Sodium-cooled Reactor Research Articles

Assessing boronized and aluminized thermal diffusion coatings in molten chloride salt and molten sodium environments

The service of structural materials, such as steels and alloys, in molten salt and sodium-cooled reactors in nuclear power generation is associated with high temperature corrosion of these materials, which leads to degradation and failures of structural components. To prevent these issues and to extend the reactor components' service life, protective coatings on steels need to be employed. In the present work, thermal diffusion protective coatings based on borides and aluminides have been tested, for the first time, in molten Na-K-Mg-chlorides of eutectic composition and in molten sodium at 500 °C in high-purity Ar glovebox where the test conditions simulated molten materials environments in nuclear reactors. The structural and compositional examination of the proposed protective coatings revealed only their minimal changing and deterioration in the corrosive environments. Thus, no or minimal formation of the corrosive scale on the top of the coatings was observed with neither structure transformation, nor reduction of the protective layer after 30-days testing in the molten materials' environments. Furthermore, no penetration of the corrosive media to the steel substrates occurred. Due to promising test results, the selected protective coatings may be highly recommended for further long-term and higher temperature testing in modeling and actual molten salts and molten sodium conditions associated with the service in nuclear reactors.

High-temperature creep and corrosion behavior of 316LN stainless steel in oxygen-saturated sodium

The high-temperature performance and sodium corrosion of in-core components are the main problems of liquid sodium-cooled reactors, and low-carbon nitrogen-controlled 316 stainless steel (316LN) is the main research and development direction of the main materials for the core of sodium-cooled fast reactors. In this work, the creep behavior of 316LN in oxygen-saturated liquid sodium at high temperatures (650 ℃, 700 ℃ and 750 ℃) was tested and compared with the results of creep tests in air at the same conditions. It is found that the creep life in the sodium environment is reduced by 63.44%, 67.93% and 40.07%, respectively. The specimens in oxygen-saturated sodium exhibit reduction in ductility, selective dissolution as well as oxidative corrosion. The corrosion mechanism of 316LN under external loading in oxygen-saturated sodium is illustrated.

Multiphysics and multiscale simulation for an experimental sodium-cooled fast reactor

This work aims to explore a new upscaling methodology to describe heat transfer simultaneously coupled with the neutronic process, in the core of a sodium-cooled reactor. The upscaling method, as demonstrated in this work, allows the development of models that describe reactor-scale phenomena, starting at the scale of a cell, composed of a nuclear fuel pin and liquid metal. In principle, this methodology allows high fidelity in the description of these multiphysics and multiscale phenomena. The main result is the two-upscaled energy no-equilibrium model to describe the temperature distribution of the fuel and coolant. These upscale models are described in terms of the upscaled heat transfer coefficient, which is presented and developed in this work, for the Experimental Sodium-cooled Fast Reactor. The use of the new proposed models/methodology leads good results within lower computational times compared with CFD codes, which is useful when computations have to be done a large number of times, for instance for sensitivity analyses to parameter choices in the conception phase of the reactor, and other applications such as plant analyzers, and for operator training in full-range simulators.

Validation of Pronghorn’s subchannel code using EBR-II shutdown heat removal tests: SHRT-17 and SHRT-45R

Pronghorn-Subchannel, referred to as Pronghorn-SC throughout this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Pronghorn-SC was initially designed to model flows in water-cooled, square-lattice, bare subassemblies. Its capability has been extended to model flows in liquid-metal-cooled, triangular-lattice, wire-wrapped subassemblies. To ensure the accuracy of Pronghorn-SC in predicting the behavior of liquid sodium-cooled reactors, the code was validated by comparing calculations with experimental data, obtained from the experimental breeder reactor (EBR-II), Shutdown Heat Removal Tests (SHRT) 17 and 45R. The steady-state calculation at the beginning of the transients was validated using temperature measurements taken at different axial elevations in the instrumented subassembly XX09. The validation exercise was performed in successive stages. First, a comparison between the measured temperature profiles and standalone Pronghorn-SC simulations using a uniform pin power profile was made. The pin power profile was then refined using a Serpent-2 model of the reactor core. Finally, the radial temperature profile was further corrected considering the inter-assembly heat transfer. A Pronghorn Finite-Volume (FV) thermal hydraulic simulation of XX09 and its six neighboring subassemblies; calculated the heat flux on the inner duct surface of the XX09 subassembly to inform the Pronghorn-SC model. Last, a transient validation calculated the peak temperature evolution during the SHRT tests.

