Advanced modularity design for the MIT pebble bed reactor

The future of all reactors will depend on whether they can be economically built and operated. One of the major impediments to new nuclear construction is the capital cost due in large part to the length of construction time and complexity of the plant. Pebble bed reactors offer the opportunity to reduce the complexity of the plant because the number of safety systems required is significantly reduced due to the inherent safety of the technology. However, because of its small size, the capital cost per kilowatt is likely to be large if traditional construction approaches are followed. This strongly suggests the need for innovative construction concepts to reduce the construction time and cost. MIT has proposed a modularity approach in which the plant is pre-built in space-frame type modules which are built in factories. These space frames would contain all the equipment contained in a given volume. Once equipment in the space frame is installed, the space frame would then be shipped to the site and assembled “lego-style.” Studies presently underway have demonstrated the feasibility of the concept. Thermal stress analysis has been performed and an integrated design with the space frames has been developed. It is expected that this modularity approach will significantly shorten construction time and expense. This paper proposes a concept for further development, not a final design for the entire plant.     

Optimization of a radially cooled pebble bed reactor

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained.The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal.The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as 1 cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling.Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop ( bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (C) can be achieved compared to present axially cooled designs.     

Experimental investigation of the overall residence time of pebbles in a pebble bed reactor (PBR) using radioactive pebble

The granular flow of pebbles in a pebble bed reactor (PBR) under the influence of gravity is a dense granular flow with long-lasting frictional contacts. The basic governing physics is not fully understood and hence the dynamic core of a PBR and non-idealities associated with pebbles flow inside the reactor core are of non-trivial significance from the point of view of safety analyses, licensing, and thermal hydraulics. In the current study, overall and zonal pebbles residence time investigation is carried out by implementing noninvasive radioisotope-based flow visualization measurement techniques such as residence time distribution (RTD) and radioactive particle tracking (RPT). The characteristics of overall pebble residence time/transient number, zonal residence time, and the z-component of average zonal velocities at different initial seeding positions of a tracer particle have been summarized. It is found that the overall pebbles residence time/transient number increases (the z-component of average zonal velocities decreases) from the center towards the reactor wall. Also, pebbles’ zonal residence time results (the whole core is divided into three zones) which provide more insight and understanding about PBR core dynamics have been reported. The benchmark data provided could be used for assessment of commercial/in-house computational methodologies related to granular flow investigations.     

Individual pebble temperature peaking factor due to local pebble arrangement in a pebble bed reactor core

Scientists at the German AVR pebble bed nuclear reactor discovered that the surface temperature of some of the pebbles in the AVR core were at least 200 K higher than previously predicted by reactor core analysis calculations. The goal of this research paper is to determine whether a similar unexpected fuel temperature increase of 200 K can be attributed solely or mostly to elevated power production resulting from exceptional configurations of pebbles. If it were caused by excessive pebble-to-pebble local power peaking, there could be implications for the need for core physics monitoring which is not now being considered for pebble bed reactors. The PBMR-400 core design was used as the basis for evaluating pebble bed reactor safety. Through exhaustive Monte Carlo modeling of a PBMR-400 pebble environment, no simple pebble-to-pebble burn-up conditions were found to cause a sufficiently high local power peaking to lead to a 200 K temperature increase. Simple thermal hydraulics analysis was performed which showed that a significant core coolant flow anomalies such as higher than expected core bypass flows, local pebble flow variation or even local flow blockage would be needed to account for such an increase in fuel temperature. The identified worst case scenarios are presented and discussed in detail. The conclusion of this work is that the stochastic nature of the pebble bed cannot lead to highly elevated fuel temperatures but rather local or core-wide coolant flow reductions are the likely cause.     

Survey of dust production in pebble bed reactor cores

Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.     

Fuel elements of the high temperature pebble bed reactor

This work is devoted to spherical fuel elements for the high temperature pebble bed reactor, their manufacture and the conditions which they must satisfy for use in a process-heat reactor with an average gas outlet temperature T_{G, out} of 950°C. The positive results known from the operation of the AVR with T_G,out = 950°C and from extensive irradiation tests of the THTR-300 element with BISO coated mixed-oxide particles, even beyond the range of design specifications, and possible damage mechanisms are described in detail. They show that a spherical fuel element already exists, for which only a short-term development is needed to produce a coolant temperature of 950°C in a process-heat reactor. Further developments will be characterized by the use of a pebble bed HTR for high conversion rates (c ≈ 0.95) or for average gas outlet temperatures of more than 950°C. At higher temperatures the increased demands, mainly with regard to the release of fission products, can be fulfilled through the application of TRISO-coated fuel particles and the doping of the fuel kernels with . The reprocessing programme for fuel elements in the Federal Republic of Germany is mentioned briefly.     

