BOR-60 Fast Reactor Research Articles

Emulation of neutron-irradiated microstructure of austenitic 21Cr32Ni model alloy using dual-ion irradiation

In this study, the capability of heavy-ion irradiation to emulate neutron irradiation was demonstrated on an austenitic 21Cr32Ni type ternary model alloy. The model alloy used in this study is chemically analogous but compositionally simpler than of alloy 800H, which is a candidate austenitic Ni alloy which has been proposed for use in Generation IV reactors.The microstructure of the 21Cr32Ni model alloy irradiated in the BOR-60 fast reactor to 17.1 dpa and 35 dpa at ∼380 °C was characterized using transmission electron microscopy (TEM). The 17.1 dpa BOR-60 irradiated microstructure was then compared with the microstructure of the same material developed under dual-ion (DI) irradiation using various He/dpa ratios between 0.1 and 16.6 appm He/dpa in the temperature range of 430 °C-500 °C. The results showed that both neutron and DI irradiation of 21Cr32Ni model alloy produced dislocations in the form of a dislocation network as well as {111}-type faulted dislocation loops, cavities, and radiation-induced Ni enrichment at radiation-induced sinks. When the dose and the He/dpa ratio were kept similar to those in neutron irradiation, DI irradiation of the 21Cr32Ni model alloy at 460 °C resulted in over-nucleation of small cavities and in a high density of faulted dislocation loops compared to those observed in the fast-neutron irradiated alloy of the same heat irradiated at ∼380 °C. The optimal condition for reproducing the neutron-irradiated microstructure was DI irradiation at 460 °C and 0.1 appm He/dpa. In that case, the faulted loop and cavity size distributions in the BOR-60 irradiated 21Cr32Ni model alloy samples closely matched with those measured in the DI irradiated 21Cr32Ni model alloy sample. The fact that the He/dpa is an order of magnitude smaller than the helium generation rate for fast neutron irradiation, stops over nucleation and allows for the development of a similar microstructure as for neutron irradiation.

Emulation of fast reactor irradiated T91 using dual ion beam irradiation

Dual ion irradiations using 5 MeV defocused Fe2+ ions and co-injected He2+ ions were conducted on a ferritic-martensitic steel alloy, T91, in the temperature range of 406 °C–570 °C over a damage range of 14.6–35 dpa followed by characterization of the microstructure using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). Dislocation loops were observed to increase in diameter and decrease in density with temperature until only network dislocations were observed at the highest temperatures of 520 °C and 570 °C. Swelling exhibited the expected bell-shaped trend with temperature following the number density of cavities, peaking at 460 °C and with a bimodal size distribution except at 520 °C and 570 °C. Nickel- and silicon-rich clusters formed under dual ion irradiations near the surface at all but the highest temperatures of 520 °C and 570 °C. Very little Cr and Si segregation was observed at lath boundaries while Ni enriched at all temperatures examined. Segregation of Cr and Ni appeared to saturate by 17 dpa, while Si enriched up to 35 dpa. The dislocation and cavity microstructures of dual ion irradiated T91 and T91 irradiated in the BOR-60 fast reactor matched extremely well using a temperature shift of +60–70 °C. However, segregation to grain boundaries and formation of nickel-silicon rich clusters were minimal in the dual ion irradiated T91 and less than that in T91 irradiated in the BOR-60 fast reactor.