Simulation of Melt Behavior in the Sodium-Cooled Reactor Core Catcher Using the EUCLID/V2 Integrated Computer Code HEFEST-FR Module

Simulation of Melt Behavior in the Sodium-Cooled Reactor Core Catcher Using the EUCLID/V2 Integrated Computer Code HEFEST-FR Module

Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison

The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.

Machine learning informed visco-plastic model for the cyclic relaxation of 316H stainless steel at 550 °C

Among the structural alloys for this fast reactor, 316H stainless steel has emerged as a promising candidate. Because the operating temperature of Sodium-cooled reactor is specifically designed to be 550 °C, this operating temperature triggers material inelastic behavior depends more on the coupling of fatigue and creep, which complicates the constitutive model. By introducing static recovery terms, previous studies could capture some experimental features, but failed to describe the interaction by fatigue and creep. In this work, in order to describe the fatigue and creep during cyclic relaxation of 316H stainless steel at 550 °C, we propose a modified visco-plastic constitutive model within the framework of unified Chaboche model. In the proposed model, the parameters related to the static recovery items are coupled, and thus cannot be identified from experiments using the traditional trial and error. To address this issue, we employed the Bayesian approach to identify these parameters. The parameter identification involves two steps: (i) constructing a Gaussian Process surrogate model using data generated from the finite element method, and (ii) obtaining the value of parameters through Markov Chain Monte Carlo sampling under the Bayesian framework. The proposed procedure, is demonstrated by the using experimental results of 316H stainless steel at 550 °C. Under the coupling of fatigue-creep, the material exhibits a cyclic-dependent accelerated stress relaxation before reaching the saturated stage and a steady state of relaxed stress after a long holding time. These mechanical responses are well predicted by the proposed model. Further, we conducted two kinds of multi-axial cyclic test, tensile test of notched bar and coupled tensile-torsion test, to validate the proposed constitutive model for the cyclic behavior under the multi-axial stress state.

Structural materials selection for the sodium-cooled reactor plant steam generator

Steels and alloys of various structural classes are considered from the point of view of the required level of service characteristics necessary for the safe operation and manufacture of a sodium-cooled reactor plant steam generator. After the analysis, it was found that the optimal structural material for the steam generator of a high-power reactor plant with a sodium coolant is martensitic stainless steel with a chromium content of 12 wt.%.

A Study on Creep-Fatigue Evaluation of Nuclear Cladded Components by ASME-III Division 5

In this paper, a study on creep-fatigue evaluation on the cladded nuclear component subjecting to low pressure and high temperature services is carried out. To do this, the codes and standards presented by ASME-III Division 5 are reviewed, and a detailed evaluation procedure is presented step by step for practical applications. As an example of practical design application, a molten salt reactor vessel with a cladding thickness of 10% of the base material thickness is designed and four representative operation cycle types are established. The stress cycle types based on finite element stress analysis are determined from the operation cycle types having coolant temperature and pressure time history loads, and results of the creep-fatigue evaluation are described step by step according to the evaluation procedure. From the result of the creep-fatigue evaluation, it is found that the creep-fatigue evaluation for reactors such as molten salt reactor, sodium-cooled reactor, and so on, which are operated at low pressure and high temperature, is dominated by thermal loads. In this study, the effects of the cladding material and the thermal stresses on the creep-fatigue evaluation are investigated. In addition, as one of the design options to reduce the thermal stresses, the thickness of the exampled vessel is reduced, and the calculated creep-fatigue values are compared with the acceptance creep-fatigue envelope criteria of the ASME-III Division 5.

Symmetric Heat Transfer Pattern of Fuel Assembly Subchannels in a Sodium-Cooled Fast Reactor

The method outlined in this paper is convenient and effective for studying the thermal performance of fuel assemblies cooled with sodium fast reactors using the subchannel procedure. To initially study an optimization model for a symmetric single fuel assembly heat transfer pattern analysis in a fast sodium-cooled reactor based on subchannel calculations, this paper innovatively proposes a subchannel heat transfer analysis method with the entransy dissipation theory, which can solve the limitations and inaccuracies of the traditional entropy method such as poor applicability for heat transfer processes without functional conversion and the paradox of entropy production of heat exchangers. The symmetric distributions of the thermal-hydraulic parameters such as coolant flow rate, coolant temperature, cladding temperature, and fuel pellet temperature were calculated, and the entransy dissipation calculation method corresponding to the fuel assembly subchannels was derived based on the entransy theory. The effect of subchannel differences on the thermal-hydraulic parameters and the symmetric distribution pattern of entransy dissipation during the cooling process of the fuel assembly was analyzed and compared from the symmetrical arrangement of subchannels in the axial and radial directions.

