SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.

In typical pressurized water reactors, zirconium alloys are used as cladding material for the fuel. However, zircalloy is known to face problems with the high temperature steam, due to the chemical process of oxidation, the oxygen molecules will be separated from the water molecules of the coolant leading to hydrogen gas releases. Recently, a research team at KAIST, South Korea suggested a methodology to fabricate nanoporous oxide layer with the aim of preventing the zircalloy outer surface from reacting with the coolant. Although, this new proposal offers a better solution to prevent the potential hydrogen gas generation, it is still not well understood how the nanoporous-layer is going to affect the convective heat transfer rates between the coolant and the fuel. In fact, on one hand the low conductivity of the oxide layer is expected to reduce the conduction heat transfer within the cladding material; but on the other hand, the nanopores on the oxide layer might act as an effective surface roughness, hence affecting both the hydrodynamic and thermal fields within the coolant channels. In this study, a CFD analysis is carried out to investigate the influence of this nanoporous layer on the convective heat transfer rate and pressure drop coefficient. A detailed 2-D steady-state numerical analysis on single-phase model is performed using Star-CCM+ code. The study is conducted using pores with a diameter of 30 to 100 nm. The results obtained from these predictions are then compared with the ones obtained in the case of the smooth surface. Therefore, the main objectives of the present study are to examine the effect of this nanopourous layer on the thermal hydraulic parameters and to produce the corresponding correlations to be used in the system scale thermal-hydraulic codes.

Breakaway oxidation is a potential fuel failure mechanism during a loss-of-coolant accident (LOCA), especially small-break LOCA. Two kinds of new advanced zirconium alloy, CZ1 and CZ2, were developed by China Nuclear Power Research and Technology Institute of China General Nuclear Power Group. The breakaway oxidation behavior of CZ1 and CZ2 was studied. The outer surface of all samples was examined visually and photographed. After 2730s oxidation in steam, the outer surface of CZ1 alloy sample and CZ2 alloy sample remained lustrous-black. The outer surface of CZ1 sample oxidized in steam at 1000 °C for 4181s was grey, but under the same experimental conditions the outer surface of CZ2 sample was still lustrous black. The hydrogen pickup content of different oxidation time was measured. The samples with grey appearance showed significant hydrogen pickup. The microstructure was observed by optical microscope. Evolution of oxide structure was described, and the mechanism was discussed.

During a severe accident or Beyond Design Basis Accident (BDBA), the reaction of water with zirconium alloy as fuel clad, radiolysis of water, molten corium-concrete interaction (MCCI) and post-accident corrosion can generate a source of hydrogen. In the present work, hydrogen distribution due to in-vessel reaction (between zircaloy and steam) has been simulated inside a WWER-1000 reactor containment. In the first step, the thermal hydraulic parameters of containment have been simulated for a DECL (Double Ended Cold Leg) accident (DBA phase) in both short and long time and the effects of spray as Engineering Safety Features (ESFs) on mitigating the parameters have been studied. In the second step, it has been assumed that the accident developed into an in-vessel core melting accident. While in pre-phase of core melting (severe accident phase), hydrogen will be produced as a result of zircaloy and steam reaction (BDBA phase), the hydrogen distribution has been simulated for 23 cells inside the reactor containment by using CONTAIN 2.0 (Best estimate code) and MELCOR 1.8.6 codes. Finally, the results have been compared to FSAR results. As it can be seen from the comparisons, both CONTAIN and MELCOR codes can predict the results in good agreement with FSAR (ANGAR code) results. CONTAIN shows peak pressure around 0.36 MPa in short-term and this amount is about 0.38 and 0.4 MPa for MELCOR and ANGAR (FSAR) results respectively. All these values are under design pressure that is around 0.46 MPa. Cell 20 has the maximum mole fraction of hydrogen in long-term about 9.5% while the maximum amount of hydrogen takes place in cell 22. The differences between the results of codes are because of different equations, Models, Numerical methods and assumptions that have been considered by the codes. The simulated Hydrogen Distribution Map (HDM) can be used for upgrading the location of HCAV systems and Hydrogen Mitigator features (like the recombiners and ignitors) inside the containment to reduce the risk of hydrogen explosion.

