Boiling Water Reactor Operating 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.

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.

Electrochemical Corrosion Potential Abatement of 304 Stainless Steel Coated with ZrO2 Nanoparticles under Boiling Water Reactor Operation Conditions

Electrochemical Corrosion Potential Abatement of 304 Stainless Steel Coated with ZrO2 Nanoparticles under Boiling Water Reactor Operation Conditions

Phenomenological model of disturbance waves in annular two-phase flow

A new, phenomenological, model of disturbance waves in annular two-phase flow is presented. The model builds upon the three-field (gas, drops and film) one-dimensional model of annular two-phase flow by considering the disturbance waves as an additional fluid field, traveling over a thin liquid base film, along with its own mass, momentum and energy conservation equations. The waves are modeled as fluid particles characterized by basic macroscopic parameters such as the wave shape factor (amplitude/width) and the wave number density (spatial frequency of waves). Phase change is considered and relevant closure models are developed for the field mass and momentum exchange rates (drop entrainment/deposition and waves/base film interactions). A Boltzmann transport equation of wave number density is introduced to model the hydrodynamic non-equilibrium effects related to the wave creation, merging and splitting processes, characteristic of developing annular two-phase flows (due to e.g., inlet effect, phase change, geometrical change or transient). Finally, the base film mass and wave number density sink and source rates are modeled using a relaxation time approximation. The new model is motivated and fully documented in this paper and has been implemented in the latest version of the transient sub-channel analysis code MEFISTO-T. While the proposed model framework is generic for coherent disturbance waves, the required wave characteristic correlations are specifically developed and calibrated using high pressure steam/water experiments under equilibrium conditions relevant of Boiling Water Reactor operation. The model relaxation times are then estimated by comparing the model predictions against measured wave parameters downstream various flow perturbations (evaporation and film suction), along the channel recovery length toward equilibrium. For all considered experiments, the model predicts the averaged wave and base film parameters typically within +/-20% and is able to capture the axial development of disturbance wave characteristics. The consideration of non-equilibrium effects is found fundamental to the accurate simulation of axially developing annular two-phase flows.

Nanoscale chemistry of Zircaloy-2 exposed to three and nine annual cycles of boiling water reactor operation — an atom probe tomography study

Atom probe tomography was used in this work to study the metal close to the metal/oxide interface in the zirconium alloy Zircaloy-2 exposed to three and nine annual cycles of operation in a commercial boiling water reactor. The two exposure times correspond to before and after the onset of acceleration in corrosion, hydrogen pickup, and growth.The alloying elements Sn, Fe, Cr, and Ni were observed to be redistributed after exposure. After both three and nine cycles, clusters containing Fe and Cr and typically of a spheroidal shape with an approximate diameter of 5 nm were observed to be located in layers presumed to be layers of <a>-loops. On average, the cluster number density was slightly higher after nine cycles, with larger and more Cr-rich clusters. However, there were large grain-to-grain variations, which were larger than the differences between the two exposure times. Ni was only occasionally observed in the clusters. Sn was observed to be slightly enriched in the Fe–Cr clusters, but the Sn concentration was higher between than inside the layers of clusters. After nine cycles, clusters of Sn were detected in regions that were depleted of Fe and Cr. Enrichment of Sn, Fe, and Ni at features that appeared to be <c>-component loops was observed after nine cycles, whereas no such features were observed after three cycles. Enrichment of Sn and Fe, and small amounts of Cr and Ni, was observed at grain boundaries after both exposure times. After three cycles, a partially dissolved second phase particle of Zr(Fe,Cr)2 type that contained about ten times more Cr than Fe was observed.

