Parametric evaluation of large-scale high-temperature electrolysis hydrogen production using different advanced nuclear reactor heat sources

High-temperature electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800-950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an intermediate heat exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies.The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO₂ reactor concepts). Sensitivity studies were then performed with the objective of evaluating the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency of the integrated plant design for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.     

Economic efficiency of power generation units based on a modular helium reactor

Economic aspects and the feasibility of introducing modular helium reactors (MHR) for power generation and for production of hydrogen under competitive market conditions are examined. A dynamic balance model is used, which makes it possible to estimate the equilibrium (market) costs of electrical energy and hydrogen, rates of growth of production, and the characteristics of the resulting profit. It is shown that using a gas-turbine modular helium reactor for production of electrical energy is clearly economically efficient for the current price of natural gas. The efficient production of hydrogen in a MHR-fuel system is possible only for significantly higher natural gas prices, which have already been reached in the world market. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 291–296, November, 2008.     

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.     

Methane conversion using a dielectric barrier discharge reactor at atmospheric pressure for hydrogen production

In the present paper,we carried out a theoretical study of dielectric barrier discharge (DBD) filled with pure methane gas.The homogeneous discharge model used in this work includes a plasma chemistry unit,an electrical circuit,and the Boltzmann equation.The model was applied to the case of a sinusoidal voltage at a period frequency of 50 kHz and under a gas pressure of 600 Torr.We investigated the temporal variation of electrical and kinetic discharge parameters such as plasma and dielectric voltages,the discharge current density,electric field,deposited power density,and the species concentration.We also checked the physical model validity by comparing its results with experimental work.According to the results discussed herein,the dielectric capacitance is the parameter that has the greatest effect on the methane conversion and H2/CH4 ratio.This work enriches the knowledge for the improvement of DBD for CH4 conversion and hydrogen production.     

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.     

Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

The design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346?°C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.     

Investigation of a gas turbine-modular helium reactor using reactor grade plutonium with 232Th and 238U

Utilization of natural uranium (nat-U) and thorium as fertile fuels has been investigated by in a Gas Turbine – Modular Helium Reactor (GTMHR) using reactor grade plutonium as driver fuel. A neutronic analysis for the full core reactor was performed by using MCNP5 with ENDF/B-VI cross-section library. Different mixture ratios were tested in order to find the appropriate mixture ratio of fertile and fissile fuel particles that gives a comparable k_eff value of the reference uranium fuel. Time dependent calculations were performed by using MONTEBURN2.0 with ORIGEN2.2 for each selected mixture. Different parameters (operation time, burnup value, fissile isotope change, etc.) were subject of performance comparison. The operation time and burnup values were close to each other with nat-U and thorium, namely 3205 days and 176 GWd/MTU for the former and 3175 days 181 GWd/MTU for the latter fertile fuel. In addition, the fissile isotope amount changed from initially 6940.1 kg–4579.2 kg at the end of its operation time for nat-U. These values were obtained for thorium as 6603.3 kg–4250.2 kg, respectively.     

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.     

聚变高温制氢反应堆锂铅包层结构力学分析

在高温液态锂铅包层结构设计、热工水力学设计和中子学计算基础上,建立包层的三维有限元分析模型,应用商用有限元软件ANSYS对高温液态包层进行热-力结构耦合分析与应力评定。经计算第一壁材料ODS RAFM钢最高温度635℃,最大Von Mises应力379 MPa;包层结构材料RAFM钢最高温度508℃,最大Von Mises应力175 MPa;FCI材料最高温度950℃,最大Von Mises应力218 MPa。初步的分析结果表明结构设计方案是合理、可行的。     

CANDU reactor core simulations using fully coupled DRAGON and DONJON calculations

The operating CANDU-6 reactors are refueled on-power to compensate for the reactivity loss due to fuel burnup. In order to predict the core behavior, fuel bundle burnups and local parameter information need to be tracked. The history-based approach has been developed to follow local parameter as well as history effect in CANDU reactors.     

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima.     

Adapting the deep burn in-core fuel management strategy for the gas turbine – modular helium reactor to a uranium–thorium fuel

In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine – modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium–thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: ²³⁵U, which represents the 20% of the fresh uranium, ²³³U, which is produced by the transmutation of fertile ²³²Th, and ²³⁹Pu, which is produced by the transmutation of fertile ²³⁸U. In order to compensate the depletion of ²³⁵U with the breeding of ²³³U and ²³⁹Pu, the quantity of fertile nuclides must be much larger than that one of ²³⁵U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of ²³⁵U. At the same time, the amount of ²³⁵U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k_eff and mass evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium–thorium fuel.     

Simple formula to evaluate helium production amount in fast reactor MA-containing MOX fuel and its accuracy

A simple formula is developed for the evaluation of the helium production amount in fast reactor minor actinide (MA) containing uranium–plutonium mixed oxide (MOX) fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, it is found that the formula evaluates the helium production amount with the difference of less than about 10% from the detailed calculation and the PIE results, when the MA isotope content is less than 5 wt.%. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.     

Project of the GT-MHR high-temperature helium reactor with gas turbine

In February 1995, MINATOM of Russia and General Atomics (USA) signed the Agreement for the development and design of the GT-MHR facility with a modular helium reactor and a gas turbine intended to be constructed in Russia. This Agreement was subsequently expanded by the participation of Framatom and Fuji Electric. The GT-MHR facility is designed for burning weapons-grade plutonium and utilization of the heat produced in the direct gas-turbine cycle with electricity production efficiency of about 50%. In future such facilities with uranium fuel will be proposed for use as commercial NPPs. A GT-MHR prototype and fuel production facility are intended to be constructed at the Siberian Chemical Combine in Seversk (Tomsk-7). In accordance with the Agreement, a conceptual design of the GT-MHR should be developed in September 1997. As a part of the conceptual design, a reactor module with a power conversion system is being designed and plutonium fuel is being developed.     

Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.     

Mass spectrometers for determining the isotopic composition of hydrogen and helium

A. A. Bochvar All-Union Scientific-Research Institute of Standardization in Machine Building. Translated from Atomnaya énergiya, Vol. 76, No. 2, pp. 120–124, February, 1994.