BOR-60 Reactor Research Articles

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe3+ and 3.42 MeV energy degraded He2+) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. The comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

Radiation-induced structure of austenitic steels with different nickel content under neutron irradiation in SM-3 and BOR-60 reactors

Comparative studies of the radiation-induced structure of austenitic steels with a nickel content of 10, 20 and 25 wt.%, irradiated sequentially in the SM-3 and BOR-60 reactors, as well as to higher damaging doses in the BOR-60 reactor, have been carried out. The phase composition, dislocation structure, pores, and radiation-induced segregations at grain boundaries were studied by high-resolution analytical methods of transmission electron microscopy, scanning electron microscopy, and atomic probe tomography. The formation of radiation-induced phase precipitates based on nickel has been established, and its volume fraction correlates with the level of radiation-induced segregations, and increases, the higher the nickel content in the steel. The values of barrier strength factors for radiation-induced structural elements in the studied steels are adjusted by calculation and experiment, which makes it possible to determine their contribution to radiation hardening.It is shown that the largest contribution to radiation hardening as a result of neutron irradiation in BOR-60 at high irradiation temperature up to 29 dpa is made by large radiation-induced precipitates of (G + γ') phases. It is shown that with an increase in the damaging dose, the main factor limiting the performance of internal devices will be radiation swelling, since the contribution to the change in properties from radiation-induced phases and radiation defects will not increase due to their density reaching saturation. Steel with 25 wt.% Ni exhibits the lowest level of swelling at high radiation doses, which makes it possible to consider it as a material-candidate for internals for promising VVER reactors with higher temperatures and longer service life.

Effect of reactor irradiation on the elemental composition, sizes and crystal structure of second phase precipitates in zirconium alloys E110 and E635

Using transmission electron microscopy, the authors of this paper looked at the effect of reactor irradiation in the temperature ranges of 40 to 70 and 290 to 340 oC on the elemental composition, crystal structure and particle sizes of the second phases in zirconium alloys E110 and E635. Regularities have been established behind redistribution of elements between the second phase particles and the solid solution of the surrounding matrix in the above alloys as a function of a rising dose and a changing irradiation temperature. The authors note that the redistribution of elements between the structural components of the above zirconium alloys follows a similar pattern in both the experimental reactor BOR-60 and the power reactor VVER-1000. A difference is shown in the effect produced by irradiation on the morphology and elemental composition of β-Nb phase and Laves phase precipitates in the above zirconium alloys in the low temperature region of 40 to 70 oC and in the temperature range of 290 to 340 oC. It is noted that the Laves phase particles become amorphous as the result of low-temperature irradiation, whereas the β-Nb particles retain their BCC structure. The authors offer their interpretation with regard to the staged transformation of the Laves phase HCP monocrystal structure into the BCC polycrystal structure induced by rising irradiation dose. The role of iron is shown as it leaves the Laves phase and enters the surrounding matrix during crystal lattice transformation. The paper describes stages when fine dispersed irradiation-induced particles appear in alloys E110 and E635 and when they change their size and concentration as the irradiation dose keeps rising. The difference between the particles are also noted in the said alloys.

Phase instabilities in austenitic steels during particle bombardment at high and low dose rates

Disruption of phase stability by energetic particle bombardment is a major challenge in designing advanced radiation-tolerant alloys and ion beam processing of nanocomposites. Particularly, ballistic dissolution susceptibility of different solute nanocluster species in alloys is poorly understood. Here, low dose rate neutron irradiations were conducted on a Fe-Cr-Ni based austenitic steel in the BOR-60 reactor (9.4 × 10−7 dpa/s, 318 °C) followed by accelerated dose rate ion irradiations at multiple temperatures (~ 10−3 dpa/s, 380 – 420 °C). Using atom probe tomography, the stability of radiation-enhanced Cu-rich and radiation-induced Ni-Si-Mn-rich nanoclusters was evaluated. During neutron irradiation, Cu-rich clusters nucleated with their core concentrations progressively increasing with dose, while Ni-Si-Mn-rich clusters formed and evolved into G-phase precipitates. Ion irradiations dramatically altered the nanoclusters. Cu-rich clusters were ballistically dissolved, but Ni-Si-Mn-rich clusters remained stable and coarsened with dose at 400 and 420 °C, highlighting that different nanocluster species in a single microstructure can have innately distinct ballistic dissolution susceptibilities. Solute-specific recoil rates were incorporated into the Heinig precipitate stability model, which shows that in addition to radiation-enhanced diffusion, recovery from ballistic dissolution depends on solute concentration gradient near cluster interfaces. The combined experimental-modeling study quantified the critical temperatures and damage rates where ballistic dissolution dominates for each cluster species.