Microstructure evolution of T91 irradiated in the BOR60 fast reactor

Microstructures of T91 neutron irradiated in the BOR60 reactor at five temperatures between 376 °C and 524 °C to doses between 15.4 and 35.1 dpa were characterized using transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT). Type a<100> dislocation loops were observed at 376–415 °C and network dislocations dominated at 460 °C and 524 °C. Cavities appeared in a bimodal distribution with a high density of small bubbles less than 2 nm at irradiation temperatures between 376 °C and 415 °C. Small bubbles were also observed at 460 °C and 524 °C but cavities greater than 2 nm were absent. Enrichment of Cr, Ni, and Si at the grain boundary was observed at all irradiation temperatures. Radiation-induced segregation (RIS) of Cr, Ni and Si appeared to saturate at 17.1 dpa and 376 °C. The temperature dependence of RIS of Cr, Ni and Si at the grain boundary, which showed a peak Cr enrichment temperature of 460 °C and a lower peak Ni and Si enrichment temperature of ∼400 °C, was consistent with observations of RIS of Cr in proton irradiated T91, suggesting that the same RIS mechanism may also apply to BOR60 irradiated T91. G-phase and Cu-rich precipitates were observed at 376–415 °C but were absent at 460 °C and 524 °C. The absence of G-phase at 524 °C could be related to the minimal segregation of Ni and Si in that condition.

Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

Investigation of high temperature annealing effectiveness for recovery of radiation-induced structural changes and properties of 18Cr–10Ni–Ti austenitic stainless steels

A complex study of structural state and properties of 18Cr–10Ni–Ti austenitic stainless steel after irradiation in BOR-60 fast research reactor (in the temperature range 330–400 °С up to damaging doses of 145 dpa) and in VVER-1000 light water reactor (at temperature ∼320 °С and damaging doses ∼12–14 dpa) was performed. The possibility of recovery of structural-phase state and mechanical properties to the level almost corresponding to the initial state by the recovery annealing was studied. The principal possibility of the recovery annealing of pressurized water reactor internals that ensures almost complete recovery of its mechanical properties and microstructure was shown. The optimal mode of recovery annealing was established: 1000 °C during 120 h.

Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320°C to doses between approximately 4 and 12dpa or by a 3.2MeV proton beam at 360°C to a dose of 5.5dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni–Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

Prediction of void swelling in the baffle ring of WWER-1000 reactors for service life of 30–60 years

Major internal components of WWER-type nuclear reactors are made from annealed 18Cr10NiTi steel, a close analog to AISI 321. Void swelling of the baffle ring in particular could be a major factor limiting operation of the reactor beyond the current 30years license. A predictive swelling equation is needed to forecast the spatial variation of swelling so as to identify those areas requiring additional attention. Available data on the swelling of this steel arising from irradiation in the BOR-60 fast reactor was combined with data from a heavy ion accelerator at higher displacement rates to formulate a predictive equation of swelling for the WWER-relevant range of temperature, irradiation dose and dose rates. This equation was used to estimate the swelling distribution over a cross-section of the baffle ring of a WWER reactor during a service life up to 60years, reaching a local maximum of ∼30% swelling. It was shown that void swelling extends over a larger portion of the baffle ring than previously expected and exhibits a very complex and irregular distribution in response to complex distributions of temperature and dpa rate. Most importantly, as operation is extended beyond 30years several areas of the ring may experience swelling in excess of 10% where void-induced embrittlement begins and one area may exceed 20% after 45years, a swelling level where a complete loss of ductility is known to occur.

Atom probe tomography of nanoscaled features of oxide-dispersion-strengthened ODS Eurofer steel in the initial state and after neutron irradiation

Atom probe tomography has been used to investigate nanoscale features in the yttrium oxide dispersion strengthened steel ODS Eurofer, which is a perspective structural material for the reactor cores. In the initial material, a large number (∼2 × 1024 m−3) of ultrafine (∼2.5 nm in diameter) clusters enriched in yttrium, oxygen, nitrogen, and vanadium have been revealed. The investigation of the ODS Eurofer steel irradiated at 330°C to 32 dpa in the BOR-60 fast reactor has also revealed a large number of ultrafine (1–3 nm in diameter) nanoclusters significantly enriched in yttrium, oxygen, manganese, and chromium. In the irradiated material, an increase in the concentration of clusters and changes in the chemical composition of the clusters and matrix have been noted. The irradiation by fast neutrons leads to a partial transition of vanadium from the clusters into the surrounding matrix and to a general increase in the concentrations of yttrium and oxygen in the volumes under investigation.

