theory and the concept of a pseudo cross section; the latter takes into account, when the medium consists of moderator and small fuel grains, the probability, per unit path length, that a neutron either collides with a moderator nucleus or hits a fuel surface. The above method possesses a general validity, and it is suitable for the treatment of spheres (fuel pebbles), cylinders (fuel pins), or cuboids (fuel prisms) filled by moderator and small fuel grains.The predictions of the analytical method well match the results of the MCNP code; nevertheless, since in the case of prismatic cores the mathematical model involves the calculation of complicated double integrals, the CPU time required by the two different methods becomes comparable.

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF 2 (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.

Investigation of bounds on particle packing in pebble-bed high temperature reactors

Models and methods are presented for determining practical limits of the packing density of TRISO particles in fuel pebbles for a pebble-bed reactor (PBR). These models are devised for designing and interpreting fuel testing experiments. Two processes for particle failure are accounted for: failure of touching particles at the pressing stage in the pebble manufacturing process and failure due to inner pressure buildup during irradiation. The second process gains importance with increasing fuel temperature, which limits the particle packing density and the corresponding fuel enrichment. Suggestions for improvements to the models are presented.

ENRICHED URANIUM TECHNOLOGY AT THE SILLAMÄE OIL SHALE PROCESSING PLANT

The shale oil factory at Sillamae, Estonia was founded in 1928 by the ”Estonian Oil Consortium”, belonging to Swedish capital. It is the only factory that has processed both the brown (kukersite) oil shale and the black (dictyonema) shale that are both abundant in Estonia. The much older black shale contains, besides the oil-bearing organic material, significant amounts of various heavy metals, including uranium, and is therefore radioactive. Uranium-based nuclear technology has dominated the second half of the last century, but now faces potential competition from the emerging hydrogen-based energy technologies. It is little known that the Sillamae Oil Shale Plant has played a significant role in the development of the gas cooled very high temperature pebble bed reactor for direct production of either hydrogen or electricity. We analyze here the timeline of this development effort.