Innovative Reactor Designs Research Articles

Impact of new evaluated nuclear data libraries on core characteristics of innovative reactor designs

The impact of new evaluated nuclear data libraries (ENDF/B, JENDL and JEFF) on core characteristics of thermal and fast spectra innovative reactor designs was investigated. The innovative reactor design with thermal neutron spectrum is represented by small-sized, long-life CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactors) HTGRs (High Temperature Gas-cooled Reactors) with uranium and thorium fuel cycles, while the one with the fast neutron spectrum is represented by compact, sodium-cooled B&BRs (Breed-and-Burn Reactors). The CANDLE HTGR core characteristics (i.e. kinf, keff, discharge burnup, burning region moving velocity, core life time and axial power peaking factor) were evaluated by a dedicated deterministic neutronics and depletion code based on few-group neutron diffusion theory in 2-D RZ geometry, while the required microscopic cross sections were prepared by using SRAC2006 with JENDL-4.0, ENDF/B-VI and JEFF-3.1 based new SRAC2006 libraries. The B&BR core characteristics were evaluated using a continuous energy Monte Carlo neutron transport code, SERPENT, with JENDL-4.0, ENDF/B-VII.0, ENDF/B-VII.1 and JEFF-3.1.1 nuclear data based new libraries. The impact of new evaluated nuclear data libraries for innovative CANDLE HTGR core characteristics was found significantly larger for thorium fuel cycle than the one for uranium fuel cycle, especially in the discharge burnup, burning region moving velocity which affected the core life time significantly. The findings indicated the needs of more accurate nuclear data libraries for nuclides involved in thorium fuel cycle. The impact of new evaluated nuclear data libraries for innovative B&BR core characteristics was found in a slightly large variation for keff evolution during burnup which in turn affected the estimated core life time. However, a good agreement within the statistical error on the integral kinetic parameters, Doppler reactivity coefficients, coolant density reactivity coefficients, axial and radial expansion reactivity coefficients, power density distributions can be observed among the libraries. In addition, the buildup of fission products and minor actinides were also similar.

Theory and implementation of nuclear safety system codes – Part II: System code closure relations, validation, and limitations

This is Part II of two articles describing the details of thermal-hydraulic system codes. In this second part of the article series, the system code closure relationships (used to model thermal and mechanical non-equilibrium and the coupling of the phases) for the governing equations are discussed and evaluated. These include several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models between the fields. These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models.Code verification and validation, including Separate Effects Tests (SETs) and Integral effects tests (IETs) is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities, but some of the more advanced reactor designs will push the limits of validation for the codes.Lastly, the limitations of the codes are discussed by considering next generation power plants, such as Small Modular Reactors (SMRs), analyzing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, innovative fuel types, liquid metal coolants, and many other design features that may not be fully analyzed by current system codes.This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations, validation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν¯e) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of 235U, 239Pu and 241Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

The Detection of Reactor Antineutrinos for Reactor Core Monitoring: an Overview

There have been new developments in the field of applied neutrino physics during the last decade. The International Atomic Energy Agency (IAEA) has expressed interest in the potentialities of antineutrino detection as a new tool for reactor monitoring and has created an ad hoc Working Group in late 2010 to follow the associated research and development. Several research projects are ongoing around the world to build antineutrino detectors dedicated to reactor monitoring, to search for and develop innovative detection techniques, or to simulate and study the characteristics of the antineutrino emission of actual and innovative nuclear reactor designs. We give, in these proceedings, an overview of the relevant properties of antineutrinos, the possibilities of and limitations on their detection, and the status of the development of a variety of compact antineutrino detectors for reactor monitoring.

A novel extension of chord length sampling method for TRISO-type fueled reactor applications

A new chord length distribution model is proposed to characterize the stochastic distribution of TRISO fuel particles in nuclear reactor systems and is used in the chord length sampling (CLS) method to analyze the neutronic behavior of TRISO fuel systems. In this model, the coating layers of fuel particles are homogenized with the background matrix region. The probability density function (PDF) of the chord length between fuel kernels, instead of fuel particles, is developed and is used in the CLS method. We first apply the new CLS model to solving one-group eigenvalue problems in a simplified 3-D stochastic medium system. Good accuracy is obtained in predicting the multiplication factor and fission density distribution. The relative differences are within 1.0% for both the multiplication factor and the total fission density in all the studied scenarios. We then apply the new CLS model to analyzing three realistic reactor designs: two Very High Temperature Gas-cooled Reactor unit cells and one fuel pin unit cell of an innovative light water reactor design with accident tolerant fuel. Infinite multiplication factor and intra-Dancoff factor are evaluated for the three unit cells respectively. Compared with the reference results, predictions from CLS simulations show a relative difference of less than 0.26% for infinite multiplication factors and less than 1.0% for intra-Dancoff factors in all the studied cases. Meanwhile, a significant improvement in the computational efficiency has been observed for the CLS new model compared with the old model, with a speedup of at least 1.70. The CLS method with the new chord length distribution model is verified to be an efficient Monte Carlo method, superior over the old model and without sacrificing the accuracy.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Factors affecting the microwave coking of coals and the implications on microwave cavity design

