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The corrosion behavior and mechanism of 316 stainless steel (316SS) immersed in liquid Pb were investigated in Ar at 650 and 750 °C, respectively. The oxidation corrosion first dominated at 650 °C. The outer layer of the corroded 316SS surface was the loose iron oxide layer, and the inner layer was dense Fe-Cr spinel. However, the selective dissolution of Ni and Cr was dominant at first at 750 °C, and the O diffused to the interface increased with time, then an inner oxide layer is formed with Ni, Cr, and Fe in inner layer of the matrix. The sizes of pores formed by the dissolution gradually increased with the corrosion time increase, and the dissolved area was also partially transformed into the whole. The selective dissolution corrosion of the corroded 316SS was dominant at 750 °C with the time increase. The interaction mechanism between the liquid lead and 316SS was revealed. [ABSTRACT FROM AUTHOR]

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

225Ac is a promising nuclide for targeted alpha therapy. At present, the nuclide is mainly extracted from 229Th generators, but its total radioactivity of the world is only 63 GBq/y. Therefore, it is unclear whether this method alone can meet future global medical demand. In this study, we analyzed the production amount of 225Ac and its uncertainty due to the neutron flux and nuclear reaction cross-sections while analyzing the generation process of 225Ac using the experimental fast reactor Joyo operated by the Japan Atomic Energy Agency (JAEA). The analysis showed that irradiating 1 g of 226Ra for 45 d at the core center of Joyo can produce 15.4 ± 6.2 GBq of 225Ac. Therefore, with six irradiations per year (45 d × 6 = 270 d/year), Joyo can produce 55.3 to 129.1 GBq of 225Ac, equivalent to 86 to 200% of the current world supply. [ABSTRACT FROM AUTHOR]

Fast reactor plays an important role in the sustainable development of nuclear energy, and it is a key reactor type that be committed to develop all over the world. The reactor physics calculation is the basis of fast reactor core design, and the research of related methods and the development of corresponding computing codes are very important in the R&D of fast reactor theory and numerical simulation technology. In this paper, the developments of reactor physics analysis methods for the fast reactors, especially the developments in the past 20 years are reviewed. By reviewing the dedicated and general computing codes developed at home and abroad for the fast reactors, the progress of fast reactor physics analysis theory is introduced. Inside, the consistency among different methods and the trend of the developments in recent years are discussed. Furthermore, the suggestions for the development of fast reactor physics analysis methods and for the progress of independent software development in our country are provide.

The deterministic two-step method for the fast reactor neutronics calculation, composed of cross-section homogenization and diffusion or transport core calculations, was widely applied in the fast reactor engineering design and analysis field. The homogenized cross-section calculation method based on Monte Carlo with continuous-energy and fine geometry can provide high-precision cross-sections for advanced fast reactors. The current development status and trends of coupling Monte Carlo-generated homogenized cross-sections with diffusion and transport core calculations were briefly reviewed in this paper. The methods discussed include the Monte Carlo flux-volume homogenization method, the superhomogenization equivalence technique (SPH), and the Monte Carlo flux-moment homogenization method (MHT). The MET-1000 metal fuel fast reactor is used as a benchmark. The SPH equivalent techniques are widely used to preserve the reaction rates of a reference heterogeneous model and a homogenous model. In this paper, the SPH was applied to the control rods' cross-section address to improve the diffusion core calculations. This equivalence technique reduces the overestimation of the control rod worth using the diffusion core solver from 13.5% to 0.35% and improves power distribution prediction accuracy. With SPH correction, the MC/diffusion in this work exhibits about

As the main type of fast reactor in the future development of nuclear energy, neutron calculation is very important for core design and safety analysis. The fast spectrum critical experimental facility is mainly used for the study of fast spectrum core design, core neutron behavior and nuclear data acquisition. At present, there is some progress in the research of fast spectrum critical experimental facility at home and abroad, but there is little research on the exact neutron simulation of fast spectrum critical experimental facility with square component plate fuel. The fast spectrum and few group cross section calculation software TULIP and the few group transport calculation software LAVENDER were developed by the Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University and integrated into LoongSARAX. LoongSARAX can be used to study its square assembly plate fuel fast spectrum critical experimental facility. LoongSARAX is a neutronics calculation program system developed by Xi'an Jiaotong University for liquid metal fast reactors. Based on the benchmark of the SNEAK fast spectrum criticality experimental facility, the modeling calculation analysis and preliminary verification of LoongSARAX were carried out. Firstly, based on the Monte Carlo program OpenMC, the homogenization modeling method of the critical experimental facility with plate fuel was analyzed, and the simplified model suitable for LoongSARAX was determined. Based on the two-step method system of TULIP and LAVENDER in LoongSARAX, the calculation model of the SNEAK criticality experimental facility was established, and the effective multiplication factor of the whole reactor, the reactivity coefficient of the core, the power distribution and the micro-section were analyzed, and the results were compared with OpenMC conducted a comparative analysis. The analysis shows that the calculation results of LoongSARAX for the plate fuel criticality experimental facility are in good agreement with the MC method calculation results in terms of effective proliferation factor and power distribution. Due to the complex structure of the MZA fuel assembly, the maximum deviation of the effective breeder factor after processing is greater than that after SNEAK fuel assembly processing. When the spatial weighted mixing of the axial plate fuel square component is carried out, smaller errors will be introduced for the component with high repeatability and weak non-uniformity, while larger errors will be introduced for the component with strong heterogeneity. The maximum deviation of the calculation results of safety rod and control rod value and the maximum deviation of the calculation results of typical component reactivity are small. At the same time, the calculation results of core power distribution show that the maximum error is within the reasonable range. LoongSARAX is suitable for the plate fuel criticality experimental facility and has the ability to calculate for the plate fuel criticality experimental facility. In addition, LoongSARAX has higher computational efficiency than MC software OpenMC in terms of computation time.