Solute segregation and precipitation across damage rates in dual-ion–irradiated T91 steel

Dual-ion irradiations using 5.0 MeV defocused Fe2+ ions and co-injected energy degraded 2.00 to 2.85 MeV He2+ ions were conducted on a Fe9CrMo ferritic-martensitic steel T91 to 17 dpa at a damage rate range of 5 × 10−5 dpa/s to 3 × 10−3 dpa/s at 445°C, followed by characterization of the microstructure using conventional and scanning transmission electron microscopy. Radiation induced Ni/Si clusters and radiation induced segregation were quantified using energy dispersive X-ray spectroscopy at each condition and were compared with the same material irradiated in the BOR-60 reactor and in the as-received condition. No significant Cr segregation was found at lath boundaries after dual-ion irradiation, while Ni and Si enrichments both decreased with increasing damage rate leading to a sharp decrease in the density of Ni/Si clusters with damage rate. Increased point defect recombination at higher ion damage rates likely reduced the Ni/Si cluster density compared with BOR-60. Although the overall vacancy concentration and diffusion are enhanced by the irradiation damage rate, the lack of time for thermal diffusion and ballistic displacements of solutes are significant limiting factors for Ni/Si cluster formation. This work demonstrates the effect of irradiation damage rate on elemental segregation and clustering when using ion irradiation to simulate reactor irradiation.

Investigation of the microstructure of the base and weld metal of steel Fe–0.08C–18Cr–10Ni–Ti after low-temperature irradiation in the BOR-60 reactor in the range of damage doses from 40 to 101 displacements per atom

The microstructures of irradiated samples of Fe–0.08C–18Cr–10Ni–Ti steel were studied by transmission electron microscopy (TEM). Specimens of the base metal (BM) and the weld metal (WM) were irradiated in the BOR-60 reactor at temperatures from 330 to 350°C to doses of 43 dpa (BM) and 40 dpa (WM), 96 dpa (BM) and 101 dpa (WM). New data on the quantitative microstructural characteristics of dislocation loops, titanium carbonitrides, fine G-phase, and voids in irradiated specimens in the range of damage doses from 40 to 101 dpa were obtained during the studies of specimens of the base metal and the weld metal.

Status on performance study of mixed nitride fuel pins of BREST reactor type

Within the framework of the “PRORYV” project the “The comprehensive program for the calculation and experimental justification of fuel pins with mixed uranium-plutonium nitride fuel of BN-1200 and BREST reactors for the period up to 2020” has been developed Tpoянoв et al., 2015 [1]. The purpose of the program is to judge fuel pins performance, to ensure stability and required quality of the experimental technology for manufacturing of mixed nitride, fuel pins and experimental fuel assemblies (EFA) for testing in BOR-60 and BN-600 reactors, to optimize the design and technology of fuel pins with mixed nitride for the developed BN-1200 and BREST-OD-300 reactors based on the results of reactor testing. Currently, the program aims at justifying the starting operation stages of BREST-OD-300 and BN-1200 reactors. The following results were obtained to judge the fuel performance at the starting operation stage of BREST-OD-300 reactor (the maximum mixed nitride burn-up of 6 at%, the maximum damage dose of 89 dpa): 1. A set of out-of-pile investigations has been performed in order to obtain the necessary data on the properties of fuel and cladding materials, which are required for the development of calculation models of mixed nitride behavior under irradiation and design justification of their performance, including in design accidents. 2. To March 2021 the irradiation of eight BN-600 EFAs with EP823-Sh cladding was finished. In EFA-11 the maximum burn-up 9.0 at% and the maximum damage dose 107.6 dpa have been reached. PIE of fuel pins of two OU irradiated in BOR-60 and four EFAs irradiated in BN-600 with maximum burn-up of 5 at% and damage dose of 74.7 dpa are completed. Experimental data required for verification of calculation results of temperature and stress-strain state of mixed nitride pins with EP823-Sh steel cladding, as well as for evaluation of fuel pin performance and establishment of allowable lifetime of the BREST-type fuel pins with mixed nitride have been obtained. 3. The KORAT, DRAKON and BERKUT-U codes were modified and verified. Certification of the DRAKON code has been completed and the BERKUT-U code has been submitted for certification. Obtaining the permission to extend the irradiation of EFA beyond the parameters planned in the Program was possible due to the fact that independent concurrent calculations with the use of KORAT, DRAKON and BERKUTU codes were performed during the development of technical designs of fuel pins with mixed nitride in justification of the possibility to extend the irradiation of EFA-11, EFA 14, EFA-15 in BN-600 reactor. All codes confirmed this possibility. Positive results of irradiation of BREST-type fuel pins in BOR-60 reactor were also the most important argument for obtaining permission. In order to improve fuel pins with (U, Pu)N and further burn-up increase, the technologies of their production were developed and optimized, the design of the pellets was changed (chamfer) and also fuel pins with liquid-metal sub-layer have been considered.

