Reactor Vessel Internals Research Articles

Verification and Validation of CFD analysis for 1/7th scale model of APR1400 for CVAP

This paper presents a methodology for analyzing the internal flow of the APR1400 Reactor Vessel Internals (RVI) using Computational Fluid Dynamics (CFD). The Comprehensive Vibration Assessment Program (CVAP) is performed to verify the structural integrity of reactor internals for flow-induced vibrations prior to the commercial operation of nuclear power plants. To classify the APR1400 as a valid prototype, it is necessary to demonstrate through successful CVAP that there are no adverse effects. In this study, CFD is used to calculate the local velocity and flow distribution to generate the forcing functions caused by turbulent flow for analysis programs in the CVAP. The Verification &Validation (V&V) procedure recommended by the American Society of Mechanical Engineers (ASME) is performed to verify the accuracy of CFD based on reliability, and it is compared with tests on a scale model. This paper provides valuable insights into the analysis and V&V process for CVAP using CFD, which can help ensure the safe and reliable operation of nuclear power plants.

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

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

Impurity sensitivity analysis for activated structure materials of OPR-1000

Various kinds of radioactive wastes are expected to be generated resulting from the decommissioning of nuclear power plants. For the safe management of decommissioning wastes, the preliminary classification of radioactive wastes such as very low level, low level, intermediate level, and regulatory clearance has to be performed. In this regard, we intend to conduct an activation analysis and radioactive waste level estimation using a computational simulation code for the structure materials of an OPR-1000, a standard nuclear power plant in South Korea. In addition, the structure material that constitutes the reactor vessel, reactor vessel internals, and bio-shield concrete of the nuclear power plant contains small amounts of unintended impurities. The results of the activation analysis may vary depending on the presence of those impurities. In this study, a detail modeling of nuclear reactor and the activation analysis are carried out combined the MCNP code with the ORIGEN-S code. Based on this methodology, the sensitivity test is done for impurities of structure materials. From the result of the specific activity of activated structure material, the cobalt impurity is found to be the most important contributor in the level determination of radioactive wastes. However, the bio-shield concrete has the highest specific activity when Li impurity is contained. From this study, some impurities affect radioactive waste level, thus it is recommended to consider structure material impurities for decommissioning the nuclear power plants.

Refinement MRP-135 constitutive equation model for tensile properties of austenitic stainless steel

This study focuses on analyzing and enhancing MRP-135 constitutive models, which are crucial for assessing the tensile properties of irradiated stainless steels in the reactor vessel internals of pressurized water reactors. The newly developed model, named M1, incorporates parameters from MRP-135, while minimizing modifications to ensure high compatibility with existing models. A significant improvement in M1 is the resolution of the inversion phenomenon observed between the yield strength and ultimate tensile strength at low temperatures in Type 316 stainless steel, a challenge previously identified in MRP-135 Rev. 2. M1 integrates the temperature effects from the earlier version, MRP-135 Rev. 1, to address this issue effectively. Additionally, for Type 316, M1 refined the equations to represent the cold work factor more accurately, enabling precise interpolation of strength variations across different levels of cold work from solution-annealed to cold-worked conditions. A comparative analysis of the M1 and the MRP-135 models against actual irradiated tensile test data for Type 304 and 316 stainless steel demonstrated the superior accuracy of M1, as indicated by a lower root-mean-square deviation. This study highlights the importance of continuous improvements in constitutive models, with future work aimed at further refining these models based on comprehensive irradiation test data.

Westinghouse fluence methodology uncertainty estimates for reactor vessel internals fast neutron exposure evaluation in subsequent license renewal period

The United States Nuclear Regulatory Commission (USNRC) has expressed concerns about the calculated fast neutron fluence exposure uncertainties for the reactor vessel internals (RVI) of light water reactors for extended operation [1, 2]. The current guidance for calculating neutron fluence, Regulatory Guide (RG) 1.190 [3], was developed with a focus on the reactor pressure vessel (RPV) region surrounding the active core, which may not be adequate for the RVI above and below the active core during the license renewal periods, especially the subsequent license renewal (SLR) period. An updated version of ANSI/ANS 19-10 [4] may fill the regulatory guidance gap on neutron fluence determination methodology for RVI but still faces the challenges of lacking RVI fluence measurement benchmark data and analytical uncertainty analysis, which are two cornerstones for qualifying a neutron fluence determination methodology. The Pressurized Water Reactor Owners Group (PWROG) has responded to this regulatory concern and have performed analytical uncertainty analysis and measurement benchmark work to address these two technical challenges for the nuclear power industry.

