Search Results (13)

This study presents a comprehensive analysis of the microstructural characteristics and chemical composition of base and weld materials from reactor pressure vessels in the first (units 1 and 2) and second (unit 8) generations of Russian VVER 440 reactors at the Greifswald nuclear power plant. We measured the specific activities of ⁶⁰Co and ¹⁴C in activated samples from units 1 and 2. ⁶⁰Co, with its shorter half-life (t_1/2 = 5.27 a), is a key dose-contributing radionuclide during decommissioning, while ¹⁴C (t_1/2 = 5700 a) plays an important role in a geological repository for low- and intermediate-level radioactive waste. Our findings reveal differences in the proportions of trace elements between the base and weld materials as well as between the two reactor generations. Microstructural analysis identified Mo-rich precipitates and (Mn, S)-rich inclusions containing secondary micro-inclusions in the unit 1 and 2 samples. Raman spectroscopy confirmed iron oxides (γ-Fe₂O₃, Fe₃O₄), silicates (Mn-SiO₃), and Cr₂O₃/NiCr₂O₄ in the base metal as well as MnFe₂O₃ in the weld metal. X-ray photoelectron spectroscopy identified Mn inclusions as MnS, MnS₂, or mixed Mn, Fe sulfides, and the Mo precipitates as MoSi₂. These findings offer valuable insights into the speciation of elements and the potential release of radionuclides through corrosion processes under repository conditions. Full article

Low carbon dioxide (CO₂) emissions make nuclear energy crucial in decarbonizing the economy. In this context, nuclear safety, and especially the operation of nuclear power plants, remains a critical issue. This article presents a new fractal cluster method of control that improves the quality of assessing fuel element cladding integrity, which is critical for nuclear and environmental safety. The proposed non-destructive testing method allows for detecting defects on the inner and outer cladding surfaces without removing the elements from the nuclear reactor, which ensures prompt response and prevention of radiation leakage. Studies have shown that the fractal dimension of the cladding surface, which varies from 2.1 to 2.5, indicates significant heterogeneity caused by mechanical damage or corrosion, which can affect its integrity. The density analysis of defect clusters allows quantifying their concentration per unit area, which is an important indicator for assessing the risks associated with the operation of nuclear facilities. The data obtained are used to assess the impact of defects on the vessel’s integrity and, in turn, on nuclear safety. The monitoring results are transmitted in real time to the operator’s automated workstation, allowing for timely decision making to prevent radioactive releases and improve environmental safety. The proposed method is a promising tool for ensuring reliable quality control of the fuel element cladding condition and improving nuclear and environmental safety. While the study is based on VVER-1000 reactor data, the flexibility of the proposed methodology suggests its potential applicability to other reactor types, opening avenues for broader implementation in diverse nuclear systems. Full article

As the cornerstone of severe accident management strategy, either in-vessel or ex-vessel retention of core melt (IVR or EVR) plays a pivotal role in the stabilization and termination of a severe accident and ultimately secures the safety goal of “Practical elimination of large radioactive release” for light water reactors. In contrast to the IVR measures that are more or less identical in reactor designs, the EVR measures are quite different from design to design. This study intended to give a critical review on the EVR measures adopted in the reactor designs of VVER-1000, EPR, ESBWR, EU-APR1400 and APWR. The review study began with a general description of the existing EVR measures, including their principles, operational procedures and research efforts. We then focused our discussions on the pros and cons of each EVR measure through the comparisons with the IVR and with the others in terms of simplicity, reliability and economy. We finally tried to identify the remaining issues and uncertainties in the qualification of the EVR measures, based on which potential design improvements and future research needs were recommended. Full article

