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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.

Boiling water reactors have internal components manufactured with stainless-steel type 304. Some relevant cracks have been detected through the periodic inspections of their internal components. Re...

Kuosheng Nuclear Power Plant (NPP) is a BWR/6 plant in Taiwan. First, this study focuses on the establishment of the TRACE/SNAP model for Kuosheng. To check the system response of the Kuosheng TRACE/SNAP model, the Final Safety Analysis Report (FSAR) and startup tests data are used to assess the TRACE/SNAP model. The TRACE predictions are consistent with the FSAR and startup tests data. This indicates that there is a respectable accuracy in the TRACE/SNAP model of Kuosheng NPP. Second, this study also focuses on the application of the TRACE/SNAP model in the core shroud leakage. The Kuosheng NPP TRACE/SNAP model is used to perform the analysis of core shroud leakage. The TRACE results imply that Kuosheng NPP is in a safe situation when the core shroud leakage transient occurs.

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.

Improvement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant. In this study, the FSI effect on dynamic characteristics of RVIs in a typical 1,000 MWe nuclear power plant was investigated. Modal analyses of an integrated assembly were conducted by employing the fluid-structure (F-S) model as well as the traditional added-mass model. Subsequently, structural analyses were carried out using design response spectra combined with modal analysis data. Analysis results from the F-S model led to reductions of both frequency and Tresca stress compared to those values obtained using the added-mass model. Validation of the analysis method with the FSI model was also performed, from which the interface between the upper guide structure plate and the core shroud assembly lug was defined as the critical location of the typical RVIs, while all the relevant stress intensities satisfied the acceptance criteria.

This paper presents investigation results on effect of neutron irradiation embrittlement on cyclic hardening behavior of core shroud assembly in pressurized water reactor under beyond-design basis earthquake (BDBE). In order to simulate the cyclic hardening behavior considering the neutron irradiation embrittlement, the previous user subroutines developed for monotonic hardening behavior were updated based on some cyclic hardening test results of irradiation-simulated specimens. Finite element (FE) elastic-plastic seismic analysis was performed by using Chaboche combined hardening models considering cyclic hardening or cyclic softening and a bi-linear kinematic hardening model. As a result, it is found that the bi-linear kinematic hardening model derived the most conservative results, and the Chaboche combined hardening model considering cyclic softening derived the analysis results close to those of the bi-linear kinematic hardening model. In addition, variation of the FE elastic-plastic seismic analysis results vs. neutron irradiation dose was investigated using the bi-linear kinematic hardening model. Finally, it is identified that von-Mises effective stress and plastic strain energy increase and equivalent plastic strain decreases with increasing the neutron irradiation dose.This paper presents investigation results on effect of neutron irradiation embrittlement on cyclic hardening behavior of core shroud assembly in pressurized water reactor under beyond-design basis earthquake (BDBE). In order to simulate the cyclic hardening behavior considering the neutron irradiation embrittlement, the previous user subroutines developed for monotonic hardening behavior were updated based on some cyclic hardening test results of irradiation-simulated specimens. Finite element (FE) elastic-plastic seismic analysis was performed by using Chaboche combined hardening models considering cyclic hardening or cyclic softening and a bi-linear kinematic hardening model. As a result, it is found that the bi-linear kinematic hardening model derived the most conservative results, and the Chaboche combined hardening model considering cyclic softening derived the analysis results close to those of the bi-linear kinematic hardening model. In addition, variation of the FE elastic-plastic seismic analysis...

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.

Crack growth rate (CGR) testing in BWR normal water chemistry was performed on compact tension (CT) specimens of two different sizes (B = 8 mm and B = 19 mm), machined from a Type 304 SS core shroud at a dose of ~1 dpa. The objectives were to study the effect of specimen size on the CGR, and to determine the K validity limit for a CT specimen dimension used in previous studies. The results show that for materials with significant strain hardening capacity remaining, there is no effect of specimen size on the CGR when testing is conducted at stress intensity factors valid according to ASTM E399 using the flow strength. For materials at higher dose in which the strain hardening capacity is lost or greatly reduced, a different K validity criterion might be applicable.