Corrosion resistance of 12% chrome steel under the operation conditions of a steam generator of a reactor plant with sodium coolant

The influence of an aqueous medium and superheated steam on the corrosion resistance and resistance to corrosion-mechanical destruction of 07Kh12NMFB steel in various operating modes of a steam generator of a promising high-power sodium-cooled reactor plant has been studied. Steel of this grade meets the requirements for the operation of heat exchange pipes and vessel elements of direct-flow steam generators of a reactor plant in terms of corrosion resistance and corrosion-mechanical strength.

Sodium-cooled Small Modular Reactors

Amid expectations for nuclear power as a carbon-free energy source are growing to prepare for climate change, deploying efforts for small modular reactors (SMRs) are being actively carried out for the application area to which existing large nuclear power plants are not proper to contribute. Sodium-cooled SMRs are also being developed as one of the candidates to lead the SMR market in the future. This manuscript introduces the background and history of the development of sodium-cooled fast reactors, and the current status of development of Sodium-cooled SMRs and microreactors that are being developed based on them.

ARKADIA—For the Innovation of Advanced Nuclear Reactor Design

Abstract This paper describes the outline and development plan for “Advanced Reactor Knowledge- and Artificial Intelligence (AI)-aided design integration approach through the whole plant lifecycle (ARKADIA),” which the Japan Atomic Energy Agency has begun developing to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize AI-aided integrated numerical analysis to offer the best possible solutions for any challenge that could arise in the design and operation of a nuclear plant, including optimization of safety equipment as well as structures, systems, and components. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-design and ARKADIA-safety will be constructed individually, with the first target of a sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity that is technology inclusive and applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.

Numerical Investigation of Special Heat Transfer Phenomenon in Wire-Wrapped Fuel Rod of SFR.

Sodium-cooled reactors (SFR) have always been recognized as one of the most promising candidates for the fourth-generation nuclear systems as announced by the Generation-IV International Forum. In the design of SFR, helical wire-wrapped rod is applied to stabilize the structure of the rod bundle and enhance coolant mixing. Although there has been considerable research on SFR in computational fluid dynamics (CFD), the phenomenon of heat transfer has rarely been paid attention to. This article discovered that there exists reversed heat flux from coolant to wrapped wire, which is contrary to our usual understanding. This phenomenon has not been reported in previous CFD calculations. Hence, a solid heat conduction model is proposed to prove this phenomenon and analyze the heat transfer process. The simulation results show that the wrapping wire embedding depth, the shape of the calculation domain and the physical properties of all components have great influence on the magnitude of the reversed heat flux. The present findings will have strong influence on the temperature field and maximum value of the fuel rod as well as profound reference value for future flow calculation, especially in grid generation and treatment of the junction between the winding wire and fuel rod.

Possibility of simulating natural circulation in fast neutron reactors using a light water test facility

The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. The sodium-cooled BN-1200 reactor was selected as the reactor installation to be modeled. The development of novel reactor designs must be based on the results of experimental studies. Some problems of modeling thermohydraulic processes in BN type reactors are studied by using sodium test facilities. Experimental studies of natural convection processes using light water test facilities can be considered as a good alternative to those using sodium test facilities. To validate the model, the similarity theory and the “black box” method were used and their principles and applicability were analyzed. Using the “black box” method makes it possible to avoid detailed modeling of such components as the reactor core and heat exchangers, replacing them by a simplified representation of these components to simulate the integral characteristics of the existing real life equipment. The paper considers the basic criteria which determine the similarity of the thermohydraulic processes under study. The governing criteria of similarity were estimated based on the fundamental differential equations of natural convection heat transfer. Based on these criteria, a set of dimensionless values was obtained which show the correlation between the model parameters and the characteristics of the reactor facility. Besides, generalized relationships were derived which can be used to estimate the scaling factors for calculating the key values of the reactor facility based on the model parameters. These relationships depend on the thermal-physics parameters of the working fluids, the geometrical scale value and the ratio of the thermal power of the model to that of the reactor facility, i.e., model-to-reactor thermal power ratio. The conditions under which it is possible to model sodium coolant by light water with adequate accuracy were analyzed. An example is given of the numerical values of the scaling factors for one of the reference light water test facilities. The paper uses the experience of a number of foreign researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. According to the available estimates, the assumptions used do not result in considerable losses in accuracy. Thus, the natural circulation of the sodium coolant in the upper plenum of the fast-neutron reactor can be simulated with adequate accuracy by using light water test facilities.