BackgroundThe accuracy of transient thermal hydraulic parameter prediction of reactor cores under various working conditions directly affects reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which are often modeled as time-series prediction problems.PurposeThis study aims to solve the accuracy problem of continuous prediction of core thermal hydraulic parameters under instantaneous conditions and to test the feasibility of a gated cycle unit based on the attention mechanism in core parameter prediction.MethodsThe 1/2 full core model of China Experimental Fast Reactor (CEFR) core was taken as the research object. The subchannel SUBCHANFLOW program was employed to generate the time series of transient core thermal hydraulic parameters. The gated recurrent unit (GRU) model based on soft attention was used to predict the mass flow and temperature time series of the core.ResultsThe results show that, compared with the adaptive radial basis function (RBF) neural network, the GRU network model with soft attention offers better prediction results. The average relative error of temperature is

The loading and unloading elevators are the primary equipment in the refueling system, used for transferring fuel assemblies in the sodium-cooled fast reactors. The guideway friction pairs are the critical components of these elevators in the refueling system. With the excellent hardness and wear resistance in air, nitrided W18Cr4V steel is a promising material for the guideway friction pairs. In order to assess the feasibility of using nitrided W18Cr4V steel, it is essential to understand the aging behavior of nitrided W18Cr4V steel in high-temperature sodium. Aging tests were conducted on nitrided W18Cr4V steel in sodium and in argon environments at various temperatures for different exposure times. The results showed that the nitrogen atoms in the nitrided layer exhibited bidirectional diffusion behavior in the sodium or argon environment at 540 °C. Compared to the argon environment, cracks formed within the nitrided layer and the diffusion of nitrogen into the sodium was accelerated in the nitrided layer. As a significant number of nitrogen atoms had diffused into the sodium, there was little difference in the hardness between nitrided W18Cr4V steel and non-nitrided W18Cr4V steel after long-term exposure to 540 °C sodium. [ABSTRACT FROM AUTHOR]

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Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Application of computation tools resting on contemporary physical and mathematical models for substantiating the design solutions adopted for various heat-transfer equipment components helps save time, manpower, and financial resources of design institutions. The variety of both existing reactors and those being designed, which differ from one another both in design and type of coolants calls for the availability of a versatile thermal hydraulic computer code suited for a wide range of applications. The new-generation HYDRA-IBRAE/LM thermal hydraulic module of the EUCLID integrated code, which has been developed as part of the Proryv (Breakthrough) Project, meets these requirements. The operation of this thermal hydraulic module as part of the integrated code opens the possibility to simulate an essentially wider range of reactor plant operation modes and, as a consequence, those of individual heat-transfer equipment components. The developed thermal hydraulic module, which has been certified at the Scientific and Engineering Center for Nuclear and Radiation Safety (SEC NRS), offers the possibility to analyze the thermal hydraulics of sodium, lead, lead–bismuth, gas, and water coolants in various NPP equipment items. Reactor plant steam generators (SGs) belong to the category of equipment components most complex for modeling since they may contain two types of coolants. The article presents study results demonstrating the code's abilities to analyze in a correct way the processes in the steam generators of only sodium cooled reactor plants, because these plants exist and are actively operated in Russia and around the world. The data presented in the article allow a conclusion to be drawn that the thermal hydraulic module developed at IBRAE RAS is an efficient tool for numerically analyzing complex heat-transfer processes in reactor plants. By using an extended system of closing correlations implemented in the module, it is possible to perform substantiation of design thermal engineering solutions as applied to individual heat-transfer equipment components. [ABSTRACT FROM AUTHOR]

