Reactor Pressure Vessel Research Articles

Validation of EPG/SAG Rev. 4 of the Kuosheng Nuclear Power Plant Using the MAAP5 Code

The Boiling Water Reactor Owner’s Group released Emergency Procedure Guidelines and Severe Accident Guidelines Revision 4 (EPG/SAG Rev. 4) in 2018. The major improvement to EPG/SAG Rev. 4 was Contingency 1 (Alternate Level/ Pressure Control). Contingency 1 coordinates the reactor pressure vessel (RPV) water level and RPV pressure control action to prolong the availability of steam-driven injections and optimize the transfer to motor-driven systems. In this study, the effectiveness of the EPG/SAG Rev. 4 Contingency 1 strategy was compared with those of EPG/SAG Revision 2 Contingency 1 and Specific Major Incident Guidelines (SMI) using the Modular Accident Analysis Program, Version 5 (MAAP5). SMI was developed by the Taiwan Power Company to mitigate a Fukushima-like accident. The surrogate plant that analyzed is the Kuosheng Nuclear Power Plant (NPP). The Kuosheng NPP is BWR-6 Mark-III containment. MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic risk assessments and for verification and validation of mitigation actions specified in severe accident management guidelines. The simulation scenarios were extended loss of alternative-current power and loss of ultimate heat sink. The low-capacity, motor-driven portable pump was the only available system for RPV injection in the first hour of the accident. In this time period, the RPV water level and pressure were controlled by reactor core isolation cooling and safety relief valves. After this study, the strategy of EPG/SAG Rev. 4 Contingency 1 was successfully validated, and the effectiveness of minimum pre-depressurization RPV water level and the low-capacity, motor-driven portable pump were also demonstrated in this study.

Numerical simulation study of boiling Critical Heat Flux characteristics of graphene nanofluids

IVR-ERVC (In-Vessel Retention – External Reactor Vessel Cooling) is a critical accident management method for ensuring the integrity of the reactor pressure vessel (RPV) lower head. One of the most crucial aspects within this method is to enhance the CHF (Critical Heat Flux) on the outer surface of the reactor pressure vessel (RPV) lower head. This paper explores the application of graphene nanofluids in IVR-ERVC. This paper uses computational fluid dynamics to numerically simulate the CHF characteristics of graphene nanofluids under different undercooling, mass flow rate, concentration, tilt angle, and gap size conditions and analyzes the impact of different factors on CHF and the coupling effects between different factors. The undercooling range is 5 K–100 K, the mass flow rate range is 150 kg/(m2∙s) to 3000 kg/(m2∙s), the concentration range is 0–1%, the inclination angle range is 0°–90°, and the gap sizes are 10 mm and 20 mm. The results show that the CHF can be effectively improved by adding nano-graphene into the base solution, and the CHF increases with the increase of subcooling degree, mass flow rate, concentration, dip Angle and gap size. Within the simulation range, the strengthening effect on CHF weakens when raising the undercooling and mass flow rate. And the larger the concentration and inclination angle within the simulation range, the better the enhancement effect on CHF. The maximum increase was 290%, while the average increase was 75.6%.This article studies the coupling effects between different parameters through numerical simulation. It can be concluded that increasing the mass flow rate weakens the enhancement of subcooling and concentration on CHF, increasing the subcooling weakens the enhancement effect of mass flow rate and concentration on CHF, and increasing the concentration enhances the enhancement effect of subcooling and mass flow rate on CHF. Enhance the mass flow rate will weaken tilt angle effect, while increasing the concentration will enhance the strengthening effect of tilt angle on CHF.

Pearson Correlation and Chi-Squared Methods Applied to the Sensitivity and Uncertainty Analysis of Hydrogen Generation During a Boiling Water Reactor STSBO Using MAAP5 and AZTUSIA