Assessing the embrittlement and hardening of reactor pressure vessel steels is critical for the extension of the service lifetime of nuclear power plants. This paper summarises the tensile test results on the irradiation behaviour of realistic VVER-1000 welds from the STRUMAT-LTO project. The welds were irradiated at the HFR (Petten, the Netherlands) to a fluence of up to 1.087 × 10²⁰ n·cm⁻², and their irradiation hardening was studied by means of tensile testing. The four grades, with different Mn and Ni contents, show different hardening behaviours. The highest degree of irradiation hardening is observed for the weld that has the highest combined Ni + Mn content. The results show that there is a synergetic effect of Mn and Ni on the irradiation hardening behaviour of the VVER-1000 welds. Besides irradiation hardening, the effectiveness of post-irradiation annealing treatments on the recovery of the tensile properties is studied in the present work. Post-irradiation annealing treatments conducted at 418 °C and at 475 °C proved to be effective for three of the four investigated welds. For the realistic weld with the highest combined Ni + Mn, only the annealing at 475 °C led to the complete recovery of the tensile properties. Full article

This paper presents the results obtained from numerical evaluations for the possibility of large-scale ²³⁸Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of ²³⁸Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor. Full article

The two-phase emission detector RED-100 with 130 kg of liquid xenon as a working medium has been exhibited at a distance of 19 m from the core of the VVER-1000/320 nuclear power reactor at the fourth power unit of the Kalinin Nuclear Plant Power in 2021–2022. Due to the high sensitivity of the detector for weak ionization signals (down to single electrons), the detector has been used to search for the elastic coherent scattering of reactor electron antineutrinos off xenon nuclei. However, the observation of ~30 kHz single-electron noise did not quite allow for an effective selection of the useful events. The next experiment with the RED-100 detector is considered to be arranged with 62 kg of liquid argon as a working medium. The advantages of this approach are discussed in this paper. Full article

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 10²² n/m² at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (>1.5 wt.%) and Mn (>0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ₂-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C). Full article

Specimens of 15Kh2MFAA steel used for reactor pressure vessels V-213 (VVER-440 reactor) were studied by positron annihilation techniques in terms of their radiation resistance and structural recovery after thermal treatment. The radiation load was simulated by experimental implantation of 500 keV H⁺ ions. The maximum radiation damage of 1 DPA was obtained across a region of 3 µm. Radiation-induced defects were investigated by coincidence Doppler broadening spectroscopy and positron lifetime spectroscopy using a conventional positron source as well as a slow positron beam. All techniques registered an accumulation of small open-volume defects (mostly mono- and di-vacancies) due to the irradiation, with an increase of the defect volume ΔV_D ≈ 2.88 × 10⁻⁸ cm⁻³. Finally, the irradiated specimens were gradually annealed at temperatures from 200 to 550 °C and analyzed in detail. The best defect recovery was found at a temperature between 450 and 475 °C, but the final defect concentration of about ΔC_D = 0.34 ppm was still higher than in the as-received specimens. Full article

This paper considers the possibility for large-scale production of plutonium isotope ²³⁸Pu in the light-water nuclear power reactor VVER-1000. ²³⁸Pu is a unique source of long-term autonomous energy supply in various devices for remote regions of the Earth and in outer space. The design of the irradiation device with ²³⁷NpO₂ as a starting material is proposed, which is placed in the central zone of the VVER-1000 reactor core and makes it possible to achieve 8% of the specific Pu production (Pu/²³⁷Np) by optimizing the pitch of NpO₂-rod lattice. The computations showed that the scale of ²³⁸Pu production in the irradiation device was remarkably larger (2 ÷ 7 times more) than similar values in research reactors. At the same time, the use of heavy neutron moderators with low neutron absorption (natural lead or lead isotope ²⁰⁸Pb) around the NpO₂ fuel assembly (FA) made it possible to obtain high-purity ²³⁸Pu with the content of ²³⁶Pu below 2 ppm. The paper also shows that if the irradiation device is placed in central zone of the VVER-1000 reactor core, then the displacement damage dose in the reactor vessel remains low enough to conserve its strength properties throughout the entire period of the reactor operation (60 years). Full article