For numerically simulating the melt behavior in the core catcher of a sodium-cooled reactor, the HEFEST-FR module—a software tool based on the SAFR computer code elaborated at the Nuclear Safety Institute, Russian Academy of Sciences, for simulating the meltdown and destruction of liquid metal-cooled fast reactor core components—has been developed and incorporated into the EUCLID/V2 integrated computer code. This module is intended for numerically simulating the melt retention and cooling-down processes in the reactor vessel with taking into account the heat transfer from the vessel's internal structures to the coolant. For this purpose, a 2D-problem (implying that there is no dependence of temperature on the azimuthal angle) of unsteady heat conduction for materials located in the reactor's core catcher tray is solved in a cylindrical coordinate system. The heat-conduction equation coefficients depend on time, coordinates, and temperature, the latter being the solution of equation. Boundary conditions of the first, second, and third kind are used, and heat loss by radiation at the boundary is specified. The total or volumetric decay heat power is given inside of the melt. For numerically solving the 2D heat-conductivity problem, an enthalpy approach based method is applied. The formulations used in the method make it possible to overcome, in a natural manner, the problem relating to discontinuity of material specific melting point enthalpy in solving the heat-conduction problem with taking phase transitions into account. The solution yields the temperature field in melting/freezing of materials (steels of various grades and fuel) in the fast reactor's core catcher tray. The results of verifying the HEFEST-FR module against the solution of an analytical problem have been demonstrated. Using the module, methodical computations of the fuel and fuel pin cladding melt behavior in the sodium-cooled reactor's core catcher have been carried out. [ABSTRACT FROM AUTHOR]

Buckling is a critical failure mode of the Fast Reactor Main Vessel (FRMV) subjected to seismic load. Post-buckling stability of FRMV is a crucial safety issue during excessive earthquakes. Our prior study revealed that global response becomes stable after buckling by phase-inverse mechanism. However, local fatigue cracks can initiate, penetrate and propagate under cyclic load. In the present study, shaking table experiments using cylindrical models are carried out. The post-buckling crack propagation process is experimentally observed and its mechanism is analyzed. It is shown that dominant cracks always propagate in circumference direction along the diagonal of the diamond-shape buckling dimple. Rapid collapse due to unstable crack propagation never occurs. Instead, a stable propagation mode is observed, where crack growth rate declines with crack circumferential angle. This stability is owing to the displacement-controlled characteristic of dynamic load, which results from the increasing frequency ratio due to crack propagation. Furthermore, a simplified evaluation method based on the estimation of J-integral under displacement-controlled condition is applied to predict the crack growth rate. The comparison with experimental data shows a satisfactory agreement. As a complementary study on global stability of FRMV, the present study confirms a local stability mechanism after buckling, contributing to a more comprehensive understanding on the post-buckling behavior of FRMV.

Rhenium-186 (Re-186) has attracted attention as a medical isotope. The feasibility of producing Re-185, the raw material for Re-186, using a fast reactor was evaluated using a continuous energy Monte Carlo code. The irradiation of natural tantalum (Ta) in the fast reactor can produce Re-185 with an isotopic purity of 99%. A two-step irradiation process with different moderators was found to improve the production rate of Re-185. Specifically, this can be achieved by using zirconium hydride (ZrH1.7) as a moderator in the first transmutation process from natural Ta to tungsten (W), and then zirconium deuteride (ZrD1.7) as a moderator in the second transmutation process from W to Re-185. Due to the two-step irradiation, the production rate of Re-185 from Ta can be increased up to a maximum of 470 times compared with irradiation without a moderator, and 2.3 g of Re-185 can be obtained from 1571 g of Ta in 1 year of irradiation. The proposed isotope production method is a new method that is different from the conventional electromagnetic enrichment process.

How to generate the precise broad group cross section is important for the fast reactor design. In this study, a fast reactor multi-group cross-section generation code MGGC2.0 are developed in-house for processing ultrafine group MATXS format library. Validation and verification are performed for MGGC2.0 code by applying the benchmarks of ICSBEP handbook, and the results of MGGC2.0 agree well with that of MCNP. The consistent PN method with critical buckling search is in good agreement that condensed with TWODANT flux and flux moment for the inner core and outer core region. For the radial blanket and reflector, two region approximation method has been applied in MGGC2.0 by using collision Probability Method neutron flux solver. The RBEC-M benchmark was used to verify the power distribution calculation, and the relative error of power distribution comparison with the reference are less than 0.8% in the fuel region and the maximum relative error is 5.58% in the reflector region. Therefore, the precise broad cross section can be generated by MGGC2.0 for fast reactor.