System levelized fuel cost of electricity generation in a two-component nuclear energy system with a closed uranium-plutonium NFC

The authors propose an approach to the calculation of the levelized unit fuel cost (LUFC) of electricity generation for a fast reactor in a two-component nuclear energy system (NES) with regard for plutonium production. The approach is based on taking into account the additional economic effect, which can be achieved through the sale at the market price of the natural uranium released due to the substitution of thermal reactors by fast reactors with MOX fuel based on the plutonium bred in a fast reactor. This requires considering simultaneously the reactor parts of the fuel cycle for fast and thermal reactors. Relationships have been obtained which connect the key neutronic and fuel characteristics with the NPP and fuel cycle economic performance. The described methodology was used for the computational study of the LUFC for a fast sodium-cooled reactor. Calculations have shown that, in the considered case, taking into account the plutonium production leads to the LUFC reduction by nearly half and, therefore, to a major decrease in the total unit cost of electricity generation (levelized cost of electricity or LCOE).

Reassessment of Computational Tools for the Modeling of Heat Transfer in a Molten UO2 Pool in Natural Convection

Abstract This work, performed within the European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) H2020 European project, is part of a larger framework intending to reassess the modeling of heat transfer in molten pools on SCARABEE available experimental results. This paper presents simulation results of the in-pile Bain Fondu (BF1) test, performed within the SCARABEE-N program, using accident source term evaluation code (ASTEC), Sn method implicit multifiel multiphase Eulerien recriticality (SIMMER) III, and SIMMER V simulation tools as well as comparison with its available experimental data. This program was performed in the 1980s in the frame of the Safety Assessment studies of Superphenix sodium-cooled reactor. This test was dedicated to verify the stability of a pure molten UO2 pool under decay heat conditions within natural convection and the long-term resilience of the peripheral fuel crust. The pool was generated in a stainless steel crucible by a progressive heating of a fuel pellet stack through six successive power plateaus. For the benchmark purposes, only the molten pool behavior k the last power plateau, where the pool was the largest and the fuel temperatures the highest, was investigated. Experimental data such as the axial profile of radial heat fluxes and heat transfer from the pool to the surrounding interassembly coolant as well as the axial profile of the peripheral fuel crust thickness were used for the reassessment of the simulation tools. In addition, other variables of interest not measured during the test, such as the radial and axial velocities in the pool, were also benchmarked. Finally, a critical analysis of the correlations and models used in the different simulation tools for the BF1 test modeling is also provided in the paper.

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Abstract Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

Evolution of a low-alloy steel/nickel superalloy dissimilar metal weld during post-weld heat treatment

The design of the dissimilar metal weld investigated here is aimed at applications in the steam generator of a sodium-cooled nuclear reactor, with a multi-decade lifespan in demanding operational conditions. It consists in a narrow-gap joint between 2.25Cr-1Mo low-alloy steel and an austenitic alloy using a nickel-based alloy 82 as filler material. This study focuses on understanding the microstructural and micromechanical evolution in the near fusion boundary region between the low-alloy steel and the nickel alloy filler metal during post-weld heat treatment, using notably electron probe micro-analysis and nano-indentation. The difference in matrix phase and chemical composition between the two alloys leads to a large difference in chemical potential for carbon, which is mobile at the post-weld heat treatment temperature. A number of fine-scale characterization techniques were used to assess the gradient of composition, hardness, and microstructures across the fusion boundary, both as-welded and after post-weld heat treatment. This complete analysis permits to highlight and understand the main microstructural and micromechanical changes occurring during post-weld heat treatment and opens the way to their long term study in service conditions.