Light Water Reactor Pressure Vessel Research Articles

The Effect of Operational Factors on Phase Formation Patterns in the Light-Water Reactor Pressure Vessel Steels

This paper presents the results of atom probe tomography studies on radiation-induced phase formation in light-water reactor pressure vessel steels after neutron irradiation under various conditions in comparison with the literature data. The given irradiation conditions are fluence (10–100) × 1022 m−2, flux (5–2700) × 1014 m−2s−1 and irradiation temperature (50–400) °C. The composition of the studied steels varies in a wide range for the elements significantly affecting radiation and thermal resistance of steels: Ni in the range of 0.2–6.0 wt.%, Mn–0.03–1.1 wt.%, Cu–0.01–0.16 wt.%, and P–0.01–0.03 wt.%. The number density, volume fraction, size, composition, and nucleation sites of precipitates are determined. The regularities of the effect of various operational factors on the phase formation in these steels have been analyzed and revealed. The study shows that in materials with high copper content, Cu-rich precipitates are formed by a radiation-enhanced mechanism. In materials with low copper content, their formation upon irradiation at 300 °C occurs by a radiation-induced mechanism since the main nucleation sites are point defect clusters formed in cascades. At the same time, the density, volume fraction, and composition of the precipitates depend on the steel composition (Ni and Mn content). In the steel with increased Ni content up to 5 wt.% but with ultra-low Mn content ≤ 0.03 wt.%, it is possible to suppress the formation of Ni-Si-Mn precipitates under irradiation.

The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels

Embrittlement of light water reactor pressure vessel (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. The steels were irradiated to a fluence of ∼ 1.38×1020 n/cm2 at ∼ 292 °C in the UCSB ATR-2 experiment. In nominally Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promote precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.

Machine learning predictions of irradiation embrittlement in reactor pressure vessel steels

Irradiation increases the yield stress and embrittles light water reactor (LWR) pressure vessel steels. In this study, we demonstrate some of the potential benefits and risks of using machine learning models to predict irradiation hardening extrapolated to low flux, high fluence, extended life conditions. The machine learning training data included the Irradiation Variable for lower flux irradiations up to an intermediate fluence, plus the Belgian Reactor 2 and Advanced Test Reactor 1 for very high flux irradiations, up to very high fluence. Notably, the machine learning model predictions for the high fluence, intermediate flux Advanced Test Reactor 2 irradiations are superior to extrapolations of existing hardening models. The successful extrapolations showed that machine learning models are capable of capturing key intermediate flux effects at high fluence. Similar approaches, applied to expanded databases, could be used to predict hardening in LWRs under life-extension conditions.

Variation of Mechanical Characteristics and Microstructural Evolution in AISI 316 Austenitic Stainless Steel Subjected to Long-Term Thermal Aging at Elevated Temperature*

The objective of this study was to investigate the mechanical degradation of AISI 316 austenite stainless steel as light-water reactor pressure vessel materials in the primary system of nuclear power plants. The influences of long-term aging on the microstructural and mechanical characteristics were studied. The long-term aging tests were interrupted at various stages to obtain different levels of degraded specimens. The test specimens were heat-treated and exposed to an acceleration heat-treatment at 600°C for up to 10,000 hours. AISI 316 steel showed polygonalshaped grains with many annealing twins, and these tended to become gradually more circularshaped grains during long-term aging. In the initial material degradation, twins were distributed uniformly within most grains, but they all recovered and disappeared after a prolonged 10,000-hour aging time. Delta ferrite along austenitic grain boundaries transformed into sigma phases and Cr23C6 precipitates during long-term aging, and the area fraction on the grain boundary increased. The peak strength appeared at a 100-hour aging time and then decreased up to 1000 hours. With further aging time, the strength increased to a higher level than the initial state. However, the elongation and toughness decreased continuously, demonstrating the material embrittlement during long-term thermal aging.

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65 Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65 Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237 Np(n,f) in the cavity capsule using BUGLE-B7.

Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300°C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

Thermodynamic models of low-temperature Mn-Ni-Si precipitation in reactor pressure vessel steels

Abstract

Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

This paper is illustrating the potential of the well-known low alloyed clean steels, extensively used for the current light water Reactor Pressure Vessels (RPV) steels, for a likely use as a structural material also for the new generation nuclear systems. This option would provide, especially for large components, affordable, easily accessible and a technically more convenient solution in terms of manufacturing and joining techniques.A comprehensive comparison between several sets of surveillance and research data available for a number of RPV clean steels for doses up to 1.5 dpa, and up to 12 dpa for 9%Cr steels, is carried out in order to evaluate radiation stability of the currently used RPV clean steels even at higher doses. Based on the numerous data available, positive preliminary conclusions are drawn regarding the eventual use of clean RPV steels for the massive structural components of the new reactor systems.

Effect of Chloride on the Oxides on Low-Alloyed Steel in Conditions of a Light Water Reactor Pressure Vessel Cladding Flaw

Strain-induced corrosion cracking in low alloyed steels has been extensively investigated during the last decades. One finding from recent investigations is that small amounts of chlorides tend to increase the cracking susceptibility of such materials. The present paper investigates the effect of chloride addition on oxide films on low-alloyed steel in conditions simulating those in a cladding flaw of a light water reactor pressure vessel under oxygenated high-temperature water conditions. The films were studied by in-situ electrochemical impedance spectroscopy and mixed potential measurements, coupled to ex-situ characterization of the formed oxides by microscopic and surface analytical techniques. The impedance data were quantitatively interpreted by the mixed-conduction model for oxide films to yield estimates for the main kinetic and transport parameters of oxide growth and restructuring. On the basis of the experimental data and calculation results, it can be concluded that the effect of chloride on low-alloyed steel oxide is moderate, concerns mostly the processes at the inner layer/coolant interface and is to a major extent reversible.

Effects of alloying elements on radiation hardening based on loop formation of electron-irradiated light water reactor pressure vessel model steels

Electron irradiations using a high voltage electron microscope were conducted on several reactor pressure vessel model alloys in order to investigate the effects of alloying elements on the formation and development of defect clusters. In addition, the effects of alloying elements on yield stress change after irradiation were considered, comparing the mean size and number density of dislocation loops with the irradiation-induced hardening. High Cu alloys formed Cu and Mn–Ni–Si rich clusters, and these are important in determining the yield stress increase. High Ni alloys formed a high density of small dislocation loops and probably Mn–Ni–Si rich cluster, which have the effect of increasing the yield stress. High P enhanced radiation-induced segregation on grain boundary, helping prevent dislocation movement.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation

This paper, and its companion paper [Tran C.T., Dinh, T.N. The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application. Progress in Nuclear Energy (companion paper), in preparation] document the development, validation and applications of a simulation platform for computationally-effective, sufficiently-accurate numerical predictions of core melt-structure-water interactions in the light water reactor lower head during a postulated severe core-melting accident. The centerpiece of this work is the Effective Convectivity Model (ECM) for description of energy splitting in a core melt pool. Built on the concept of characteristic velocities in Effective Convectivity Conductivity Model and supported by the key findings obtained from Computational Fluid Dynamics (CFD) simulations of turbulent natural convection, heat transfer and phase changes in volumetrically heated liquid pools, the ECM is refined and extended to three-dimensions and phase changes to enable simulations of melt pool formation and corium coolability in complex geometry such as a Boiling Water Reactor (BWR) lower plenum.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper ( Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

Use of CPXSD for Generation of Effective Fast Multigroup Libraries for Pressure Vessel Fluence Calculations

Abstract Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contributon and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine- and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (∼1 %–2 %) with the fine-group library from which they were derived, in reaction rate calculations.