Исследование радиационных характеристик корпуса реактора и ВКУ остановленного для подготовки к выводу из эксплуатации блока № 3 Нововоронежской АЭС с РУ ВВЭР-440

The paper presents computational and experimental studies focused on the radiation characteristics of the reactor vessel and reactor internals performed to estimate the volume and the characteristics of activated intermediate- and high-level waste expected to be generated from the decommissioning of unit No. 3 at the Novovoronezh NPP. Two software products, namely, MCU-6 and Serpent were used in the calculations by the Monte Carlo method. To verify the calculation results these were compared with data on metal samples of the reactor internals and reactor vessel obtained from their laboratory studies.

Irradiation-induced age-related degradation mechanisms effect on integrity of baffle former bolts

For long-term operation of nuclear power plants, integrity of structures, systems and components should be ensured rigorously. In particular, since the complicated reactor vessel internals (RVIs) have been exposed to high temperature and neutron irradiation environment, materials reliability programs on the RVIs to resolve anticipated issues were conducted by several research institutes. The objectives of this research are to examine evaluation procedures derived from the previous works and to apply them for RVIs in two operating reactors with baffle former assembly from different electricity powers. Finite element (FE) analyses were performed employing five user-subroutines developed by the authors for reflection of multiple age-related degradation mechanisms (ARDMs) as well as pre- and post-processing. Temperature distributions due to gamma heatings, fuel cycles, typical normal operating temperature and pressure histories were considered with relevantly varying material properties. As results, effects of irradiation were quantified in terms of stresses, strains and irradiation assisted crack sensitivity parameters. Further assessment was carried out through not only investigating contribution of each ARDM but also comparing the present FE analysis results with those obtained from a different reactor type in detail.

The role of nickel in forming a structure providing increased service properties of reactor structural materials

Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.

Feasibility of Safe Operation of WWER-440-Type Nuclear Power Plants for Up to 60–70 Years

Most WWER-440 reactors are operated over the planned operational lifetime of 30 years. Now, the owners/operators are preparing for the second phase of extending the operation. The paper presents an overview of the most important aspects of ageing of the main components of the WWER-440-type reactors based on the operational and ageing management experiences and primary research efforts supporting the operation. The paper aims to demonstrate that the expectations for the operability of these reactors for up to 60–70 years are realistic. The scope of the review is limited to the ageing of the reactor pressure vessel, vessel internals, and steam generator, which are the lifetime-limiting components besides the containment structure. Some of the corrective actions, improvements of in-service inspections, maintenance, and ageing management are also considered in the paper. It is shown that the selection of materials and unavoidable material degradation processes do not limit the operation of the WWER-440 reactors for up to 60–70 years under the established operational environment and ageing management practice.

Investigation on effect of neutron irradiation on welding residual stresses in core shroud of pressurized water reactor

This paper presents the results of investigating the change in welding residual stresses of the core shroud, which is one of subcomponents in reactor vessel internals, performing finite element analysis. First, the welding residual stresses of the core shroud were calculated by applying the heat conduction based lumped pass technique and finite element elastic-plastic stress analysis. Second, the temperature distribution of the core shroud during the normal operation was calculated by performing finite element temperature analysis considering gamma heating. Third, through the finite element viscoelastic-plastic stress analysis using the calculated temperature distribution and setting the calculated residual stresses as the initial stress state, the variation of the welding residual stresses was derived according to repeating the normal operation. In the viscoelastic-plastic stress analysis, the effects of neutron irradiation on mechanical properties during the cyclic normal operations were considered by using the previously developed user subroutines for the irradiation agings such as irradiation hardening/embrittlement, irradiation-induced creep, and void swelling. Finally, the effect of neutron irradiation on the welding residual stresses was analysed for each irradiation aging. As a result, it is found that as the normal operation is repeated, the welding residual stresses decrease and show insignificant magnitudes after the 10th refueling cycle. In addition, the irradiation-induced creep/void swelling has significant mitigation effect on the residual stresses whereas the irradiation hardening/embrittlement has no effect on those.

햅틱 디바이스를 이용한 Unity 기반 가상 로봇 매니퓰레이터 조작 시스템 구현

The digital twin has been attracting attention in various industrial fields, including decommissioning nuclear power plants. In this study, an environment including a multi-axis robot manipulator for dismantling nuclear power plants and reactor vessel internals (RVIs), one of the major nuclear power plant facilities, is implemented virtually using Unity, and a multi-axis manipulator with complex and non-linear characteristics is evaluated for the disturbance effects. Integral sliding mode control (ISMC) that enables robust control without the robot dynamic model information was applied to the virtual RVIs environment. It demonstrated strong performance against disturbance compared with the existing PID control even if the dynamic model informations were not used. Moreover, the repulsive force is transmitted to the haptic device using the haptic feedback algorithm when the robot manipulator collides with the surrounding environment.