Following the recent surge in harnessing clean energy sources to fast-track carbon neutrality, renewable and nuclear energies have been the best-rated sources of clean energy. Even though renewable energy presents an almost insignificant risk to public health and the environment, they are insufficient to support the growing demand for the high energy required for industrialization. Despite the competitive potential of nuclear energy to meet these demands, public concerns about its safety have significantly hindered its mass deployment in developing countries. Therefore, one of the primary considerations in commissioning a nuclear power plant is the establishment of emergency planning zones based on the reactor type and other siting criteria. Based on Ghana’s reactor type assessment (RTA), four reactor designs were considered in this study which are APR1400, HPR1000, VVER1200, and Nuscale Power Module. Using the NRC’s SNAP/RADTRAD and RASCAL codes, this research sought to investigate radionuclide doses released at the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), Control room (CR), and the 16 km recommended public safe zone during Fuel handling Accidents (FHA), Rod Ejection Accident (REA), and Long-Term Station Blackout (LTSBO). The results revealed that reactors’ power contributed to the source term activities and offsite consequences during REA and LTSBO, while FHA was predominantly affected by the number of fuel assemblies and a fraction of damaged fuel. Additionally, the accidents considered in this study followed a similar trend of impact in decreasing order of reactor power and the number of fuel assemblies; APR1400 < VVER1200 < HPR1000 < Nuscale. Nevertheless, all the doses were within regulatory limits. Full article

Targeting ultimate fidelity reactor physics calculations the Dynamic Monte Carlo (DMC) method simulates reactor transients without resorting to static or quasistatic approximations. Due to the capability to harness the computing power of Graphics Processing Units, the GUARDYAN (GpU Assisted Reactor DYnamic ANalysis) code has been recently upscaled to perform pin-by-pin simulations of power plant scale systems as demonstrated in this paper. A recent rod drop experiment at a VVER-440/213 (vodo-vodyanoi enyergeticheskiy reaktor) type power plant at Paks NPP, Hungary, was considered and signals of ex-core detectors placed at three different positions were simulated successfully by GUARDYAN taking realistic fuel loading, including burn-up data into account. Results were also compared to the time-dependent Paks NPP in-house nodal diffusion code VERETINA (VERONA: VVER Online Analysis and RETINA: Reactor Thermo-hydraulics Interactive). Analysis is given of the temporal and spatial variance distribution of GUARDYAN fuel pin node-wise power estimates. We can conclude that full core, pin-wise DMC power plant simulations using realistic isotope concentrations are feasible in reasonable computing times down to 1–2% error of ex-core detector signals using current GPU (Graphics Processing Unit) High Performance Computing architectures, thereby demonstrating a technological breakthrough. Full article

This work outlines an approach for localizing anomalies in nuclear reactor cores during their steady state operation, employing deep, one-dimensional, convolutional neural networks. Anomalies are characterized by the application of perturbation diagnostic techniques, based on the analysis of the so-called “neutron-noise” signals: that is, fluctuations of the neutron flux around the mean value observed in a steady-state power level. The proposed methodology is comprised of three steps: initially, certain reactor core perturbations scenarios are simulated in software, creating the respective perturbation datasets, which are specific to a given reactor geometry; then, the said datasets are used to train deep learning models that learn to identify and locate the given perturbations within the nuclear reactor core; lastly, the models are tested on actual plant measurements. The overall methodology is validated on hexagonal, pre-Konvoi, pressurized water, and VVER-1000 type nuclear reactors. The simulated data are generated by the FEMFFUSION code, which is extended in order to deal with the hexagonal geometry in the time and frequency domains. The examined perturbations are absorbers of variable strength, and the trained models are tested on actual plant data acquired by the in-core detectors of the Temelín VVER-1000 Power Plant in the Czech Republic. The whole approach is realized in the framework of Euratom’s CORTEX project. Full article

The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA. Full article