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The operating environment of nuclear reactors imposes extreme challenges on the materials from which the structures within and surrounding the reactor are constructed. Understanding the effects of exposure to this environment is critical for ensuring the safe long-term operation of these reactors. The Grizzly and BlackBear codes are being developed to model the progression of aging mechanisms and their effects on the integrity of critical structures. These codes take advantage of the capabilities of the MOOSE framework to solve the wide range of coupled physics problems that are encountered in predictive simulation of structural degradation. This paper provides an overview of these codes, with a specific focus on two capabilities relevant for light water reactor applications: reactor pressure vessel embrittlement and concrete degradation. [ABSTRACT FROM AUTHOR]

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant. As-welded and reheated regions mainly consist of acicular and polygonal ferrite, respectively. Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions (0.5–2.5 µm) at the brittle fracture primary initiation sites. High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch. The density, geometry, and chemical composition of the primary initiation inclusions were investigated. A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the de-bonded interfaces between the uncracked inclusions and weld metal matrix. Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance. [ABSTRACT FROM AUTHOR]

Featured Application: In this paper, the application of a nondestructive magnetic testing method is analyzed for neutron irradiation-generated embrittlement of nuclear pressure vessel materials. It was found that in spite of the big scatter of measurement points, the degradation of reactor pressure vessel steel can be successfully predicted by this technique. Furthermore, it turned out that the embrittlement highly depends on the local material conditions, which was found to be very different even within the same block of reactor steel material. This implicates different trajectories of their neutron irradiation induced degradation, as well as identifiable differences in the outcomes of their nondestructive evaluation. The magnetic descriptor results add reliable and precise determination of local embrittlement in the material than the traditionally used destructive mechanical parameters, so the results can be helpful for the future practical application of this technique for regular inspection of nuclear reactors. Neutron irradiation-generated embrittlement of nuclear pressure vessel steel was inspected by a nondestructive magnetic method, called magnetic adaptive testing (MAT). This method is based on systematic measurement and evaluation of minor magnetic hysteresis loops. Result of MAT measurement was compared with the result of the traditional Charpy measurement. Good correlation was found between these parameters. One of the main findings of the present work is that the considerable part of scatter of points obtained by magnetic measurement can be attributed to local material inhomogeneity. Another important conclusion is that the embrittlement highly depends on the initial local material conditions, i.e., the initial microstructure, which are very different even within the same block of reactor steel material. By taking this into account, the magnetic descriptors obtain more precise determination of the local embrittlement than the traditionally used destructive mechanical parameters from Charpy data. [ABSTRACT FROM AUTHOR]

Featured Application: In this study, the influence of neutron irradiation on the mechanical properties of nuclear pressure vessel materials is investigated using two independent methods of nondestructive magnetic testing. A correlation was found between magnetic characteristics and neutron irradiation-induced damage, regardless of the applied measurement technique. Additionally, by merging the outcome of both testing methods and applying a calibration/training procedure, the damage to reactor steel was successfully predicted. The results are helpful for the potential future practical application of these techniques to the regular inspection of nuclear reactors. The embrittlement of two types of nuclear pressure vessel steel, 15Kh2NMFA and A508 Cl.2, was studied using two different methods of magnetic nondestructive testing: micromagnetic multiparameter microstructure and stress analysis (3MA-X8) and magnetic adaptive testing (MAT). The microstructure and mechanical properties of reactor pressure vessel (RPV) materials are modified due to neutron irradiation; this material degradation can be characterized using magnetic methods. For the first time, the progressive change in material properties due to neutron irradiation was investigated on the same specimens, before and after neutron irradiation. A correlation was found between magnetic characteristics and neutron-irradiation-induced damage, regardless of the type of material or the applied measurement technique. The results of the individual micromagnetic measurements proved their suitability for characterizing the degradation of RPV steel caused by simulated operating conditions. A calibration/training procedure was applied on the merged outcome of both testing methods, producing excellent results in predicting transition temperature, yield strength, and mechanical hardness for both materials. [ABSTRACT FROM AUTHOR]

