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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.

This paper presents the assessment of the ex-core detectors in the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) in response to the movement of single primary control assembly, to evaluate the detectors’ locations and to determine a proper number of detectors. The detectors are located below the containment vessel and distributed in different ranges from the center and different azimuthal angles. Another position considered is in the radial direction of the core, at the sodium coolant gap between the outermost assembly and core shroud. To calculate the detection rate, the contribution of each fuel region to the detector and the core power distribution are required. Initially, the aforementioned contribution is generated by performing a continuous-energy forward fixed-source Monte Carlo calculation using MCNPX. The reactor kinetics code FREK simulates the power change due to an inadvertent withdrawal of a primary control assembly up to 10% from its critical position followed by full insertion of secondary control assemblies. The response from each detector location was evaluated, and consequently, the detectors located below the containment vessel, which can “see” all the fuel assemblies, responded better to the core power increase with a minor error of less than ∼0.6%. In addition, the impact of locating the detector below the containment vessel at different locations from the center or at different azimuthal angles was insignificant but several detectors should be distributed at even azimuthal angles for effective core monitoring and redundancy purposes.

The Advanced Boiling Water Reactor (ABWR) was adopted as a physical model in this work to investigate the influence of the type of PI controller combined with MAAP 5.03 code on the Core Shroud Water Level (CSWL). For all cases, the proportional gain ( k p ) is equal to 1. The numerical results reveal that when the integral gain ( k i ) and k p are the same, the feedwater flow modified by PI controller is sensitive. Accordingly, CSWL is always larger than the Level-2, which does not incur a scrammed reactor. Besides, CSWL modified by the PI controller also consists with the designed water level. However, k i ≪ k p causes the feedwater flow modified by PI controller to be insensitive. To avoid a scrammed reactor, a P controller (k ≫ 1 ) is added in front of the PI controller to increase the response speed of CSWL, which improves the tardy speed of feedwater injected into the reactor. Although k i ≫ k p does not cause a scrammed reactor, a P controller (k ≪ 1 ) added in front of the PI controller is able to reduce the vibration of CSWL; hence, CSWL is close to the designed water level. To exactly simulate the steady-state performance of a nuclear power plant under the normal operation, our results can offer MAAP code users an important reference to design the PI controller.

As limited by the capacity of the spent fuel pool (SFP), the last discharged full core will remain in the core for storage for an undetermined period. As a result, an extended refilling mode operation will be expected after the permanent shutdown of a BWR/4. During the extended refilling operation, the SFP, refilling cavity and reactor vessel are all connected. In the early transition phase, the residual heat removal (RHR) system will provide cooling to the reactor core, while the enhanced SFP cooling system will provide cooling to the SFP. While after the transition phase, the enhanced SFP cooling system will provide cooling to both the SFP and reactor core. An evaluation was performed to determine the duration of the transition phase containing both RHR and SFP cooling systems. Furthermore, systematic safety analysis was also performed to ensure the safety of the extended refilling mode operation. In this paper, advanced modeling techniques for RELAP5 and MELCOR were successfully developed to analyze the thermal hydraulic response of the extended storage pool. A postulated loss of cooling accident of the extended storage pool without any emergency mitigation was analyzed by RELAP5 and MELCOR, and good agreement between these two codes was observed. According to the steady state analysis of the extended storage pool, the early transition phase with RHR system cooling to the core will be three days, to assure the maximum water temperature in the core below the technical specification limits. Moreover, an assessment of a LBLOCA with rupture at recirculation line was performed by RELAP5 model. It was indicated that the existing makeup system cannot provide effective makeup coolant to the reactor core, unless the makeup coolant could be directed into the core shroud. It was also demonstrated by RELAP5 simulation that only the core spray can provide effective cooling to the reactor core, even under the worst recirculation line break scenario.

The development of analysis models for the Swiss reactors is a key objective of the STARS project at the Paul Scherrer Institut (PSI). Within this context there is a need for the development of computational fluid dynamics (CFD) models of the Swiss reactors in support of future high fidelity investigations of steady-state and transient scenarios. This article presents initial results for the CFD analysis of a Siemens KWU PWR with a focus on the flow behaviour and heat transfer in the gap between the core shroud and core barrel. Temperatures and densities in this region of the reactor are important, for example, for accurate estimations of fast neutron fluence and activation in the steel structures of the core shroud, core barrel and reactor pressure vessel. The flow behaviour in this region may also be relevant for better understanding of ex-core detector responses. The flow conditions in the core bypass region were found to be in the transition-to-turbulence regime, with vortex shedding taking place downstream of the core formers as a result of flow instabilities. The non-stationary nature of the flow presented a challenge in terms of obtaining a solution within a reasonable time period. Two approaches were proposed to address this challenge: time-averaging of the flow-field information before solving the conjugate heat transfer problem; time-averaging of surface heat fluxes in order to derive detailed surface heat transfer coefficients. Both approaches yielded similar results with similar computational effort. Several characteristics and features of the core bypass flow are discussed. Updated Monte Carlo simulation results show that the influence of the core bypass temperatures on the neutron fluence predictions is non-negligible. This highlights the importance of including accurate bypass temperatures in future Monte Carlo simulations focused on ex-core regions.

