Pressurized Water Reactor Conditions Research Articles

Interface stability of ultrasonic additively manufactured Zircaloy-4 during hydrothermal corrosion

Simulated pressurized water reactor conditions (330 °C, 15.6 MPa, ∼20 ppb oxygen) without irradiation were used to investigate the hydrothermal corrosion behavior of ultrasonic additively manufactured Zircaloy-4 up to 1000 h. X-ray computed tomography allowed for visualization of defects from processing and their progression after corrosion experiments. The specimens were found to have clear variability in the mass change data, compared to typical wrought Zircaloy-4 specimens. The variation in the mass change after exposure was attributed to weld defects connected to the specimen surface which allowed ingress of oxidant into the samples. Defects visualized by computed tomography were found via metallography and characterized. Ultrasonic additively manufactured Zircaloy-4 was found to have comparable corrosion behavior as wrought Zircaloy-4 for specimens which did not have clear surface defects along weld interfaces.

A Monte Carlo simulation study for designing collimators for a CZT-based spent nuclear fuel characterization system

As an attempt to develop a spent nuclear fuel characterization system that can acquire both high-resolution gamma-ray energy spectrum and neutron flux simultaneously at multiple positions around the spent nuclear fuel assembly and provide more information compared to conventional equipment, simulation study was conducted to design a collimator around a gamma-ray detector to obtain high-resolution gamma-ray spectra in a high-flux gamma-ray environment. The ORIGEN depletion code was used to evaluate the source terms of spent nuclear fuel, and MCNP 6.2, the Monte Carlo simulation code, was used to simulate the detector response. Considering the realistic operating conditions of Korea's pressurized light water reactor, the source term for the ACE7 17×17 nuclear fuel assembly was derived under the conditions of initial enrichment ranging from 3 to 5%, burnup ranging from 30 to 60 GWd/MTU, and maximum cooling time of 30 years. MCNP 6.2, the Monte Carlo simulation code, was used to design a spent nuclear fuel characterization system using a CZT detector as a gamma-ray detector to obtain a high-resolution gamma-ray energy spectrum. In particular, the research focused on designing collimators to be used in high-flux radiation environments, such as spent nuclear fuel inspection, with CZT detectors. A simulation model of the CZT detector to be installed internally was made and the validity of the detector simulation model was confirmed by comparing simulation results with the measurement experiment results. Finally, the response of the CZT detector was calculated by changing various setting parameters of the collimator. By comparing the calculated detector response with the design standard count rate, the design of a collimator that can utilize the CZT detector in a spent nuclear fuel inspection environment was derived.

Experimental investigation and prediction of CHF in wire-wrapped rod bundles

The wire-wrapped tight lattice fuel assembly is one of the options for small modular reactors. To clarify the characteristics of critical heat transfer for departure from nucleate boiling, experiments were conducted to investigate the critical heat flux in a wire-wrapped 19-rod bundle. The critical heat flux coincident with associated thermal–hydraulic parameters was obtained under typical pressurized water reactor conditions. Validation of data accuracy was performed to provide confidence in the experiment results. It was found that the critical heat flux in the wire-wrapped rod bundle is sensitive to the change in steam quality at some steam quality regions. Besides, the critical heat flux in wire-wrapped rod bundles has a similar reflection under the effects of local pressure, local mass flux, and steam quality. However, the current correlations tend to overestimate the critical heat flux for a wire-wrapped rod bundle. Hence, further development of KfK-3 correlations was proposed for better prediction of the critical heat flux in wire-wrapped bundles. This work is helpful for rod bundle designs of advanced pressurized water reactors.

