Boiling Water Reactor Research Articles

Formation of pure zirconium islands inside c-component loops in high-burnup fuel cladding

High-burnup Zr-based nuclear fuel claddings exhibit accelerated irradiation growth, corrosion and hydrogen pick-up, all correlated with the emergence of c-component dislocation loops. We made use of sub-nm-resolution atom probe tomography to characterize the nanoscale chemistry of c-loops in fuel cladding from boiling water reactor operation. We found segregation of Fe, Ni and Sn to dislocation lines and depletion of Sn and O inside the loops, resulting in nearly pure Zr islands. We also observed nucleation of suboxide inside one c-loop, pointing to a possible mechanism of accelerated in-reactor corrosion. The Zr islands might also promote hydride precipitation and associated degradation.

Fluid-Solid Interaction Analysis at the Slip Joint of a Boiling Water Reactor Type 5 Jet Pump

In this paper, the fluid-solid interaction of a jet pump of a boiling water reactor type 5 (BWR/5), with its riser subjected to a leakage flow through its slip joint, is reported. This is a fluid-elastic instability problem. A methodology is proposed for the evaluation of the velocity of the fluid at the slip joint with and without a labyrinth seal. It is calculated with computational fluid dynamics. The results show that such a seal reduces the velocity of the fluid and produces a stable and linear behavior between the inlet and the outlet fluid velocities at the slip joint. Then the first five natural frequencies of the jet pump assembly are evaluated. The range is between 24.74 Hz and 60.21 Hz. The mass of water inside and outside of such an assembly is considered. With these data and the dimensions of the slip joint, a finite element mesh is developed and the time step (∆t = 0.001 s) is determined. The fluid and structure mesh are coupled. The fluid flow through the slip joint without a labyrinth seal is evaluated with a two-way fluid-structure interaction under normal conditions of operation. Accelerations up to 8 g can be developed at the bottom of the mixer. The fluid flow is estimated during the first 0.25 s. Flow-induced vibration can be exacerbated in resonance conditions. These values are similar to those obtained in the experimental analyses reported in the open literature. One of the excitation frequencies caused by the interaction between the fluid and the structure was close to the third natural frequency of this assembly (46.99 Hz). If the integrity of the labyrinth seal is maintained, the jet pump will not present high-amplitude oscillations. Therefore, an adequate management of seal degradation is required and failures of the jet pump can be avoided.

Evaluations of radionuclide activity releases into environment during loss of coolant accidents using the ASTEC code in pressurized water reactors within design basis and design extension conditions

The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). An increase of the level of Nuclear Power Plant (NPP) safety by consolidated and more realistic evaluations of the Radiological Consequences (RC) of Design Basis Accidents (DBA) and a strengthening of the assessments of the NPP safety levels by considering accidental situations more severe than those integrated in plant designs (i.e belonging to Design Extension Conditions domain) were the two main motivations behind this project.More specifically, the project aims at consolidating and/or refining the assessments of the radiological consequences of explicit accidental scenarios within Design Basis Accidents (DBA) and Design Extension Conditions (DEC-A conditions without significant fuel degradation) in Light Water Reactors (LWR) through the improvements of existing code predictability; the upgrading of calculation chains and methodologies; the development/refinements of models. Within the Work Package 2 of the project, coordinated by TRACTEBEL and dedicated to calculation methodologies, existing methodologies or calculation chains and simulation tools have been applied to run a first batch of calculations dealing with different reactor types: PWR (Pressurized Water Reactor), BWR (Boiling Water Reactor), VVER (Water-Water Power Reactor) and EPR (European Pressurized Reactor). Loss Of Coolant (LOCA) and Steam Generator Tube Rupture (SGTR) accidents have been selected for the exercise and bounding scenarios of the DBA and DEC-A domains have been analysed. The results of this first set of calculations will be used as a reference to quantify the gains obtained by the updated methodologies/simulation tools developed within the project.This paper describes the results of the first batch of calculations, performed with the ASTEC integral code, simulating LOCA scenarios (DBA and DEC-A categories) in a PWR 900 MWe with a focus on the predicted number of failed fuel rods and Source Term (ST) in the environment governing the RC. Limitations of the used approach are outlined, as well as the needs for further upgrading the calculation chains are proposed in the light of the improvement that was planned within the project in Work Package 3 (WP3: LOCA – Loss of Cooling Accidents), coordinated by IRSN and dedicated to the improvement of code models dealing with LOCA scenarios.