In 1990s the first generation of PWR fuel assembly FA300 was developed in China. With constantly design improvements over the following 20 years, three types of FA300 fuel assemblies have been developed with max fuel assembly average burnup reach 40000 MWd/tU and successfully used in 300MWe PWRs. By now over 1100 fuel assemblies of FA300 series have been successfully operated in three different 300MWe PWRs with satisfied operation reliability. The CAP1400 fuel assembly development program was launched in 2010 which includes new zirconium cladding material development, UO 2 pellet development, high performance fuel assembly mechanical structure development, fuel rod performance code and fuel assembly seismic analysis code development, out-of-pile fuel assembly test facility construction and in core irradiation program. The main purpose of CAP1400 fuel assembly development program is to meet the needs of self-sufficient fuel supply to CAP1400 reactor. Based on FA300 and CAP1400 fuel assembly development, a full series of PWR Fuel R&D technology and test system has been successfully established by SNERDI which will continuously support the fuel assembly improving and new type of fuel development. This paper will mainly describe the PWR fuel technology including cladding technology, pellet technology, fuel assembly mechanical design technology, fuel assembly test facility and test technology, fuel rod and fuel assembly code development progress. Furthermore, the development progress of CAP1400 fuel assembly will also be introduced. By now two types of new zirconium alloys have been selected as candidate alloys, all of the out-of pile performance tests have been finished. Both two alloys show good corrosion resistance. Through full series of fuel assembly components performance tests, the fuel assembly has finished design finalization. Most of the fuel assembly mechanical and hydraulic tests will be finished by the end of 2016. The test reactor irradiation program and commercial PWR irradiation program are also on the schedule. After the LTA program and commercial application licensing, the CAP1400 fuel assembly is anticipated to provide adequate burnup capability and operation reliability to CAP1400 reactor in China.

The hydrogen may be introduced into the fuel rod during the process of production and manufacture. During the operation in reactor, the irradiated fuel pellets also produce radioactive isotopes of hydrogen and tritium. Under the operating condition in pile, the hydrogen in fuel rod will enter the zirconium alloy cladding tube forming hydride, lead to the hydrogen brittleness of cladding tube, and severe cases can lead to the cladding tube broken. The radioactive tritium inside fuel rod has high activity, and it possibly goes through the cladding tube by diffusion penetration into the reactor coolant. With the reactor in waste water or steam waste emissions to the environment, such as lead to tritium radiation safety problems of environmental pollution. Thus, reduce the hydrogen source and tritium pressure in fuel rod, is the way to reduce the hydrogen absorption effect and the release of tritium to coolant. By conducting the Zr-4 alloy nickel-plated hydrogen-absorption device design research, through nickel plating process on the surface of Zr-4 alloy structure parts, eliminating the influence of the oxide film to maintain its excellent absorbing hydrogen isotope activity. During the design operating temperature conditions of fuel rods, the reaction of zirconium hydride has lower hydrogen balance pressure, while the gas cavity kept low pressure hydrogen isotope, can significantly reduce the hydrogen pickup of fuel rod zirconium alloy cladding tube and reduce the tritium permeation emissions by cladding tube. Through nickel-plated hydrogen-absorption device structure design, manufacture, performance testing, analysis and evaluation, demonstrates that the flat plate and cross nickel-plated hydrogen-absorption device can meet the expected effect.