Prediction of swelling of 18Cr10NiTi austenitic steel over a wide range of displacement rates

The internal components of pressurized water reactors of Russian types WWER-440 and WWER-1000 are constructed of annealed 18Cr10NiTi steel, a close analog to AISI 321. Void swelling of the internals is a concern for plant life extension and predictive equations are required to assess the potential of swelling in critical components such as the baffle ring or reflection shield that surrounds the WWER core. The only previously available swelling data for this steel were derived at higher than PWR-relevant displacement rates in the BOR-60 fast reactor. The swelling equation previously developed from these data does not incorporate the effect of displacement rate on swelling. Using heavy-ion irradiation at very high dpa rates (10 � 2 and 10 � 3 dpa s � 1 ) and doses (5–100 dpa) and coupling the results to available neutron data a swelling equation has been developed that specifically incorporates the effect of dpa rate on void swelling. Experimental results allow description of the swelling peak, the incubation period and the steady-state swelling rate over a wide range of irradiation temperature. For the first time it appears possible to describe both ion and neutron data on this steel within the framework of a single empirical model. Swelling maps constructed from this model permit forecasting of the behavior of the steel in WWERs under the required irradiation conditions, not only at already attained exposure doses, but more importantly to higher dose levels that will be reached following plant life extension.

Study of helium embrittlement in boron doped EUROFER97 steels

To simulate helium effects in Reduced Activation Ferritic/Martensitic steels, experimental heats ADS2, ADS3 and ADS4 with the basic composition of EUROFER97 (9%Cr-WVTa) were doped with different contents of natural boron and separated 10B-isotope (0.008–0.112 wt.%) and irradiated in High Flux Reactor (HFR) Petten up to 16.3 dpa at 250–450 °C and in Bor-60 fast reactor in Dimitrovgrad up to 31.8 dpa at 332–338 °C. The embrittlement and hardening are investigated by instrumented Charpy-V tests with subsize specimens. Complete burn-up of 10B isotope under neutron irradiation in HFR Petten led to generation of 84, 432 and 5580 appm He and partial boron-to-helium transformation in Bor-60 led to generation of 9, 46, 880 appm He in ADS2, ADS3 and ADS4 heats, respectively. At low irradiation temperatures T irr ⩽ 340 °C the boron doped steels show progressive embrittlement with increasing helium amount. Irradiation induced DBTT shift of EUROFER97 based heat doped with 1120 wppm separated 10B isotope could not be quantified due to large embrittlement found in the investigated temperature range. At T irr ⩽ 340 °C helium induced extra embrittlement is attributed to material hardening induced by helium bubbles and described in terms of phenomenological model.

Very high swelling and embrittlement observed in a Fe–18Cr–10Ni–Ti hexagonal fuel wrapper irradiated in the BOR-60 fast reactor

The highest void swelling level ever observed in an operating fast reactor component has been found after irradiation in BOR-60 with swelling in Kh18H10T (Fe–18Cr–10Ni–Ti) austenitic steel exceeding 50%. At such high swelling levels the steel has reached a terminal swelling rate of ∼1%/dpa after a transient that depends on both dpa rate and irradiation temperature. The transient duration at the higher irradiation temperatures is as small as 10–13 dpa depending on which face was examined. When irradiated in a fast reactor such as BOR-60 with a rather low inlet temperature, most of the swelling occurs above the core center-plane and produces a highly asymmetric swelling loop when plotted vs. dpa. Voids initially harden the alloy but as the swelling level becomes significant the elastic moduli of the alloy decreases strongly with swelling, leading to the consequence that the steel actually softens with increasing swelling. This softening occurs even as the elongation decreases as a result of void linkage during deformation. Finally, the elongation decreases to zero with further increases of swelling. This very brittle failure is known to arise from segregation of nickel to void surfaces which induces a martensitic instability leading to a zero tearing modulus and zero deformation.