New insights into stress corrosion cracking mechanism of Alloy 600 in boiling water reactor (BWR) environment

Several mechanisms have been proposed to explain the intergranular stress corrosion cracking (IGSCC) of Alloy 600 in BWR, but slip-dissolution at grain boundaries remains the most recognized cause. However, it is still uncertain whether the above degradation is accompanied by selective internal oxidation (SIO) as a precursor. Following new insights into the creviced bent beam (CBB) test, relevant evidence was discovered. Additionally, the results of First Principle Calculation indicate interstitial O actively competes for chromium in IG carbides to form chromium oxides during the SIO process, instead of passively waiting for the oxidation/decomposition of carbides to provide a chromium source.

Microstructure and mechanical properties evolution of the Cr coatings on SiC substrates under He irradiation

Environment coatings are frequently considered, to improve the corrosion resistance of silicon carbide (SiC) in pressurized water reactor (PWR) and boiling water reactor (BWR) surroundings. Chromium (Cr) is one of the main candidate materials for coatings on zirconium-based alloys for accident-tolerant fuel (AFT) cladding. In this work, the magnetron sputtered Cr coatings on SiC substrates were irradiated by the 500 keV He+ ions at room temperature. Moreover, the microstructural characterization was conducted on the irradiated specimens by using scanning electron microscope (SEM), and transmission electron microscope (TEM). The scratch test showed that the coating-substrate adhesion decreased from 41 N to 28 N (∼31.7%) before and after the irradiation, maybe due to that the irradiation produced tensile strain in the coating and mesophase along the coating-substrate interface. It revealed that irradiation accelerated the migration and aggregation of C elements on the surface of the coating. After the 800 °C annealing for 20 min, massive Cr atoms diffused into the SiC matrix and reacted with Cr to form a chromium silicide interlayer. Some cracks were observed at the interface, being attributed to the Kirkendall effect. Therefore, it is very necessary to avoid the formation of chromium silicide interlayer and decrease swelling mismatch between the metal coatings and the SiC substrate for the actual application in nuclear systems.

Multifractal detrended fluctuation analysis of boiling water reactors

The aim of this work is to determine a relationship between multifractal parameters and Boiling Water Reactor (BWR) stability; therefore, a Multifractal Detrended Fluctuation Analysis (MF-DFA) of the data series obtained from the Average Power Range Monitor (APRM) and the Local Power Range Monitors (LPRM) during stable operation and during an instability event of a BWR-5, which concludes with the SCRAM, was performed. First, a much larger number of data than recommended was considered to calculate the multifractality for the instability event (4,200 data points) and for stable operation (5,700 data points) and determine if there was any difference among these analyses. Subsequently, the data were divided into periods of 50 s (500 data points) to detect the moment when the reactor begins to operate in an unstable manner, here it was not necessary to have enough information to calculate the multifractal parameters because it was only of interest to detect any trend or difference between the multifractal parameters that indicated instability in the operation. The results show that the data from the stable operation present a behavior with higher multifractality compared to the data from the unstable operation. In fact, the results indicate that the data from the unstable operation of the reactor tend to monofractality. This can be seen more clearly through the width of the multifractal spectrum, it is larger for the data corresponding to the stable operation. The proposed methodology was successfully applied to a challenging case study (Forsmark stability benchmark) as it is difficult to detect unstable reactor operation.