基于球床密实实验装置的熔盐堆分流板设计

在熔盐球床堆设计中,为实现堆芯内部燃料球堆积结构的稳定性,需保证堆芯内部的流场均匀分布。研究基于模拟熔盐球床堆堆芯水力特性的球床密实实验装置(Pebble bed dense experiment facility,PBDE),通过设计不同形状和不同孔道分布的分流板,运用计算流体力学(Computational fluid dynamics,CFD)方法使用FLUENT软件对其堆芯内部的流场分布进行数值模拟,目的是保证实验中堆芯的流场分布均匀稳定。模拟结果表明,平板形分流板较锥形分流板能更好地使堆芯内流场均匀分布;且增加分流板的孔道数目或减小孔径能使堆芯内部的流场更加均匀稳定;比较设计的6种分流板模拟结果,最终给出满足PBDE堆芯流场均匀分布的分流板,为PBDE实验提供了基础,也为熔盐球床堆的堆芯流量分配设计提供技术方案与选型参考。     

Thermal-hydraulic analysis techniques for axisymmetric pebble bed reactor cores

A mathematical model for the analysis of coupled thermal-hydraulic problemsin steady-state pebble bed nuclear reactor cores is presented. The bed has been treated macroscopically as a generating, conducting porous medium. The model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, and new coefficients of the viscous and inertial loss terms are presented. The remaining equations in the model make use of continuity and thermal energy balances on the solid and fluid phases. None of the usual simplifying assumptions such as constant properties, constant velocity flow or negligible conduction and/or radiation are used. A computer program based on this model has been constructed; it has been validated by comparing predictions with measured values of previous experiments. Validation of the nonlinear fluid flow model is reported in a companion paper.     

Numerical treatment of pebble contact in the flow and heat transfer analysis of a pebble bed reactor core

This paper studies the numerical treatment of the inter-pebble regions in the modeling of a packed bed geometry for the computational fluid dynamics (CFD) analysis of a pebble bed reactor (PBR) core, where the pebbles are physically in contact with each other. In some studies, the inter-pebble regions have been approximated with gaps, in consideration of the problems on mesh quality or economy of the CFD calculation. To examine such a methodology, a sensitivity analysis for the gap size was conducted with two spherical pebbles, where the inter-pebble region was modeled by means of two kinds of inter-pebble gap and two kinds of direct contact. The cases of direct contact showed numerous differences in the results of the flow regime around the pebbles as well as in the wake, compared to the cases of the inter-pebble gap. No large differences were found between the two cases of direct contact. Based on the result of the sensitivity analysis, the two cases of inter-pebble modeling, i.e., the 1-mm gap and area-contact, were applied to the PBR simulation. It was concluded that the flow regimes and their relevant flow-induced local heat transfer were significantly dependent on the modeling of the inter-pebble region.     

Core physics and fuel cycles of the pebble bed reactor

A high degree of flexibility for the application of the pebble bed HTR results from the size of the fuel elements and the continuous loading of fuel during operation. The OTTO loading scheme (once-through-then-out) allows the gas outlet temperature to be increased up to 1190°C as desired. Different fuel cycles can be used in the same reactor. The highly enriched uranium/thorium fuel cycle allows favourable utilization of the fissile reserves. For a conventional design the net consumption is 467 g/GWd(th), and this figure can be reduced to 58 g/GWd(th) for the advanced variant, achieving the breeding ratio 0.95. The advantage of the low enrichment uranium fuel cycle is the direct in situ utilization of the bred plutonium, being as high as 87%, which reduces the demand for short-term recycling. The reactor allows change over between the different cycles under full power operation.     

Stochastic techniques for the control and surveillance of a modular pebble bed reactor

This paper represents the first of a series of publications describing work in progress on the research, design and testing of a control and surveillance system for a Modular Pebble Bed Reactor.

The scope of the project involves the design of a simple state of the art control system for the reactor and a surveillance system based on smart instrumentation and expert system logic. It is noted that there are some physical and experimental problems unique to the MPBR family. These problems are connected with long neutron lifetimes, the need for a new evaluation of kinetic parameters and reactivity effects, and the need for very high sensitivity counting channels.     

