Advanced Nuclear Energy Systems Research Articles

Effect of dislocation on helium irradiation defects in low-activation martensitic steel

Reduced-activation martensitic steel is a main candidate structural material for key components of advanced nuclear energy systems because of its good mechanical properties at room temperature, low neutron activation characteristics and satisfactory radiation resistance. In this work, we study the interaction between dislocations and helium-irradiation-induced defects in the steel and the effect of dislocations on the behavior of helium migration and desorption. The well-annealed samples are pre-deformed to 10% and 20% reductions in thickness by using cold rolling mill. After pre-deformation, samples are heat-treated to remove vacancy defects and retain dislocation defects. Then, the samples with reserved dislocations are irradiated with helium at room temperature (50 keV, 1×10<sup>17</sup> He/cm<sup>2</sup>). After irradiation, the samples are characterized by synchrotron radiation grazing incident X-Ray diffraction, positron annihilation Doppler broadening spectroscopy, and thermal desorption spectroscopy. The results show that dislocations hinder the diffusion of helium and helium-vacancy complexes, and reduce the accumulation of radiation damage. Such an effect becomes more significant with the increase of dislocation density. The BCC → FCC phase transition of low activation martensitic steel occurs at 1179 K. The increase of dislocation density will lead to the forward shift of helium desorption peak induced by phase transition. The retention of helium in the undeformed sample, 10% deformed sample and 20% deformed sample is 10.3%, 15.7% and 17.9%, respectively, indicating that high density dislocations promote the retention of helium.

Effects of He-ion irradiation on the microstructures and mechanical properties of the novel Co-free VxCrFeMnNiy high-entropy alloys

High-entropy alloys (HEAs), as a promising candidate structural material for advanced nuclear energy system, have been widely studied recently due to the combination of the outstanding irradiation tolerance and enhanced mechanical properties at elevated temperatures. In this study, an FCC-structured CrFeMnNi2 HEA and a BCC-structured VCrFeMn HEA without Co element were designed and prepared. The effects of He-ion irradiation on the microstructures and mechanical properties were investigated, so as to preliminarily evaluate the irradiation tolerance of the novel Co-free HEAs. At fluences of 3 × 1016 and 6 × 1016 ions/cm2, the CrFeMnNi2 and VCrFeMn HEAs showed good structural stability at 1023 K; the transmission electron microscope results revealed no amorphization/precipitations. The average sizes and number densities of He bubbles increased as the fluence improved for the two alloys. At each fluence, the average size of the He bubbles in the VCrFeMn HEA was larger than that in the CrFeMnNi2 HEA, while the number density and volume fraction were lower. The CrFeMnNi2 HEA possessed a higher hardening fraction (77%), and an irradiation-induced hardening saturation was observed in the VCrFeMn HEA with a lower hardening fraction of 33%. The experimental results indicated that the novel low-activation VCrFeMn HEA might be a competitive candidate structural material for advanced reactors. The mechanisms of the irradiation tolerance of Co-free HEAs were also discussed.

Establishment of multi-beam irradiation facility at Wuhan University

A multiple ion beam irradiation facility has been established at Wuhan University to simulate severe irradiation conditions to materials in advanced nuclear energy systems. The platform consists of a 2 × 1.7 MV tandem accelerator, a 200 kV ion implanter, and a 50 kV ion implanter. A single high-energy irradiation chamber and a triple-beam irradiation chamber were set up. China Low Activation Martensitic Steel (CLAM) samples were irradiated with single beam of iron ions and simultaneous dual-beam of iron ions and helium ions at 450 °C. A large number of dislocation loops were observed in the samples after both irradiation conditions. The number density and average size of dislocation loops in the dual-beam irradiated samples were significantly larger than those in the single-beam irradiated samples. Irradiation hardening in dual-beam irradiated samples is more severe than that in single-beam irradiated samples.

Development and preliminary verification of a Monte Carlo neutron transport code IMPC-Neutron

A Monte Carlo neutron transport code IMPC-Neutron has been developed for Accelerator Driven Advanced Nuclear Energy System (ADANES) in Chinese Academy of Sciences. Based on the geometry module and initialization module provided by open source Monte Carlo code OpenMC, IMPC-Neutron adopts Unionized Grid method and Delta-tracking method to improve the speedof simulation. Besides, in order to reduce the memory consumption,a low-memory mode is developed in IMPC-Neutron to satisfy different computing conditions. A critical benchmark of ICSBEP and three full reactor core model benchmarks are used to verify the reliability andacceleration effect of IMPC-Neutron. The results show that the calculation results of IMPC-Neutron are in good agreement with OpenMC, and the acceleration efficiency is higher.