Abstract After the nuclear accident at the Fukushima Daiichi nuclear power plant, hydrogen generation has resurfaced as a key issue, with respect to its quantification in the simulation of severe accidents (SAs), because of the potential risk of deflagration or detonation, which would compromise the integrity of the plant facilities. Severe accident simulation codes may have different models to predict cladding oxidation rates and the consequent hydrogen generation. Thus, sensitivity and uncertainty analysis can be applied to determine which input variables of the code have a significant impact in the implemented models, and consequently on the overall predicted values of key figures of merit. In this work, the generation of hydrogen in the reactor core during a short-term station blackout (STSBO) is studied as a figure of merit for a boiling water reactor (BWR). The scope of the analysis is until the failure of the reactor pressure vessel (RPV). This study shows a sensitivity and uncertainty analysis through the complementary calculations of two methods: Chi-squared parameter and Pearson simple correlation coefficient (PSCC). The severe accident simulation code was MAAP 5.0.3 (Modular Accident Analysis Program, EPRI, USA) and the AZTUSIA code (AZtlan Tool for Uncertainty and SensItivity Analysis, Mexico) was used as the statistical tool for the sensitivity and uncertainty analysis. Ten uncertain parameters were chosen, of which two are MAAP models options, and the values were generated via the simple random Monte Carlo technique. The results show that the Chi-squared parameter in conjunction with correlation coefficients provides a more comprehensive understanding of the impact of both quantitative and qualitative input variables on the figures of merit.

A study on modeling of boiling heat transfer in core debris bed of SFR

In case of a hypothetical severe accident in a Sodium-cooled Fast Reactor (SFR), coolability of the debris bed in the post-accident phase plays a vital role in mitigating the accident and ensuring the structural integrity of the reactor vessel. Few numerical studies are reported in literature, in which the boiling heat transfer in debris bed is expressed as equivalent heat conduction using similarity law between heat conduction and two-phase heat transfer. However, these studies assumed steady state mass conservation for the boiling zone and neglected the gravity force. Hence, a detailed study has been carried out for various particle sizes and porosities of SFR debris to investigate the influence of above considerations. The effect of gravity on debris bed coolability is studied using steady state model of Lipinski, which showed that gravity has a non-negligible effect, for particle size of 0.3 mm and porosity of 0.5. However, the gravitation force was found to have a negligible effect in dryout heat flux estimation for the bottom cooled configuration. A transient numerical model is developed for simulating the boiling phenomena in debris beds and validated with the published experimental results. The assumption of steady state mass conservation is verified by carrying out transient analysis, which indicated early prediction of the dryout inception. For time dependent heat generation case, the unsteady mass conservation predicted higher DHF compared to constant heat generation.

Research on the Design of Pressure Vessel Fixtures for Nuclear Reactors

Nuclear energy, as a clean, pollution-free, almost zero emission of clean energy, is in line with the world's energy development direction. With the development and maturity of science and technology, the development of the nuclear industry has promoted all walks of life more and more obviously. Nuclear reactor pressure vessel is one of the key components, and its design and manufacture is directly related to the operation of nuclear power equipment. The production and manufacturing of nuclear reactor pressure vessel cannot be separated from convenient and reliable working fixture. This paper carries out the tooling design of the fixture, which effectively solves the support and positioning problems to be solved in the overall processing and lifting process, and makes contributions to the development of nuclear power industry.

Experimental investigation of countercurrent flow limitation (CCFL) in a 1/19 scale of VVER-1000 geometry

Countercurrent flow limitation (CCFL) has been experimentally considered in a VVER-1000 hot-leg. The proposed test facility consists of a hot-leg pipe, a reactor vessel simulator (RVs), and a steam generator simulator (SGs). It was designed and constructed based on power-to-volume scaling approach by preserving the Froud number in the hot leg. The scaling ratio was considered 1/19 of the Bushehr NPP (VVER-1000) geometry. The hot leg of the proposed facility provides good optical observation of the gas/liquid interface. The SG simulator has also a side view for optical observation. Experimental results were captured at atmospheric pressure with air/water as fluid. CCFL was obtained by analyzing the pressure differences between RV and SG simulators.This CCFL data was compared with empirical data and correlations available in the literature. With increasing liquid superficial velocity, the gas superficial velocity curve is continuously decreased, which is consistent with Navarro's results. In this study, an experimental correlation is proposed and compared to other correlations. It was very close to the Navarro and Richter correlations.The results of experimental investigations clarified the onset of the CCFL phenomenon in a Bushehr NPP (VVER-1000) hot-leg geometry. Moreover, the achieved patterns and behaviors of the air/water mixture would be used in computational fluid dynamics (CFD) code validations.