An atom probe tomography study of the chemistry of radiation-induced dislocation loops in Zircaloy-2 exposed to boiling water reactor operation

This study is complementary to previous atom probe tomography (APT) studies of irradiation effects in the zirconium alloy Zircaloy-2. Using APT in voltage pulse mode, a difference in morphology was observed between clusters of Fe and Ni and clusters of Fe and Cr in Zircaloy-2 exposed to a high fast neutron fluence in a commercial boiling water reactor. The Fe–Ni clusters were disc-shaped with a diameter of 5–15 nm, whereas the Fe–Cr clusters were spheroidal with a diameter of approximately 5 nm. Both types of clusters appeared to be located at irradiation-induced <a>-type dislocation loops aligned in layers normal to the <c>-direction. The concentration of Fe was higher in the Fe–Cr clusters than in the Fe–Ni clusters. The dilute Fe–Ni clusters, which seem to be segregation of Fe and Ni inside the loops, had formed on all three families of first-order prismatic planes with some deviation from perfect <c>-axis alignment. The Fe–Cr clusters might be very small precipitates with a nucleation associated with the loops.

Experimental characterization of boiling two-phase flow structures under BWR core operating conditions

Local two-phase flow parameters have been measured at the Westinghouse FRIGG facility in a test fuel bundle under Boiling Water Reactor (BWR) core operating conditions, over a wide range of flow and power (up to critical power). The measurements were performed at end of heated length using sapphire optical probes and Pitot tubes placed in various radial positions of the bundle. By using both instruments simultaneously, the void fraction, mixture velocity, dispersed phase (bubble or drop) characteristic lengths and volumetric interfacial area can be measured. In addition, the identification of two-phase flow regimes can be attempted. The measurements are of particular interest since there is very limited available data in the literature at high pressure relevant to BWR operation.Experimental measurements are reported in this paper covering a range of inlet flow and bundle power relevant to BWR operation, at a system pressure of 70 bar. The dynamic void fluctuations are used to establish two-phase flow regime maps, one based on global parameters (for 1-D applications) and the second based on local parameters (for CFD applications). However, it is discovered that transition flow regimes typically observed at low pressure (cap, slug, etc) are not similarly observed at high pressure. In addition, the measured characteristic lengths of the dispersed phase are reported in bubbly and annular two-phase flow regimes. In view of the results, recommendations are provided for further experimental work and modeling improvements.

Large eddy simulation of thermal mixing with conjugate heat transfer at BWR operating conditions

Thermal fatigue occurs in most metals under cyclic heat loads and can threaten the structural integrity of metal parts. Detailed knowledge of these loads is of utter importance to prevent such issues. In this study, a large eddy simulation (LES) with wall-adapting local eddy viscosity (WALE) subgrid model is performed to better understand turbulent thermal mixing in an annulus with a pair of opposing cold inlets at a low axial level (z=0.15m) and with a pair of opposing hot inlets at a higher axial level (z=0.80m). Each inlet pair is 90° from each other in the azimuthal direction. Conjugate heat transfer between fluid and structure is accounted for. The geometry simplifies a control-rod guide tube (CRGT) in a boiling water reactor (BWR). LES results are compared with measurement data. This is one of the first times BWR conditions are met in both experiments and LES: pressure equals 7.2 MPa, while the temperature difference between hot and cold inlets reaches 216 K. LES temperatures at the fluid-structure interface are fairly correlated with their experimental equivalents, with regard to mean values, local variances, and dangerous oscillation modes in fatigue-prone areas (z=0.65-0.67m). An elastic analysis of the structure is performed to evaluate stress intensities there. From them, cumulative fatigue usage factors (CUFs) are estimated and used as screening criteria in the subsequent frequency analysis of temperature time series at the fluid-structure interface. The likelihood of initiating a fatigue crack is linked to the maximum CUF, which is 3.2 × 10−5 for a simulation time of ~10s.