Code development of Single-phase 2-D pressure wave propagation

Obtaining accurate capture of the pressure wave propagation process is very important for structural load calculation and stress analysis. For traditional nuclear reactors, one-dimensional system analysis codes such as RELAP5 and TRACE are recognized to be the effective analysis tools for pressure wave propagation simulating. However, advanced nuclear energy systems have a highly integrated and complicated geometric configurations. It is obvious that the system analysis codes are difficult to obtain an elaborate simulating of pressure wave propagation process. It is significant to develop multi-dimensional pressure wave propagation simulation codes. In this paper, a 2-D single-phase pressure wave propagation calculating code is developed. Furthermore, code verifications are carried out by using two typical single-phase shock tube benchmarks. Finally, two-dimensional propagation characteristics of pressure wave are investigated. Conclusions can be drawn that the new code proposed in this paper can reasonably reproduce the behaviors of multi-dimensional pressure wave propagation.

Hierarchical-interface design for stable helium storage and structural evolution in homogeneous nano-multilayered Al1.5CoCrFeNi high entropy alloy films

Hierarchical interfaces and nanometer grain boundaries (GBs) were synchronously designed inside the homogeneous Al1.5CoCrFeNi high entropy alloy (HEA) nano-multilayered films, in which each lamellae was composed of nano-crystals to construct helium (He) ion radiation-tolerant materials used for future advanced nuclear-energy system. The depth profile of He behavior and phase-structure evolution in the irradiated [Al1.5CoCrFeNi (10 nm)]30 and [Al1.5CoCrFeNi (30 nm)]10 nano-multilayered films were investigated systematically. Results indicated that the He bubble size in multilayered film exhibited a reduction tendency and a wide distribution feature as the individual-thickness increased from 10 nm to 30 nm, demonstrating inverse results with previous heterogeneous-interface modeling. Reduced bubble pressures were identified in the two kinds of nano-multilayered HEA films, indicating that the coexistence of nanocrystals and homogeneous-interfaces was beneficial to manage He distribution and prolonged bubble growth effectively by tailoring each-lamellae size in these two kinds of multilayered Al1.5CoCrFeNi HEA films. Partial phase transformation from body-centered-cubic (BCC) to face-centered-cubic (FCC) structure was induced through promoting the structural evolution between the irradiation-induced amorphous regions and compositional reorganization in these nano-multilayered HEA films upon irradiation. The interfaces and GBs promoted the annihilation of these defects and further facilitated the self-healing behavior. Such hierarchical-interface in the HEA provides a novel strategy for the effective He entrapment in improving radiation tolerance and a potential possibility of real-application.

The Solitary Wave in Advanced Nuclear Energy System

The solitary wave naturally arises in many areas of mathematical physics, including in nonlinear optics, plasma physics, quantum field theory, and fluid mechanics. In the past few years, for an advanced nuclear energy system, a particular class of traveling wave reactor called the Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy production (CANDLE) reactor has been proposed, and an analytical solution has been desired since it could reveal the global characters of the solution. In this study, from the perspective of the solitary wave, the analytical solution of this advanced nuclear energy system is demonstrated through coupling the one-group neutron diffusion equation with the burnup equation. The tanh-function method is applied to solve that nonlinear partial differential equation. The relationship between the velocity of the solitary wave, wave amplitude, or neutron flux and the evolution of the nuclide is revealed by the analytical method. The results demonstrate that the neutron flux is proportional to the wave velocity. The results also imply that the amplitude of the neutron flux is proportional to the square root of the diffusion coefficient but is inversely proportional to the initial 238U density.

Transmutation of long-lived fission products in an advanced nuclear energy system

Disposal of long-lived fission products (LLFPs) produced in reactors has been paid a lot attention for sustainable and clean nuclear energy. Although a few transmutation means have been proposed to address this issue, there are still scientific and/or engineering challenges to achieve efficient transmutation of LLFPs. In this study, we propose a novel concept of advanced nuclear energy system (ANES) for transmuting LLFPs efficiently without isotopic separation. The ANES comprises intense photoneutron source (PNS) and subcritical reactor, which consist of lead–bismuth (Pb-Bi) layer, beryllium (Be) layer, and fuel, LLFPs and shield assemblies. The PNS is produced by bombarding radioactive cesium and iodine target with a laser-Compton scattering (LCS) γ-ray beam. We investigate the effect of the ANES system layout on transmutation efficiency by Monte Carlo simulations. It is found that a proper combination of the Pb-Bi layer and the Be layer can increase the utilization efficiency of the PNS by a factor of ~ 10, which helps to decrease by almost the same factor the LCS γ-beam intensity required for driving the ANES. Supposing that the ANES operates over 20 years at a normal thermal power of 500 MWt, five LLFPs including 99Tc, 129I, 107Pd, 137Cs and 79Se could be transmuted by more than 30%. Their effective half-lives thus decrease drastically from ~ 106 to less than 102 years. It is suggested that this successful implementation of the ANES paves the avenue towards practical transmutation of LLFPs without isotopic separation.

