Two-phase flow features inherent randomness and discontinuity when observed instantaneously and locally. Such phenomena require averaging over sufficient samples to achieve the deterministic and continuous description commonly adopted in engineering modeling. However, this sampling sufficiency is challenged when flow measurement faces demand in both spatial and temporal resolutions. Specifically, there may not be a window that is (1) wide enough to cover sufficient local samples of interfaces while also (2) narrow enough to resolve a fast transient of interest. Such a dilemma for measurements at one point in time necessitates multiple realizations and ensemble averaging, which are nominally achievable through costly repeating experiments. In practice, when data beyond steady states are of interest, a more affordable alternative is setting up periodic conditions and continuously measuring multiple oscillation cycles. This approach was effectively conducted in a past project where traversable four-sensor probes were able to quantify instability-induced flow oscillations. This paper revisits the data from this past practice and presents more details on uncertainty quantification. An example case study is included to demonstrate the procedure for quantifying random errors from sampling insufficiency, based on estimators whose validity is verifiable through a posterior convergence study. By sharing this experience, the authors intend to inspire their colleague experimentalists with interest in transient two-phase data featuring resolution and precision comparable to those achievable under steady states.

The primary function of the reactor cavity cooling system (RCCS) for the high-temperature gas-cooled reactor pebble module (HTR-PM) is to cool the reactor cavity and the reactor pressure vessel, especially under accident conditions. During the commissioning of the HTR-PM, the operational parameters of the RCCS were measured. The results showed that both the atmosphere temperature and the cold helium temperature can affect the operational parameters of the RCCS. The heat removal power of the RCCS increases with the increase in the cold helium temperature and decreases with the increase in the atmosphere temperature. Moreover, a thermohydraulic code has been developed to predict the operational parameters of the RCCS, which is characterized by two coupled natural circulations. Compared to the measurement results obtained during the commissioning of the HTR-PM, the calculation accuracy of this code has been verified, demonstrating a certain conservative margin. The calculation results show that the hot water temperature (109.5°C) remains below the saturation temperature (133.3°C) at an operational pressure of 0.3 MPa, even under the combined conditions of a pressurized loss-of-forced cooling accident, extreme atmosphere temperature, and the shutdown of one RCCS unit in the HTR-PM.

Dynamic probabilistic risk assessment (PRA) is a computational approach to risk assessment that offers certain advantages over the more conventional PRA methods. Advantages of dynamic PRA include automatic generation of potential accident scenarios and explicit consideration of complex system dynamics; however, dynamic PRA typically requires computationally costly simulations, which have been a significant barrier to its widespread adoption. To improve the computational performance of dynamic PRA, this paper proposes a novel algorithm for sampling-based dynamic PRA, the exploratory nuclear tree sampler (ENTS). ENTS utilizes importance sampling and Monte Carlo tree search to reduce the computational cost and increase the number of unique scenarios generated during a dynamic PRA. To demonstrate the computational improvements offered by ENTS, its performance is assessed using two case studies from the dynamic PRA literature. The computational performance of a comparable Monte Carlo method is assessed and used as a benchmark for comparison. This comparison shows that the ENTS approach requires fewer overall simulations to estimate the system risk and generates more unique accident scenarios than the Monte Carlo approach.

U-bend geometries are commonly used flow restrictions in nuclear reactor systems. Two-phase flows through U-bends are quite different from those in straight pipes, yet systematic modeling of inverted U-bend effects is lacking. A separate-effects air-water two-phase flow test facility has been used to study the two-phase flow transport from vertical upward to vertical downward across a vertical U-bend (25.4-mm inner diameter; curvature-to-diameter ratio = 9). Detailed data, including void fraction, gas velocity, bubble diameter, and pressure loss, were obtained. Using the obtained experimental data, models and correlations were developed to characterize the U-bend effects, which include models and correlations for variance of void fraction σ 2, U-bend dissipation length, bubble velocity, and pressure loss. The U-bend strength can be represented by the variance of the void fraction, which dissipates exponentially in the U-bend dissipation region. The dissipation lengths of U-bend effects under different test conditions are determined by the dissipation rate β . The bubble velocity correlations are related to the development of σ 2. The Lockhart-Martinelli’two-phase flow frictional loss correlation can be used to predict the experimental two-phase pressure drop across the U-bend with some modifications. Experimental data also suggest a strong correlation between σ 2 and bubble interaction covariance terms (covariance of random collision, CO V RC ) in the U-bend and U-bend dissipation region. A modified Froude number, F r m , derived from the two-fluid model momentum equation is used as a fundamental parameter in developing the correlations for σ 2, β , U-bend dissipation length, bubble velocity, and CO V RC . To model the void fractions in the interfacial area transport equation across the U-bend, a continuity equation was used, while conventional drift-flux models were used in the straight pipe sections. Model coefficients of different bubble interaction terms were determined by evaluating each region (i.e. vertical upward, U-bend, U-bend dissipation, vertical downward) using experimental data individually. The one-group interfacial area transport from vertical upward to vertical downward two-phase flow across a vertical U-bend is then evaluated using all the above developed models and correlations. The evaluation shows that the models predict a i development effectively, with deviations generally within ±15%.