The material properties of the core shroud and supporting tie rod components are important factors for Boiling Water Reactor (BWR) life extension considerations. Given the variability of nickel and boron content found in 304 stainless steel shroud plates, and XM-19 and Inconel X-750 tie rod materials, this paper combines the latest spectral information in the vicinity of the core shroud to calculate hydrogen and helium gas production limits as well as atomic displacement damage spanning 80 full-power years (FPY) using the SPECTER code. Results indicate that, whereas boron has little effect on the atomic displacement damage, helium production from boron can be important and must be included in weldability assessments. Since the boron concentration is uncertain, helium generation is calculated assuming a composition range based on specifications and reported measurements. The results show that helium production caused by transmutations in the 304 SS shroud material are dominated by the concentration of boron and will have a significant impact on weldability after only a few years of service. For tie rods, the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) has been assessed by determining helium production in Inconel X-750 IASCC specimens irradiated in the Advanced Test Reactor (ATR) reactor and correlating with the helium produced in a BWR. As a tie rod material in a high thermal neutron flux environment, Inconel X-750 is more likely to be IASCC sensitive during service in a BWR due to helium than the XM-19 stainless steel alloy because of a greater nickel content and boron concentration. Weldability of the core shroud plate material, IASCC of the tie rod materials, and the gamma photon contribution to radiation damage are discussed.

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.

The continued operation of aging Boiling Water Reactors (BWRs) worldwide requires gradually increasing cost of inspection, maintenance, and repair. Intergranular Stress Corrosion Cracking (IGSCC) in sensitized austenitic stainless steel piping first became a major issue for BWRs in the 1980s, resulting in recognition of the susceptibility of reactor internals to IGSCC. Shroud cracking identified in 1993–1994 confirmed that IGSCC of internals is a significant issue for BWRs. IGSCC is a time-dependent, material degradation process, which is caused and accelerated by the presence of residual stresses, material sensitization, irradiation, cold work, elevated temperature, and corrosive environments. This paper emphasizes the importance of accounting for corrosive environments, or more exactly, electrochemical phenomena in modelling IGSCC and predicting the service life of BWR in-vessel components, and stresses the necessity of performing such modeling not just for a single state point under full power conditions, but for the whole operating history of the reactor, including startups and shutdowns. This paper demonstrates that ignoring electrochemical considerations may result in underestimating component lifetime and lead to unnecessary expenses for inspection and repair.

An empirical crack growth rate correlation model and a predictive model based on the slip–oxidation mechanism for Stress Corrosion Cracking (SCC) were used to calculate the crack growth rate in a BWR core shroud. In this study, the crack growth rate was calculated by accounting for the environmental factors related to aqueous environment, neutron irradiation to high fluence and the complex residual stress conditions resulting from welding. In estimating the SCC behavior the crack growth measurements data from a Boiling Water Reactor (BWR) plant are referred to, and the stress intensity factor vs crack depth throughout thickness is calculated using a generic weld residual stress distribution for a core shroud, with a 30% stress relaxation induced by neutron irradiation. Quantitative agreement is shown between the measurements of SCC growth rate and the predictions of the slip–oxidation mechanism model for relatively low fluences (5 × 10 24 n/m 2 ), and the empirical model predicted better the SCC growth rate than the slip–oxidation model for high fluences (>1 × 10 25 n/m 2 ). The relevance of the models predictions for SCC growth rate behavior depends on knowing the model parameters.