Zr-Nb zirconium alloy was developed at CNPRI and designated as CZ2. In this paper, the composition, size distribution and crystal structure of second phase particles (SPPs) in CZ2 have been analyzed under transmission electron microscopy (TEM). Two types of SPPs were observed in CZ2, which are bcc β-Nb and hcp Zr-Nb-Fe. To study the radiation effects on SPPs, CZ2 in two different heat treatment conditions CZ2-SRA (stress relieved) and CZ2-RXA (recrystallized) were irradiated with 3 MeV Zr ++ up to 1.14×10 16 Zr/cm 2 (50 dpa at the damage peak) at three different temperatures 320°C, 360°C and 400°C in Texas A&M University. Commercial low-tin Zr-4 was also irradiated at the same condition as a reference. The results show that the SPPs of Zr-4 become completely amorphous after irradiation at 320°C and 360°C, while retain crystalline at 400 °C. SPPs of CZ2-RXA are partially disordered only after irradiation at 320°C. The crystal structure of SPPs in CZ2-SRA survived at all temperatures.

During the development of zirconium alloys, the irradiation in the test reactor is a critical step to comparison the irradiation properties of candidate alloys, such as corrosion, creep and irradiation growth. In this paper, a small scaled fuel assembly for test reactor irradiation is designed, which meets the needs of new zirconium alloys development. The irradiation fuel assembly (IFA) can be easily disassembled, and the test fuel rods or irradiation specimen can be easily replaced, which makes it possible to do the further post-irradiation examination in the hot cell to obtain the irradiation performance data. Now the IFA has finish fabrication and the test reactor irradiation program is planned to launch in 2017.

Two Zr-Sn-Nb alloys with minor Germanium or silicon additions were prepared by traditional manufacturing process to meet the design requirements. Transmission electron microscope (TEM) and scanning electron microscope (SEM) were utilized to characterize the detail microstructure of base alloys. Corrosion resistance was examined by the weight gain in static autoclave with different water chemistry environments. The mechanical properties at room temperature and elevated temperature were evaluated by conventional tensile testing. Thermal creep resistance was evaluated by an internally-pressurized creep test at 385 °C with hoop stresses of about 108 MPa and 150 MPa (during 24 h). It was found that SZA-6 and SZA-4 alloys consisted of partially recrystallized grain structures with uniformly distributed fine second phase particles (SPPs) located within grain interior and at grain boundaries. Both SZA-4 and SZA-6 alloys exhibited excellent corrosion resistance in two water chemistry conditions. The corrosion resistance of SZA-6 was better than the reference commercial alloy, and SZA-4 was slightly better than SZA-6. The mechanical properties of two new zirconium alloys were comparable, and both of them can meet the design criterion. Moreover, the thermal creep resistance of SZA-4 and SZA-6 alloys was equivalent to existing commercial alloy. Considering the outstanding corrosion resistance, satisfied mechanical properties and thermal creep resistance, SZA-4 and SZA-6 alloys were suggested as promising alloys used for CAP1400 fuel assembly in the future.

Delayed hydride cracking (DHC) is one of the major factors that affect the safety of zircaloy tubes. In order to effectively simulate DHC, the coupling process including hydrogen diffusion, hydride precipitation, energy flow and mechanical fields should be correctly simulated. In this study, a cohesive model considering irradiation embrittlement and damage accumulation is used to simulate the cracking behavior of zircaloy, and in the finite element model its traction-separation relation is converted to the boundary condition between the traction and normal displacement component in the surfaces to be cracked. The coupling calculation procedures to simulate the DHC behavior using the automatic finite element program generator (FEPG) are obtained and validated, in which the three-dimensional constitutive relation involves plastic strain hardening, irradiation hardening and hydride-induced anisotropic expansion. We find it effective to study the DHC behavior. And we draw the conclusion that irradiation embrittlement and irradiation hardening could visibly accelerate the cracking process. This study improves the numerical simulation method for DHC, provides a reference basis for theoretical study and lays a foundation for further researches on DHC of zircaloy tubes.