Prediction of stress corrosion crack growth rate of cold worked stainless steels in BWR conditions by the slip oxidation model

Stress corrosion cracking (SCC) is a critical issue for boiling water reactor (BWR) components, and their cold working is considered to be an accelerating factor for SCC crack growth rates (CGRs). However, no method has been proposed to quantitatively evaluate the SCC CGRs of cold worked stainless steels. In this study, the changes in material properties due to cold working were reflected in the input parameters of the Hashimoto-Koshiishi model, a mechanistic model of SCC growth based on the slip oxidation mechanism, to enable a quantitative evaluation of the SCC crack growth rate (CGR) of cold worked SSs. In addition, a simplified equation as a function of the yield strength of cold worked 316LSSs was developed for ease of practical use in the integrity assessment of components. The CGRs calculated by the model and/or the simplified equation were in good agreement with the experimentally obtained CGRs for cold worked SSs and the heat affected zone (HAZ) of Type 316L SSs.

Integrating core physics and machine learning for improved parameter prediction in boiling water reactor operations

This study introduces a novel method for enhancing Boiling Water Reactor (BWR) operation simulations by integrating machine learning (ML) models with conventional simulation techniques. The ML model is trained to identify and correct errors in low-fidelity simulation outputs, traditionally derived from core physics computations. These corrections aim to align the low-fidelity results closely with high-fidelity data. Precise predictions of nuclear reactor parameters like core eigenvalue and power distribution are crucial for efficient fuel management and adherence to technical specifications. Current high-fidelity transport calculations, while accurate, are impractical for real-time predictions due to extensive computational demands. Our approach, therefore, utilizes the standard two-step simulation process-assembly-level lattice physics calculations followed by whole-core nodal diffusion computations-to generate initial results, which are then refined using the ML-based error correction model. The methodology focuses on improving simulation accuracy in regular BWR operations rather than developing a universal ML predictor for reactor physics. By training an advanced neural network model on the difference in high-fidelity and low-fidelity simulations, the model can reduce the nodal power error from low-fidelity simulations to around 1% on average and the core eigenvalue down to under 100 pcm. This result is under the condition of the normal variations of control rod pattern and core flow rate changes in standard BWR operations used in the training and evaluation of the machine learning model. This work suggests a promising approach for achieving more accurate, computationally feasible simulation solutions in nuclear reactor operation and management.

Evaluated Material Attractiveness of Plutonium Composition from Economic Simplified Boiling Water Reactor (ESBWR)

The potential for misuse of nuclear fuel has prompted many to make efforts to prevent a country from acquiring nuclear explosive devices (weapons). Weapon utility equated with nuclear material attractiveness when applied to protection and security. Material attractiveness evaluated using an equation first developed by Masaki Saito, namely ATTR. ESBWR is one of the reactors of the BWR reactor type that is still operating. The purpose of this study is to evaluate Material Attractiveness based on the composition of plutonium in the ESBWR reactor, during reactor operation and after reactor operation. The ATTR concept utilizes the plutonium isotope composition to evaluate the material attractiveness aspect. Parameters such as BCM, Rossi-alpha, and neutron prompt life become additional aspects in evaluating material attractiveness. In the research results, the ATTR value decreased from the beginning of reactor operation until the reactor stopped operating. At the beginning of the reactor operation, it was categorizing as a weapon-grade with an ATTR value of 0.19 and when the reactor stopped operating it was categorizing as an unusable grade with an ATTR value of 0.012. At the time of cooling time up to 1x106 (one million) cooling time, the ATTR value increases by 0.019 which was categorizing as an unusable grade. After a cooling time of 1x106 (one million) years up to 1x107 (ten million) years, the ATTR value was 0.001 which categorized as an exempt level. This states that the increasing burnup value will further reduce the ATTR value. Likewise with the Rossi-Alpha and Bare Critical Mass values. Conversely, it will decrease the value of neutron lifetime. The ATTR value after the reactor operates up to one hundred years of cooling time increases and then decreases significantly after 1x106 (one million) years due to the half-life of plutonium isotopes. Likewise, the Rossi-Alpha and Bare critical mass values. However, it is different with the neutron fast lifetime.