Synthesis and multi-scale properties of PuO2 nanoparticles: recent advances and open questions.

Due to the increased attention given to actinide nanomaterials, the question of their structure-property relationship is on the spotlight of recent publications. Plutonium oxide (PuO2) particularly plays a central role in nuclear energetics and a comprehensive knowledge about its properties when nanosizing is of paramount interest to understand its behaviour in environmental migration schemes but also for the development of advanced nuclear energy systems underway. The element plutonium further stimulates the curiosity of scientists due to the unique physical and chemical properties it exhibits around the periodic table. PuO2 crystallizes in the fluorite structure of the face-centered cubic system for which the properties can be significantly affected when shrinking. Identifying the formation mechanism of PuO2 nanoparticles, their related atomic, electronic and crystalline structures, and their reactivity in addition to their nanoscale properties, appears to be a fascinating and challenging ongoing topic, whose recent advances are discussed in this review.

Monte-Carlo study of pre-neutron emission mass and energy for neutron-induced <sup>232</sup>Th fission

The development of fourth-generation reactors and advanced nuclear energy systems require high-precision, multi-nuclide, and wide-energy-area neutron nuclear data. However, the current nuclear energy-related nuclear fission data in the China Nuclear Data Evaluation Library (CENDL library) are incomplete and cannot meet the current need. It is extremely important to establish the reliable calculation methods and tools for the neutron nuclear data. Based on the Monte-Carlo method, a model for calculating the pre-neutron fission fragment is established in this work. The mass and kinetic energy distribution of <sup>232</sup>Th(n,f) reaction at the medium- and low- incident neutron energy are studied. The calculations of the mass distribution with the different values of incident energy are compared with the experimental results. The maximum deviation of this work from the experimental data is ~1%, which is advantageous compared with the GEF and TALYS code (maximum deviation from the experimental value is ~2%). The calculation of the pre-neutron fission fragment kinetic energy also shows good agreement with experimental result. The results indicate that this model can well describe and predict the characteristics of pre-neutron fission fragment for <sup>232</sup>Th(n,f) reaction at the medium- and low- incident neutron energy. It also provides a new idea for calculating the neutron-induced actinide fission reactions.

Studies on the design and properties of FeCrVTix medium-entropy alloys for potential nuclear applications

The materials' issue is one of the main bottlenecks restricting the development of nuclear energy systems. Based on the requirements of low activation, high-temperature thermal stability, high strength, high ductility, high thermal conductivity, irradiation resistance, corrosion resistance and other characteristics, the FeCrVTix medium-entropy alloys (MEAs) were designed for the working under extreme environments of the advanced nuclear energy systems. Here, thermodynamic calculation and coupling with the selection of low activation and functional elements and phase composition estimation, to MEAs are shown in the effort to design this kind of modern advanced alloy materials. Meanwhile, a kind of FeCrVTix (x=0, 0.05, 0.1, 0.2, 0.3) MEAs, as a model material, was prepared by arc melting and its above-mentioned properties were investigated. The results show that FeCrVTix MEAs have excellent properties and quite match the previous expected design, which have great potential for nuclear applications as structural materials. Most importantly, this novelty and proven method of alloy design based on the target requirements and configuration entropy has been proven to be correct and feasible, providing a new way for the design and rapid screening of new materials.

Transient analysis of the main heat-transfer system of the pool-type sodium-cooled fast reactor

<p indent="0mm">Sodium-cooled fast reactors are the first choice of the fourth generation of advanced nuclear energy systems. They have attracted significant attention in the development of advanced nuclear energy technology owing to their advantages of proliferative nuclear fuel, transmutation of radioactive waste, and high thermal efficiency power generation. The design and construction of the secondary sodium-cooled fast reactor are ongoing in China. The physical characteristics of fast neutrons and the special thermal, physical, and chemical properties of liquid sodium metal cause special and complex scientific problems in the design of sodium-cooled fast reactors. In this study, taking the energy transfer process of sodium-cooled fast reactors as the object, we establish mathematical and physical models for analyzing large-scale transients in various energy transfer and conversion processes. The model is verified and validated by the natural circulation test of the Phoenix reactor and the international benchmark of the fast flux test facility loss in flow tests. The trend of the calculated values is consistent with that of the test data, and the results are in good agreement. For a typical pool-type sodium-cooled fast reactor, four types of transient events in the power plant are determined by engineering evaluation and deductive analysis. Among them, there are nine normal operation transients, 23 expected transients, 10 abnormal transients, and 7 emergency transients. Based on the established mathematical and physical models, the analysis results of each type of typical transient are given, which provide an important basis for the design of the system and equipment/components and lay a scientific and technological foundation for completing the design of the transient condition and operation mode of the sodium-cooled fast reactor as well as safety evaluation.