Results of post-irradiation examinations of mixed nitride pins with gas and liquid metal sub-layers

Today the investigations have been completed on helium-bonded fuel pins with mixed uranium-plutonium nitride of BN-600/BN-800, BN-1200 and BREST reactors types after irradiation as a part of ten EFAs of BN-600 reactor up to maximum burn-up of 7.5; 6.0 and 4.5at%, respectively. Also the PIE of mixed nitride pins of BREST type with helium and lead sub-layers after intermediate tests to maximum burn-up of 4,8 and 3,9at% as part of dismountable EFAs of BOR-60 reactor are over. As a result of postirradiation examinations, the main intra-fuel pin processes that affect on materials properties change and fuel pins state, their relationship with the initial nitride state were identified. The regularities and quantitative characteristics of nitride swelling, the behavior of fission products, cladding corrosion and mechanical properties change are revealed. The average for the irradiation time rate of fuel swelling in cross sections near the core midplane in gas-bonded pins of different EFAs irradiated in BN-600 reactor is equal to (1.6 - 2.0) %/at%. According to the results of irradiation in BOR-60 reactor, the fuel swelling rate in lead-bonded pins is lower than in helium-bonded pins: 1.4 ± 0.2 and 1.7 ± 0.2 %/at%, respectively. The presence of carbon and oxygen impurities in the fuel can lead to local areas of claddings carburization and oxidation, randomly distributed along the height and perimeter of its inner surface. The key characteristics of the fuel that determine the cladding carburization and oxidation, and ways to prevent them, are determined. Cladding corrosion in lead-bonded pin is caused by selective dissolution of cladding steel components. The maximum depth of the corrosion zone is located in the upper part of the fuel column and is equal to 40 microns. The results of cladding mechanical tests obtained using longitudinal segment samples and tubular samples loaded by internal gas pressure showed the significant margin of strength and ductility of cladding materials along the entire height of fuel pins both at nominal operating temperatures and at room temperature. High values of strength characteristics indicate a weak influence of corrosion on the strength of studied claddings materials. The mechanical cladding properties of a lead – bonded pins have the same pattern of change along the pin height depending on irradiation and testing temperatures, as of fuel pins with helium sublayer, keeping enough ductility in the area of low-temperature irradiation embrittlement. So, for example, at irradiation and test temperatures of 350-380 °C, the values of the strength limit from 1070 to 1200 MPa were obtained with the values of the total elongation of not less than 2 %. Thus, the results of post-irradiation examinations confirmed the performance of mixed nitride pins at the achieved test parameters and the possibility of their irradiation prolongation to higher fuel burn-up.

Unification of the BN-1200 reactor core designs with MOX and MNUP fuel

Two types of mixed uranium-plutonium fuel are considered for BN-1200 reactor (Matveev and Khomyakov, 2012) within the framework of Proryv Project: conventional, well-mastered MOX fuel and advanced nitride fuel (MNUP fuel) having higher density. Decision about the expediency of use of this or that type of fuel in the first BN-1200 reactors at the initial stage of their operation will be made taking into account the results of MNUP fuel pins tests carried out in BOR-60 and BN-600 reactors. In this report the information on basic engineering decisions taken in the design of BN-1200 reactor core (Vasiliev et al., 2010; Shepelev et al., 2013; Vasiliev et al., 2017) is given, features of cores with MOX and MNUP fuels are analyzed, possible variants of cores with unified design of fuel pins and FSA are presented, estimates of core characteristic change with increased fuel burnup are given, problems of transition from MOX fuel to MNUP fuel are considered.