Radiation damage in beryllium at 70–440°C and neutron fluence (0.3–18)·1022 cm−2 (E n &gt; 0.1 MeV)

The results of investigations of the damage to samples of five brands of beryllium (TV-56, TV-400, TV-30, TIP, and DIP), prepared using the hot-pressing and extrusion technology and also hot isostatic pressing, are presented. The beryllium samples were irradiated in the channels of the core of an SM reactor and as part of the photoneutron source of the BOR-60 fast reactor at 70–440°C and neutron fluence (0.3–18)·1022 cm−2 (En > 0.1 MeV). The experimental program included mechanical tests for stretching and compression, measurements of the microhardness, swelling and specific thermal conductivity, as well as a study of the microstructure. The experimental results are used to construct the dose dependences of the change of the physicomechanical characteristics of beryllium at different irradiation temperatures and the trends in these dependences are analyzed.

Radionuclide production using a fast flux reactor

The production of 89Sr, 32,33P, 35S via the ( n, p)-reaction, and 117mSn, 153Gd via the ( n, γ)-reaction using the BOR-60 fast flux reactor was experimentally studied. Test samples were irradiated in the active core of the BOR-60 reactor with fast neutron flux of 1 × 10 15 cm −2s −1 for 40–100 effective days. Gadolinium-153 was produced in a radial blanket cell, characterized by a modified (“softened”) neutron spectrum. Data on target materials, procedure of irradiated target reprocessing, radionuclide yield, and specific activity are summarized in the report. The results of the experiment showed that large-scale production of the radionuclides listed above is possible using a fast-flux reactor.

Isotopic composition and residual heat release of oxide fuel with a high americium content after irradiation in bor-60

Calculations of the isotopic composition, residual heat release, and helium content in mixed oxide fuel with a high americium content during a run and during prolonged cooldown were performed for a BOR-60 fast reactor as part of the preparation for the experimental program RETSIKL. The fuel composition consisted of 75% enriched uranium and 20 or 50% by mass of americium-curium fraction and rare-earth elements of spent VVER-1000 fuel (burnup 40 GW-daylton, 10 year holding period). The americium transmutation rate is about 16.5% for the reactor run, the mass of the accumulated curium is about 13% of the mass of the transmuted americium, the plutonium mass is 35–45% of the latter (almost all on the plutonium consists of nonfissioning isotopes). The total helium accumulation can reach 1 kg/ton of heavy metals. The residual heat release due to α decay242Cm in about 1 month is ∼1 MW/ton, which has important consequences for planned or emergency shutdown of a reactor. 1 figure, 5 tables, 3 references.

Change in the properties of FeCrNi and FeCrMn austenitic steels under mixed and fast neutron irradiation

Detailed investigations are performed on mechanical properties, swelling and structure of different types of FeCrNi and FeCrMn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800°C and the maximum achieved level of damage doses is 60 dpa for FeCrMn steel (with 4–5% of Ni) and 30 dpa for steels of the C12Cr20MnWT type. Presented are dose dependencies of swelling and mechanical properties of FeCrNi and FeCrMn steels. It is shown that at temperatures below 530°C the investigated FeCrMn steel systems are less susceptible to swelling as compared to FeCrNi ones. FeCrMn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400°C and much higher embrittlement after irradiation from 350 up to 400°C in the fast spectrum in comparison with FeCrNi steels. Higher hardening rate of FeCrMn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in FeCrMn steels under irradiation are considered taking into account austenite stabilization in the initial state.

Effects of fast neutron irradiation on the mechanical properties of precipitation-hardened copper alloys

Samples of three copper alloys were irradiated in the BOR-60 fast reactor up to 1.6 × 10 2 n/cm 2. Irradiation temperatures were 350–460°C. Mechanical properties and the structure of the irradiated materials were studied. The precipitation-hardened alloys exhibited a decrease of the yield strength with increasing irradiation temperature and dose. After irradiation to (1.5−1.6) × 10 22 n/cm 2 at about 450°C large precipitates were observed in the alloys, which were accompanied by softening.

Radioactivity in the primary circuit of the BOR-60 fast reactor

Radioactivity in the primary circuit of the BOR-60 fast reactor