This paper contributes to a TEM examination on the oxide films formed at three locations along a crack path in Alloy 182 weld from a BWR core shroud support leg, namely, the crack mouth, the midway between the mouth and the crack tip, and the crack tip. In the crack mouth the oxide film was approximately 1.6 μm in thickness and consisted of relatively pure NiO. The midway oxide film was mainly a nickel chromium oxide with a film thickness of 0.3 μm. At the crack tip the oxide film was a nickel chromium iron oxide with a film thickness of 30 nm. In all studied locations the main oxides had the similar rocksalt structure and the cracks were much wider than the thicknesses of the oxide films. It probably suggests that the corroded metal was largely dissolved into the coolant. The different dissolution rates of nickel, chromium and iron cations in the oxide films are clearly displayed with the compositions of the residual oxides. The oxide stability under different redox potentials along the crack path is briefly discussed.

Stress corrosion cracking (SCC) of the welded joints in a reactor core shroud is the primary result of the residual stresses caused by welding, corrosion and neutron irradiation in a boiling water ...

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.

Mexico has a nuclear power plant with two BWR reactors. Their internal components were manufactured with type 304 stainless steel. Since the initiation of their commercial operation, some relevant cracks have been detected through the periodic inspections of their internals components. Specifically, there have been crack found at the core shroud and jet pumps. The causes of such cracks have been: excessive vibration, wear, stress corrosion cracking and residual stresses, acting alone or simultaneously. There are several aspects that have to be considered in the evaluation of the structural integrity of the core shroud, for instance: its shape is a cylinder and its R/t ratio is in the range between 60 and 100, as well as the water environment and the neutron irradiation. The last one has modified the mechanical properties of the material. Besides, each circumferential weld has a different non-symmetrical and irregular distribution of cracks with various depths. These cracks are monitored every 18 months, only during the plant shutdowns. Furthermore, the potential unstable cracking of the lower circumferential welds could induce important safety consequences if a loss of coolant accident and an earthquake event are postulated. Due to these reasons, it is very important to apply or develop adequate analytical procedures that can provide fast and reliable results. These procedures must cover elastic-plastic fracture analysis and plastic limit load analysis. In this paper, a methodology for the plastic limit load analysis was developed. The critical limit load for any crack distribution was calculated with a great degree of confidence. Operating conditions were considered. The shape of the cracks varies and the evaluation depends on the geometry of the cracked cylinder. In the problem at hand, the neutral axis varies, as it translates and rotates simultaneously at every cracked cross section, depending on the loading condition. This situation was also taken into account.

Сделан анализ рассчитанного флюенсанейтронов для выгородки реактора ВВЭР-1000. Определены текущие и прогнозные механические свойства материала выгородки – стали 08Х18Н10Т. Представлена математическая модель радиационной ползучести материала элемента ВКУ. Реализована математическая модель распухания аустенитной стали 08Х18Н10Т, учитывающая напряженно-деформированное состояние, температуру, влияние пластических деформаций и радиационной ползучести. Даны оценки формоизменению и напряженному состоянию выгородки в течение 25, 40 и 60 лет работы реактора ВВЭР-1000.Показано, что за период 60 лет металл выгородки удовлетворяет всем предъявленным требованиям., роблено аналіз розрахованого флюенсанейтронів для вигородки реактора ВВЕР-1000. Визначені поточні та прогнозні механічні властивості матеріалу вигородки – сталі 08Х18Н10Т. Представлена математична модель радіаційної повзучості матеріалу елемента ВКП. Реалізована математична модель розпухання аустенітної сталі 08Х18Н10Т, яка враховує напружено-деформований стан, температуру, вплив пластичних деформацій та радіаційної повзучості. Дані оцінки зміни форми та напруженому стану вигородкипротягом 25, 40 та 60 років роботи реактора ВВЕР-1000. Показано, що за період 60 років метал вигородки задовольняє всім пред’явленим вимогам., The calculations of neutron fluence on the WWER-1000 reactor’s core shroud were made. The current and predicted mechanical properties of the main material (Steel 08Х18Н10Т) of the core shroud were determined. Represented mathematical model for the radiation creep at the main material of the vessel internals. The radiation intumescence mathematical model for the austenitic steel that considers different properties (stress stain condition, temperature, the influence of radiation creep and plastic deformation) was implemented. The forming and stress condition evaluations of the core shroud were made for the operational term of 25, 40 and 60 years. It is shown that for a period of 60 years metalcore shroud meets all requirements.