The irradiation resistance of pre-deformed FeCu alloy was studied using a 3 MeV Fe ion irradiation experiment at room temperature in comparison with that of the as-received sample. Nanoindentation and atom probe tomography (APT) were used to characterize the mechanical properties and solute distribution. The stress–strain curve obtained by nanoindentation shows that the yield strength (σ0.2) of the pre-deformed sample is unexpectedly reduced with an increase in the irradiation dose to five displacements per atom (dpa). We suggest that it results both from the decrease in the dislocation density and the suppression of defects during irradiation. APT shows that the nucleation of the Cu cluster is suppressed; however, its growth is promoted in the pre-deformed sample, resulting in the formation of sparse and coarse clusters at 1 dpa irradiation. These coarse Cu clusters were then unexpectedly refined to finer grains with an increase in the irradiation dose to 5 dpa. Theoretically, the improvement in the resistance to irradiation in the pre-deformed sample is attributed to the dense point-defect sinks, that is, the dislocations and grain boundaries introduced by pre-deformation. In addition, the contributions of the dislocations and grain boundaries to the sink strength are estimated for both the as-received and pre-deformed samples. The results indicate that dislocations, rather than grain boundaries, play a major role after deformation. [ABSTRACT FROM AUTHOR]

In situ irradiation in transmission electron microscope following pre-irradiation was performed on Fe-0.6Ni alloy to investigate the growth of dislocation loops under irradiation. The observations showed two types of loop growth behaviour. One type was almost linear growth with irradiation dose, possibly caused by loop absorbing freely migrating defects. The other was abrupt growth with rapid increase in diameter. It was found that nanometre-sized loop might migrate one-dimensionally, following which the large loop could absorb the small loop. This process can result in the abrupt growth of loop and broadening of loop size distribution. Image, graphical abstract [ABSTRACT FROM AUTHOR]

• A high-precision measurement method of refractive index matching temperature was significantly improved. • Sources of PIV random errors were evaluated by synthetic particle images generated by Monte Carlo method. • The turbulent flow in a scaled-down model of reactor pressurized vessel was obtained. Understanding and control of turbulent flow is desired for the sake of mixing enhancement or flow distribution in vessels of energy system, such as reactor pressure vessel (RPV) in pressurized water reactor (PWR). However, optical measurement of turbulent flow, e.g., particle image velocimetry (PIV), suffers from refraction at solid–liquid interfaces, especially when the vessels contain complicated internals. In order to reduce negative effects of refraction to negligible level, high precision (10−4) of matched index of refraction (MIR) is desired for turbulent flow measurement in vessels. The refraction method was modified and improved for precisely measuring the MIR temperature of the MIR system of sodium iodide (NaI) aqueous solution and acrylic for application in RPV model with complex geometry. The uncertainty of the refractive index difference between NaI solution and acrylic was reduced to 3 × 10−5. To obtain high-fidelity experimental data in complex RPV model, the detailed uncertainty analysis of PIV measurements should be quantified. To estimate uncertainty of the adaptive PIV algorithm, randomly distributed tracer particles in an assumed fluid domain are generated with Mento Carlo method. And sources of random errors were evaluated and fitted as empirical correlations to estimate PIV uncertainty. The high-fidelity PIV methodology was employed to realize high-resolution measurement of turbulent flow in a scaled-down RPV model. Flow in inlet pipe, nozzle and lower plenum were measured. The horizontal jets behind the holes on flow skirt were well-resolved. Current results have demonstrated promising potential for the developed MIR method in flow measurement in complex geometries. [ABSTRACT FROM AUTHOR]