Calculation and Discussion for Crack Driving Force of Dissimilar Metal Welded Joint of Nuclear Power Equipment Nozzle and Safe End

This study investigates the effects of the material mechanical properties heterogeneity and strength mismatch variation on crack driving forces for cracks in a dissimilar metal welded joint (DMW) of a reactor pressure vessel inlet nozzle to the safe end. Linear elastic and elastic-plastic analyses are carried out to calculate the stress intensity factor (SIF) and J-integral for cracks in the DMW. The effects of crack locations, crack depths, and strength mismatch factors on the SIF and J-integral are studied. The SIF results obtained by finite element analysis are compared with those obtained by the influence function method adopted in the nuclear power code. Results show that the effect of crack location on the SIF can be ignored. The SIFs calculated by the influence function method for cracks in the DMW are basically reasonable, although its conservatism is slightly insufficient. Moreover, with moving the crack location from the nozzle through buttering and weld metal to the safe end, the J-integral increases. The effect of the crack location and strength mismatch factors on the J-integral is essentially caused by variation of the plastic zone and the material properties of the crack front. The crack sizes affect the level of influence of the crack location and the strength mismatch factors on the J-integral. The J-integral increases with decrease of the strength mismatch factor.

Vibration response analysis of mobile liquid Pb-Bi micro-reactors under transportation with liquid sloshing

Mobile micro-reactors, which used the liquid lead–bismuth eutectic (LBE) as coolants, providing energy to remote regions had become a popular topic in nuclear energy due to their desirable characteristics, including inherent safety and modularity. However, the liquid sloshing effect in a reactor may exhibit fluid–structure interaction (FSI) coupling effects, which could have a significant impact on the relevant internal systems. And the ratio of coolant in liquid to solid form was correlated with the duration of heat dissipation. This paper aimed to investigate the liquid sloshing effect in a micro-reactor that was subject to vehicle vibrations, the objective was to provide recommendations for transportation scenarios involving various liquid sloshing. To achieve this, a full-vehicle dynamic model of the system was developed using multi-body dynamics. Additionally, MATLAB simulation software was utilized to simulate the road roughness loadings that the micro-reactor may encounter during transportation. Moreover, a finite element numerical model of the Mobile Micro-reactor had been developed in order to simulate the sloshing effect that occurred during transportation, the numerical models presented in this study were validated by comparing them with a previously-verified Housner model that involved a cylindrical tank. This study investigated the sloshing effects of the Mobile Micro-reactor under different liquid–solid ratios of coolant, internal component lengths, and liquid surface heights. The results indicated that the internal components within the reactor vessel can alleviate the sloshing effect. The solid ratios of liquid lead–bismuth coolant and the liquid surface height during transportation had an impact on the sloshing effect. These findings provided valuable insights for studying the sloshing response of on-board reactors during liquid sloshing transportation.

CFD Analysis of Coolant Flow Characteristics in Reactor Pressure Vessel: A Comprehensive Review

This comprehensive review explores the application of Computational Fluid Dynamics (CFD) in analyzing coolant flow within Reactor Pressure Vessels (RPVs) of nuclear power plants. By synthesizing existing research, methodologies, and advancements specific to RPVs, the paper offers in-depth insights into critical aspects such as boundary conditions, turbulence modeling, heat transfer mechanisms, and validation techniques. Examining a range of studies encompassing various reactor types from Pressurized Water Reactors (PWRs) to Integral Pressurized Water Reactors (IPWRs), the review underscores CFD's pivotal role in enhancing safety, efficiency, and performance optimization in nuclear reactors. Through systematic exploration, this study underscores the critical importance of precise modeling in facilitating safety assessments, operational optimization, and design enhancements across various reactor systems. Accurate modeling serves as a cornerstone for informed decision-making processes aimed at maximizing reactor performance while ensuring the highest standards of safety and reliability. The paper navigates through challenges such as computational limitations and turbulence modeling intricacies, while also discussing emerging trends like the porous media method aimed at improving computational efficiency. By offering a comprehensive understanding of thermal-hydraulic behavior in nuclear reactors, the review underscores CFD's contribution to enhancing safety and reliability in nuclear power generation. Overall, this review underscores the indispensable role of CFD in advancing our understanding of nuclear reactor dynamics, thereby contributing significantly to the overarching goals of improved safety and reliability in nuclear power generation.