Enhancing Nuclear Reactor Safety through ICE Digital Systems and Cyber Integration

This article examines the imperative of enhancing nuclear reactor safety through the Integration of Instrumentation, Control, and Electrical) Digital (ICE) Systems and cybersecurity measures. Focused on advancements within Boiling Water Reactors (BWRs), particularly the BWRX-300, the discussion emphasizes the transformative impact of digital technology on safety protocols. ICE Digital Systems contribute to real-time monitoring, predictive maintenance, and remote operation, optimizing reactor performance. Simultaneously, robust cybersecurity measures are integrated to safeguard against potential cyber threats. The article highlights the synergistic effect of ICE Digital Systems and cybersecurity in fortifying the safety framework of nuclear reactors, ensuring a secure, efficient, and sustainable future for nuclear power.

Numerical study on the effect of the PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code

Abstract In this work, the boiling water reactor (BWR) was adopted as a physical model to study the influence of PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code. The designed reactor pressure can be simulated through the open area of a valve modified by PI controller. The proportional gain (k p) is equal to one in all cases. For the integral gain (k i) smaller than or equal to 100 (i.e., k i = 0.1, 1, 10 and 100), a scrammed reactor incurred by the high reactor water level (Level-8) that is due to the large reactor pressure drop does not occur in the simulation. Compared to k i = 1, 10 and 100, for k i = 0.1, the reactor pressure modified by the PI controller is more close to the designed reactor pressure; however, the time to meet the designed reactor pressure is longer. The reason is that k i = 1, 10 and 100 incur a larger overshoot in the modified reactor pressure through the feedback system; although the time to meet the designed reactor pressure can be shortened, the amplitude of the reactor pressure varying with time is obvious. To exactly simulate the reactor pressure of a nuclear power plant under the normal operation, our results can offer MAAP code users an important reference to design the PI controller.

Leakage and loss of contact/adjustment effects on slip-joins of a jet-pump assembly of BWR nuclear reactors by dynamic fluid-structure interaction modelling

This work presents a methodology for the evaluation of self-excited vibrations of jet-pump assemblies that make up the water recirculation system of boiling water nuclear reactors. Therefore, for the analysis three different type of models were developed: finite element, computational fluid dynamics and fluid-structure interaction models of a jet-pump assembly. The models were subjected to pressurized water flow in order to analyse leakage effect in the gaps between slip-joints and the diffuser of the assembly, and how the vibrations level can affect the integrity of jet-pump assemblies were also evaluated. Analyses of loss of adjustment of structural components in the wedge system were also performed, in order to study their contribution on the vibrations level under water flow conditions, since this effect has been identified as a potential failure mechanism in jet-pump assemblies. Three different gap sizes were modelled along with two critical loss of mechanical/adjustment components conditions, and the vibrations level of the jet-pump assembly under each modelled condition was obtained. The results showed the presence of self-excited vibrations of the jet-pump assembly under water flow, however little effect of leakage occurred in the slip-joint, on the vibration levels of the jet-pump assembly. Nonetheless, for the cases of loss of adjustment/contact components in the wedge system, high natural frequencies were obtained, and the vibrations were more severe when the loss of adjustment occurred in two mechanical/adjustment components. The results were also used to identify the mechanical components that may experience accelerated material degradation in the mixing-diffuser section of jet-pump assemblies, which occurred under loss of adjustment of mechanical components.

Experimental investigation on local flow structures of upward cap-bubbly flows in a vertical large-size square channel