Development of technologies for production of fuel with minor actinides

The main results and plans for the future in the framework of R&D on the involvement of americium, neptunium (MA) in mixed nitride fuel of fast reactors are presented. Depending on the method and amount of MA introduced into the mixed nitride, R&D is carried out in three directions: homogeneous introduction of minor actinides; intra-fuel pin heterogeneity and intra-fuel assembly (FA) heterogeneity of minor actinides. At the first stage, the technology of homogeneous transmutation is being tested using dismountable FA (OU) in the BOR-60 reactor. Each OU includes 7 fuel pins. An intermediate unloading of fuel pins is planned with the corresponding additional loading of new ones and irradiation prolongation. The irradiation parameters of subsequent experimental fuel pins with MA will be specified after determining the characteristics of fuel pellets of mixed nitride with MA. In order to study the effect of MA additives on the cladding corrosion from the first batches of pellets, at the end of 2018, three experimental fuel pins with Np and Am with pellet density of ~10 g/cm3 in OU-16 were manufactured and loaded in BOR-60. Upon reaching the maximum fuel burn-up of 1.36 at% and a dose of 10.7 displacements per atom (dpa) OU-16 was unloaded for intermediate examination of fuel pins with MA. To minimize the loss of Am at the stage of fuel pellets sintering, the method of high-voltage electric impulse consolidation for pellets with Am is considered.

Design and first operation of the MACARON irradiation experiment of Sodium Fast reactor absorber pins in the BOR-60 reactor

The development of the next generation of Sodium-cooled Fast Reactors (SFRs) needs to integrate innovative options regarding the core design in order to comply with the objectives of the 4th generation reactors in terms of safety, cost and performance. In this framework, the aim to design fast core with economy improvements leads to limit the number of control and shutdown systems and to extend their lifetime. These objectives require choosing absorber material with maximal efficiency in terms of neutronic absorption.The French concept of absorber pin is based on the use of boron carbide (B4C) pellets with 10B various enrichments and the absorber element consists of a vented non-leaktight pin with a permanent sodium bond between the pellet column and the stainless steel clad.In order to test an innovative concept of absorber pin and with the aim to improve in one hand, the in pile behavior understanding and in the other hand, the absorber rod modelling codes, an experimental program named MACARON is undergone. This program consists in an irradiation test in the BOR-60 reactor following by a complete NDE and PIE program. Two short absorber pins are irradiated currently. This program is performed by both CEA and RIAR teams.A brief description of the experiment is first provided, in terms of objectives. The main characteristics of absorber pin with the main design criteria are detailed. The manufacturing steps and the characterization of the experimental absorber pins are given with as manufactured results.The design studies of the Irradiation Rig and the neutronic and thermal hydraulic calculations are described in details. The irradiation test has begun in September 2018 and the first experimental results are provided. The huge program of Non-Destructive and Destructive Examinations is detailed with experimental objectives and methods that will be used.

Determining the Specific Activity of Tritium in Samples of Spent Nuclear Fuel

Procedure for determining the specific activity of tritium in samples of spent nuclear fuel (SNF) was developed and methodologically certified. The procedure was used to determine the specific activity of tritium in samples of mixed nitride uranium-plutonium nuclear fuel irradiated in BOR-60 and BN-600 reactors to be 8.5×106 ± 1.1×105 and 2.6×106 ± 2.0×105 Bq/g, respectively. The content of tritium in the shell of a nuclear fuel element irradiated in a BN-600 reactor was estimated, and the efficiency of the voloxidation procedure of the mixed nitride uranium-plutonium spent nuclear fuel for removing tritium was evaluated.