Zirconium alloys remain as the main cladding materials in most water reactors. CZ zirconium alloy is recently developed for advanced PWR fuel assembly (STEP) designed by China General Nuclear Power Group (CGN). There are two kinds of zirconium alloys, designated as CZ1 and CZ2 respectively. It is well known that creep property is one of the most important characteristics to evaluate the integrated performance of new commercial alloys. In this study, we have collected the creep date of CZ alloys heat-treated at different temperatures. The result shows that the creep rates of the zirconium alloys enlarge with the creep stress in the temperature of 375 °C. The creep resistance is improved with the increasing annealing temperature. The creep resistance of CZ-RXA (high temperature) is better than CZ-SRA (low temperature). The creep rate of CZ alloys at relatively lower annealing temperature is much larger than that at higher annealing temperature. At the given stress condition, the creep behaviors of different annealed CZ alloys are found to have almost the same tendency. In addition, the creep rate of CZ1 is smaller than CZ2 at the same annealing temperature.

Advanced Thermal Reactor (ATR) FUGEN is the heavy water-moderated, boiling light water-cooled, pressure tube-type reactor. The commercial operation of FUGEN started on Mar. 1978 and terminated on Mar. 2003 and the decommissioning of FUGEN has been carried out since the decommissioning plan was approved in 2008. In order to perform the decommissioning work such as dismantling and decontamination safely and reasonably, technology development for the decommissioning has been promoted actively. This paper describes a part of technology development as follows. (1) Technology development on reactor dismantling The reactor of FUGEN is made of various materials such as stainless steel, carbon steel, zirconium alloy and aluminum which have relatively high activity concentration by operation for 25 years. With consideration of these characteristics, the reactor will be dismantled under water remotely in order to shield the radiation and prevent dust from migrating from water to air generated by the cutting considering the usage of zirconium alloy which is likely to be oxidized. In addition, laser cutting method whose features are fast cutting speed and less secondary waste in cutting will be applied for reactor dismantling. However, laser cutting method has no experiences to be applied to dismantlement of reactor facilities. Therefore, laser cutting for actual dismantled objects in air was demonstrated in controlled area in FUGEN using laser cutting system composed of articulated robot and laser cutting head. As a result, safety and applicability of laser cutting system was confirmed. From now on, primary cutting work in air, cutting demonstration with a relatively high dose rate and mock-up test in water for dismantling the actual reactor will be carried out. (2) Technology development on investigation of contamination It is necessary to evaluate radioactive inventory in the facilities accurately in order to reflect the evaluated data to dismantling plan appropriately. Therefore, the investigation of the contamination for the facilities has been carried out for safe and reasonable decommissioning work. The in-situ simple investigation method for the contamination of inner pipes which is mostly dominated by Co-60 is started to develop using the portable NaI(Tl) spectrometer. This method complements conventional investigation method to take samples from the pipes and to analyze them by radiochemical method to figure out the contamination of the whole facility.

The motivation for exploring the potential development of accident tolerant fuels in PWR to replace existing Zr alloy clad monolithic oxide fuel is outlined. The assessment includes a brief review of core degradation processes under transient conditions and introduces the enhancement of the accident tolerance of PWR by the development of fuels/cladding that can tolerate loss of active cooling in the core and loss of coolant water for a considerably longer time period while maintaining or improving the fuel performance during accident conditions. probabilistic risk assessment are presented that illustrate the impact of these new candidate fuel/cladding materials on the reduction of core damage frequence.

Based on lessons learned from the Fukushima Daiichi nuclear power plant accident, pursuit of accident tolerant fuel (ATF) has been discussed by many institutions in the world. Toshiba identified a silicon carbide (SiC) ceramic as the most promising material for accident tolerant fuel. Since SiC has less active characteristics in the presence of high temperature water steam (H 2 O) and is expected to be tolerant of severe accident conditions. Moreover, SiC has a smaller neutron absorption cross-section which is advantageous feature in terms of neutron economy. Zirconium alloys (Zry) are one of the main structural materials in LWR core. In high temperature H 2 O environment under severe accident conditions, Zry rapidly reacts with H 2 O and oxidation reaction accompanied by release of hydrogen gas occurs. Since SiC may inhibit the progress of oxidation reaction compared to Zry metal alloys, hydrogen and heat generation is expected to decrease in the case of core uncovered accident conditions. In order to confirm the advantage of SiC over Zry as core materials, transient analysis and safety analysis are carried out. For transient analysis, analyses of temperature behavior of cladding at plant transient condition are carried out with best-estimate transient analysis code. This analysis confirmed the effect of physical properties differences between SiC and Zry on cladding temperature behavior. Moreover to indicate the effectiveness of SiC under the core uncovered condition with oxidation reaction, safety analysis with latest “MAAP” code is carried out and the whole plant behavior during severe accident sequence is simulated. This analysis showed the effectiveness of SiC to mitigate the oxidation reaction. As the result of these analyses, the advantage of SiC over Zry can be perceived. And also, future challenges of SiC application as ATF can be clarified through these analyses.