The development of fast reactors with complex designs and operation status requires more accurate and effective simulation. The Monte-Carlo method can generate multi-group cross-sections in arbitrary geometry without approximation on resonances treatment and leads to good results in combination with diffusion codes. However, in previous studies, the coupling of Monte-Carlo generated multi-group cross-sections (MC-MGXS) and transport solvers has shown relatively large biases in fast reactor problems. In this paper, the main contribution to the biases is proved to be the neglect of the angle-dependence of the total cross-sections. The flux-moment homogenization technique (MHT) is proposed to take into account this dependence. In this method, the angular dependence is attributed to the transfer cross-sections, keeping an independent form for the total sections. For the MET-1000 benchmark, the multi-group transport simulation results with MC-MGXS generated with MHT are improved by 700 pcm and an additional 120 pcm with higher order scattering. The factors that cause the residual bias are discussed. The core power distribution bias is also significantly reduced when MHT is used. It proves that the MC-MGXS with MHT can be applicable with transport solvers in fast reactor analysis.

In order to reduce Minor Actinides (MA), the transmutation of MA in a fast reactor has been studied in Japan. In the transmutation in the fast reactor, MAs are mixed in the ordinary MOX fuel and they are burned up. For the design of a fast reactor with inherent safety characteristics, it is necessary to secure negative void reactivity and cooling via a natural convection under the situation of the severe accident. Therefore, this study aims to fundamentally design the fast reactor for the MA transmutation with inherent safety characteristics. First, in order to clarify the design range of the core design parameter, the parametric study was conducted using four design parameters determining the core configuration. As the result of this parametric study, the feasible combination of each four parameters satisfying the negative void reactivity was clarified. Among parameters satisfying the negative void reactivity, the condition that results in the maximum reduction of MA amount was obtained. Secondly, the CFD simulation focused on the sodium coolant channel between fuel pins was conducted. In this CFD simulation, as the first step for the design activity, the confirmation of the occurrence condition of a natural convection due to the decay heat after shifting to the transient situation was confirmed. Temperature and velocity distribution were evaluated. And the heat transfer efficiency was summarized using Ra number and Nu number.

In this paper, a single-equation drift-flux type model for gas-lead-bismuth eutectic (LBE) two-phase flow analysis in the LBE fast reactor steam generator tube rupture (SGTR) accident is established. As the area-averaged superficial gas velocity increases, the calculated exponent P in the Bankoff model increases, and the distribution parameter C0 decreases. The reliability of the model is confirmed by comparing the predicted results with a total of 76 sets of open experimental data from a bubble-column case and a gas lift pump case. The relative deviation between the predicted and experimental values of the void fraction is within 20 %. The model proposed in this paper helps to realize the early warning and analysis of SGTR accidents in LBE fast reactors, and can provide theoretical and technical guidance for the safety design and evaluation of LBE fast reactors. [ABSTRACT FROM AUTHOR]

Rhenium-186 (Re-186) has attracted attention as a medical isotope. The feasibility of producing Re-185, the raw material for Re-186, using a fast reactor was evaluated using a continuous energy Monte Carlo code. The irradiation of natural tantalum (Ta) in the fast reactor can produce Re-185 with an isotopic purity of 99%. A two-step irradiation process with different moderators was found to improve the production rate of Re-185. Specifically, this can be achieved by using zirconium hydride (ZrH1.7) as a moderator in the first transmutation process from natural Ta to tungsten (W), and then zirconium deuteride (ZrD1.7) as a moderator in the second transmutation process from W to Re-185. Due to the two-step irradiation, the production rate of Re-185 from Ta can be increased up to a maximum of 470 times compared with irradiation without a moderator, and 2.3 g of Re-185 can be obtained from 1571 g of Ta in 1 year of irradiation. The proposed isotope production method is a new method that is different from the conventional electromagnetic enrichment process. [ABSTRACT FROM AUTHOR]

Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR reactor series. A large scope of R&D has been undertaken, and large-scale nonnuclear experimental facilities are under construction. International Euro-US consortiums for the development of the ALFRED, PLFR and MYRRHA reactors do not expect any unsolvable technical issues either and are currently formulating requirements to the test facilities and candidate materials and technologies required for further activities.

The mechanical consequences of core disruptive accidents (CDAs) are a major safety concern in sodium-cooled fast reactors. Once core disruption occurs, liquefied core materials rapidly disperse vertically and radially. The dispersed materials penetrate the pin bundles and control rod guide tubes (CRGTs) before freezing at the edge of the penetration zone as heat is transferred to surrounding structures. Such freezing phenomena can suppress the negative reactivity feedback of fuel dispersion. The discharge of core materials can be impeded, resulting in a molten core pool formation when tight blockages occur inside CRGTs due to frozen material. Accordingly, freezing phenomena of core materials play a key role in governing the mechanical consequences of a CDA. To validate a freezing model implemented in our CDA analysis code, ASTERIA-SFR, a preliminary simulation of the THEFIS RUN#1 test, was performed. The calculation results show that freezing on the structural wall and crust formation were key phenomena affecting the penetration behavior, and the structural heat transfer is an important parameter. A remarkable reduction of the heat transfer coefficient was required to reproduce the penetration length observed in the experiment. This suggests that the momentum exchange and flow regime at the leading edge as well as heat transfer should be well modeled to predict the freezing phenomena in rapidly evolving CDAs.