The Thermal-Hydraulic model for the pebble bed modular reactor (PBMR) plant operator training simulator system

This paper provides a discussion of the model development status and verification efforts for the Reactor Core Thermal-Hydraulic model developed for the full-scope plant Operator Training Simulator System of the Pebble Bed Modular Reactor (PBMR). Due to the First of a Kind Engineering nature and lack of reference plant data, model verification has mainly been focused on benchmarking the model configurations against test cases performed by PBMR design analysis codes, i.e. TINTE, VSOP and FLOWNEX.As a first step, due to the symmetrical physical nature of the PBMR core, a two-dimensional (2D) model configuration in radial and axial directions (axial-symmetry) was developed. The design was subsequently extended to a three-dimensional (3D) configuration. Through the use of cross-flow and cross-conduction links, three nearly identical 2D configurations were glued together to form this 3D model configuration. To date, the 3D configuration represents the most comprehensive model to simulate the PBMR core thermo-hydraulics. This paper concludes with the verification of thermodynamic and heat-transfer properties of two steady state (100% and 40% power) conditions between the 3D Reactor Core Thermal-Hydraulic model and the available FLOWNEX and TINTE design code analysis. The transient operations between these two power levels are also discussed.     

Comparison of two models for a pebble bed modular reactor core coupled to a Brayton cycle

The pebble bed modular reactor (PBMR) plant is a promising concept for inherently safe nuclear power generation. This paper presents two dynamic models for the core of a high temperature reactor (HTR) power plant with a helium gas turbine. Both the PBMR and its power conversion unit (PCU) based on a three-shaft, closed cycle, recuperative, inter-cooled Brayton cycle have been modeled with the network simulation code Flownex.One model utilizes a core simulation already incorporated in the Flownex software package, and the other a core simulation based on multi-dimensional neutronics and thermal-hydraulics. The reactor core modeled in Flownex is a simplified model, based on a zero-dimensional point-kinetics approach, whereas the other model represents a state-of-the-art approach for the solution of the neutron diffusion equations coupled to a thermal-hydraulic part describing realistic fuel temperatures during fast transients. Both reactor models were integrated into a complete cycle, which includes a PCU modeled in Flownex.Flownex is a thermal-hydraulic network analysis code that can calculate both steady-state and transient flows. An interesting feature of the code is its ability to allow the integration of an external program into Flownex by means of a so called memory map file.The total plant models are compared with each other by calculating representative transient cases demonstrating that the coupling with external models works sufficiently. To demonstrate the features of the external program a hypothetical fast increase of reactivity was simulated.     

Development of a method of evaluating an inventory of fission products for a pebble bed reactor

The design of the high temperature gas-cooled reactor (HTGR) has evolved and the relevant safety requirements have been defined; accordingly, the source term to be used as the basis for licensing must also be developed. However, analysis of the source term in the HTGR has not been adequately investigated and there has not been definite improvement in this respect. Because radioactivity in normal operation must be well understood, the purpose of this study is to establish a method for activity evaluation by the code combination MCNP-ORIGEN-MONTEBURNS-MOTEX. The sophisticated method, which constructs the HTR-10 core by using the unit lattice of a hexagonal prism, is developed for core modeling. The MCNP modeling is used to simulate the generation of fission products with an increase of burnup, and ORIGEN is utilized for depletion calculation of each fission product. Continuous fuel management is divided into five discrete periods for the feeding and discharging of fuel pebbles. MONTEBURNS is used for discrete fuel management. In short, this work by aid from MOTEX traces 41 isotope nuclides, the results of which seem highly probable. In addition, the inventory of actinides at the end of each cycle is also investigated. It would be informative when the waste management of spent fuel of HTGRs would be taken into account. This article lays the foundation for future work on the analysis of the source term in HTGRs and will hopefully serve as a platform from which the safety assessment of radioactive material release during accidents can be undertaken in future.     

Flow visualization in a pebble bed reactor experiment using PIV and refractive index matching techniques

In the advanced gas-cooled pebble bed reactors for nuclear power generation, the fuel is spherical coated particles. The energy transfer phenomenon requires detailed understanding of the flow and temperature fields around the spherical fuel pebbles. Detailed information of the complex flow structure within the bed is needed. Generally, for computing the flow through a packed bed reactor or column, the porous media approach is usually used with lumped parameters for hydrodynamic calculations and heat transfer. While this approach can be reasonable for calculating integral flow quantities, it may not provide all the detailed information of the heat transfer and complex flow structure within the bed. The present experimental study presents the full velocity field using particle image velocimetry technique (PIV) in a conjunction with matched refractive index fluid with the pebbles to achieve optical access. Velocity field measurements are presented delineating the complex flow structure.     