Зроблено аналіз розрахованого флюенсанейтронів для вигородки реактора ВВЕР-1000. Визначені поточні та прогнозні механічні властивості матеріалу вигородки – сталі 08Х18Н10Т. Представлена математична модель радіаційної повзучості матеріалу елемента ВКП. Реалізована математична модель розпухання аустенітної сталі 08Х18Н10Т, яка враховує напружено-деформований стан, температуру, вплив пластичних деформацій та радіаційної повзучості. Дані оцінки зміни форми та напруженому стану вигородки протягом 25, 40 та 60 років роботи реактора ВВЕР-1000. Показано, що за період 60 років метал вигородки задовольняє всім пред’явленим вимогам. The calculations of neutron fluence on the WWER-1000 reactor’s core shroud were made. The current and predicted mechanical properties of the main material (Steel 08Х18Н10Т) of the core shroud were determined. Represented mathematical model for the radiation creep at the main material of the vessel internals. The radiation intumescence mathematical model for the austenitic steel that considers different properties (stress stain condition, temperature, the influence of radiation creep and plastic deformation) was implemented. The forming and stress condition evaluations of the core shroud were made for the operational term of 25, 40 and 60 years. It is shown that for a period of 60 years metalcore shroud meets all requirements Сделан анализ рассчитанного флюенса нейтронов для выгородки реактора ВВЭР-1000. Определены текущие и прогнозные механические свойства материала выгородки – стали 08Х18Н10Т. Представлена математическая модель радиационной ползучести материала элемента ВКУ. Реализована математическая модель распухания аустенитной стали 08Х18Н10Т, учитывающая напряженно-деформированное состояние, температуру, влияние пластических деформаций и радиационной ползучести. Даны оценки формоизменению и напряженному состоянию выгородки в течение 25, 40 и 60 лет работы реактора ВВЭР-1000.Показано, что за период 60 лет металл выгородки удовлетворяет всем предъявленным требованиям.

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.