The loading and unloading elevators are the primary equipment in the refueling system, used for transferring fuel assemblies in the sodium-cooled fast reactors. The guideway friction pairs are the critical components of these elevators in the refueling system. With the excellent hardness and wear resistance in air, nitrided W18Cr4V steel is a promising material for the guideway friction pairs. In order to assess the feasibility of using nitrided W18Cr4V steel, it is essential to understand the aging behavior of nitrided W18Cr4V steel in high-temperature sodium. Aging tests were conducted on nitrided W18Cr4V steel in sodium and in argon environments at various temperatures for different exposure times. The results showed that the nitrogen atoms in the nitrided layer exhibited bidirectional diffusion behavior in the sodium or argon environment at 540 °C. Compared to the argon environment, cracks formed within the nitrided layer and the diffusion of nitrogen into the sodium was accelerated in the nitrided layer. As a significant number of nitrogen atoms had diffused into the sodium, there was little difference in the hardness between nitrided W18Cr4V steel and non-nitrided W18Cr4V steel after long-term exposure to 540 °C sodium.

This paper deals with a conceptual study on a plutonium (Pu) and minor actinide (MA) burning fast reactor core for the distant future phaseout of a fast reactor fuel cycle after it is commercialized and used for a long time. This burning core aims to reduce the Pu and MA inventories contained in the fuel cycle through multiple recycling. A key point for the core design is the degradation of Pu and MAs during multiple recycling. This degradation affects the feasibility of the nuclear design by increasing the sodium void reactivity and decreasing the absolute value of the Doppler constant. A feasible core concept was found by incorporating the following three factors to improve the reactivity coefficients: core geometry flattening, fuel burnup reduction, and use of silicon carbide (SiC) in the cladding and wrapper tubes. Notably, softening the neutron spectrum using the SiC structural material not only improved the reactivity coefficients but also indirectly mitigated the degradation of Pu and MAs. Consequently, the designed core allowed for multiple recycling to continue until the Pu and MAs reduced significantly, particularly by about 99% in a phaseout scenario starting from a fast reactor fleet of 30-GWe nuclear power capacity. Fast reactors were found to have the potential to become self-contained energy systems that can minimize the inventories of Pu they produced themselves, as well as long-lived MAs. Fast reactors can be among the important options for environmental burden reduction in the future. [ABSTRACT FROM AUTHOR]

A transition to a two-component nuclear power structure with a reactor fleet consisting of thermal and fast reactors as envisioned in the Russian nuclear power development strategy to 2050 and outlook to 2100 will require optimal spent nuclear fuel and radioactive waste management solutions. A core issue in this regard is managing the long-lived minor actinide (MA) inventory that affects overall nuclear power ecological safety. The study examines several options for homogenous MA (Am and Np) transmutation using modern calculation codes with MA transmutation rate and material balances taken into account. Results demonstrate that if fast reactor installed capacity reaches 92 GWe by 2100 there would not be any need for dedicated MA-burners as the MA issue would be gradually resolved within the two-component nuclear energy system by the end of the century.

The article presents the results of work aimed at solving the problem of measuring the coolant level in miscellaneous tanks of liquid-metal-cooled reactor plants, mainly of an integral layout with a free level of the primary coolant. The choice of relevant measuring means and methods is limited by the extreme parameters of the liquid metal coolant (LMC) and operating conditions. Traditional measuring means are practically unsuitable; therefore, measuring the HLMC level is a complex technical task. Based on this review, they propose and describe a method of pulsed microwave reflectometry as the most promising in terms of combining the characteristics of reliability, accuracy and ease of use. The results of the experimental study demonstrated the efficiency of the level gauge, which worked according to this method, for measuring the level of lead-bismuth coolant in the control tank under conditions close to natural ones. An analysis of the results confirmed the possibility of using this method to control the level of melts of various metals as applied to HLMC reactor plants. Using the device for measuring the level, which works according to the proposed method, it is possible to control the level of melt of various metals in tanks in real time without the need to move various parts of the sensitive element of the level gauge while maintaining the tightness of the circuit. This device is applicable for various nuclear power plants, accelerator-controlled systems, research reactors and experimental facilities with liquid metal coolants.