A Study on Boiling Heat Transfer Performance of TiO2-CNTs Hybrid Nanofluid on a Downward-Facing Heating Surface

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a strategy used to respond to nuclear reactor accidents. One of the key performance indicators determining its feasibility is critical heat flux (CHF). Our focus is on simulating real-world scenarios through surface pool boiling to improve the implementation of the IVR-ERVC strategy with hybrid nanofluids. Two groups of TiO2/COOH-CNTs hybrid nanofluids were prepared: group 1 with different concentrations at the same proportion and group 2 with different proportions at the same total concentration. Researchers compared the improvement of the two groups’ CHF and heat transfer coefficient (HTC), and analyzed the potential mechanism of heat transfer enhancement through roughness of surface, hydrophilicity, and scanning electron microscopy observations. The results showed that a mass concentration of 8 mg:8 mg per liter exhibited the best heat transfer performance, with a CHF enhancement up to 28.21% and an improvement in HTC as well. Meanwhile, correlations between alterations in surface roughness, hydrophilicity, and enhancements in CHF were observed. Finally, by detecting the deposition surface, the possible mechanism of TiO2/COOH-CNTs hybrid nanofluids in enhancing heat transfer was inferred.

Gamma noise to non-invasively monitor nuclear research reactors

Autonomous nuclear reactor monitoring is a key aspect of the International Atomic Energy Agency’s strategy to ensure nonproliferation treaty compliance. From the rise of small modular reactor technology, decentralized nuclear reactor fleets may strain the capacities of such monitoring and requires new approaches. We demonstrate the superior capabilities of a gamma detection system to monitor the criticality of a zero power nuclear reactor from beyond typical vessel boundaries, offering a powerful alternative to neutron-based systems by providing direct information on fission chain propagation. Using the case example of the research reactor CROCUS, we demonstrate how two bismuth germanate scintillators placed outside the reactor vessel can precisely observe reactor criticality using so called noise methods and provide core status information in seconds. Our results indicate a wide range of applications due to the newly gained geometric flexibility that could find use in fields beyond nuclear safety.

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO2) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. The reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

In nuclear reactors, the structural materials of reactor pressure vessel (RPV) are damaged by the radiation that produced by fission reactions. Neutron irradiation damage is one of the most critical factors for the degradation of the reactor pressure vessel (RPV). A displacement per atom (dpa) is a standard measurement for calculating neutron-and gamma-induced radiation damage in materials. The aim of this paper was to evaluate the neutrons radiation damage in the reactor pressure vessel of VVER-1200. In this paper, neutron damage factors, including fast neutron fluxes above 1 MeV, above 0.5 MeV, above 0.1 MeV, the damage rate (dpa/s), and Integrated damage [DPA for one effective full power year (EFPY)] were investigated for VVER-1200/AES-2006 reactor pressure vessel (RPV). The calculations are performed by using NJOY code together with the MCNP6 code at beginning of Cycle (BOC). The cross section library used in present calculations is based on ENDF/B-VII.1 library (ENDF71x). The NJOY code was used for generating DPA cross sections in an ACE formatted file that suitable for using in the MCNP6 code. The MCNP6 code was used for simulating the VVER-1200 reactor core and to calculate neutron DPA rates. MCNP6 simulation of the VVER-1200 reactor core identified the area where the RPV neutron radiation is maximized. The integrated damage for a full effective year was evaluated to be 2.128E-04 DPA/EFPD for the VVER-1200 RPV.

Effect of helium ion irradiation on the microstructure, mechanical properties and surface morphology of Inconel 625 alloy

Inconel 625 alloy is mainly used for reactor pressure vessel, fuel cladding, nuclear fuel elements and cryogenic channels for cooling high temperature superconductors in Tokamaks. This study focuses on deciphering the impact of helium ion irradiation on the properties of Inconel 625 alloy. Helium-ions interaction with material leads to blistering, void formation, swelling, and degradation of mechanical properties. To understand these effects, the samples of Inconel 625 alloy were irradiated by helium-ion fluence beams at three different helium-ion beam fluence (1 × 1013, 1 × 1014, 1 × 1015 He/cm2). A decrease in grain size was observed from 34.4 to 19.0 nm with an increase in the ion fluence. According to the results, the residual strain is increased, from 1.8 to 43.0 micro-strain with the increase in helium-ion fluence, however structure is relaxed with decrease in micro-strain at higher helium-ion fluence (1 × 1015 He/cm2) and also there is a prominent increase in grain size. The surface morphology as observed by scanning electron microscope (SEM) illustrates the fuzz formation at low fluence (1 × 1013 He/cm2), while samples irradiated with high helium ion fluence (1 × 1015 He/cm2) showed grain growth, voids and surface blistering (deep valleys). An increase in hardness value at low helium-ion beam fluence is observed due to the presence of large population of defects. However, at higher helium-ion beam fluence, a decrease in hardness value is observed. This decrease might be attributed to the dislocation clustering and grain growth.