Fuel assembly bow in pressurized water reactor cores is largely triggered by lateral hydraulic forces together with creep processes generated by neutron flux. A detailed understanding of the flow induced bow behaviour is, therefore, an important issue. The experimental feedbacks and laboratory tests on fuel assembly bow show that it is characterized to a high degree by fluid–structure interaction (FSI) effects, therefore, consideration of FSI is essential and indispensable in full comprehension of the bow mechanism. In the present study, coupled computational fluid dynamics (CFD) and finite element simulations are introduced, calculating fuel assembly deformation under different conditions as a quasi-stationary phenomenon. The aim has been, on the one hand, to develop such a simplified fuel assembly CFD model, which allows set up of fuel assembly rows without loosing its main hydraulic characteristic; on the other hand, to investigate the bow pattern of a given fuel assembly row under different boundary conditions. The former one has been achieved by comparing bow shapes obtained with different fuel assembly (spacer grid) modelling approaches and mesh resolutions with experimental data. In the second part of the paper a row model containing 7.5 fuel assemblies is introduced, investigating the effect of flow distribution at inlet and outlet boundary regions on fuel assembly bow behaviour. The post processing has been focused on the bow pattern, lateral hydraulic forces, and horizontal flow distribution. The results have revealed importance of consideration of FSI, and that an inhomogeneous inlet/outlet flow condition causes significant ‘C’ shaped deformation of fuel assemblies towards the core shroud. The important effect of the core shroud has been also shown.

Tremendous growths of computer and computational fluid dynamic (CFD) technology enables the extensive analysis for the fuel and component integrity problems, the core inlet/exit flow distributions, the grid-to-rod fretting performance adjacent to the core shroud geometry, etc., by the use of large-scale computer model and resources. In this study, a full model is constructed to obtain the 3-dimensional, global and localized flow distributions of PWR reactor vessel upper plenum for Maanshan Nuclear Power Plant located in Taiwan and to characterize the important flow features based on a comprehensive flow model with extensive computational cells. Through the extensive computations, the lateral/axial flows in individual guidetube assembly, pressure and flows in the entire domain in the upper plenum region and the effect on the fuel integrity as well as the distributions of the shear stress are obtained for the upper plenum internal design improvement. The detailed flow patterns are important references to justify the patterns by subcomponent analysis or much simplified geometries due to the limitations of aged computer.

Stress corrosion cracking (SCC) in the heat affected zone is the primary damage form due to weld residual stress, corrosion and neutron irradiation environment in the core shroud of a boiling water reactor. The distribution of weld residual stress around a weld is necessary to be clarified to evaluate the structural integrity of core shroud for SCC. Moreover, studying the effects of welding parameters on residual stress on reducing the residual stress is very important to suppress the initiation and propagation of SCC. In this paper, we used a finite element method (FEM) to clarify the distribution of weld residual stress around the sixth horizontal weld (H6a) between the lower ring and the cylinder in the core shroud. The simulation results of axial stress were consistent with the experimental results at the inside and outside surfaces of the core shroud, respectively. The effects of thermal loads and cooling conditions were also investigated with the same model. We simulated the welding progress with water cooling on the inside and outside surfaces of the core shroud in order to study the influence of cooling conditions on the residual axial stress around the weld. The simulation results indicated that water cooling decreased the residual axial stress at the same side due to changing the temperature-affected fields. Moreover, with fixing the peak temperatures of weld passes, the simulation results of the distribution of residual axial stress by the thermal loads with different heating time were compared. The simulation results suggested that the heating time was expected to be longer and the heat flux to be smaller for reaching the small tension residual axial stress or even compression stress around the H6a weld.

This study deals with the design and development of calculational techniques and evaluation of key neutronic parameters of a typical PWR core having a total reactor power of 2652 MWt (890 MWe). The PWR core consists of 157 fuel assemblies containing a total of ∼72 tons of uranium arranged vertically in a concentric square array within the core shroud. Each fuel assembly contains 264 UO2 fuel pins with various enrichments (2.1, 2.6 and 3.1%), 24 control rods of Gd2O3 and one central water channel and all are arranged in a 17 × 17 array of matrix. Different computer codes including WIMS, TWOTRAN, CITATION and MCNP have been employed to develop a versatile and accurate reactor physics model of the PWR core. The computational methods, tools and techniques, customization of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardized and established/validated for the overall core analysis. The analyses were performed in 3 steps: firstly for fuel pincells, then for the fuel assemblies and finally for the whole core. The WIMS and MCNP calculated infinite multiplication factors for fuel pincells having 2.1% enriched 235U were found to be 1.23393 and 1.23654, for 2.6% enrichment 1.28635 and 1.28887, and finally for 3.1% enrichment 1.32481 and 1.32812, respectively. For fuel assembly, WIMS and MCNP calculated infinite multiplication factors having 2.1% enrichment were found to be 1.24853 and 1.25445, for 2.6% enrichment 1.30372 and 1.30992, and for 3.1% enrichment 1.34424 and 1.35041, respectively. The effective multiplication factor calculated by CITATION, TWOTRAN and MCNP for whole core were found to be 1.25580, 1.25909 and 1.26382, respectively. The peak thermal neutron flux in the core calculated by MCNP was found to be 5.0298 × 1014 neutrons/cm2 s and the average core power density was 17.1 kW/cm3. The calculated results from different codes were found to be very good agreement for different moderator conditions. The choice of computer codes like WIMSD, TWOTRAN, CITATION and MCNP which are being used in nuclear industry for many years were selected to identify and develop new capabilities needed to support PWR analysis. The ultimate goal of the validation of the computer codes for PWR applications is to acquire and reinforce the capability of these general purpose computer codes to perform the core design and optimization study.