Experimental investigation and modeling of void drift in modern BWR fuel designs

Abstract Modern fuel designs for Boiling Water Reactor (BWR) include complex features, such as part-length rods (PLRs) and spacer grids with mixing vanes that make the flow very heterogeneous and challenging to simulate accurately (e.g. using sub-channel analysis codes or CFD tools). Due to the lack of reliable local experimental data, the codes have thus far not been validated for the simulation of two-phase flow parameters under BWR operating conditions across the resulting heterogeneous fuel lattice. In order to improve the thermal-hydraulic code predictions, local annular two-phase flow parameters (void and velocity) have been recently measured in a test fuel bundle (including PLRs and spacer grids with mixing vanes) at the Westinghouse FRIGG facility under prototypical BWR core operating conditions. The instruments (local optical probes and Pitot tubes), have been positioned both in an open region downstream PLRs and in a neighboring region surrounded by full length rods (FLRs) at the outlet of the test section. The measurements show higher void fraction and velocity in the region downstream PLRs, thus clearly illustrating a steam crossflow phenomena towards open regions, also known as “void drift”. The measurements are used to support the development and validation of a new void drift crossflow model in the Westinghouse sub-channel analysis code VIPRE-W/MEFISTO-T, applicable to modern fuel rod bundles with PLRs. The results show significant improvements of the code predictions.

Application of sub-channel modeling to BWR core analysis

The current trends in nuclear science and engineering, related to modeling and simulation, are towards high-fidelity multi-physics multi-scale simulations to address industry challenge and high impact problems. These trends stimulate the utilization of sub-channel modeling approaches to nuclear reactor cores for both operation applications (core follow and cycle depletion evaluations) and safety applications (transient and accident analysis). While sub-channel modeling of Pressurized Water Reactor (PWR) cores has advanced significantly in last few years, the application of sub-channel modeling to Boiling Water Reactor (BWR) cores is under development. CTF is an improved version of COBRA-TF being developed and maintained by the North Carolina State University (NCSU) in cooperation with Oak Ridge National Laboratory (ORNL), and with support of the US Department of Energy (DOE) Consortium for Advanced Simulation of Light Water Reactors (CASL) as well as from the members of the CTF User’s Group. CTF uses a two-fluid, three-field representation of the two-phase flow, which makes it capable of modeling the high-void flow conditions expected in BWR operation. This paper focuses on applications of CTF to mini- and whole-core BWR calculations on assembly/channel and pin-cell/sub-channel resolved levels as well as on demonstrating that CTF can properly model bypass flow. To increase the confidence in CTF’s BWR modeling capabilities, simulations have been performed using the international Organization for Economic Cooperation and Development (OECD)/US Nuclear Regulatory Commission (NRC) Oskarshamn-2 benchmark, including modeling of a single assembly and a mini-core of 2 × 2 assemblies on a pin-by-pin/sub-channel level, and a full core model on an assembly/channel level. Each model is varied with an increasing amount of detail. Key parameters such as pressure losses and void fraction distribution were analyzed to determine the impact of different levels of detail within a thermal-hydraulic model on the simulation results. The results demonstrated that CTF is capable of modeling BWR core on different spatial resolution levels. The Oskarshamn-2 core simulations was used to further verify and demonstrate CTF’s capabilities of modeling BWRs.

Non-linear boiling water reactor stability with Shannon Entropy

The most common indicator currently used to assess boiling water reactor (BWR) instability is the Decay Ratio (DR), which is an indicator that assumes stationarity and linearity of BWR signals. However, it is well known that BWRs are complex dynamical systems that may exhibit chaotic behavior when an instability event occurs, jeopardizing the DR validity and reliability when the reactor is working at a specific operating point. Thus, it is required to study new stability indicators that satisfy as much as possible this complex dynamics of BWR systems. With this latter fundamental idea in mind, in this work, the non-linear Shannon Entropy (SE) is explored to study BWR instability. The SE measures the uncertainty of BWR signals to appraise for system stability, a low SE estimation indicates a predictable BWR operation (stable behavior) whereas a high SE estimation indicates an unpredictable BWR operation (unstable behavior). The SE estimation was validated with artificial signals from a non-linear Reduced Order Model (ROM), that represents qualitatively the dynamic behavior of a BWR system. The result comparison proves that the SE satisfies the BWR complex dynamics whereas the DR does not during the chaotic behavior. The SE was also compared with the Largest Lyapunov Exponent (LLE), which represents adequately the chaotic behavior of a BWR but, from the practical point of view, it cannot be applied to an online stability monitor, while the methodology presented in this work based on the SE, is a good candidate.

BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

After 40years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved.To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience.This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system.

Air–water downscaled experiments and three-dimensional two-phase flow simulations of improved steam separator for boiling water reactor

Reducing the pressure losses in steam-separator systems in boiling water reactor (BWR) plants is useful for reducing the required pump head and enhancing the design margins to ensure core stability. We need to reduce the pressure losses while maintaining the gas–liquid separation performance. In this study, we improve a steam separator with air–water downscaled experiments and two-phase flow simulations.First, we confirm the effectiveness for the separator performance prediction by adjusting the quality and the two-phase centrifugal force between the air–water downscaled experiments and the steam-water mockup tests, and we design the improved steam separator, which moves the swirl-vane section from diffuser section to the first-barrel section. From the air–water downscaled experiments, the improved separator can decrease pressure loss in the swirler more than 50% around the BWR normal operating conditions compared to the conventional separator, and the carryover of the improved separator is almost the same level as the conventional separator.Next, we evaluate the improved steam separator performance under the BWR operating conditions by means of a two-phase flow simulation, and we have the prospects of the improved separator for reducing the total separator pressure losses by about 30% compared to the conventional separator, while maintaining carryover characteristics.

Three-dimensional time-averaged void fraction distribution measurement technique for BWR thermal hydraulic conditions using an X-ray CT system

We have developed a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray computed tomography (CT) system to understand two-phase flow behavior inside a fuel bundle for boiling water reactor (BWR) thermal hydraulic conditions of 7.2 MPa and 288 °C. As a first step, we measured the 3D void fraction distribution in a vertical square (5 × 5) rod array that simulated a BWR fuel bundle in the air–water test. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer showed satisfactory agreement within a difference of 0.03. Thus, we confirmed that the developed system could be used to get 3D imaging of the vertical square rod array used in the test under the BWR operating pressure condition. In the next step, we did a verification test using the vertical pipe (11.3 mm ID) for BWR thermal hydraulic conditions. A comparison of the cross-sectional-averaged void fractions evaluated by the X-ray CT system with those evaluated by the drift-flux model showed good agreement within a difference of 0.05. We confirmed that the evaluated void fraction distribution forms in the horizontal cross section changed with the quality in response to the flow regime transition.

Preliminary test of an ultrasonic liquid film sensor for high-temperature steam–water two-phase flow experiments

A prototype liquid film sensor for high-temperature steam–water experiments has been developed. The sensor shape simulates a boiling water reactor (BWR) fuel rod. The pulse-echo method can be utilized to measure the thickness of the liquid film covering the sensor surface. A piezoelectric element is soldered onto the inside of the sensor casing which consists of two curved casing pieces. After the piezoelectric element is attached, the two casing pieces are laser welded together. It is confirmed that the temperature rise at the time of the laser welding does not influence soldering of the piezoelectric element. The pressure proof test shows that the sensor can be used at a high-pressure condition of 7 MPa. Simple air–water experiments are done at atmospheric pressure to confirm the liquid film thickness can be measured with the sensor. The fluctuation of the liquid film thickness is satisfactorily captured with the sensor. The minimum and maximum thicknesses are 0.084 and 0.180 mm, respectively. The amplitude of the waveform at 286 °C is predicted by the calculation based on the acoustic impedance. It is expected that the sensor is able to measure the liquid film thickness even at BWR operating conditions.