Influence of prior austenite grain size on martensite-austenite islands morphology and decomposition characteristics during two-step tempering in Mn-Ni-Mo steel

Martensite-austenite (M/A) islands are typical in the quenched microstructure of Mn-Ni-Mo steel, frequently utilized for fabricating reactor pressure vessels (RPV). These islands can decompose into harmful long rod-shaped Fe3C carbides during tempering above 600°C, quantifying the negative effect of M/A island decomposition on mechanical properties and indicating a potential issue in RPV fabrication. To address this issue, the present study investigates a two-step tempering process on three different prior austenite grain size (PAGS) specimens (5 µm, 47 µm, and 119 µm). The effect of PAGS on M/A island decomposition characteristics during sub-600°C tempering and their impact on precipitated carbide morphology is investigated. A preliminary pre-tempering treatment at 200°C, 350°C, and 500°C for 2 hours was followed by a 3-hour final tempering step at 650°C. The findings revealed unique temperature-dependent decomposition characteristics of M/A islands. For example, at 200°C, partial M/A island decomposition took place, resulting in long rod-shaped carbides during final tempering. Meanwhile, the 350°C pre-tempering specimen revealed aggregated carbide precipitates of varying sizes in former M/A island locations. In contrast, the 500°C pre-tempering specimen showed coarse globular carbides and the absence of carbide precipitate clustering. This result can be attributed to the decrease in surface energy associated with the carbide-matrix interface as size increases. After tempering, all specimens produced three types of precipitates: rod-shaped Fe3C carbides, rhombus-shaped Mn7C3, and globular-shaped Mn23C6 carbides. The formation of long rod-shaped carbides during M/A island decomposition reduces the critical cleavage stress for micro-crack initiation, jeopardizing steel mechanical properties. Aggregated carbides cause the simultaneous formation of numerous micro-voids in close proximity, which eventually merge to form a large crack, resulting in material failure. When compared to other pre-tempered and directly tempered specimens, pre-tempering at 500°C produced the most favorable microstructure in terms of impact properties.

Physical simulation of the underclad heat affected zone in a reactor pressure vessel to study intergranular cracking

Underclad cracking in nuclear pressure vessels was of significant concern in the 1970s and 1980s before mitigating adjustments were made both to steel compositions and manufacturing practice. Unfortunately, the cracking mechanisms are still not well understood, and this can undermine confidence when changes to cladding operations are under consideration. In this work, Charpy-sized test coupons of 18MND5 steel were subject to a range of thermal cycles that can be experienced by the substrate immediately adjacent to the interface with the overlay. The Charpy test results are considered in combination with analysis of the samples through field emission gun (FEG) scanning electron microscopy (SEM) and wavelength dispersive X-ray spectroscopy (WDXS). The findings suggest that the heat affected zone subject to the coarse grained plus intercritical thermal cycle, before post weld heat treatment, has a peculiar and potentially brittle microstructure. However, in a steel with no obvious macrosegregation, no clear evidence of intergranular cracking in the early stages of the post weld heat treatment could be found. The second part of the work focuses on the effects of segregation regions in large forgings, more specifically ghost lines, on the tendency for underclad cracking. Ghost line regions that subsequently form a coarse grained heat affected zone (CGHAZ), and are then heated to just under the A1 temperature, seem to be the most prone to intergranular cracking. Cracking susceptibility appears to be associated with elevated concentrations of alloying and impurity elements at grain boundaries.