In nuclear power plants, stress corrosion cracking (SCC) has been observed near the weld zone of the core shroud and primary loop recirculation (PLR) pipes made of low-carbon austenitic stainless steel Type 316L. The joining process of pipes usually includes surface machining and welding. Both processes induce residual stresses, and residual stresses are thus important factors in the occurrence and propagation of SCC. In this study, the finite element method (FEM) was used to estimate residual stress distributions generated by butt welding and surface machining. The thermoelastic-plastic analysis was performed for the welding simulation, and the thermo-mechanical coupled analysis based on the Johnson–Cook material model was performed for the surface machining simulation. In addition, a crack growth analysis based on the stress intensity factor (SIF) calculation was performed using the calculated residual stress distributions that are generated by welding and surface machining. The surface machining analysis showed that tensile residual stress due to surface machining only exists approximately 0.2 mm from the machined surface, and the surface residual stress increases with cutting speed. The crack growth analysis showed that the crack depth is affected by both surface machining and welding, and the crack length is more affected by surface machining than by welding.

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are also presented.

To decrease the timescale for the installation of a nuclear reactor, a reduced-scale model system is proposed for Korea's third generation nuclear reactor, the APR1400 (advanced power reactor 1400). The construction period of a nuclear power plant is one of the most important factors to make a company competitive in international nuclear energy markets. Our study is related to the modularization of reactor internals to reduce the construction period of nuclear power plants. Generally, reactor internals comprise three components: the core support barrel (CSB), the lower support structure (LSS)/core shroud (CS), and the upper guide structure (UGS). The existing method of assembly is very complicated and requires approximately 8–10 months to complete. The installation of the reactor vessel (RV) is a critical process during the construction period. The proposed method for the modularization of reactor internals can shorten the construction period by a minimum of 2 months compared to the existing method. In order to modularize the reactor internals, gaps between the CSB snubber lug and RV core-stabilizing lug must be measured using a remote method from outside the RV. Therefore, the development of a remote measurement system was necessary. In this paper, we select a suitable sensor and develop a reduced-scale model system for the physical simulation of gap measurement.

Recent studies suggest that material hardening should cause enhancement in stress corrosion cracking (SCC) growth rate of stainless steels (SS). In this study, SCC growth rates of SS irradiated up to 1.2×10 25 n/m 2 and of un-irradiated SS were measured in order to obtain the reasonable estimation for SCC growth behavior in the core shrouds, using 0.5, 0.7, 1.0T-CT specimens prepared from the actual BWR components (a core shroud made of 304SS and a top guide made of 316SS) as well as 0.5T-CT specimens prepared from H3 and H4 shroud weld mock-ups made of 316L. In irradiated SS, SCC growth rate in actual core shroud of high fluence was estimated later to be 10 −10 m/s. In un-irradiated SS, all SCC growth rates were below the K -d a /d t disposition curve of the JSME NA1-2002 standard, and this fact suggests that the degree of hardening assumed in the actual shrouds’ heat-affected zone (HAZ) should bring little enhancement effect in SCC growth rates.