Numerical investigation on boiling crisis in a full-length 5 × 5 fuel assembly under typical pressurized water reactor conditions

Critical heat flux (CHF) presents a research priority in the Pressurized Water Reactor (PWR) safety, as it can lead to the severe accidents and release radioactive materials. In this study, a wall heat flux partition model, proposed by Demarly [1], was employed within the two-fluid model by Computational Fluid Dynamics (CFD) methodology to investigate the characteristics of CHF under typical PWR conditions. A multi-scale interface model was coupled to consider the interfacial transfers in bulk flow during flow regime transition. The models were validated by using experimental data on subcooled flow boiling, boiling crisis in vertical tubes, and rod bundles. The PWR sub-channel and bundle test (PSBT) benchmark, comprising a full-length 5 × 5 fuel assembly with three different types of spacers, was simulated in this study. The CHF value was evaluated, demonstrating an acceptable accuracy compared to the experimental data, with a deviation of less than 10 % in the selected conditions. The CHF location in the cross section was successfully predicted by CFD approach, while the axial location of CHF was underestimated than experimental measurement. Besides, the impacts of mixing vanes with different deflection angles on boiling heat transfer characteristics were analyzed, revealing the crucial role of mixing vanes in the distribution of vapor and dry-patch area fraction in both the circumferential and flow directions.

Decreasing Hydrogen Content within Zirconium Using Au and Pd Nanoparticles as Sacrificial Agents under Pressurized Water at High Temperature.

Decreasing hydride-induced embrittlement of zirconium-based cladding is a significant challenge for the successful dry storage of spent nuclear fuel. Herein, to radically minimize hydride-induced embrittlement, we used nanoparticles as sacrificial agents with a greater affinity than zirconium for hydrogen. Corrosion experiments in the presence of gold (Au) and palladium (Pd) nanoparticles under simulated pressurized water reactor (PWR) conditions revealed that the hydrogen content of the zirconium samples was remarkably reduced, with a maximum decrease efficiency of 53.9% using 65 nm Au and 53.8% using 50 nm Pd nanoparticles. This approach provides an effective strategy for preventing hydride-induced embrittlement of zirconium-based cladding.

Electrochemical and modelling study of ZrNbO alloys aged under high temperature and high pressure PWR simulated conditions

This work proposes a kinetic study of Zirconium alloy when working under controlled temperature and pressure conditions to simulate pressurized water reactor (PWR) conditions. A detailed analysis of the experimental results allowed to describe each time constant observed on the impedance diagrams. A model based on the PDM including the anodic and cathodic reactions and the description of the interfacial capacitance is proposed and allows to fit the experimental results. Even if the overall oxide film formed on the alloy can reach several micrometers in thickness, only a compact oxide sub-layer of a few nanometers contributes to the impedance response.

The DISNY facility for sub-cooled flow boiling performance analysis of CRUD deposited zirconium alloy cladding under pressurized water reactor condition: Design, construction, and operation

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

Statistical determination of stress threshold and survival analysis on IASCC initiation data of stainless steels in pressurized water reactor conditions

This paper presents the first attempt to perform a rigorous statistical analysis of the compiled IASCC initiation data of 304 and 316 stainless steels available in the existing literature. First, the stress threshold for IASCC initiation was determined statistically as a function of the neutron dose. The 95% lower bound or 2.5% cumulative failure probability line appeared to be a conservative stress threshold, although it could be adjusted according to the required conservatism. Survival analysis was conducted using the filtered data set for 316 stainless steel IASCC initiation times. A proportional hazards model revealed statistically significant covariates in order of greatest impact: stress level as a percentage of irradiated yield strength, neutron dose, and component. Unsurprisingly, the failure rate was found to increase with increasing stress level and neutron dose. It was also found that FTT materials had higher failure rates than BFB materials. The effects of these covariates on the failure rate were constant with time. An accelerated life model based on a Weibull distribution was then used to model the IASCC initiation time. Two different link functions were used for the model: the link function with a power-law form for the stress level and neutron dose fitted the observed data better than a log-linear type link function.

Effect of surface machining on the environmentally-assisted cracking of Alloy 182 and 316L stainless steel in light water reactor environments: results of the collaborative project MEACTOS

Abstract The main objective of the EU-funded project mitigating environmentally-assisted cracking through optimisation of surface condition (MEACTOS) was to gain knowledge on the ability of different surface machining procedures to mitigate environmentally-assisted cracking (EAC) in typical light water reactor structural materials and environments. Surfaces of cold-worked (CW) type 316L austenitic stainless steel and nickel-based weld metal Alloy 182 flat tapered tensile specimens were machined using different processes. EAC initiation susceptibility of these specimens was evaluated using constant extension rate tensile (CERT) tests under simulated boiling water reactor (BWR) and pressurized water reactor (PWR) conditions and assessed using constant load experiments. More than a hundred tests were performed covering about 10 years of autoclave testing time. Only minor or no measurable improvements in EAC initiation susceptibility as a function of surface treatments (grinding or advanced machining) compared to the standard industrial face milling were demonstrated. In most cases, the stress thresholds for EAC initiation determined in constant load tests confirmed the trend obtained from CERT tests. This paper summarises the most important results and conclusions concerning the EAC initiation behaviour for the CW 316L and Alloy 182 under reducing PWR and oxidizing BWR conditions.