In this work, the pressure-temperature (P-T) limit curves considering unified constraint for a typical large-sized reactor pressure vessel (RPV) with axial surface cracks have been investigated and developed based on the K J - A d * two-parameter and the Master Curve (MC) method. The ASME fracture toughness K Ic curve and the MC with and without considering unified constraint for irradiated RPV steel were used in the constructions of the P-T limit curves. The effects of crack sizes (constraint levels) and fracture toughness curves on the P-T limit curves have been analyzed for the RPV under heatup and cooldown conditions. The results further show that the P-T limit curves based on traditional ASME K Ic curve has the highest degree of conservatism and the use of the MC method can significantly increase the safety margin of P-T limit curves. With increasing crack sizes (increasing constraint levels), the safety region in P-T limit curves reduces, which may lead to the decrease of operating flexibility for nuclear power plants. The accuracy of the P-T limit curves can be improved by considering unified constraint. For smaller size cracks with lower constraint in the RPV, the consideration of constraint can reduce conservatism of the P-T limit curves, and for larger size cracks with higher constraint, it can avoid the non-conservatism. It is recommended to obtain more accurate P-T limit curves based on the MC with considering unified constraint and finite element analysis for RPVs under various operating conditions. • Pressure-temperature (P-T) limit curves considering unified constraint for reactor pressure vessel was developed. • Effects of crack sizes and fracture toughness curves on P-T limit curves were analyzed. • Safety region in P-T limit curves reduces with increasing crack sizes (increasing constraint levels). • The accuracy of the P-T limit curves can be improved by considering unified constraint. [ABSTRACT FROM AUTHOR]

The reactor pressure vessel (RPV) is one of the most important elements of a nuclear power plant (NPP). The RPV determines the plant operational lifetime since it is not replaceable economically. The purpose of the RPV steel study and enhancement to increase the NPP’s (Nuclear Power Plants) operation lifetime from the original 30–40 years up to 60–80 years or even beyond. The RPV lifetime limited by ageing of the RPV steels. RPV ageing highly depends on the main environmental effects: fast neutron radiation, thermal effects causing thermal ageing and low-cycle fatigue. Firstly, the chemical composition via aged mechanical properties was studied. Efforts to increase the toughness against the radiation embrittlement was enhanced by the appearance of the modern microstructural testing devices such as APFIM (atom probe field ion microscopy), SANS (small-angle neutron scattering) positron annihilation spectroscopy (PAS), transmission electron microscopy (TEM) and Mössbauer spectroscopy (MS). The information on the effect of alloying and polluting elements for the microstructure allowed us to produce increased ageing toughness of the RPVs, and to enhance the safety and lifetime calculations of them, supporting long-term safe operation (LTO).

Spring energized metal C-ring is an important basic component which is crucial to ensure the safety of nuclear power plant operation. The thickness of alloy lining which is one of the component part of metal C-ring will affect the sealing performance of C-ring. This article has built C-ring three-dimensional elastic-plastic finite element analysis model. Based on the reliability of the theoretical model which is verified by the experimental study,thickness of alloy lining on the mechanical behavior and sealing performance of C ring are discussed. The results show that under the same amount of compression,the thicker the alloy lining,the larger C ring stiffness slightly. But it has no obvious effect on the springback characteristic of C ring.When alloy lining thickness is small,the plastic strain value is higher which is adverse for the sealing performance. On the contrary,the greater thickness of alloy lining,the more uniform contact stress distribution,but too thick alloy lining can lead to sealing face wrinkles. So for millions of kilowatt nuclear reactor pressure vessel（ RPV） with a metal C ring,in order to ensure that metal C ring has good sealing performance,alloy lining thickness value of 0. 5-0. 7 mm should be recommended in the project,in order to ensure the sealing performance of C ring.

The 3 D FE models of reactor pressure vessel（ RPV） with an axial crackwere built. The aspect ratio of the crack is a/c = 1/2（ a = 30 mm,c = 60 mm）. The elastic-plastic fracture mechanics analysis of reactor pressure vessels with different cladding thickness（ h = 0 mm,4 mm and 8 mm） subjected to pressurized thermal shock（ PTS） transients was studied. The influence of cladding thickness on integrity has been discussed by using J-A2 method and J-Qmethod. The result shows that the Jintegral on the front of crack decreases with the increasing thickness of cladding. The Q and A2 valueincreases near the interface between cladding and base material. But the maximum of Q and A2 ofcladding on the front of the crack is near the deepest point of the crack and decreases with the increasing cladding thickness.