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of keff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

Austenitic stainless steels are selected as candidate materials for in-core and out-of-core components of Generation-IV fast reactors due to their excellent operating experience in light-water reactors over several decades. However, the performance of conventional austenitic stainless steels proves to be inadequate through operation feedback in fast reactors. To withstand the demands for material performance exposure to the extreme operating environment of fast reactors, modified austenitic stainless steels for in-core and out-of-core components have been developed from the first-generation 300-series steels. The design of an appropriate microstructure becomes a top priority for improving material performance, and key metallurgical features including δ-ferrite content, grain size and secondary phase precipitation pertinent to austenitic stainless steel are focused on in this paper. δ-ferrite content and grain size are closely correlated with the fabrication program and their effects on mechanical properties, especially creep and fatigue properties are critically assessed. Moreover, the impacts of some major elements including nitrogen, stabilization elements (Nb, Ti, V), phosphorus and boron on secondary phase precipitation behaviors during aging or creep are reviewed in detail. Based on the role of the aforementioned metallurgical features, the recommended specification of nitrogen content, stabilization ratio, phosphorus content, boron content, δ-ferrite content and grain size are put forward to guarantee the best-expected performance, which could provide reactors designers with attractive options to optimize fast reactor systems. [ABSTRACT FROM AUTHOR]

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in keff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

This article addresses the development of approaches to numerically analyzing the processes of interaction between liquid metal sodium coolant and destructed fuel-pin components (fuel and steel in solid and liquid states). Such processes may occur during a severe accident involving core destruction, and also when fuel-pin components (fuel or cladding) heated to a high temperature release into the flow of relatively cold liquid coolant or when the molten fuel begins to melt the corium catcher. A dramatic growth of power caused by self motion of pins and stoppage of forced coolant circulation without actuation of the reactor plant's active and passive safety systems are among possible events leading to accidents with such consequences. For simulating the thermal interaction, it is proposed to use a multicomponent thermally nonequilibrium model based on the solution of a system of mass, energy, and momentum conservation equations with the relevant relationships that take into account the specific features of thermal and mechanical interaction between the melt and coolant. Simulation of the processes is very important for determining pressure jumps in the reactor plant caused by release of destructed fuel-pin components into the coolant flow. Thermal interaction of fuel-pin components with the coolant may cause intense coolant evaporation and, as a consequence, the occurrence of drastic pressure jumps determined by the intensity of heat transfer from components to coolant and the amount of vapor produced. To find the rate of heat transfer between various components, a chart of heat-transfer modes and closing relationships corresponding to each mode are used. [ABSTRACT FROM AUTHOR]

The mission of the U.S. Department of Energy (DOE) Office of Nuclear Energy is to advance nuclear power to meet the nation's energy, environmental, and national security needs. Advanced nuclear technology development, as well as support for the current nuclear power industry, requires a robust infrastructure for experimentation, testing, design evolution, and component qualification. The current lack of fast neutron spectrum testing capabilities has been identified as a significant gap in the U.S. infrastructure that impedes the development of next-generation nuclear reactors—many of which require a fast neutron spectrum for operation—and equally impacts the United States' ability to regain global technology leadership in this arena. To close the gap and support advanced technology development, the DOE has established the Versatile Test Reactor (VTR) project to provide high-performance testing capability, specifically, a fast neutron source to develop, test, and qualify advanced fuels and materials for the next generation of advanced reactors and existing commercial reactors. This paper describes the establishment of the project, the identification of its mission and requirements, and the design approach and status. [ABSTRACT FROM AUTHOR]

The Fuel Performance Design Basis for the Versatile Test Reactor begins with requirements to maintain safe and efficient reactor operation. For the metal-fueled Versatile Test Reactor, this means a limited number of fuel rod breaches, no fuel melting under steady-state operation and anticipated transients, and continuity of the fuel rod and assembly configuration to avoid impacts to operations of safety systems, maintain expected coolant flow, and allow for efficient fuel handling. Using a large database gathered from previous testing, data were examined to identify and establish preliminary limits on fuel operating conditions. Fuel performance aspects important to fuel operating limits have been identified, including cladding creep, which is addressed with a cladding deformation limit to ensure a limited cladding breach. In addition, fuel-cladding chemical interaction is addressed through limits on cladding temperature and time-at-temperature for steady-state operation, transients, and accidents to mitigate effects leading to cladding breach or fuel melting. Through the implementation of these limits, cladding breach, fuel melting, and deleterious fuel rod and assembly dimensional changes will be prevented. [ABSTRACT FROM AUTHOR]

Thermal response time is an important parameter for the control of fast reactors. Modern thermal hydraulic codes allow for the modeling of transient responses and can also be used to understand the dominant factors that affect them. However, simulations can be computationally expensive, particularly for performing parametric analyses of how thermophysical properties affect transient behavior. Here, we present a method for using linear stability analysis to estimate thermal response time and determine the key parameters that affect transient behavior without performing a forward simulation. The approach can also be used to corroborate simulation results and is tested against simulation results produced with a 2D finite difference model. The results show that this approach produces time-dependent temperature profiles that are within 2 × 10−5–0.1% of the numerical results for a single node perturbation. Changes in temperature have the greatest effect on thermal response time, followed by changes to thermal conductivity. [ABSTRACT FROM AUTHOR]