Cost reduction and reliability increase are systematically pursued systems and components; this requires, amongst other, the availability of sophisticated computer programs and detailed analysis models. As an example, the core shroud, the structure having the function to maintain the core centered on its axis, is being designed in the Westinghouse AP1000, differently from previous plants, as a highly heterogeneous structure. Its thermal-mechanical sizing must take into due account accuarately determined internal heat generation rates. The latter, if determined by combining 2D and 1D neutron and γ-ray calculations which imply the separation of spatial variables and are mainly applicable for fluxes in the reactor beltline region, may include overly conservative margins. On the other hand, Monte Carlo methods do not allow an easy quantification of the uncertainties related to overall calculation. Three-dimensional deterministic models, based on the discrete ordinate transport theory, have the potential to provide accuarate design data; they can be also effective provided that the well-known difficulty to create and tune a complex geometrical model in a reasonable time is overcome and adequate computer resources are available to perform the calculation (until few years ago [Botta et al., 1996. Three-Dimensional Reactor Pressure Vessel Fast Neutron Fluence Calculations for the AP600 Using TORT, 3-D Deterministic Radiation Transport Computer Programs: Features, Applications and Perspectives, NEA/NSC/DOC, OECD/NEA. Paris, France], massive parallel computers (i.e. Cray Computers) available only to large national laboratories and selected industrieshad to be used). ANSALDO is acting as Westinghouse subcontractor and it supported Westinghouse in all AP1000 Licensing Process to NRC from 1999 to nowdays for the internal heating rate generation rate and RPV fluence calculations. As computer power growths up ANSALDO refined its calculation methodology in order to improve the design itself. The recent development, by ENEA Bologna, of a pre-post processor based on the combinatorial geometry (the computer program BOT3P, [ORSI, 2002. BOT3P: Bologna Transport Analysis Pre-Post-processors Version 1.0, Nuclear Science and Engineering, no. 142. American Nuclear Society, USA, pp. 349–354; ORSI, 2004. BOT3P: Bologna Transport Analysis Pre-Post-processors Version 3.0, Nuclear Science and Engineering, no. 146. American Nuclear Society, USA. pp. 248–255; ORSI, 2005. BOT3P: Bologna Radiation Transport Analysis Pre-Post-Processors Version 4.0, Nuclear Science and Engineering, no. 150. American Nuclear Society, USA, pp. 368–373; ORSI, 2003. BOT3P: A Mesh Generation Software Package for the Transport Analysis Codes DORT, TORT, TWODANT, THREEDANT and MCNP, International Conference on Supercomputing in Nuclear Applications SNA’2003. Paris, France; ORSI, 2004. NEA-1678/05 BOT3P4.1. BOT3P4.1 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results”, OECD/NEA Data Bank. Issy-les-Moulineaux, France]) and the ongoing increment of computer power make the three-dimensional deterministic program TORT (part of the DOORS 3.2 Package, [RSICC Computer Code Collection DOORS-3.2, 1998, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA]) an attractive choice to perform the determination of the core shroud heating rate distribution as a 3D continue functions. This paper illustrates the computational approach followed by ANSALDO Nucleare to determine the heating rate distribution into the W AP1000 Core Shroud (see Fig. 1). The calculation has been performed at 67 energy groups (47 neutrons plus 20 gammas) in P3 or in P5 Legendre approximation and 2,000,000 meshes have been used. The computational power required is massive: 2 GB of RAM and up to 50 GB of swap files peak storage for each run with an S8 full-symmetric angular quadrature and 120 GB of swap files peak storage with an S16 full-symmetric angular quadrature. By using a latest generation workstation based on an INTEL P-IV 3.2 GHz processor under LINUX operating system (both SUSE 9.2 or Red Hat 9.1 and better are suitable), each run requires less than a week of full computer time with the P3 Legendre approximation with an S8 full-symmetric angular quadrature. The code BOT3P has been used both for the preparation of the geometrical model and as post processor of the TORT program outputs. BOT3P permitted to cut the model development time by a significant factor and to retrieve efficiently large amount of data from the TORT binary files, allowing a graphical review of the input and output data and, finally, resulting in a unique tool for Quality Control for complex 3D models. This computational experience can be considered a good working example for the 3D deterministic transport method revival by considering that this method is widely accepted by licensing authorities.

The water cooled shield blanket made of Type 316L SS for the international thermonuclear experimental reactor (ITER) has potential issues related to stress corrosion cracking (SCC). Shroud mock-ups and boat samples taken from the core shroud of the boiling water reactor (BWR) with SCC were investigated from the viewpoint of microstructures and nano-hardness. Fine grains and deformation bands were observed in the hardened surface thin layers of the shroud mock-up, where hardness profiles in the ground portion was different from those in the milled portion. In the fine grain region, crevices were found only in the ground surface. In the core shroud, hardened surface regions were also found. Results showed that the crevices found on the ground surface could be one possible factor for SCC initiation.

This paper deals with the seismic analysis and fracture evaluation of a stabilized core shroud in a boiling water reactor of nuclear power plant. To study the adequacy of original seismic loadings, the dynamic behaviors of core shrouds with cracks, without cracks and with stabilizers are analyzed. Seismic analysis of a lumped-mass model of reactor internals is then performed to obtain the seismic loadings around various weldments of the repaired core shroud. The interaction between the core internals and this repaired core shroud is thus taken into account in this study. Further, fracture analysis of the stabilized core shroud is performed to obtain the stress intensity factors along the crack front of horizontal welds based on these seismic loadings. The computed results show that the influence of existing cracks in the stabilized core shroud is insignificant on the overall structural integrity.