Recently intergranular stress corrosion cracking (IGSCC) in austenitic stainless steel 316L of Primary Loop Recirculation (PLR) piping and also a nozzle, and several core shrouds (CS) have been experienced in BWRs [1-4]. PWSCC incidences in nickel base alloys of Vessel Head Penetrations (VHP), Bottom Mount Instrumentation (BMI) nozzle and Steam Generator (SG) inlet nozzles in PWR were also reported [5-7]. Therefore, a great concern has been focused and an extensive research and development have been done on these SCC phenomena from a point of view of structural integrity of reactor components. In addition to those SCC incidences in BWR and PWR, there is one unique SCC incidence in 316 stainless steel in PWR plants [8-9]. The mode of this cracking is intergranular and cracks were found in weld HAZ of 316 stainless steel welded to a low alloy steel of SG nozzle with nickel base alloy 82 where PWSCC were also observed. Preliminary root cause analysis implied that the cracked 316 HAZ has a highly hardened not only at surface by machining but also at the region of a far distant from an inner surface of 316 pipe, probably due to a weld shrinkage strain as have been observed in PLR piping and Core shroud cracks in BWR. In laboratory tests [10, 11], there are many well-known evidences that strain hardened austenitic stainless steels such as 304, 304L, 316, 316L and 347, show SCC susceptibility without any sensitization of grain boundary when they are hardened by any sources during fabrication or under an operational condition. An extensive metallographic and TEM analysis has been performed on this cracked 316 HAZ sample as a part of failure analysis to characterize the microstructure and oxides morphology, compositional profiles as well as their crystal structures near the cracked surface of 316 HAZ [8, 9]. Typical strain hardening was measured on the sample due to weld shrinkage and also due to surface grinding. In addition to these analyses, an extensive surface oxides analysis has been performed by Raman Spectroscopy (RS) on a break-opened IGSCC surfaces and a cross-section sample where oxides can be observed in the cracks of concerns. In order to identify the oxides by RS, an extensive laboratory analysis of oxides formed in well-known environmental conditions in the laboratory was performed to generate a set of reference database of oxides by Raman spectroscopy. Based upon the observation of these oxides on/in the cracks, the crack growth history during the plant operation will be discussed in connection to a residual stress distribution at HAZ and oxides variation on/in the cracks. Possible time dependence of crack growth rate with crack growth in components will be discussed based upon the evidences observed about oxides. Finally, an importance of surface integrity assessment in SCC initiation and propagation will be emphasized from a point of view of oxidation localization which can be promoted by strain (dislocation density), straining and stress, which play an crucial role in an oxidation due to accelerated mass transfer in oxides as well as underlying metallic materials. Especially, preferential oxidation along slip bands suggests that oxygen diffusion in such a region with a high dislocation density is faster than the other region. This fact implies that grain boundary can also be a preferential path of oxidation as has been observed by TEM, TOFSIMS and 3D-APT. This localization of oxidation and acceleration will be discussed based upon an analysis of an profile development at an oxide/metal interface which is recently proposed by Y. Takeda et al. [12] based upon some statistical analysis of an oxide/metal interface profile and a concept of surface integrity will be emphasized for an quantitative prediction of SCC initiation.

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.

In the innovative design of the advanced boiling water reactor (ABWR), conventional recirculation loops are removed and replaced by multiple reactor internal pumps. Therefore, there is no major penetration of the reactor pressure vessel (RPV) below the elevation of the top of active fuel. As a result, an ABWR loss-of-coolant accident (LOCA) can have a decreased impact on reactor safety. Moreover, in the new RPV design the injection points of all the conventional low-pressure emergency core cooling (ECC) systems (ECCSs) are shifted out of the core shroud to the downcomer andfeedwater line as a new low-pressure ECCS, namely, a low-pressure flooder (LPFL). Consequently, the net hydraulic head built inside the downcomer will be the only driving force to bring the low-pressure ECC water into the core shroud during a large-break LOCA. In the analysis of a feedwater line break with RELAP5-3D/K, it was occasionally found that the hydraulic head built in the downcomer might not be great enough to bring the ECC water into the core shroud, and when the mixture water column ascends above the elevation of the feedwater rings, all the water injected by the LPFL will be directly driven to the break on the feedwater line. Fortunately, the capacity of the remaining high-pressure ECC flow directly injected above the core is great enough, and this ECC low-pressure injection bypass phenomenon can be terminated once the high-pressure ECC injection is manually turned off. This phenomenon of low-pressure ECC injection bypass is unexpected in the ABWR design, and it is worth further investigation.

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.

Neutron irradiation is known to have a considerable impact on the mechanical characteristics and the behaviour of materials and components. The distribution of residual stresses is one of the properties affected by irradiation. However, because of the difficulties in performing measurements in radioactive components, not many experimental data have to date been collected. At the High Flux Reactor (HFR) of the European Commission’s Joint Research Centre (JRC), a facility has been developed for residual stress measurements in steel specimens subjected to longterm irradiation. The objective of this development was to establish the neutron radiation induced changes in the residual stresses around welds in test pieces representative of the core shroud of boiling water reactors. Residual Stress measurements on such double-V butt welds in stainless steel plates after irradiation exposure have been performed by neutron diffraction using this facility. The comparison with measurements in non-irradiated companion specimens [1] showed that irradiation changed the distribution of residual stresses. The results suggested that the impact of irradiation varied with the distance of the test location from the specimen surface. On the basis of currently measured data we could not draw definitive conclusion about the influence of neutron dose and irradiation duration upon the stress modification.