Gas–liquid two-phase flows are typical in engineering systems involving heat and mass transfers. The cross-sectional geometries and sizes of these flow channels vary, exerting a notable impact on the thermofluid behavior. To develop thermofluid models, it is imperative to obtain local measurements of two-phase flows and gain insights into their characteristics from these measurements. Numerous experimental studies have investigated the local flow characteristics in circular pipes and rectangular channels; however, few studies on square channels have been conducted, particularly for flow regimes beyond bubbly flows, such as cap-bubbly flows. Given the importance of two-phase flows in large square channels for advanced nuclear reactors, such as the economic simplified boiling water reactor (ESBWR), we experimented with upward cap-bubbly flows in a large square channel. Detailed cross-sectional distributions of the void fractions, axial gas velocities, and interfacial area concentrations for two bubble-size groups were obtained at three axial locations using local measurements of a four-sensor optical probe. Based on the database, the cap-bubbly flow characteristics were understood, including flow development in a large square channel. In addition, the existing drift-flux correlations for large circular pipes predicted the measured void fraction with an accuracy of approximately ± 10 %, whereas the existing interfacial area concentration correlations reasonably correlated with the measured interfacial area concentration with an accuracy of approximately ± 30 %. Furthermore, the database was used to calculate the covariances of the void fractions. The correlations for large circular pipes significantly underestimated the void fraction covariance of the larger bubble group, probably because of the peculiar void fraction distribution of the group in a large square channel. The other covariances were well predicted.

High temperature electrochemical reaction parameters affecting electrochemical corrosion potential of nickel-base Alloy 82 weld metal

ABSTRACT The electrochemical corrosion potential (ECP) data of nickel-base alloy 82 weld metal (Alloy 82) in both O2 and H2O2 environments were measured. Although Alloy 82 is used at various welds in boiling water reactors (BWRs), the ECP data of Alloy 82 under BWR conditions are limited. The ECP data measured in this study revealed that Alloy 82 had higher ECPs than 304 and 316 L stainless steels and nickel-base alloy 182 weld metal (Alloy182) at low oxidant concentrations in both O2 and H2O2 environments. The ECPs of Alloy 82 calculated with an analytical ECP model that was developed in this study agreed with measured values in the cases that it was acceptable for the model to have an error range of ± 0.1 V in the tested range of O2 and H2O2 concentrations. The possible reason for the ECPs of Alloy 82 to be higher than those other materials at low oxidant concentrations is given as follows: the anodic polarization curve of Alloy 82 exhibits a lower current density at low potentials because of its higher Ni and Cr contents. The developed ECP model validated by the ECP data will contribute to evaluations of the effectiveness of mitigation technologies for stress corrosion cracking environments.

Determination of Mo-93 inventory in irradiated BWR tie plate using triple quadrupole ICP-MS

The inventory of radionuclides is a crucial input parameter for the safety assessment of irradiated metal waste. Among the radionuclides, Mo-93 is remarkable radionuclide owing to its long half-life (4,000 years) and poor sorptive property (anionic species). A method for determining Mo-93 by measuring characteristic X-rays was proposed previously; however, this method requires complex sequential and/or chromatographic separation of Mo-93. In the present study, triple quadrupole inductively coupled plasma mass spectrometry (ICP-MS/MS)—a leading-edge method—was used to determine the inventory of Mo-93 in an irradiated boiling water reactor (BWR) tie plate (stainless steel) for the first time. Furthermore, an activation calculation for Mo-93 was conducted using ORIGEN2. The results indicated that the inventory of Mo-93 in the BWR tie plate was 690 ± 83 Bq/g. This study demonstrates that triple quadrupole ICP-MS allows the simple quantitative analysis of Mo-93 in irradiated stainless steel without complex chemical separation.Graphical abstract

Initial Verification and Validation of a New CASMO5 JENDL-5 Nuclear Data Library for Typical LWR Applications

A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expanded for general-purpose applications. At the same time, many essential nuclides for conventional light water reactor (LWR) analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5–based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1–based library as possible with JENDL-5. This study performs initial verification and validation of the new library under typical LWR conditions. Verifications were performed based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark phase III-C, and the calculated kinf and fuel compositions of the boiling water reactor fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, a tank-type critical assembly critical experiment and Takahama-3 postirradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1–based library in predictions of reactivity and fuel compositions for typical LWR systems.