Development and preliminary verification of a Monte Carlo photon transport code IMPC-Photon

The Monte Carlo photon transport code IMPC-Photon is developed for serving the software development of designing advanced nuclear energy system such as Accelerator Driven Advanced Nuclear Energy System (ADANES) in Chinese Academy of Sciences. For photon physics, four main photon-atomic interactions and three secondary processes including atomic relaxation, electron-positron annihilation, and thick-target bremsstrahlung are implemented. For code calculations, the unified energy grid method is applied to improve the speed of searching for the cross section. Under the same computational conditions, the unified energy grid method is more time-efficient and the search efficiency is significantly improved. In addition, the “point gamma-ray source in infinite medium” and “gamma-ray skyshine experiment” benchmarks are employed to verify the reliability of the code. The results show that IMPC-Photon has a good agreement with reference codes and experimental measurement, which indicate that IMPC-Photon is reliable and credible for photon transport problem.

Effect of helium bubbles on irradiation hardening of additive manufacturing 316L stainless steel under high temperature He ions irradiation

To evaluate the effects of irradiation on additive manufacturing materials which are expected to be applied in advanced nuclear energy systems, selective laser melting 316L stainless steel (SLM 316L SS) and traditionally manufactured 316L stainless steel (TM 316L SS), were irradiated with 500 keV He ions at 700°C, respectively. SEM, TEM and nanoindentation measurements were performed to survey the effects of helium (He) bubbles on the irradiation hardening. It was found that the degree of irradiation hardening of the two materials increased with an increase of ions fluences. The hardening range of SLM 316L SS was 6-47% while that of TM 316L SS was 32-75%. Moreover, the hardening saturation phenomenon was more pronounced in TM 316L SS than in SLM 316L SS. These results suggested that SLM 316L SS owned higher resistance to irradiation hardening than TM 316L SS. TEM observations showed that the density of He bubbles in TM 316L SS was 7.06 × 1023/m3 while that in SLM 316L SS was 3.33 × 1023/m3. Further, the average sizes of bubbles were 2.2 and 2.4 nm, respectively. The differences of the number density might be why SLM 316L SS had a slighter irradiation hardening than TM 316L SS, which was consistent with Orowan model. It was suggested that the numerous intrinsic microstructures such as entangled dislocation network walls and subgrain boundaries in SLM 316L SS could greatly reduce the density of He bubbles.

An adaptive optimization reverse regulation method for neutron spectrum based on standardized response module

• An adaptive optimization reverse regulation method for neutron spectrum based on standardized response module was developed. • 50 standardized response modules with different thickness and materials were designed for neutron spectrum regulation. • The differential evolution and artificial neural network algorithms are used to optimize the arrangement of response modules. • Verifications on ITER TBM and RBEC-M have confirmed the accuracy and universality of the developed method. The neutron spectrum is one of the essential parameters in the design of advanced nuclear energy systems. Neutron spectra of different nuclear energy systems have wide differences so that accurate regulation of a specific neutron spectrum is of great significance to its development. In this research, an adaptive optimization reverse regulation method for neutron spectrum based on standardized response module was developed. The proposed method designed different types of standardized response modules and calibrated it with their corresponding multi-dimensional response matrix calculated by the Monte Carlo program. Combined with an intelligent optimization algorithm and artificial neural network algorithm, the regulation scheme of the target neutron spectrum could be constructed and modified accurately and efficiently. The 14 MeV single-energy fusion neutron source was used to verify the regulated results of the proposed method on the tritium breeder blanket of a fusion reactor (ITER) whose results showed that the accuracy of the proposed method was improved by 7.69 % compared to the forward manual experiment. Also, a verification of the regulated spectrum on a fission reactor RBEC-M was done to ensure universal application of the method.