РАСЧЁТНЫЙ АНАЛИЗ ТЕПЛОГИДРАВЛИЧЕСКОЙ УСТОЙЧИВОСТИ ОБРАТНОГО ПАРОГЕНЕРАТОРА ИССЛЕДОВАТЕЛЬСКОЙ ЯДЕРНОЙ УСТАНОВКИ МБИР

The stability of the circulation in the steam-water circuit, that is, the preservation of the thermohydraulic parameters within acceptable limits for perturbations characteristic of the operation, largely determines the reliability of the steam generator itself. Under certain conditions, pulsating pressure, flow and temperature pulsations may appear in boiling apparatus. For the first time, ripples were encountered during the creation of once-through boilers, and the harmful effect of flow instability on the apparatus design led to the need for a detailed study of the problem. In this work, a computational analysis of the thermal-hydraulic stability of the reverse steam generator (OPG) of the MBIR nuclear research facility (NRF) is carried out. The need for such a study is due to the fact that the OPG consists of three parallel connected steam generating modules. With such a design, water flow rate pulsations may appear, including with overturning flow rates in separate steam-generating channels. To avoid this, it is necessary to conduct throttling at the inlet to each OPG module, therefore, in this study, the hydraulic resistance of the chokes is determined. The research methodology is based on the use of a numerical thermohydraulic model of a reverse steam generator, made by means of a certified thermohydraulic design code HYDRA-IBRAE/LM/V1. The OPG RNU MBIR at the economizer and superheating sections has twisted heat exchange intensifiers that direct water and steam both in the longitudinal and transverse directions relative to the bundle of heat exchange tubes. This circumstance required a special approach in calculating the hydraulic resistance and heat transfer coefficients along the path of the OPG modules. After the creation of the computational model, it had to be verified, and since the design of the OPG of the NRF MBIR is similar to the design of the OPG of the BOR-60 reactor, the experimental data obtained on the OPG BOR-60 were used for verification. The calculations of the normal operation modes of the OPG NRF MBIR after the throttling of the modules confirmed the absence of fluctuations in the flow rate in these modes.

Irradiation Creep Behavior of Steels 08Х18Н10Т and 07Х16Н12М3Т Irradiated in the BOR-60 Reactor in a Temperature Range of 330 to 350°С

The paper describes the results of post-irradiation examinations for pressurized specimens that varied in design and were made of austenitic steels 08Х18Н10Т and 07Х16Н12М3Т. They were under irradiation testing in the BOR-60 reactor to attain different damage dose values. The contribution of irradiation creep to the overall change in diameter was studied. Irradiation creep moduli data were obtained for austenitic steels 08Х18Н10Т and 07Х16Н12М3Т at the stage of steady-state creep in the damage dose range of 5–90 dpa.

Characterisation of nano-scale precipitates in BOR60 irradiated T91 steel using atom probe tomography

Atom probe tomography has characterised the microstructural changes in T91 steel after BOR60 reactor irradiation at five temperatures between 376 °C and 524 °C to doses between 14.6 dpa and 35.1 dpa. Irradiation-induced precipitation and segregation to carbide/matrix interface induced by neutron irradiation has been characterised. Atom probe tomography characterisation shows that Mn, Ni, Si -rich (MNS-rich) clusters form in T91 steel irradiated in BOR60 reactor at temperatures between 376 °C and 415 °C, which is not observed at higher temperatures 460 °C and 524 °C. The number density, volume fraction and composition of MNS-rich clusters have been characterised. Ni, Mn, Si and P is found to segregate at carbide matrix interface after irradiation at lower temperature and only P segregation is observed at 524 °C.

Development of Fuel Pin with Uranium-Plutonium Nitride Fuel and Liquid-Metal Sublayer

The status of development work on fuel pins with fast-reactor nitride fuel and a liquid-metal sublayer, which is being considered in Project Breakthrough (Project Proryv) together with fuel pins with a helium sub-layer, is presented. Such fuel pins are shown to have the potential to improve the core characteristics of the advanced high-power reactors BR-1200 and BN-1200. The basic results for the development and fabrication of experimental nitride-fuel pins with different variants of the lead sublayer and the results of their tests in the BOR-60 reactor and post-reactor studies are presented. The problems of fabricating fuel pins with a lead sublayer are noted, possible solutions are discussed, and the required additional studies are indicated.

Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: Effect of temperature

The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457 °C) to doses between ∼14.6–18.6 displacements per atom (dpa). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the influence of irradiation temperature on microstructure evolution. Mn/Ni/Si-rich (G-phase) and Cr-rich (αʹ) precipitates were observed within martensitic laths and at various defect sinks at 650 and 690 K (377 and 417 °C). For both G-phase and αʹ precipitates, the number density decreased while the size increased with increasing temperature. At 730 K (457 °C), within martensitic laths, a very low density of large G-phase precipitates nucleating presumably on dislocation lines was observed. No αʹ precipitates were observed at this temperature. Both a <100> and a/2 <111> type dislocation loops were observed, with the a <100> type being the predominant type at 650 and 690 K (377 and 417 °C). On the contrary, very few dislocation loops were observed at 730 K (457 °C), and the microstructure was dominated by a/2 <111> type dislocation lines (i.e., dislocation network) at this temperature. Small cavities (diameter < 2 nm) were observed at all three temperatures, whereas large cavities (diameter > 2 nm) were observed only at 690 K (417 °C), resulting in a bimodal cavity size distribution at 690 K (417 °C) and a unimodal size distribution at 650 and 730 K (377 and 457 °C). The highest swelling (%) was observed at 690 K (417 °C), indicating that the peak of swelling happens between 650 and 730 K (377 and 457 °C).

РАСЧЕТНО-ЭКСПЕРИМЕНТАЛЬНОЕ ОПРЕДЕЛЕНИЕ ПРОСТРАНСТВЕННО-ЭНЕРГЕТИЧЕСКОГО РАСПРЕДЕЛЕНИЯ НЕЙТРОНОВ В БОКОВОМ ЭКРАНЕ РЕАКТОРА БОР-60. ЧАСТЬ 2

The BOR-60 reactor is actively used for conducting irradiation experiments in support of nuclear reactors both under operation, construction and design. Most of these experiments are conducted in the reactor core using various types of irradiation rigs. However, in recent years there has been keen interest in conducting reactor tests in the BOR-60 blanket. This is associated with several factors: the availability of blanket cells, weak influence on the reactor neutronic characteristics and safety, wide range of irradiation conditions (temperature, neutron flux density, damage doses, neutron spectrum), and cost of irradiation. The experience in BOR-60 blanket computational and experimental studies showed an ill agree-ment between the calculated and experimental neutronic characteristics. In addition to no orientation of the samples under study relative to the core center, there is a significant radial non-uniformity in the neutronic characteristics distribution in the BOR-60 blanket that limits its application for experimental studies. Therefore, detailed computational and experimental studies of the spatial-energy distribution of neutrons in the BOR-60 steel blanket is an important and urgent task.

Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: effect of dose

A ferritic/martensitic steel, HT9, was irradiated in the BOR-60 reactor to ∼17.1 and ∼35.1 displacements per atom (dpa) at 650 ± 23 K (377 ± 23 °C). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the role of irradiation dose on the microstructure evolution. Radiation-induced Mn/Ni/Si-rich (G-phase) and radiation-enhanced Cr-rich (αʹ) precipitates were observed within the martensitic laths at all doses. In addition, the G-phase was also observed to precipitate heterogeneously at various defect sinks. The number density for these second-phase precipitates decreases while the size increases with increasing dose, resulting in an increase of the volume fraction. Both a <100> and a/2 <111> type loops were observed with the a <100> type being the predominant type at both doses. The proportion of a <100> loops is consistent with that previously observed in HT9 ion irradiated to similar doses at ∼693–743 K (∼420-470 °C). Only small cavities (diameter < 2 nm) were observed at ∼17.1 dpa whereas both small and large cavities were observed at ∼35.1 dpa, resulting in a bi-modal cavity size distribution at this dose. Alloying elements, Ni and Si, were observed to segregate to the cavity surface, forming Ni/Si-rich shells around the cavities. The swelling at ∼17.1 dpa is evaluated at ∼0.02% while the swelling at ∼35.1 dpa is found to be ∼0.07% with variations from grain to grain. attributed to the spatial variation of the density of large cavities (in different grains). The swelling data obtained in this study was compared with the neutron data of F/M steels available in the literature.

Results of Investigations of BREST-Type Reactor Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel, Irradiated in BOR-60 and BN-600

The complex program of computational and experimental validation of dense fuel for fast reactors in Project Breakthrough provides for testing of fuel rods with mixed uranium-plutonium nitride fuel in the BOR-60 and BN-600 reactors. The irradiation parameters and post-reactor studies of the fuel rods of a BREST reactor are presented. Testing of fuel rods in BOR-60 to maximum burnup 5% h.a. and damaging dose 74.8 dpa and in BN-600 to maximum burnup 4.5% h.a. and damaging dose 53 dpa did not result in depressurization of the rods. The results of post-reactor tests of irradiated fuel rods – swelling of fuel, gas release from the fuel, mechanical and corrosion properties of the cladding – are presented.