The work carried out in this paper assessed how processing conditions and feedstock affect the quality of the coke produced during microwave coke making. The aim was to gather information that would support the development of an optimised microwave coke making oven. Experiments were carried out in a non-optimised 2450MHz cylindrical cavity. The effect of treatment time (15–120min), power input (750W–4.5kW) and overall power input (1700–27,200kWh/t) on a range of coals (semi-bituminous–anthracite) was investigated. Intrinsic reactivity, random reflectance, strength index and dielectric properties of the produced cokes were compared with those of two commercial cokes to assess the degree of coking produced in the microwave system.Overall energy input and coal rank were found to be the major factors determining the degree of coking following microwave treatment. The dependency on coal rank was attributed to the larger amount of volatiles that had to be removed from the lower ranked coals, and the increasing dielectric loss of the organic component of the coal with rank due to increased structural ordering. Longer treatment times at lower powers or shorter treatment times at higher powers are expected to produce the same degree of coking.It was concluded that microwave coke making represents a potential step-change in the coking industry by reducing treatment times by an order of magnitude, introducing flexibility and potentially decreasing the sensitivity to quality requirement in the feedstock. The main challenges to development are the energy requirements (which will need to be significantly reduced in an optimised process) and penetration depth (which will require an innovative reactor design to maximise the advantage of using microwaves). Understanding and quantifying the rapidly changing dielectric properties of the coal and coke materials is vital in addressing both of these challenges.

Effects of fuel particle size distributions on neutron transport in stochastic media

This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (keff) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron transport. As high as 1.00% relative difference in keff and ∼1.50% relative difference in peak fission power density are observed. As the packing fraction and optical thickness increase, the effect gradually dissipates. Neutron channeling between fuel particles is identified as the effect most responsible for the different neutronic results. Different size distributions result in the difference in the average number of fuel particles and their average size. As a result, different degrees of neutron channeling are produced. The size effect in realistic reactor unit cells is also studied and, from the predicted values of infinite multiplication factors, it is concluded that the fuel particle size distribution effects are not negligible.

Energy from thorium—An Indian perspective

Nuclear energy as a sustainable resource in India has been very clearly formulated in the three stage nuclear programme. The role of thorium as a potential fuel in the third stage of this programme has also been elucidated. With this aim there have been pioneering research efforts in all aspects of the thorium fuel cycle. Thorium being fertile and not accompanied by the fissile species requires the use of a fissile topping. There have been several studies in India on the use of thorium in different reactor systems from thermal to intermediate and fast spectrum, molten salt reactors, high temperature reactors, compact nuclear power packs and even Subcritical systems. In this paper, we present some of the research studies on use of thorium in thermal reactor systems. We give an overview of the neutronic properties of thorium and the bred fissile material and then proceed to show the performance potential in different reactor systems. We also present the innovative Indian reactor designs which utilize thorium, namely the Advanced Heavy Water Reactor (AHWR) and the Indian High Temperature Reactors.

The reactor dynamics code DYN3D and its trigonal-geometry nodal diffusion model

Abstract The reactor dynamics code DYN3D is a three-dimensional best-estimate tool for simulating steady states and transients of light-water reactors and innovative reactor designs. An overview of the DYN3D features is provided. This paper further focuses on the recently developed trigonal-geometry diffusion model DYN3D-TRIDIF including a description of the underlying nodal approach and the characteristics of trigonal geometries. Via a mesh refinement study by means of a VVER-1000-type core benchmark using a fine-mesh diffusion reference solution, DYN3D-TRIDIF shows spatial convergence. Furthermore, the performance of DYN3D-TRIDIF is verified by means of a single-assembly problem on pin-cell level. Good agreement between DYN3D-TRIDIF and the detailed-geometry transport reference is achieved with an average deviation in power of less than 1%.

Building a dynamic code to simulate new reactor concepts

Abstract Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.