Assessing the Feasibility of Diver Access During Dismantling of Reactor Vessel Internals

Assessing the Feasibility of Diver Access During Dismantling of Reactor Vessel Internals

A comprehensive analysis of thermal–hydraulic signatures in neutron noise of WWER-type reactors

The neutron noise phenomenon, occurring in all types of nuclear reactors, provides valuable insights into the dynamic behavior of these reactors. In this work, a well-justified approach for modeling neutron noise induced by thermal–hydraulic sources in a hexagonal reactor core is presented in detail. These sources encompass fluctuations in key features of the inlet coolant, including flow rate, temperature, and boric acid concentration. Our proposed approach touches on various aspects of the distribution of thermal–hydraulic parameters among coolant loops, accounting for both uniform and non-uniform distributions. The latter allows for the asymmetric and asynchronous distribution of their fluctuating component among the coolant loops. To manage such complexity, it becomes imperative to determine the contribution of each fuel assembly based on each coolant loop. To achieve this, CFD simulations are conducted for the downcomer and lower plenum of the reactor pressure vessel. These simulations provide essential data to establish the coolant mixing matrix, which accounts for non-uniform distributions. The obtained mixing matrix is subsequently integrated into our neutron noise simulator, DYNOSIM, ensuring the faithful reproduction of neutron noise. Our study comprehensively analyzes both uniform and non-uniform scenarios. The radial and axial distributions of neutron noise amplitude and phase are examined for these scenarios. Their qualitative verifications are conducted by prior works in this field. Furthermore, we compare one of the uniform scenarios with reference results obtained using a frequency domain simulator. Notably, our CFD simulation results closely match the experimental data. Our findings demonstrate the effectiveness and efficiency of this methodology in modeling and understanding neutron noise in hexagonal reactor cores.

Enhancement of strength and ductile-brittle transition temperature of SA508 Gr.3 low-alloy steel by controlling heat accumulation in laser powder-directed energy deposition

The safety and longevity of small modular reactors are affected by reactor pressure vessels, which are complex integral components made of SA508 Gr.3 low-alloy steel. In this study, the impacts of heat accumulation on the microstructural and mechanical characteristics (tensile properties and Charpy V-notched impact energy) of SA508 Gr.3 steel fabricated using laser powder-directed energy deposition were investigated. Rectangular samples were prepared using long raster and short raster scanning strategies for changing heat buildup, and the mechanical tests were conducted depending on build direction. The time-temperature profile measured at a fixed point in the long raster sample showed a lower maximum temperature and a higher cooling rate, indicating lower heat accumulation compared to that of the short raster sample. In each build direction, the yield strength of the long raster sample was 45.8 %–60.5 % higher and its ductile-brittle transition temperature was 76.8–103.8 °C lower than that of the short raster sample. Additionally, compared to conventionally made samples and without requiring heat treatment, the long raster sample exhibited over a 45 % increase in yield strength and a 22.7 °C reduction in the ductile-brittle transition temperature. The superior combination in long raster samples is induced by smaller effective grain size, smaller cementite, and a higher pre-existing dislocation density. The results emphasize the importance of controlling heat accumulation throughout the additive manufacturing process and provide valuable insights into the use of additive manufacturing for manufacturing reactor pressure vessels in the nuclear industry.

The method of research nuclear installation conversion at the decommissioning stage

The article examines the problem of forming scientific and technical programs for the conversion of research nuclear installations at the decommissioning stage, which arose in the aspect of justifying the possibility of extending the designated service life of all existing research nuclear reactors in Russia. To solve this problem, a methodology is proposed, which includes an information model and a process description of the necessary step-by-step actions, as well as a set of appendices in the form of documents justifying the most optimal ways to form a project for managing the quality of nuclear reactor conversion processes in relation to special life cycle conditions of specific nuclear reactors. The general view of the methodology is described: principles and methods of construction, structure. An example of the application of this technique in the conversion of the Hydra salt solution pulsed research reactor for the decommissioning stage is given. The purpose of the conversion of the Hydra reactor is to extend the designated service life by replacing non-repairable equipment ­ the vessel. The main criteria for justifying the residual life of the housing are formulated, which are the presence of a safety margin of the housing material, taking into account the accumulated fluence on more vulnerable areas and the justification of the integrity and tightness of the housing. For clarity, the conceptual information model of the methodology for justifying the safety of the process of replacing the Hydra reactor vessel is presented in the form of an Ishikawa diagram. The methodology represents a series of sequential scientific and technical activities, research and a finite number of step-by-step actions to achieve the final goal – extending the service life. The stages of the methodology for justifying the safety of replacing the Hydra reactor vessel are described, such as «clarification of initial data», «performing computational studies and laboratory experiments», «clarification of the requirements of methodological documentation» and «formation of a work plan» and «registration of licensing documents for operation».