Mechanics Informed Neutron Noise Monitoring to Perform Remote Condition Assessment for Reactor Vessel Internals

Abstract The use of neutron noise analysis in pressurized water reactors to detect and diagnose degradation represents the practice of pro-active structural health monitoring for reactor vessel internals. Recent enhancements to this remote condition monitoring and diagnostic computational framework quantify the sensitivity of the structural dynamics to different degradation scenarios. This methodology leverages benchmarked computational structural mechanics models and machine learning methods to enhance the interpretability of neutron noise measurement results. The novelty of the methodology lies not in the particular technologies and algorithms but our amalgamation into a holistic computational framework for structural health monitoring. Recent experience revealed the successful deployment of this methodology to pro-actively diagnose different degradation scenarios, thus enabling prognostic asset management for reactor structures.

The preparation of segmenting the TRR reactor vessel internals with higher radiation

Based on the approved Taiwan Research Reactor (TRR) Decommissioning Plan, TRR decommissioning task must be finished before March 2029, and the reactor vessel should be fully dismantled before the end of 2028. However, Taiwan has no experience on dismantling a nuclear reactor vessel and lacks the associated containers and equipments, especially those used for handling the internals with higher radiation. To reach the goal of dismantling TRR reactor vessel in time, Institute of Nuclear Energy Research (INER) has made the dismantling strategy and developed the associated containers and equipments. Concerning the personnel safety, these internals with higher radiation will be segmented under the water. This makes the equipments and container used for segmenting these internals and storing the segmented parts are hard to be accessed from the commercial market directly and need to be developed in advance. To cope with these challenges, INER has designed the underwater disc saw and band saw equipments to segment these internals and developed INER-LRW-C2 low level radioactive waste container to store and transport those segmented parts. Currently, these equipments used for segmenting these internals are under mockup tests, and INER-LRW-C2 low level radioactive waste container has been granted the usage permit.The content of this article briefly introduces how INER prepares to face these challenges and what it has done. This includes the TRR reactor vessel dismantling strategy and segmentation plan of these internals with higher radiation, the development of INER-LRW-C2 low level radioactive waste container, and the designs of these major associated segmentation equipments.

Turbulent flow effect on integrity of pressurized RVIs

As dynamic loading caused by turbulent flow may affect the integrity of reactor vessel internals (RVIs), characterizing relevant features and establishing an appropriate assessment procedure are vital. This study addresses influence of turbulent flow as a problem obstructing the safe operation of RVIs. Numerical models are developed and optimized by comparing preliminary analysis data with theoretical and measured results. Systematic computational fluid dynamics analyses are conducted considering nuclear fuel assemblies. Fluctuating pressures are calculated and converted into power spectral densities (PSDs). Structural assessment is conducted by using modal analysis data, PSDs, and frequency domain-based fatigue methods. Vulnerable conditions and locations are derived from the resulting stress and fatigue damage values. Environmental effects due to neutron irradiation and primary water are also briefly discussed as areas of future research.

Numerical investigation of irradiation induced degradation in a welded core shroud assembly

Most of reactor vessel internals (RVIs) have been constructed by austenitic stainless steels (ASSs) to support and protect core from high temperature coolant and radiation particles. However, if they are operated in the harsh environment for a long time, ASS materials may undergo significant changes in micro-structural features, macroscopic deformation and strengths due to age-related degradation mechanisms (ARDMs). In this study, user subroutines were developed based on material constitutive models of recently revised technical reports to reflect irradiation embrittlement, irradiation enhanced creep and void swelling behaviors. Benchmark analyses for a simple rod and detailed numerical analyses for a RVI sub-components assembly with welds were conducted via coupling the subroutines. With regard to the latter analyses, not only typical normal operating sequences but also three fuel cycles associated with specific core loading patterns were taken into account with a constant initial weld residual stress distribution. Adequacy of the subroutines was verified through analytical solutions generated from temperature and radiation damage dependent equations. Complex neutron dose, temperature and pressure profiles of the assembly were successfully determined from thermal and mechanical finite element analyses. Structural integrity assessment in terms of the ductility, embrittlement and irradiation assisted stress corrosion cracking susceptibility ratio after 40-years of operation led to acceptable in spite of the ARDMs. Contributions of factors and interaction effects were further quantified and discussed via design of experiment approach.

Earthquake response of a core shroud for APR1400

The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using the ANSYS Design Parametric Language. Using the El Centro earthquake, seismic analyses are performed for the simplified and detailed core shroud models. Modal characteristics are obtained and their results are used for a time history analysis. Response spectrum analyses are also performed to access the degree of seismic excitation. The results of these analyses are compared to investigate the response characteristics between the simplified and detailed core shroud models from the time history and response spectrum analyses.