The problem with higher nuclear fuel enrichment is its high initial reactivity. It has a negative effect on the peaking factor, which is one of the license conditions. The second major problem is the ability to control the reactivity of the reactor, and thereby maintaining the multiplication factor in the core equal to 1. Long-term control of the reactivity in PWR reactors is typically conducted by the concentrated boric acid (H 3 BO 3 ) in the coolant; its highest possible concentration is determined by the requirement to maintain a negative reactivity coefficient. Another option are burnable absorbers. This work deals with usage of hafnium as an advanced type of burnable absorber. Based on the model of computing code U W B 1 for the study of burnable absorbers, a new cladding of nuclear fuel is designed with a thin protective layer made of hafnium. This cladding is used as a burnable absorber that helps reducing excess of fuel reactivity and prolongs the life of the fuel assemblies, which increases economic coefficient of the use of nuclear power plants. This cladding would also work as a protective layer increasing endurance and safety of nuclear power plants. Today zirconium alloys are exclusively used for this purpose. The main disadvantages of zirconium alloys include rapid high temperature oxidation of zirconium — a highly exothermic reaction between zirconium and water steam at temperatures above 800 °C. During this reaction hydrogen and inconsiderable amount of heat are released. Hydrogen excess, released heat, and damaged cover of fuel may deepen the severity and consequences of possible accidents. Another disadvantage of zirconium alloys is their gradual oxidation under standard operating conditions and ZrH formation, which leads to cladding embrittlement.

Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys and decrease corrosion rate of zirconium alloy during standard operation. Zirconium alloys are widely used as cladding and construction material in almost all types of nuclear reactors, where usually creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper zirconium alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations (Raman spectroscopy) were done for zirconium alloy covered with polycrystalline diamond layer before and after high temperature steam exposure. Weight increase and hydrogen release ware measured during steam exposure.

Nuclear fuel rods is mainly composed of uranium dioxide pellets and zirconium alloy cladding, there is a gap between pellets and cladding, which is filled with helium. Under the reactor operation conditions, pellets produce a lot of heat by nuclear fission reactions and at the same time also produce lots of radioactive fission products. Cladding serve as the first barrier to accommodate radioactive fission product, needs to maintain its structural integrity under the reactor operation conditions. Cladding stresses can be effectively limited by controlling power increase rates. However, pellet manufacturing defects such as missing pellet surface (MPS), may lead to cladding local stress significantly high to cause cladding failure. Simulating the impact of these defects correctly can help prevent these types of failure. MPS defects are 3D phenomenon, needs 3D modeling method to study the influence of these defects on the cladding .In this paper, stress update algorithm is derived, with the help of ABAQUS (a commercial finite element software), simulated the thermal-mechanical behaviors of the MPS defects fuel rod with a 3D FEM and completed the sensitivity analysis of MPS defects size for the fuel performance. The models included in this simulation, including pellet irradiation swelling (fission gas products induced swelling and fission solid products induced swelling), pellet densification, pellet relocation, pellet thermal expansion, pellet irradiation creep, pellet irradiation hardening, cladding irradiation growth, cladding thermal expansion, cladding thermal creep, cladding irradiation creep, cladding irradiation hardening and gap heat transfer (gas heat conduction, radiation heat transfer and contact heat conduction) etc. Furthermore, considering the effects of irradiation and temperature on the material parameters such as thermal conductivity, specific heat and young’s modulus etc. According to the simulation result, showing that MPS defects have a large impact on the performance of fuel rods, this impact will be more obvious with the size of MPS defects increase. The MPS defects cause larger gap distance between pellet and cladding, higher gap distance causes smaller gap conductance, and then causes elevated temperature at the center of the pellet and in the region of the pellet adjacent to the defect. The cladding temperature is reduced in the area immediately across from the defect, and is elevated in neighboring areas. Meanwhile, MPS defects clearly have a significant effect on stress distribution and maximum stress of the cladding, cause high tensile stresses in the inner surface of the cladding and high compressive stresses on the outer surface of the cladding at the center of the defect. Around the boundaries of the defect, the stresses are reversed, with high compressive stresses on the cladding interior and high tensile stresses on the cladding exterior.