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

Probabilistic fracture mechanics (PFM) is considered a promising methodology in assessing the integrity of structural components in nuclear power plants because it can rationally represent the influence parameters in their probabilistic distributions without over-conservativeness. In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) which enables the probabilistic integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Several efforts have been made to verify PASCAL4 to ensure that this code can provide reliable analysis results. In particular, a Japanese working group, which consists of different participants from the industry and from universities and institutes, has been established to conduct the verification studies. This paper summarizes verification activities of the working group in the past two years. Based on those verification activities, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs have been confirmed with great confidence.

The Japan Electric Association code, JEAC 4206-2016 specifies a procedure for assessment of a reactor pressure vessel integrity against pressurized thermal shock (PTS) events which postulates a semi-elliptical flaw (10 mm in depth x 60 mm in length) beneath the cladding. JEAC 4206-2016 prescribes a simple method for calculating stress intensity factor of the flaw, taking into account the effect of yielding of cladding which has been originally developed in France. However, since this method is a rather conservative in comparison with that obtained by elastic-plastic finite element analysis, JEAC 4206-2016 also allows to calculate the stress intensity factor by converting the J-integral obtained from elastic-plastic finite element analysis. Since the J-integral is only defined for a linear or nonlinear elastic behavior material, the J-integral cannot be determined for an elastic-plastic behavior material during the unloading phase of PTS events. In addition, the residual stress and plastic strain induced by clad welding and joint welding needs to be considered in calculating the stress intensity factor. In this paper, the calculation procedure for a stress intensity factor of a fully circumferential flaw beneath the cladding during PTS events using finite element analysis is investigated.

The reactor pressure vessel is made of ferritic low alloy steel, which has fracture behaviors that include ductile to brittle transition temperature and decrease of fracture toughness in the beltline region of the reactor pressure vessel steel due to irradiation embrittlement. Decrease of fracture toughness due to irradiation during plant operation needs to be taken into account in the assessment of the reactor pressure vessel against pressurized thermal shock events. A large number of fracture toughness data were accumulated in domestic PWR plant surveillance program tests and knowledge on the fracture toughness evaluations, such as scattering, were obtained through research works. In this study, fracture toughness curve for the assessment of domestic reactor pressure vessels were developed taking into consideration the latest knowledge on fracture toughness evaluations, including investigations on fracture toughness data scattering and fracture toughness curve reliability.

In Korea, decommissioning of nuclear power plants and transportation of the decommissioning waste are expected to expand in the near future. It is necessary to confirm that radiological risks to the public and workers are not significant through radiological safety assessment. The objective of this study is to assess the radiological safety for transportation of RPV waste, which is a major decommissioning waste with relatively high level of radioactivity. It was assumed that the waste would be transported to the Gyeongju disposal facility by land transportation. First, the source term and transportation method of the RPV waste were determined, and the external dose rates from the waste were calculated using MCNP. Then, transportation scenarios were assumed under both normal and accident conditions. Under the scenarios, radiation doses were calculated using the RADTRAN. Under normal operation scenarios without a transportation accident, assuming 40 shipments per year, the average individual doses for the public ranged from 6.56×10 -6 to 2.18×10 -2 mSv yr -1 . The maximum individual doses for only a single shipment ranged from 2.43×10 -6 to 3.14×10 -1 mSv. For cargo handlers and vehicle crew members, the average doses were 2.26×10 1 mSv yr -1 and 2.95 mSv yr -1 , respectively. Under transportation accident scenarios, average individual radiological risks which are product of the radiation doses and the annual accident rates ranged from 1.14×10 -11 to 1.61×10 -10 mSv yr -1 by transportation route segment when considering the transportation accident rate. Average individual doses assuming transportation accident occurrence ranged from 2.62×10 -4 to 1.42×10 -3 mSv. The maximum individual dose under accident conditions was 7.99×10 -2 mSv. The calculated doses were below the regulatory limits in Korea. However, relatively high doses were observed for cargo handlers and vehicle crew members because of conservative assumptions. This study results can be used as basic data for the radiological safety assessment for the decommissioning waste transportation in the future., (© 2024 Society for Radiological Protection. Published on behalf of SRP by IOP Publishing Limited. All rights reserved.)