Passive system reliability analysis using Response Conditioning Method with an application to failure frequency estimation of Decay Heat Removal of PFBR

Innovative nuclear reactor designs include passive means to achieve high reliability in accomplishing safety functions. Functional reliability analyses of passive systems include Monte Carlo sampling of system uncertainties, followed by propagation through mechanistic system models. For complex passive safety systems of high reliability, Monte Carlo simulations using mechanistic codes are computationally expensive and often become prohibitive. Passive system reliability analysis using recently proposed Response Conditioning Method, which incorporates the insights obtained from approximate solutions like response surfaces in simulations to obtain computationally efficient and consistent probability estimates, is presented in this paper. The method is applied to evaluate the reliability of passive Decay Heat Removal (DHR) system of Indian Prototype Fast Breeder Reactor (PFBR). The accuracy as well as efficiency of the method is compared with direct Monte Carlo simulation. The variability of the reliability values is estimated using bootstrap technique. The system abilities, to prevent critical structural damage as well as to ensure operational safety, are quantitatively ascertained. The system functional failure probabilities are integrated with hardware failure probabilities and the inclusion of passive system unreliability in Probabilistic Safety Assessment is demonstrated.

Smart solar reactor for co-production of hydrogen and industrial grade carbon under any weather conditions

Abstract The impending shortage of fossil fuels and environmental consequences of fossil fuel consumption are two of the most imperative problems in the world. This presentation is about a novel design of a solar reactor cavity system composed of a camera-like aperture and moving-wall system to house a unique thermo-fluid-chemical process known as ‘solar cracking’. Differing from typical solar powered Rankine cycles, solar cracking uses concentrated solar energy as a heat source for direct decomposition of natural gas into gaseous H2 and particulate carbon. This process offers a CO2 emission free hydrogen production method, as the carbon is collected in a high-grade and/or nanotube form. However, solar cracking reactors have two implicit major problems: The intrinsic losses in energy conversion efficiency as a result of the high internal temperatures and corresponding re-radiation losses as well as the inherently transient nature of the solar energy. Literature on solar reactors reveals a distinct focus on optimal reactor design for steady state efficiency, but little regarding transient inefficiencies. This presentation provides an advanced perception to solar cracking reactors by presenting you the latest results of our research at Sustainable Energy Research Lab on the design of a ‘smart solar reactor’ that is sensitive to variations in solar flux, and can adjust itself accordingly to maintain quasi-equilibrium internal conditions. A unique system design is presented featuring a solar-flux intensity sensitive aperture that can enlarge the aperture diameter when the flux is low and reduce the diameter when the flux is high. Carbon particle deposition on the reactor window, walls, and at the exit. Carbon deposition. particularly at the exit, causes reactor clogging. There have been many innovative reactor designs aimed to achieve increased conversion efficiencies through novel flows developed at ETH-Zurich, CNRS-France, WIS-Israel, Colorado-USA, Florida-USA, and DLR-Germany. From vortex-flow to tornado, and from fluidized bed to rotating cavity, the designs of these reactors have moved toward the goal of seeking enhanced flow conditions that result in improved overall efficiencies, but have not solved the carbon deposition problem. Our latest research results at Sustainable Energy Research Lab shows that our ‘aero-shielded cyclone solar reactor’ concept provides a laminar flow shield covering the walls as a thin layer flow with a velocity that is strong enough to sweep carbon particles away. This presentation will show you the results of our research on this concept with a 3D animation of the reactor.

Particuology and climate change

The global concern over the greenhouse gas emissions and its effect on global warming and climate change has focused attention on the necessity of carbon dioxide capture and sequestration. There are many processes proposed to capture carbon either before or after combustion and these processes invariably involve investigation and application of traditional particuology. The solids employed are of different sizes, densities, morphologies, and strengths. Their handling, transportation, recirculation, and reactor applications are the essence of ‘particuology’. Particuology can play an important and vital role in achieving cost-effective removal of carbon and minimize emissions of greenhouse gases. In this paper, the existing and developing carbon capture processes are briefly reviewed and the opportunities for application of particuology are identified. The review was not intended to be exhaustive. It is only in sufficient detail to make connection between particuology and climate change. For immediate and future challenges of reducing global warming and carbon capture and sequestration, innovative reactor design and application of particuology is imperative. Expertise and innovation in particuology can greatly enhance the speed of development of those technologies and help to achieve cost-effective implementation. Particuology is indeed intimately related to the climate change and global warming.

Methodology for the Design of Optimal Chemical Reactors Based on the Concept of Elementary Process Functions

In this contribution, a methodology for the optimal design of chemical reactors based on the best reaction route in the thermodynamic state space is proposed. This route is obtained by tracking a fluid element on its way through the reactor and manipulating the fluxes into this element. Instead of choosing the reactor design a priori and optimizing the free parameters of the chosen reactor setup, an innovative reactor design is developed based on the optimal flux profiles. Besides classical reactor concepts, this methodology is suited to investigate the potential of different process intensification options such as integration of reaction, cooling and separation in a single apparatus, or the application of high interface areas for heat and mass transfer. The methodology is exemplarily illustrated for the development of a new SO2 oxidation reactor. The residence time as an example for a meaningful objective is minimized, and a reduction of 69% compared to the optimized technical reference case is achieved.