Air/steam flow and steam wetness dependence on acoustic resonance in safety relief valves

Many experimental studies related to the flow-induced acoustic resonance closed side branches have been reported. However, few studies have reported on the effects of air/steam flow and steam wetness dependence on fluctuating pressure amplitude. Therefore, we investigated the effect of air/steam flow and steam wetness dependence on fluctuating pressure amplitude by conducting a high temperature and high pressure tests at the Hitachi Utility Steam Test Leading Facility (HUSTLE). The test section consisted of a main pipe and a side branch. The side branch was mounted on the long straight main pipe. Fluctuating pressures at the end face of the side branches were measured. The following two results were obtained; the first is that the air/steam flow had little effect on the fluctuating pressure amplitude normalized by dynamic pressure and frequency normalized by the resonance frequency; the second is that under the acoustic resonance (St = 0.41) and non-resonance (St = 0.55) conditions, fluctuating pressure and frequency changed little with steam wetness. The steam wetness during the boiling water reactor operation was less than 0.1%; thus, there was no effect of steam wetness on the acoustic pressure amplitude and the frequency under this operating condition.

Effects of temperature on stress corrosion cracking behavior of stainless steel and outer oxide distribution in cracks due to exposure to high-temperature water containing hydrogen peroxide

Cracking growth tests were conducted in high-temperature water containing hydrogen peroxide (H2O2) at 561–423K to evaluate the effects of H2O2 on stress corrosion cracking (SCC) of stainless steel (SS) at temperature lower than the boiling water reactor (BWR) operating temperature. Small compact tension (CT) specimens were prepared from thermally sensitized type 304SS. Despite the observation of only a small portion intergranular SCC (IGSCC) near the side groove of the CT specimen at 561K in high-temperature water containing 100ppb H2O2, the IGSCC area expanded to the central region of the CT specimens at 423 and 453K. Effects of H2O2 on SCC appeared intensely at temperature lower than the BWR operating temperature because of a reduction in the thermal decomposition of H2O2. To estimate the environment in the cracks, outer oxide distribution on the fracture surface and the fatigue pre-crack were examined by laser Raman spectroscopy and thermal equilibrium calculation was performed.

A nodal formulation method relating assembly power and reactivity for a boiling water reactor

A second-order nodal method is derived for the one and a half-group version of neutron diffusion theory in X-Y geometry. Based on the linear reactivity model, two reactivity relations are developed to treat fresh and depleted loading batches as a way to account for the spatial variation of moderator properties due to the substantial fraction of steam void present as well as the effect of the control rod positioning. Based on these ideas a programme was developed to predict fuel batch burnup and fuel power sharing in a Boiling Water Reactor (BWR) operating cycle. The method was tested by assessing two consecutives current BWR operating fuel cycles.

Impact of Power Uprate on the Water Chemistry in the Primary Coolant Circuit of a Boiling Water Reactor Operating under a Fixed Core Flow Rate

The approach of power uprate has been adopted by the utilities of light water reactors over the past few decades in order to increase the power generation efficiency of a nuclear reactor. Upon a power uprate, the power density of a nuclear reactor would change immediately, followed by water chemistry variations due to the enhanced radiolysis of water in the core and near-core regions. For commercial boiling water reactors (BWRs), it is currently a common practice to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal hydrogen injection rate may require a proper adjustment after a power uprate is practiced in a BWR. A DEMACE computer code was used in the current study to investigate the impact of various power uprate levels on major radiolytic species concentrations and the electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of a domestic BWR operating under either normal water chemistry or HWC. The results of our analysis indicated that the chemical species concentrations and ECP did not vary monotonically with increases in reactor power level at a fixed feedwater hydrogen concentration. In particular, the upper plenum and upper downcomer regions exhibited uniquely higher ECPs at a 102% power level than at the other evaluated power levels. The impact of power uprate on the water chemistry in the primary coolant circuit of a BWR is expected to vary from location to location and eventually from plant to plant due to different degrees of radiolysis and physical dimensions.