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). To further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

Effect of zinc injection on mitigating stress corrosion cracking initiation of structural materials in light water reactor primary water

Abstract The effect of zinc (Zn) injection on the stress corrosion cracking (SCC) initiation of an Alloy 182 weld metal and a 20% cold-worked (CW) type 316L stainless steel was investigated under simulated light water reactor conditions. Accelerated SCC initiation tests revealed that under both, boiling (BWR) and pressurized water reactor (PWR) conditions without Zn injection, Alloy 182 showed a higher SCC initiation susceptibility compared to the CW 316L stainless steel. With 40 ppb Zn injection, results indicate a tendency towards higher stress thresholds or lower crack densities for SCC initiation in both materials. In other words, Zn water chemistry seems to be able to mitigate SCC initiation. The optimized, Zn-treated oxide film is believed to be responsible for the improved SCC performance. However, no clear mitigation effects were observed when the specimens were pre-exposed to high-temperature water without Zn and then tested with Zn, which is likely related to the relatively short exposure time to the Zn-containing high-temperature water.

Using Direct Current Potential Drop Technique to Estimate Fatigue Crack Growth Rates in Solid Bar Specimens under Environmental Assisted Fatigue in Simulated Pressurized Water Reactor Conditions

The direct current potential drop (DCPD) technique may be used in crack propagation tests to measure the crack growth rate (CGR). Potential probes attached to the specimen allow the variation of the crack length to be estimated. In this research, the DCPD technique using one single potential probe was applied to solid bar specimens (i.e., without any initial notch or crack) subjected to low-cycle fatigue testing in a simulated pressurized water reactor (PWR) environment. This particular analysis had two associated difficulties, the first one being the fact that crack initiation sites are not known beforehand, and the second one consisting in the experimental difficulties and conditioning factors associated with the simulation of the PWR environment. Nine solid bar specimens were tested to fatigue failure under different strain amplitudes and frequencies, while also measuring the corresponding DCPD signal during the fatigue process. It was observed that the initiation of multiple cracks was detected by the DCPD measurements. Moreover, as fatigue continued, one of the cracks became dominant and progressed to cause the specimen failure. The DCPD technique allowed the average CGR of the dominant crack to be estimated. Finally, the obtained average CGRs were validated by comparing them with average CGRs derived from striation spacing measurements, obtained from scanning electron microscopy (SEM) and from literature values gathered in the NUREG/CR-6909 document.

Numerical analysis of the subcooled boiling phenomena in a wire-wrapped annulus

The wire-wrapped tight lattice fuel assembly is one of the options for small modular reactors. To investigate the performance of this assembly, the subcooled boiling characteristics in a wire-wrapped annular have been analyzed in the present work. The Eulerian two-fluid model coupled with the RPI wall boiling model was employed in the simulation. The mathematical models are validated by the subcooled boiling experimental data of the circular channel first, and then the simulation for wire-wrapped annular is carried out. The distributions of fluid temperature, wall temperature, vapor volume fraction, and heat flux partitions under a typical pressurized water reactor condition are gained. In the simulation conditions of the current wire-wrapped annular, the results indicate the non-uniform circumferential heat transfer can be ignored in subcooled boiling conditions, and the maximum vapor volume fraction occurred near the wire surface region. In addition, the increase of wire pitch increases the swirl intensity, at the same time, it also will increase the maximum vapor volume fraction near the rod surface region, which may cause DNB to occur earlier. Therefore, the wire pitch should be chosen carefully when the wire-wrapped assembly is applied to the pressurized water reactors. This work can provide a reference for the design of wire-wrapped tight lattice assembly applied in small modular reactors.

Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

We present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ∼15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.

Hydrothermal corrosion and steam oxidation behavior comparison of UAM and conventional Zry-4

Hydrothermal corrosion and steam exposures on ultrasonic additively manufactured (UAM) Zircaloy-4 (Zry-4) specimens were conducted to compare to wrought Zry-4 performance in both simulated pressurized water reactor (PWR) operation (330°C, 15 MPa, <50 ppb oxygen, 500h) and representative accident conditions (100% H2O(g) at 700, 900, and 1100°C for up to 1h). UAM Zry-4 specimens showed 50% higher mass gain in PWR conditions and 50-80% more mass gain in steam oxidation due to internal oxidation between build layers in both cases. UAM Zry-4 in 900°C steam showed breakaway oxidation leading to sample deformation. Titanium used as an interlayer in the UAM process was detected in the substrate after the 1100°C exposure. Additional experiments to better link as-built defects to oxidation behavior are needed.

Signal-to-Noise Ratio of CABRI Hodoscope: Monte Carlo Calculation Versus Experiments

The CABRI experimental pulse reactor is devoted to the study of reactivity-initiated accidents (RIAs), for the purpose of the CABRI International Program (CIP), managed by the French Safety and Radiation Protection Institute (IRSN). CABRI’s hodoscope equipment detects the fast neutrons emitted during a power pulse by a tested rod, positioned inside a dedicated test loop reproducing either sodium reactor or pressurized water reactor (PWR) conditions. One of the most important parameters measured by the hodoscope detectors is the signal-to-noise ratio (SNR), characterizing the fraction of neutrons directly coming from the test rod (“signal”) over neutrons coming from the core (“noise”). In this article, the method used to calculate the SNR using a 2-D model of CABRI, with the MCNP6.2 Monte Carlo code, will be detailed. Comparisons between the calculated and measured SNRs for different configurations are in quite good agreement. Another parameter of interest is the so-called “scattering coefficient,” which corresponds to the fraction of neutrons coming from the test rod and being scattered between their birth and their detection. This parameter is used to enhance the analysis of the fuel displacement that may happen during the power transient. To estimate this coefficient, an innovative method using a combination of different options available in MCNP6.2 has been used. Computed coefficients show a slight discrepancy with the measurements. Finally, the sensitivity of the SNR and scattering coefficient to technological parameters and nuclear data libraries is discussed.

Critical heat flux characteristics of flow boiling on a heater rod under inclined and rolling conditions

Offshore floating nuclear power plants have been developed to supply power to remote regions. Since they are operated in an ocean environment, the thermal–hydraulic phenomena in the reactor core may be different from those in the land-based reactors. However, the experimental study on critical heat flux (CHF) under ocean condition is limited and its mechanism is unclear. In this study, an experiment was performed to measure CHF in a single heater rod under inclined and rolling conditions. To create the required environment to simulate the conditions of inclination and rolling, a rolling platform was designed and constructed. It could provide an inclination of up to 45° and rolling of up to 45° with a minimum period of 6 s. In the CHF test loop, the experimental conditions were set up to correspond to pressurized water reactor conditions, using the fluid-to-fluid scaling method. The working fluid used was R134a. The test section was a concentric annular channel composed of a heater rod and an unheated tube housing. The result of the inclination experiment suggested that in the most cases, the CHF was enhanced when compared with the vertical condition. In a limited number of cases of low-pressure and high-flow conditions, the CHF was degraded as compared to the vertical condition. These variations were proportional to the inclination angle and inversely proportional to the mass flux. From the results of the rolling experiment, a precursor of CHF was observed with a large fluctuation in temperature before the CHF, and its trigger was related to the CHF value of the inclined case. Based on this observation, the overall relations between the vertical, inclined, and rolling CHF values were derived. It was concluded that the rolling motion weakened the CHF enhancement and degradation effects of the inclination.