The Stress Corrosion Cracking (SCC) has been observed in a component of low carbon austenitic stainless steel and in a weld of Ni based alloy in the domestic Boiling Water Reactors (BWRs) and in a weld of Ni based alloy in the Pressurized Water Reactors (PWRs). It is important for stable operation with a high level of safety to adequately perform preventive maintenance. Since the laser peening improves residual stress near the surface region by irradiating laser pulses in water, it has been applied for operating BWR and PWR plants as a preventive maintenance against the SCC. The residual stress in the weld region of core shroud support of BWR before and after the laser peening were examined by a finite element analysis. The magnitude of stress intensity factor (K) for a crack with an arbitrary depth is calculated by using the stress distribution. The results show that the K after the laser peening is not larger than that before (as-welded condition) and there is no ill effect on the crack propagation. It is also predicted that the laser peening effectively prevents the propagation of a small crack.

Slow strain rate tests for sensitized, solution annealed 304, and solution annealed cooled in air and welded, several sensitized and cold worked 304L stainless steels at 288 °C were carried out. Intergranular stress corrosion cracking IGSCC and Transgranular Stress Corrosion Cracking (TGSCC) were found in sensitized 304 stainless steel and only TGSCC was found at solution annealed 304 and 304L in all metallurgical conditions. The results are discussed based on previous results of residual strain and cracking in real core shroud welds and the qualitative role of residual strain is associated to the susceptibility to TGSCC in low sensitized stainless steel.

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.

The integrity of the reactor pressure vessel (RPV) head and reactor internals was assessed by means of fluid and fluid-structural coupled analyses to evaluate the water hammer phenomenon arising from postulated high burnup fuel failure under reactivity initiated accident (RIA) conditions. The fluid viscosity effect on the water column burst as well as the complex three-dimensional flow paths caused by a core shroud and standpipes were considered in this study. It is shown that fluid viscosity becomes an influential factor to dissipate impacting kinetic energy. Integrity of the RPV head and the shroud head was ensured with a sufficient level of margin even under these excessively conservative RIA conditions.

The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased {sup 239}Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s{sup -1}) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s{sup -1}) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greatermore » than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor.« less

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.

Internal components in nuclear reactor pressure vessels are joined to the ferritic vessel by use of dissimilar metal welds which commonly include nickel base weld material Alloy 182. It has turned out that Alloy 182 sometimes is susceptible to stress corrosion cracking (SCC) for the operating environment in reactors. Tensile residual stress has a large influence on SCC and it is important to carefully characterize the residual stresses generated at manufacturing. The manufacturing of these welds includes welding Alloy 182 to the ferritic steel to form a buttering, post-weld heat treatment (PWHT) of the buttering, and finally attachment welding between the internal component and the buttering. An experimental program was designed for measurement and numerical analysis for validation of residual stresses in a nickel base Alloy 182 weld between the core shroud support leg and reactor pressure vessel. Two full-scale mock-ups were manufactured according to the original procedures for the buttering to the ferritic steel and the final attachment weld to the core shroud support. The mock-up was also carefully designed to produce correct boundary conditions for the support leg. Measurements were performed by the deep-hole drilling technique (DHD/iDHD). The residual stress fields from welding and heat treatments were predicted by detailed numerical modelling. Comparison between the numerical results and the measurement results shows very good agreement and validates the predicted residual stresses. It was concluded that the PWHT of the vessel only partly relieve weld residual stresses in the nickel base buttering.

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.