Experimental investigation on the flow instability behavior of a multi-channel boiling natural circulation loop at low-pressures

Natural circulation as a mode of heat removal is being considered as a prominent passive feature in the innovative nuclear reactor designs, particularly in boiling-water-reactors, due to its simplicity and economy. However, boiling natural circulation system poses many challenges to designer due to occurrence of various kinds of instabilities such as excursive instability, density wave oscillations, flow pattern transition instability, geysering and metastable states in parallel channels. This problem assumes greater significance particularly at low-pressures i.e. during startup, where there is great difference in the properties of two phases. In light of this, a parallel channel loop has been designed and installed that has a geometrical resemblance to the pressure-tube-type boiling-water-reactor, to investigate into the behavior of boiling natural circulation. The loop comprises of four identical parallel channels connected between two common plenums i.e. steam drum and header. The recirculation path is provided by a single downcomer connected between steam drum and header. Experiments have been conducted over a wide range of power and pressures (1–10 bar). Two distinct unstable zones are observed with respect to power i.e. corresponding to low power (Type-I) and high power (Type-II) with a stable zone at intermediate powers. The nature of oscillations in terms of their amplitude and frequency and their evolution for Type-I and Type-II instabilities are studied with respect to the effect of heater power and pressure. This paper discusses the evolution of unstable and stable behavior along with the nature of flow oscillation in the channels and the effect of pressure on it.

Economy of uranium resources in a three-component reactor fleet with mixed thorium/uranium fuel cycles

The potential for minimizing uranium consumption by using a reactor fleet with three different components and mixed thorium/uranium cycles has been investigated with a view to making nuclear power a more sustainable and cleaner means of generating energy. Mass flows of fissile material have been calculated from burnup simulations at the core-equivalent assembly level for each of the three components of the proposed reactor fleet: plutonium extracted from the spent fuel of a standard pressurized water reactor (first component) is converted to 233U in an advanced boiling water reactor (second component) to feed a deficit of multi-recycled 233U needed for the Th/ 233U fuel of the light/heavy water reactor (third component) which has a high breeding ratio. Although the proposed fleet cannot breed its own fuel, we show that it offers the possibility for substantial economy of uranium resources without the need to resort to innovative (and costly) reactor designs. A very high fleet breeding ratio is achieved by using only currently existing water-based reactor technology and we show that such three-component systems will become economically competitive if the uranium price becomes sufficiently high (>300 $/kg). Another major advantage of such systems is a corresponding substantial decrease in production of plutonium and minor actinide waste.

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10 14 n cm −2 s −1, 3.0 × 10 14 n cm −2 s −1, and 2.0 × 10 14 n cm −2 s −1, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10 14 n cm −2 s −1 is achieved. The core life of this more compact version is estimated to be about 23 days.

A new approach to quantitative assessment of reliability of passive systems

The objective of this paper is to show how probabilistic reliability can be assessed for complex systems in the absence of statistical data on their operating experience, based on performance evaluation of the dominant underlying physical processes. The approach is to distinguish between functional and performance probabilities when dealing with the quantification of the overall probability of a system to perform a given function in a given period of time (reliability). In the case of systems where sufficient statistical operating experience data are available, one can focus the quantitative evaluation entirely on the assessment of the functional probability for a given active item (e.g. a pump) by assuming that the specification, layout, construction and installation is such that the item is providing the assigned performance, e.g. in the form of generating the required flow rate. This is how traditional probabilistic safety assessments (PSAs) focus the reliability analysis for the various safety features on the calculation of values for the availability per demand. In contrast, for various systems relevant in advanced technical applications, such as passive safety features in innovative reactor designs, it is essential to evaluate both functional and performance probabilities explicitly and combine the two probabilities later on. This is of course due to the strong reliance of passive safety systems on inherent physical principles. In practice, this means that, for example, in case of a passive cooling system based on natural circulation of a given medium, one has to evaluate and to assess the probability to have a medium condition and a flow rate such that a cladding temperature, represented by a probability distribution, can be hold at a required level. A practical example of this method is given for the case of the reliability assessment of a residual passive heat removal system. General conclusions are drawn regarding reliability estimation of complex, interconnected systems in the absence of statistical performance data, such as for infrastructures.

Advanced SiC/SiC Composite Materials for Fourth Generation Gas Cooled Fast Reactors

As one of the most important breakthrough in the field of SiC/SiC composite materials, the new process called Nano-powder Infiltration and Transient Eutectoid (NITE) Process has been developed. The outstanding total properties of the NITE SiC/SiC composites are presented. Then, the current efforts to make attractive GFR based on the NITE SiC/SiC composites and the technology R &amp; D to make reactor components with the NITE SiC/SiC composites are provided together with our efforts on innovative reactor designs.