Simulation of irradiation effects in light water reactor vessel steels – experimental validation of RPV-1

The REVE project (REactor for Virtual Experiments) was an international effort aimed at developing tools to simulate irradiation effects in materials of Light Water Reactors. In the framework of this project, a European team developed a first tool, called RPV-1, to simulate irradiation effects in light water reactor pressure vessel steels. This article is the fourth of a series dedicated to the presentation of RPV-1. It has a twofold objective: – to show quantitative comparisons between experimental and RPV-1’s simulation results; – to demonstrate that RPV-1 can already be used to complement experimental programs. To this end, RPV-1 has been used to reproduce the French experimental program ESTEREL which was aimed at quantifying the neutron spectrum effect between the surveillance capsules and the vessels of the French reactors, and to determine the best irradiation parameter to assess the behaviour of these vessels from the results of the surveillance program.

06/02131 RPV–1: a virtual test reactor to simulate irradiation effects in light water reactor pressure vessel steels: Jumel, S. and Van-Duysen, J. C. Journal of Nuclear Materials, 2005, 340, (2–3), 125–148.

06/02131 RPV–1: a virtual test reactor to simulate irradiation effects in light water reactor pressure vessel steels: Jumel, S. and Van-Duysen, J. C. Journal of Nuclear Materials, 2005, 340, (2–3), 125–148.

Verification of MultiTrans Calculations by the VENUS-3 Benchmark Experiment

Abstract The MultiTrans software has been developed at VTT Technical Research Centre of Finland for 3D radiation transport problems. Adaptive tree multigrid technique is used as a deterministic solution method. This enables local refinement of the calculation grid combined with the use of effective multigrid acceleration on tree-structured nested grids: starting from a fast solution on coarse grid, successive solutions are obtained on finer and finer grids. In the MultiTrans code, simplified spherical harmonics (SP3) radiation transport approximation is used. In order to test the applicability of the new MultiTrans code to reactor dosimetry problems, light water reactor pressure vessel steel (LWR-PVS) benchmark experiment VENUS-3 (with partial length shielded assemblies) was chosen. The results show good agreement to the experimental reaction rates of the VENUS-3 benchmark, demonstrating the applicability of the new MultiTrans code in reactor dosimetry.

Phase-Field Simulation of Phase Transformation in Fe-Cu-Mn-Ni Quaternary Alloy

The phase decomposition in α(bcc) phase and the subsequent structural phase transformation from a to γ(fcc) phase during isothermal aging of an Fe-Cu-Mn-Ni quaternary alloy, which is a base alloy of the light-water reactor pressure vessel, have been simulated by the phase-field method. At the early stage of spinodal decomposition, cu-rich α phase is formed, and the Mn and Ni, which are minor components, are partitioned to the Cu-rich phase. As the Cu composition in the precipitate is increased, the Ni atoms inside the precipitates move to the interface region between the precipitate and matrix, but Mn atoms remain inside the Cu particles. When the Cu-rich particles eventually transform to the fcc structure, the Mn atoms also move to the interface region, which results in the shell structure of the fcc Cu precipitates, where each particle is surrounded by a thin layer enriched in Mn and Ni. This microstructural change can be reasonably explained by considering the local equilibrium at the surface region of the Cu-rich particles.

Interactions between dislocations and irradiation-induced defects in light water reactor pressure vessel steels

The REVE project (REactor for Virtual Experiments) is an international effort aimed at developing tools to simulate irradiation effects in light water reactors materials. In the framework of this project, a European team developed a first tool, called RPV-1 designed for reactor pressure vessel steels. This article is the third of a series dedicated to the presentation of the codes and models used to build RPV-1. It describes the simplified approach adopted to simulate the irradiation-induced hardening. This approach relies on a characterization of the interactions between a screw dislocation and irradiation-induced defects from molecular dynamics simulations. The pinning forces exerted by the defects on the dislocation were estimated from the obtained results and some hypotheses. In RPV-1, these forces are used as input parameters of a Foreman and Makin-type code, called DUPAIR, to simulate the irradiation-induced hardening at 20 °C. The relevance of the proposed approach was validated by the comparison with experimental results. However, this work has to be considered as an initial step to facilitate the development of a first tool to simulate irradiation effects. It can be improved by many ways (e.g. by use of dislocation dynamics code).

RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called ‘Experimental Test Reactors’. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature,…) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program.