Traditional limit load analysis and fracture mechanics analysis have been applied to evaluate the integrity of the degraded nuclear power plant components. Although these methodologies are generally accepted by the regulatory authorities in the nuclear industry, conservatism introduced by the uncertainties of inspection, material property, crack geometry, applied loading, neutron environment, etc. is recognized to have great impact on the evaluation accuracy. A probabilistic analysis may overcome this shortcoming and reveal some additional insight to the problem. The purpose of the present study is to apply probabilistic methods to analyze the degraded core shroud, and to predict the quantitative risk of the cracked shroud. In the analysis, the loading condition, crack growth rate, material properties and existing defects are all considered random. A sample analytical result shows that, based on some previously observed data and under certain assumptions, the crack-through probability of the studied core shroud is in the order of 10 −7 after 13 cycles of operation. The probability will increase considerably through operation cycles or operation years if no repair action is taken.

In the past, weld-induced residual stresses caused damage to numerous (power) plant parts, components and systems (Erve, M., Wesseling, U., Kilian, R., Hardt, R., Brummer, G., Maier, V., Ilg, U., 1994. Cracking in Stabilized Austenitic Stainless Steel Piping of German Boiling Water Reactors — Characteristic Features and Root Causes. 20. MPA-Seminar 1994, vol. 2, paper 29, pp.29.1–29.21). In the case of BWR nuclear power plants, this damage can be caused by the mechanism of intergranular stress corrosion cracking in austenitic piping or the core shroud in the reactor pressure vessel and is triggered chiefly by weld-induced residual stresses. One solution of this problem that has been used in the past involves experimental measurements of residual stresses in conjunction with weld optimization testing. However, the experimental analysis of all relevant parameters is an extremely tedious process. Numerical simulation using the finite element method (FEM) not only supplements this method but, in view of modern computer capacities, is also an equally valid alternative in its own right. This paper will demonstrate that the technique developed for numerical simulation of the welding process has not only been properly verified and validated on austenitic pipe welds, but that it also permits making selective statements on improvements to the welding process. For instance, numerical simulation can provide information on the starting point of welding for every weld bead, the effect of interpass cooling as far as a possible sensitization of the heat affected zone (HAZ) is concerned, the effect of gap width on the resultant weld residual stresses, or the effect of the ‘last pass heat sink welding’ (welding of the final passes while simultaneously cooling the inner surface with water) producing compressive stresses in the root area of a circumferential weld in an austenitic pipe. The computer program feresa (finite element residual stress analysis) was based on a commercially available abaqus code (Hibbitt, Karlsson, Sorensen, Inc, 1997. abaqus user's manual, version 5.6), and can be used as a 2-D or 3-D FEM analysis; depending on task definition it can provide a starting point for a fracture mechanics safety analysis with acceptable computing times.

The core shroud and other internal component (core internals) replacements were carried out at Fukushima-Daiichi unit number three of Tokyo Electric Power Company (TEPCO) in Japan. The project is the first experience of replacement procedures that have been developed for welded core shrouds and jet pumps. The outline and experience of the project are discussed in this paper.

This paper presents a computational model to predict residual stresses in a girth weld (H4) of a BWR core shroud. The H4 weld is a multi-pass submerged-arc weld that joins two type 304 austenitic stainless steel cylinders. An axisymmetric solid element model was used to characterize the detailed evolution of residual stresses in the H4 weld. In the analysis, a series of advanced weld modeling techniques were used to address some specific welding-related issues, such as material melting/re-melting and history annihilation. In addition, a 3-D shell element analysis was performed to quantify specimen removal effects on residual stress measurements based on a sub-structural specimen from a core shroud. The predicted residual stresses in the H4 weld were used as the crack driving force for the subsequent analysis of stress corrosion cracking in the H4 weld. The crack growth behavior was investigated using an advanced finite element alternating method (FEAM). Stress intensity factors were calculated for both axisymmetric circumferential (360°) and circumferential surface cracks. The analysis results obtained from these studies shed light on the residual stress characteristics in core shroud weldments and the effects of residual stresses on stress corrosion cracking behavior.

The effect of dissolved oxygen level on fatigue life of austenitic stainless steels is discussed and the results of a detailed study of the effect of the environment on the growth of cracks during fatigue initiation are presented. Initial test results are given for specimens irradiated in the Halden reactor. Impurities introduced by shielded metal arc welding that may affect susceptibility to stress corrosion cracking are described. Results of calculations of residual stresses in core shroud weldments are summarized. Crack growth rates of high-nickel alloys under cyclic loading with R ratios from 0.2 to 0.95 in high-purity water that contains