Creating an element of the reactor vessel internals of the VVER by directed metal deposition (DMD) methods

As part of the work, the analytical justification of the choice of additive technology for the manufacture of complex large-size elements reactor vessel internals (RVI) of the VVER-TOI was performed. The VVER-TOI baffle was selected as a complex-profile element of the RVI for manufacturing using direct laser growth technology. Preliminary development of technology for producing 18Cr-10Ni-Ti steel powder was carried out. The technology of direct laser growth of the baffle for VVER-TOI has been developed. The mechanical properties of the resulting metal and its chemical composition were studied. The prospects of using the 18Cr-10Ni-Ti steel material for manufacturing complex large-size elements of VVER-TOI RVI using DED methods are determined.

Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant

The baffle jetting phenomenon, intensified by pressure difference across baffle plates, may cause excessive fuel rod vibrations and possible damage. The downward flow direction through the baffle-barrel region opposite to the upward flow through the core in pressurized water reactors affects increase of pressure gradient. Gaps between the baffle plates widen with reactor operating time and water jets become more pronounced. During the 2015 outage the NPP Krško has undergone the upflow conversion (UFC) modification in order to minimize baffle jetting and prevent fuel rod failure. The modification consisted in altering reactor vessel internals in such a way that the coolant downflow path in the baffle-barrel region was converted to an upflow path.The impact of the UFC modification on the nuclear power plant behaviour was analyzed in the paper. The double ended guillotine break of one direct vessel injection (DVI) line was selected as a representative scenario because the DVI line break is exhibiting the highest peak cladding temperature among the small break LOCA cases. The thermal-hydraulic analysis was carried out with the RELAP5/Mod3.3 code for core average channel conditions, while additional FRAPTRAN runs were performed to evaluate peak cladding temperatures for high power and high burnup fuel assemblies. Distribution of pressure and velocity fields in the reactor pressure vessel baffle-barrel region and change in bypass flow rates due to modification were determined by the ANSYS-Fluent calculation. The accident analysis has been performed for both the downflow configuration (pre-UFC) and the upflow configuration (post-UFC) to assess the influence of the modification. Additional cases with different reactor coolant pump trip times, higher core decay heat and reduced capacity of safety systems were also analyzed to evaluate the impact on reactor safety.

Microstructure degradation of austenitic stainless steels after 45 years of operation as VVER-440 reactor internals

Specimens of reactor vessel internals (RVI) made of 18Cr–10Ni–Ti steel from the decommissioned VVER-440 unit after 45 years of operation were investigated. The aim was to assess the structure degradation degree responsible for the properties changes of RVI components. Radiation-induced structural elements of RVI baffle, former, and flux thimble specimens with the damaging dose 7.9–43.0 dpa and irradiation temperature 280–315 °C were studied using TEM, SEM, and APT methods. TEM-studies showed the negligible swelling under these conditions with a slight increase of radiation swelling due to the formation of the low-density large void population in case of higher temperature irradiation. APT-studies revealed the formation of the high-density Ni–Si-rich clusters (presumably G- and γ′- phase precursors) and an increase of its volume fraction with the temperature increase. The Orowan equation was used to assess the hardening contribution of different radiation-induced structural elements. The observed yield strength increase is insignificant under the studied conditions.

Full-field polycrystal plasticity simulations of neutron-irradiated austenitic stainless steel: A comparison between FE and FFT-based approaches

We compare two full-field approaches – a crystal plasticity finite element method (CP-FEM) and crystal plasticity fast Fourier transform-based (CP-FFT) method – for a specific crystal plasticity law introduced for neutron-irradiated austenitic stainless steel SA304L currently used in nuclear reactor vessel internals. This particular law is employed to identify and quantify possible advantages and drawbacks of the two approaches when used in the large-scale simulations to predict the effect of irradiation damage (e.g., crack initiation) in stainless steel microstructures. A comparison is performed in a polycrystalline context for different periodic Voronoi microstructures deformed under tension. Special emphasis is put on studying the performance of the two approaches in terms of mesh convergence analysis using aggregate models with different spatial discretizations. In the CP-FEM approach, the performance of linear as well as quadratic tetrahedral meshes is investigated. A similar performance between the CP-FEM and CP-FFT methods is demonstrated in a smaller 2-grain aggregate. However, a slower mesh convergence is observed for the CP-FEM method when comparing tensile responses of a larger 100-grain polycrystal. A drop in the convergence rate is much more pronounced on linear than on quadratic tetrahedral meshes. As a consequence, largest (average) grain boundary stresses are shown to be overestimated with the linear mesh CP-FEM approach, thus raising a concern of possible over-conservatism employed in the CP-FEM prediction of crack initiation in irradiated stainless steels. On the contrary, the mesh convergence of the CP-FFT approach is found to be practically independent of the applied macroscopic strain and also aggregate size. Therefore, for such steels, the CP-FFT approach seems to be better justified.