Metal matrix microencapsulated (M3) fuel is one of the research directions on Accident Tolerant Fuel (ATF). In this article, it provides one of ATF design which consists of BISO (Bistructural ISOtropic) particles embedded in a zirconium alloy matrix, and the cladding coating with silicon carbon (SiC). The temperature distribution of the ATF element has been built, and then the center temperature has also been calculated based on the operation parameters of the large-scale pressurized-water reactor. Simultaneity, the several factors of fuel failure is preliminary analyzed and calculated, especially the pressure shell failure mechanism.

Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys. Zirconium alloys are used as cladding material in almost all types of nuclear reactors, where creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper Zircaloy-2 alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations were done for zirconium alloy covered with diamond layer before and after corrosion and irradiation tests - ion beam irradiation tests and high temperature steam exposure.

Hot deformation characteristics of forged and β-quenched Zr-1.0Sn-0.3Nb-0.3Fe-0.1Cr (N18 alloy) in the temperature range 625–950°C and in the strain rate range 0.005–5 s −1 have been studied by uniaxial compression testing of Gleeble 3500. For this study, the approach of processing maps has been adopted and their interpretation done using the Dynamic Materials Model (DMM). Based on a series of true stress-true strain curves on various temperatures and strain rates, the flow stress has been summarized and both the strain rate sensitivity index (m) and deformation activation energy (Q) have been calculated by the constitutive equations that flow stress and the relationship of Z parameter and flow stress have been established subsequently. Furthermore, the efficiency of power dissipation (⬜) given by [2m/(m+1)] and improved by Murty has been plotted as a function of temperature and strain rate to obtain different processing maps at different true strain rates ranging from 0.1–0.7. Subsequently, the microstructures of the specimens after compression testing were characterized by electron channeling contrast (ECC) imaging techniques used an FEI Nova 400 field emission gun scanning electron microscopy (FEG-SEM). The results showed that: (i) The hyperbolic sine constitutive equation can describe the flow stress behavior of zirconium alloy, and the deformation activation energy and flow stress equation were calculated under the different temperature stages which insists that the deformation mechanism is not dynamic recovery. (ii) The hot processing maps and its validation were analyzed, which indicated that the DMM theory was reliable and could be adopted as useful tool for optimizing hot workability of Zr. The optimum parameters for extrusion and hammer forging were revealed on the processing maps of 830–950°C, 0.048–2.141 s −1 and 916–950°C, 2.465–5 s −1 . (iii) The microstructure of the ingot exhibits a typical lamellar Widmanstatten structure. Under the different strain rates, the grains formed by dynamic recrystallization existed normally in the central zone of the compression samples while the no uniformity of grain size increased with the increasing of strain rate. Meanwhile, due to the dynamic recrystallization as a thermal activation process, the grains size and uniformity increased with the increasing of temperature. In brief, microstructure analysis showed that continuous dynamic recrystallization and geometric dynamic recrystallization operated concurrently during the isothermal compressive deformation.