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

The optimization design of fast reactors faces two major problems: insufficient design or operation experience and more complicated optimization objectives, compared with pressurized water reactor (PWR). An optimization method which is independent of application banckground and experience is required. Intelligent optimization algorithm is a strategy for the search of optimal or satisfactory solution for given design objectives and satisfies above requirements. In the process of search, multiple design objectives and constraints are considered simultaneously and the times of search and calculation are limited. Genetic algorithm is a classical and representative method in numerous of optimization algorithms, which performs well in solving multi-objective problems with constraints. Genetic algorithm has inherent parallel characteristics as well for which the searching efficiency is able to be increased. Based on a genetic algorithm module in an open source toolkit DAKOTA and a self-developed neutron transport program SARAX, an optimal design frame for fast reactors with multivariable, multi-objective and multi-constraint problems was constructed in this work. Both genetic algorithm module and neutron transport program performed as a black-box. A coding system that could transform a decimal integer and a loading pattern to each other was suggested and applied to this optimal frame. This coding strategy realized the coverage of the multi-dimensional search space, so that the optimal frame can deal with the problems of continuous and discrete variables. For the constrained multi-objective problem, the direct ranking method and the weighted sum method were both applied and compared. Two cases were analyzed to test this frame: 1) Rearrange the subassemblies of ABTR to find a scheme whose keff and power peak factor meet the given expectations, the searching result was compared with the enumeration results; 2) Optimizing the design of the loading position of the 237Np-containing subassemblies for China Experimental Fast Reactor (CEFR). The objective of this case was obtaining more 238Pu after a cycle of burnup, ensuring safety constraints. The first calculation example was used to verify the feasibility of the new coding scheme. The second case demonstrated the specific application of the optimal frame in engineering practice. According to the numerical results, the fast reactor optimal frame proposed in this work achieves the research objectives of finding the best solutions and meeting the requirements under a limited number of searches with multiple variables, multiple design objectives and multiple constraints. What’s more, the coding of variables, the optimization objectives setting and the given of constraints are all of strong flexibility. To sum up, this frame can cope with the complex requirements of fast reactor engineering design.

In order to accurately calculate the photon heating distribution of fast reactor, the photon cross section processing code NGAMMA has been developed. NGAMMA can process MATXS format library and generate multi-group forms of photo-atomic cross sections, prompt photon production cross sections, delayed photon production cross sections, and neutron-photon KERMA factors for neutron-photon coupled heating calculations. An improved 21-groups photon library containing both prompt and delayed photon data is generated using the NGAMMA code based on the 1968n × 94γ MATXS format database. The photon library was verified using the fast reactor benchmark problem ZPR-6/7 and MET-1000. The results of the calculations show that the use of the 21-groups photon library generated by energy group collapse using 94-groups photon fluxes results in a significant improvement in the accuracy of the prompt photon heating calculations, especially in the non-fuel region, resulting in a relative error of less than 1% between the photon heating calculations and the MCNP calculations, compared to an error of 7% using the previous photon library. For the ZPR-6/7 benchmark problem, consideration of the delayed photons results in a 30% improvement in photon heating in the fuel region. For the MET1000 benchmark problem, the total power results of the neutron-photon coupling calculations based on the improved photon library are in good agreement with the MCNP, with a maximum relative error of no more than 3% in the total power of the fuel assembly. For reflector assemblies where photon power is a major contribution to heat release, the maximum relative error in prompt photon power is only −5.96%. Considering delayed photon significantly affects the photon heating, and for the MET1000 benchmark problem, considering delayed photon results in an overall increase of more than 30% in the photon power of the fuel assembly. • Developed a high-precision photon library processing code, NGAMMA. • The delayed photon heating was calculated distribution for fast reactors accurately. • Neutron and photon coupled heating calculations were performed for the MET1000 benchmark problem. [ABSTRACT FROM AUTHOR]