Effect of TiN Coating on the Fouling Behavior of Crud on Pressurized Water Reactor Fuel Cladding

A titanium nitride film was applied to mitigate the deposition of fouling deposits (crud) on pressurized water reactor fuel cladding. Titanium nitride was chosen as it reduces the van der Waals force, believed to be the dominant surface force at pressurized water reactor conditions, between crud and the substrate. Crud deposition experiments were conducted in a heated, flowing pressurized water reactor-simulant loop with uncoated and titanium nitride-coated Zr-Nb-Sn alloy tubes for 1–2 weeks. Crud morphology and thickness were investigated using focused ion beam–scanning electron microscopy, and crud chemical composition was measured via energy-dispersive X-ray spectroscopy. Results demonstrate that the titanium nitride coating effectively reduced crud deposition thickness compared to uncoated Zr-Nb-Sn alloy in the 2-week experiment. In addition, the titanium nitride coating increased the Ni/Fe ratio of the crud, which may help mitigate crud-induced power shifts.

Analytical investigation on rod bundle CHF-regime criterion based on dimensionless groups

The accurate prediction of rod bundle critical heat flux (CHF) is an important task for thermal-hydraulic design of pressurized water reactor (PWR) core. Under normal and transient conditions of PWR, based on the two-phase flow pattern, there are different CHF regimes. In order to correctly distinguish the CHF regimes in rod bundle, it is necessary to analytical investigate rod bundle CHF-regime criterion. In this study, a general dimensionless rod bundle CHF correlation is presented based on dimensional analysis method. The real vapor quality, Katto number and boiling length are introduced to establish the basic form of the dimensionless CHF correlation. According to the deep analysis of the distribution trend of typical CHF experimental data and the CHF characteristics of different CHF regimes, the CHF in rod bundle are separated into DO, DNB and HP-DNB regimes. Concise dimensionless CHF correlations corresponding to each regime are developed, and the dimensionless equations between the boundaries of each regime are established. Also, a clearly defined map of characteristic regimes of CHF in rod bundle is determined. This study proposes a classification criterion for rod bundle for rod bundle CHF regimes, and it is very important for the upcoming development of rod bundle CHF mechanistic model.

Linear stability analysis of RELAP5 two-fluid model in nuclear reactor safety results

System thermal-hydraulic code RELAP5 is based on a two-fluid, non-equilibrium, and non-homogeneous hydrodynamic model for simulation of transient two-phase behavior. The code model includes six governing equations to incorporate the mass, energy, and momentum of the two fluids. In this paper, linear stability analysis is performed to check the ill-posedness of the RELAP5 specific two-fluid model (TFM) for all normal and accident conditions of a standard pressurized water reactor (PWR). The analysis gives information about the soundness of the model and identifies the range of parameters where the solutions obtained from the model will be numerically convergent. The linear two-phase fluid dynamic stability (by dispersion analysis) of the RELAP5 one-dimensional two-fluid model and numerical stability of the difference equation formulation (by Von Neumann method) is presented. The present analysis shows that the two-fluid model becomes ill-posed for some fluid conditions, where the results are less accurate, so sensitivity analysis plays an important role. The ill-posed nature implies that results thus obtained (by finite difference method) have to be interpreted carefully because of the sensitive nature of reactor safety analysis. It is also identified that the variation in various parameters (like slip ratio, system pressure, void fraction, and phasic velocities) can affect the error growth rates. It has been demonstrated that the basic system of one-dimensional two-phase flow equations, that possesses complex characteristics, exhibits unbounded instabilities in the short-wavelength limit and constitutes an improperly posed initial value problem. The semi-implicit numerical method, which is unconditionally stable for hyperbolic systems, becomes unstable for non-hyperbolic systems. For some of the fluid conditions, even after the introduction of artificial viscosity terms (in the difference equation formulation) that damp the high-frequency spatial components of the solution, are not sufficient for regularization of the two-fluid model. Thus, there is a need for the addition of newer terms, e.g. bubble collision, so that the existing incomplete model provides better results. It is also demonstrated that the basic TFM of RELAP5 with additional collision term makes the system unconditionally well-posed which are originally conditionally well-posed. Excellent agreement is obtained between the predicted and computed growth rates of harmonic disturbances. It is found that the instability associated with the two-fluid model discretized system of equations is related to the instability associated with the ill-posedness of the two-fluid model but is quantitatively different.