The life-limiting mechanisms for components and systems are physical aging and wear. Both of them are related to changes of microstructure in the bulk material or at the phase boundaries medium/material and material/material. They are triggered during operation by factors such as temperature, mechanical load, and environment. Thus, to achieve an utmost effective aging management it is necessary, to understand the underlying aging and wear mechanisms such as neutron irradiation, fatigue, corrosion, fretting, etc. Definition and qualification of suitable corrective and preventive actions against accelerated aging, requires precise knowledge of the aging processes and life-limiting situations and thresholds. It is obvious, then, that materials engineering plays a large part in effective and economical plant life management. Within this paper, the role of materials science and technology in plant aging management during the various stages within a whole life cycle of a power plant is described: (1) the correct choice of materials as part of a well-based materials concept in the design stage is very important for later plant operation. As an example steam generator materials are presented. (2) The parameters of the individual manufacturing processes during erection of components and systems must be optimally selected in order to guarantee long-term operation. As an example the reasons for core shroud cracking in a BWR NPP are discussed. (3) Aging mechanisms must be accounted for in operation of components and systems, and their effects have to be counteracted in order to prevent service-life limiting situations. Details are described with respect of corrosion and neutron irradiation. Demanding future tasks for materials science and technology are presented, which are necessary to continue to contribute to an optimized plant life management and to cost-effective operation of nuclear power plants at high safety levels.

The core shroud replacement of a boiling water reactor (BWR) was successfully completed at Fukushima-Daiichi Unit #3 (1F3) of the Tokyo Electric Power Company (TEPCO) in Japan. The core shroud and other core internal components made of type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. This project was the first application of the replacement procedure developed for the welded core shroud, and employed various advanced technologies. The outline of the core shroud replacement project and applied technologies are discussed in this paper.

As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Wurgassen Nuclear Power Station need not be feared. Although the material and corrosion–chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert's agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept.

Recent experience from early Swedish BWRs corroborate that all components in a nuclear power plant can be repaired or replaced with new ones. Oskarshamn 1 has gone through a thorough refurbishment project. A number of internals were repaired or replaced including the core shroud support which was welded to the bottom of the reactor pressure vessel. The project verifies that it is fully possible to carry out complicated inspection and repair work inside a nuclear pressure vessel which has been in operation for more than 20 years. Along with increased capacity factor, operating nuclear power plants get the financial conditions needed for extensive repair and modernization projects. Large power output leads to short pay-back times for the investments. The FENIX project at Oskarshamn 1 is such a project. There are utilities whose policy is to keep their plants in as-new condition for an unlimited length of time.

TRAC-BF1 transient analysis for a General Electric boiling water reactor (BWR/2) was performed for a main steam line break (MSLB) event and a recirculation line break (RLB) transient to determine the susceptibility of the core shroud to geometric alteration when intergranular stress corrosion cracking (IGSCC) is present at the upper H3 and lower H7 core shroud weld locations. The TRAC-BF 1 model developed for the MSLB transient utilized a 1-D vessel component model with both main steam lines rupturing at transient initiation. A detailed ANSYS model was developed to find the forces imposed at the upper H3 weld positions. The TRAC-BF1 model developed for the RLB transient utilized a 3-D four azimuthal sector vessel component. FORTRAN programs were developed to calculate the forces at the lower (H7) weld location. The results of both transients followed the typical thermal-hydraulic behavior pattern expected during the transients; however, the extreme forces exerted on the core shroud could displace or rotate the shroud during these two postulated transients. Such shroud movement could impede control rod insertion and impair the safe shut down of the reactor. The TRAC-BF 1 pressure response and the ANSYS lift forces demonstrated that when IGSCC is present at the upper H3 core shroud weld, a lift force of 95 925 Pa (13.91 psi) is generated during the double MSLB transient which is great enough to overcome the weight of the vessel internals resting on the core shroud, causing the core shroud to lift. Additionally, the finite element ANSYS stress analysis demonstrated that when IGSCC is not present in the upper core shroud weld, stresses of 1757.3 MPa (254 880.0 psi) are induced in the filler material during a main steam line break. This is great enough to overcome the ultimate strength of 1723.7 Ma (250 000 psi) of the filler material and cause failure of the weld. The TRAC-BF1 RLB transient results determined that during the RLB event a force of 401.7 kN (90 306 lbf) is generated at the lower H7 core shroud weld which is great enough to cause displacement of a segment of the core shroud support structure when IGSCC is present.