Safety assessment of fast reactors involves analyzing Hypothetical Core Disruptive Accidents (HCDA) that could result in reactor core disassembly. HCDA is generally analyzed in three distinct phases that include pre-disassembly, disassembly and post-disassembly phases. In the present work, a new coupled neutronics-hydrodynamics computer code called FA st R eactor DIS assembly (FARDIS-1) is developed in 2-D cylindrical coordinate geometry to model the super-prompt critical power excursion during the disassembly phase of HCDA in a sodium-cooled fast reactor (SFR). The modelling approach followed is essentially similar to the one used in VENUS-II code but with a few improvements. The code is benchmarked against the predictions of VENUS-II for a sample case of HCDA in the FFTF reactor model. This is followed by an analysis of the disassembly phase in a medium-sized (500 MWe) mixed oxide fuelled fast reactor. Two cases are assumed at the beginning of the disassembly phase: a sodium voided core (conservative) and a partially voided core (realistic) and the influence of several core neutronics parameters on the transients is elucidated through parametric studies. The peak power, thermal energy release, peak pressure, maximum temperature and work-potential are estimated and their trends are discussed. • A coupled neutronics-hydrodynamics computer code is developed to model the disassembly phase of HCDA in fast reactors. • Hydrodynamics module uses a Eulerian framework instead of Lagrangian approach. • Benchmarked against the predictions of VENUS-II for HCDA in FFTF reactor model. • Analysis of disassembly phase in 500 MWe MOX fuelled fast reactor for initially voided and partially voided core scenario. • Parametric studies on influence of core neutronics parameters on peak power, thermal energy and work-potential during HCDA. [ABSTRACT FROM AUTHOR]

Using dry recycled powders for uranium and plutonium mixed oxide (MOX) fuel production can reduce unnecessary storage and accountability of nuclear material in facilities. The shrinkage behavior of green compacts of dry recycled powders differs from that of conventional raw powders because the dry recycled MOX powder is obtained from the fabrication scrap of sintered pellets. The shrinkage behavior of dry recycled MOX powder has been investigated by dilatometry. Based on the shrinkage curves, sintering apparent activation energies were evaluated using the master sintering curve (MSC) and the constant rate of heating methods. The obtained values were higher than the energy evaluated for raw powder experiments. The sigmoid sintering prediction equation using the MSC function was constructed. The accumulation of data on the activation energy for various sintering conditions will lead to the wide application of this prediction formula in the future. [ABSTRACT FROM AUTHOR]

Results are presented on the validation of the second version (V2) of the EUCLID integrated code with extended capabilities in comparison with its first version. The additional models incorporated in the code include, in particular, combustion of sodium and formation and transport of sodium combustion aerosols and fission products in the premises of a nuclear power plant during severe accidents in a fast neutron reactor. This paper describes the results of the code validation to substantiate its applicability in calculating accidents with depressurization of a сoolant loop in a sodium cooled fast neutron reactor, including sodium combustion and the transport of aerosols formed from the sodium combustion products. Information is given on the models introduced in the code and the experimental data employed for the code validation. It has been demonstrated that the available experimental data are sufficient to validate the code as applicable to the above-listed phenomena. The results are outlined of the model validation against the experiments carried out in the SOLFA and CSTF experimental facilities with the combustion of a sodium puddle under different conditions: with an excess of oxygen or oxygen starvation, without or with sodium supply from a free-falling jet, without a steam source (simulation of evaporation from concrete) and with a steam source, and in a closed or ventilated room. The procedure for the analysis of uncertainty and sensitivity developed at IBRAE (Nuclear Safety Institute of the Russian Academy of Sciences) was used to substantiate the main calculated parameters, such as the temperature of sodium puddle, the rate of its combustion, and the concentration of combustion aerosols. It has been demonstrated that the deviation of these values from the experiments can be explained by the uncertainty in the input data. The applicability of the EUCLID/V2 code for modeling sodium combustion and further aerosol phenomena occurring with the sodium combustion products in the rooms of a nuclear power plant with a fast reactor has been substantiated. [ABSTRACT FROM AUTHOR]

Ac-225 has lately drawn considerable attention as a radioisotope for targeted alpha therapy treatment for certain types of prostate, blood-derived, and disseminated cancers, but its supply is limited. Therefore, we investigated the production method of Ac-225 by nuclear transmutation in a fast neutron reactor. The authors investigated irradiation of Ra-226 or Th-230 as a target nuclide in the experimental fast reactor Joyo, owned and operated by Japan Atomic Energy Agency, which has abundant fast neutrons and a large loading region with high heat removal capacity. Ra-226 is in increasing demand as a target nuclide to produce Ac-225. Therefore, as another option, we selected Th-230, which is 50 times more abundant than Ra-226 in natural uranium, as an alternative nuclide. Irradiation of Ra-226 and Th-230 with high energy neutrons above the threshold causes an (n,2n) reaction, producing Ra-225 and Th-229, respectively, which are the parent nuclides of Ac-225. The analyses showed that 47 GBq of Ac-225 can be generated annually by irradiating 1 g of Ra-226, and 6.5 GBq of Ac-225 can be semi-permanently generated every year by one-time irradiation of 50 g of Th-230 for 10 years (5 EFPY). It can be concluded that 100 MWt Joyo has potential to produce more than 70% of the current global supply of Ac-225 and/or to generate the parent nuclide Th-229, which keeps producing Ac-225 for thousands of years. [ABSTRACT FROM AUTHOR]