Research on structure selection and design of LBE-cooled fast reactor main coolant pump

Fast neutron reactor has become the preferred reactor type for advanced nuclear energy systems in the world, representing the development direction of the Generation IV nuclear energy system. The lead-cooled fast reactor (LFR) uses pure Pb/LBE as coolant, the natural characteristics of lead alloy are combined with the structural design of LFR, which constitutes the inherent safety characteristics of LFR and the system. In this paper, the main coolant pump in the Generation IV of LFR is taken as the research object. According to the design requirements of the main coolant pump in LFR fast reactor, the hydraulic characteristics of the main coolant pump with double-outlet volute type and space guide vane type under normal operation and accident operation are compared and analyzed. It is found that, under normal pump conditions the hydraulic performance of double-outlet volute type main coolant pump is a slightly better, but the space guide vane type has obvious advantages in the ability to deal with accident conditions and the better reliability in its long-term operation. Finally, the space guide vane structure is determined as the design program of the main coolant pump of CLEAR-I fast reactor.

A quantitative risk assessment methodology based on CCDF risk curve for accidental releases from fusion reactors

Abstract Fusion reactors are expected to be more safe and environmentally friendly in the development of nuclear energy systems. It is vital to fully understand the overall radiation risk level of fusion reactors in comparison with existing nuclear energy systems. The objective of this paper is to propose a risk assessment methodology for fusion reactor accidents, which is capable of quantifying the risk level comparable to other advanced nuclear energy systems. In this paper, fusion specific accident characteristics are comprehensively analyzed in comparison with ITER and fission reactors. The postulated initiating events as well as possible environmental release routes are also presented for DEMO, and then deterministic risk assessment is applied to design basis accidents, while the coupling of deterministic risk assessment and probabilistic risk assessment is suggested for design extension conditions. The societal risk of DEMO is proposed to be presented by a Complementary Cumulative Distribution Function (CCDF) risk curve.

Smart 3D Network Nanocomposites Collect Irradiation-Induced “Trash”

Summary Developing outstanding radiation-resistant materials with excellent thermal and mechanical properties and self-healing ability remains a great challenge for advanced nuclear energy systems. Here, we incorporate three-dimensional (3D) carbon nanotube (CNT) networks into iron nanocrystals (Fe NCs) to obtain functional bulk Fe-CNT nanocomposites, exhibiting much higher thermal conductivity and mechanical properties than the Fe NCs without sacrificing strength. Irradiations by energetic helium (He) and krypton ions show that 3D CNT networks act as gigantic-capacity “nano-dustbins” to collect and store He atoms and defects via a “loading-transporting-unloading” mechanism in the grain boundary-CNT (GB-CNT) configuration, thus greatly enhancing their radiation tolerance compared with the Fe NC counterparts. The energetic landscape of interactions of He/defects with the GBs and the Fe-CNT interfaces in the nanocomposites are revealed by first-principles calculations. This work offers a promising strategy for material design of high-performance structural materials for future advanced nuclear reactors.

Evaluation of Corrosion Phenomena of T91 Steel in Stagnant Liquid Lead at High Operational Temperatures

The present study aims to deepen the knowledge on the corrosion behavior of martensitic T91 steel in contact with liquid lead in view of its application in advanced nuclear energy systems. Steel samples have been analyzed after the immersion in stagnant molten lead at different temperatures (550°C to 650°C) and for an exposure time of 550 h. In order to evaluate the influence of different oxidizing conditions, the corrosion tests were performed under both Ar and Ar-H2 mixtures. Within the temperature range applied, the oxygen content in the liquid lead was Co = 6.8±2.7 × 10−2 wt% and Co = 9.2±3.6 × 10−8 wt%, respectively. When the Ar atmosphere was applied, dissolution of iron and chromium has been observed for temperature above 590°C. Conversely, under an Ar-H2 mixture atmosphere, mainly the presence of internal oxidation along the grain boundaries whose depth increases with temperature was detected.

The special nuclear data library CENDL-TMSR for thorium-uranium cycle

Thorium molten salt reactor nuclear energy system (TMSR), which is the fourth generation advanced nuclear energy system, needs more advanced and reliable physical design. The reliability and accuracy of nuclear data has very important impact on the accuracy of nuclear design. According to the requirement of Shanghai Institute of Applied Physics, Chinese Academy of Sciences, China Nuclear Data Center has developed a special nuclear data library CENDL-TMSR for thorium-uranium cycle, which includes evaluated nuclear data library and application libraries for criticality calculation and shielding design of thorium molten salt reactor. In order to verify and validate the reliability and applicability of the CENDL-TMSR, the criticality benchmarking and shielding benchmarking are carried out. The criticality testing results show that the relative errors between calculated k eff and experimental values for most benchmarks are within 0.5%, which indicates that the CENDL-TMSR can be used for physical design of thorium molten salt reactor. The shielding calculation results are consistent with the experimental ones, and the overall performances are better than other evaluated nuclear data libraries.