Analysis of Ex‐Vessel Debris Coolability of Boiling Water Reactors

Analysis of Ex‐Vessel Debris Coolability of Boiling Water Reactors

Development of scaling approach based on experimental and CFD data for thermal stratification and mixing induced by steam injection through spargers

Advanced Pressurized Water Reactors (APWRs) and Boiling Water Reactors (BWRs) employ a suppression pool as a heat sink to prevent containment overpressure. Steam can be discharged into the pool through multi-hole spargers or blowdown pipes in both normal and accident conditions. Direct Contact Condensation (DCC) creates sources of momentum and heat. The competition between these two sources determines the development of thermal stratification or mixing of the pool. Thermal stratification is of safety concern as it reduces the cooling capability compared to a completely mixed pool condition. In this work we develop a scaling approach to prediction of the thermal stratification in a water pool induced by steam injection through spargers. Experimental data obtained from large-scale pool tests conducted in the PPOOLEX and PANDA facilities, as well as simulation results obtained using validated codes are used to develop the scaling. Two injection orientations, namely radial injection through multi-hole Sparger Head (SH) and vertical injection through Load Reduction Ring (LRR), are considered. We show that the erosion rate of the cold layer can be estimated using the Richardson number. In this work, scaling laws are proposed to estimate both the (i) transient erosion velocity and (ii) the stable position of the thermocline. These scaling laws are then implemented into a 1D model to simulate the thermal behavior of the pool during steam injection through the sparger.

Optimisation of the cooling of pressurised thick-walled components operating with fluid at saturation temperature

Determining the optimum heating or cooling times for thick-walled pressure elements allows a steam boiler to reduce start-up or shutdown times. A new method has been developed to determine the optimum fluid temperature change to shorten the steam boiler start-up and shutdown. The fluid temperature curve is designated so that the total circumferential stress at the hole edge does not exceed the allowable stress. The total stress at the hole edge where it is concentrated consists of thermal and pressure stress. The example of the optimum cooling of the steam boiler drum illustrated the method used. A three-dimensional finite element method (FEM) analysis was carried out for the connection between the drum and the downcomer to determine the transient temperature and stress distributions. It was shown that the total circumferential stress due to pressure and thermal loading at the hole edge did not exceed the allowable stress when optimal fluid temperature changes were used. The method shown in this paper was compared with the boiler standard EN 12952-3, which permits the determination of allowable heating and cooling rates of boiler pressure components as a function of pressure. The comparison shows that the cooling/heating time determined by the new method can be significantly shorter than that determined by the method in EN 12952-3. By using the new method for designating the optimum fluid temperature, the flexibility of the boiler and the power unit can be improved as the heating or cooling of critical pressure elements can be more rapid. It can also be used to accidently cool PWRs (Pressurised Water Reactors) and BWRs (Boiling Water Reactors) pressure vessels.

Structural integrity analysis with crack propagation for reactor pressure vessel nozzles based on XFEM

Structural integrity analysis for reactor pressure vessel (RPV) nozzles has been a vital issue due to the irradiation embrittlement and especially the discontinuity of geometry in RPV nozzles, which may cause a higher stress concentration and influence the structural integrity of RPV nozzles. For the linear elastic fracture mechanics (LEFM) analysis of the RPV nozzle, a postulated crack with a specific crack depth is presumed to evaluate the stress intensity factors (SIFs) whose crack depth does not grow during LEFM analysis. Thus, this paper would discuss the crack propagation for the postulated crack with a specific crack depth under the different reference temperature of nil-ductility transition (RTNDT), and the SIFs on the specific crack depth are also calculated. In recent research, the extended finite element method (XFEM) was developed, which is based on the enrichment solution space to simplify the computational fracture mechanics with the crack propagation of various discontinuities, and can be used to evaluate the crack propagation of RPV nozzle with a specific crack depth under RTNDT of −23.89, −10.89, 0.89, 2.69, and 2.89 °C. To investigate the crack propagation and structural integrity of RPV nozzles, the finite element analysis of Taiwan domestic boiling water reactor (BWR) nozzles containing the postulated circular corner crack with depths of one-fourth and one-tenth thickness along the 45° path from the tip of the nozzle corner is engaged. The present result shows that the crack was growing when the RTNDT was 0.89 °C and. In addition, the XFEM calculation of the SIFs for the specific crack depth is similar to the LEFM results.