Stabilized austenitic steels are susceptible to intergranular stress corrosion cracking (IGSCC) under boiling water reactor (BWR) conditions. This important finding for the German nuclear power station industry arises from the detection of cracks during the last 3 years in reactor hot water pipes made from titanium-stabilized steel AISI 321 in six BWRs and in reactor core components made from the niobium-stabilized steel AISI 347 in one BWR. All the observed cracks had a common feature: they had their origin in the chromium carbide precipitates at the grain boundaries and in the associated chromium-depleted region near the grain boundary. These microstructural features in the heat-affected zones of the hot water pipe weldments were caused by the heat input during deposition of the root bead. The TiC partially dissolved in the region near the fusion line and the released carbon reacted to form chromium-rich M23C6. Regarding the cracks found in the core shroud and the core grid plates, it was shown that a sensitizing heat treatment of rings taken from the same heat of steel could give rise to a microstructure susceptible to IGSCC in the region of a weldment. High carbon contents coupled with low stabilization ratios led to sensitization. Residual stresses developed during welding provided the significant contributions to the tensile stress necessary for IGSCC. With regard to the service medium, the influence of the electrochemical corrosion potential (ECP) was recognized as a dominant factor, together with the conductivity. The corrosion potential was mainly determined by the radiolytic formation of H2O2; with increasing distance from the core, the H2O2 content decreased owing to catalytic decomposition. For the pipes the problem of IGSCC could be resolved by the use of optimized steel (lower carbon content with maximum allowable stabilization ratio), by production modifications (lowest possible heat input during welding with favourable residual stress development) and by improvements in water chemistry (allowing the very low limits specified for normal operation to be exceeded for only short periods). Hydrogen control is seen as an optimal measure, whose introduction in Germany BWR plants must be considered in terms of the benefits and disadvantages. The outage times required for the replacement were dependent on the technical conditions prevailing in the individual plants and were generally justifiable. However in one plant the outage time was 1 year and in another it was more than 2 years. The intended replacement of the core shroud components has been cancelled owing to the decision to decommission the reactor for commercial reasons not exclusively related to the core shroud damage.

This paper summarizes the results of analyses performed to assess the effect of a variety of design parameters and operational procedures on a station blackout severe accident at the Peach Bottom Atomic Power Station. The severe-accident melcor code, version 1.8.1 was used in these analyses. The following sensitivity studies were completed: effect of the automatic depressurization system actuation timing on the accident progression; effect of fuel and cladding porosities on vessel failure and containment failure times; effect of several parameters on the amount of in-vessel steel ejected into the cavity after vessel failure; effect of different parameters on vessel penetration failure time; vessel failure timing; and lower plenum shroud and core shroud temperatures. These sensitivity studies provided valuable insights into the melcor code behavior and into the progression of this severe accident. The most significant results are: (a) the optimum steam cooling of the core is accomplished when the automatic depressurization system is actuated when the core water level is at one-third of the active core height, delaying vessel failure by minutes and containment failure by hours, (b) vessel failure is significantly delayed (by 2 h) when lower-plenum debris quenching is included in the model, and (c) the core shroud melts during this transient.

The design of nuclear power plants includes provisions for heat removal from the reactor core in the event there is a loss of reactor coolant while shut down. Boiloff from decay heat can lead to inventory reduction and fuel heatup if no coolant makeup is available. Certain decay heat removal system failures in boiling water reactors can drain the upper vessel and downcomer. This leaves the water inside the core shroud at the same level as the top of the jet pumps. This becomes the starting point from which further inventory reduction is possible through boiloff. This study investigated the core thermal response following such a scenario. A simple model of the core was used for analysis of this sequence. The goal of the analysis was to determine the time at which the water in the core would boil down and fuel heat up to a specified temperature (1256 K). It is this interval during which the operator can take action that will mitigate the transient.

Boiling water reactors rely on the injection of soluble neutron absorbers to control the power in the event of failure in the control rod (scram) system. Typically this liquid “poison” is injected from eight small holes on a standpipe positioned vertically near the outer edge of the core shroud in the lower plenum. The achievement of control is predicated on the good mixing of this injected liquid with the coolant, which is recirculated around the core, upper plenum, and downcomer. However, because the flows are rather low ( ≈ 20% of rated value with pumps tripped as expected under such conditions) and the injected solution density is much higher than that of the primary fluid, there have been concerns raised about the efficiency and completeness of this mixing. This work provides the first openly available data addressing such concerns. To avoid potentially important scaling compromises, the data were obtained from full-scale simulations. From the experiments performed so far, we can conclude that complete boron mixing (entrainment) will occur for recirculation flow rates down to 8.2% of the rated value.

Pressure differences and the resultant dynamic load act on the core shroud when pressure waves propagate in the downcomer of a light water reactor (LWR) pressure vessel after rupture of the primary pipe has occurred. An equivalent geometry, i.e. a diverging duct is used to solve by Euler and wave equation for acceleration and velocity of the fluid behind the wave front, that the two-dimensional, time-dependent pressure distribution, induced by the wave propagation, can be calculated. The assumptions lead to an approximate but conservative value of the resultant core shroud load.