Cross-section parameterization of the pebble bed modular reactor using the dimension-wise expansion model

This paper discusses the use of the dimension-wise expansion model for cross-section parameterization. The components of the model were approximated with tensor products of orthogonal polynomials. As we demonstrate, the model for a specific cross-section can be built in a systematic way directly from data without any a priori knowledge of its structure. The methodology is able to construct a finite basis of orthogonal polynomials that is required to approximate a cross-section with pre-specified accuracy. The methodology includes a global sensitivity analysis that indicates irrelevant state parameters which can be excluded from the model without compromising the accuracy of the approximation and without repetition of the fitting process. To fit the dimension-wise expansion model, Randomised Quasi-Monte-Carlo Integration and Sparse Grid Integration methods were used. To test the parameterization methods with different integrations embedded we have used the OECD PBMR 400 MW benchmark problem. It has been shown in this paper that the Sparse Grid Integration achieves pre-specified accuracy with a significantly (up to 1–2 orders of magnitude) smaller number of samples compared to Randomised Quasi-Monte-Carlo Integration.     

Advanced pebble bed high temperature reactor with central graphite column for future applications

Design evaluations of the advanced pebble bed high temperature reactor, AHTR, with central graphite column are given. This reactor, as a nuclear heat source, is suitable for coal refinement as well as for electricity generation with closed gas turbine primary helium circuit. With this design of the central graphite column, it is possible to limit the core temperatures under the required value of about 1600°C in case of accident conditions, even with higher thermal power and higher core inlet and outlet temperatures. The designs of core internals are described. The after heat removal system is integrated in the prestressed concrete reactor pressure vessel, which is based on the principals of natural convection.Research work is being carried out, whereby the spherical fuel elements are coated with a layer of silicon carbide, to improve the corrosion resistance as well as the effectiveness of the fission products barrier.     

Comparison of coolant flow predictions with those measured on a full-scale mockup of a pebble bed reactor core

In a companion paper, a mathematical model for the analysis of coupled thermal-hydraulic problems in steady-state, axisymmetric pebble bed nuclear reactor cores was presented. In this paper, predictions by the computer code PEBBLE, which is based on the previously reported model, are compared with flow measurements from the full-scale mockup of the Oak Ridge National Laboratory Pebble Bed Reactor Experiment. The code PEBBLE is shown to adequately predict distributions and magnitudes of velocity and pressure for high Reynolds number flows in packed sphere beds. Limitations of the code and its ability to adequately calculate flow distributions in large pebble bed power reactors are discussed.     

Integrated design approach of the pebble BeD modular reactor using models

The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts.This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results.This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR.     

Fuel burnup performance of an OTTO refueling pebble bed reactor with burnable poison loading

Fuel burnup performance has been analyzed for a pebble bed reactor with a once-through-then-out (OTTO) refueling scheme and compared with a reference multi-pass scheme. A new fuel pebble was designed by adding spherical B₄C particles into its free fuel zone for controlling the infinite multiplication factor during burnup, and then reducing the axial power peak of the OTTO scheme. The objective is to maximize the fuel burnup performance of the OTTO scheme while keeping the power peak under a limit and ensuring the core criticality. Numerical calculations were performed based on the 400 MWt pebble bed modular reactor (PBMR) using the MVP code. For the fuel pebble of the PBMR containing 9 g uranium with 9.6 wt% ²³⁵U enrichment, 1600 B₄C particles with a radius of 70 μm are determined to flatten the k_∞ curve in the early burnup stage. The dependences of the neutronic properties of the core with the OTTO scheme on target fuel burnup show that the maximum target burnup of 74 GWd/t can be achieved so that the power peak is reduced to about 10.80 W/cm³ which is approximate that of the multi-pass scheme (10.85 W/cm³). This target burnup is about 22% less than that of the multi-pass scheme (95 GWd/t), i.e. the fuel utilization efficiency of the OTTO scheme is about 22% lower, which could be compensated by the construction and operation cost of the fuel handling system. This result also suggests that further investigations of the fuel burnup performance and other properties are needed in both neutronic and thermal hydraulic viewpoints to find out the optimal core performance.