RELAP5-3D has been used extensively for high-temperature gas-cooled reactor (HTGR) analysis, but such analysis lies outside the validation basis for the code. Oregon State University, the U.S. Nuclear Regulatory Commission, and the U.S. Department of Energy’s Advanced Reactor Technologies gas-cooled reactor campaign worked together to design and operate the High Temperature Test Facility (HTTF) at Oregon State University to provide validation data for prismatic HTGR thermal-hydraulic analysis. Work using a legacy RELAP5-3D model of the HTTF demonstrated the ability to reproduce trends in HTTF data, but an inability to reproduce the values measured in multiple experiments within the uncertainty. Previous work hypothesized that a new model with a finer radial nodalization would perform better for HTTF analysis. In this paper, we present the development of that more refined model and analyze HTTF experiments PG-27 and PG-29, pressurized and depressurized conduction cooldown transients, respectively. Comparing the PG-27 results to the experimental data showed a better prediction of steady-state and transient temperatures; however, a deeper dive revealed that the better transient temperature predictions are a result of better initial conditions. The transient temperature rise in the core is comparable between the two models, but the legacy model predicted temperatures in the reflectors better. The new model demonstrated the ability to predict unique temperatures in each sector of the core during PG-29, capturing the azimuthal asymmetry; however, the temperature predictions still differed from the measured temperatures. These results suggest that gaps and uncertainties in the measured boundary conditions for the HTTF tests do not allow for fully validating RELAP5-3D for prismatic HTGR analyses. We demonstrated improved validation for steady-state analysis, but significantly improved transient performance has yet to be recognized.

The nuclear field faces one of the most significant gender gaps among the science, technology, engineering, and mathematics fields. Using South Korea’s government-funded data (2007–2022) linked to Web of Science, we examine gendered patterns in project characteristics and their authorship in the nuclear field. Over this period, the number of female principal investigators (PIs) doubled, and those of female first authors and last authors tripled. Yet, there is a declining trend in the share of publications led by female PIs, implying a persistent departure of women researchers from the nuclear field. In addition, female PIs have fewer men on their teams than male PIs, receive less funding on average, lead fewer institutional collaborations, and conduct more basic-stage research. Our regression results also show that the last authorship indicative of seniority tends to be female when the PI is female as well. Our findings call for policies to support the retention of female researchers through improving gender balance across multiple dimensions of research while continuously recruiting new female researchers. With increasingly diverse applications of nuclear energy, a more gender-balanced nuclear research and development ecosystem will contribute to the greater roles of nuclear energy in addressing key challenges of our time.

The high-temperature gas-cooled reactor (HTGR) with a high outlet temperature from 700°C to 800°C to 1000°C is expected to be widely used for heat supply, hydrogen production, steelmaking, seawater desalination, thermal recovery of heavy oil, and coal liquefaction and gasification. The 10-MW High Temperature gas-cooled test Reactor (HTR-10), with an outlet temperature of 700°C, has been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. Supported by the Chinese National S&T Major Project and National Key R&D Program of China, some research on HTGR technology with much higher outlet temperatures has been carried out in the Institute of Nuclear and New Energy Technology (INET), Tsinghua University, China, including the preliminary design and analysis of a HTR-10 with an outlet helium temperature of 950°C (HTR-10H). This paper presents the simulation results of the depressurized loss-of-forced cooling (DLOFC) accident of the HTR-10H. This analysis was performed with two different codes, TINTE and SPECTRA, within a cooperation between the Nuclear Research Group of the Netherlands and the INET. In the past, comparisons between TINTE and SPECTRA have also been performed for the Chinese 200-MW(electric) High Temperature gas-cooled Reactor Pebble bed Module (HTR-PM). The calculation results showed that both the fuel temperature and the reactor pressure vessel temperature remained below the acceptable limits during the DLOFC accident. The inherent safety of the HTR-10 can be guaranteed in the accident condition. Both the TINTE results and SPECTRA results for the maximum and average fuel temperature agreed well. In both models, the heat carrying capacity of the Reactor cavity cooling system (RCCS) had been underestimated, especially in the SPECTRA model. The RCCS performance needs to be investigated and the models need to be further improved in the future. The work introduced in this paper shows the feasibility of increasing the outlet helium temperature of the HTR-10 from the current 700°C to 950°C, if further design optimization can effectively reduce the fuel temperature during steady-state operation, or more irradiation tests can be carried out to prove the retention performance of the TRISO particles for the radioactive fission products at temperatures above 1200°C.

Ensuring the efficient removal of decay heat from the reactor vessel is essential for the safety of advanced reactor technologies. Several Generation-IV concepts incorporate variations in the reactor vessel cooling systems to achieve this objective. High-temperature gas-cooled reactors utilize a reactor cavity cooling system (RCCS), a passive ex-vessel system designed to operate without active components or external power during accident conditions. The RCCS removes decay heat primarily through radiative and convective heat transfer mechanisms. This study presents a comprehensive validation of a computational fluid dynamics Reynolds-averaged Navier-Stokes model for the University of Wisconsin-Madison air-cooled RCCS facility. Validation was conducted for both high- and low-power natural convection cases under a uniform heating profile. Near-wall resolution was found to be critical for accurately modeling natural convection in the RCCS; employing an all- y + wall treatment resulted in wall temperature discrepancies exceeding 50 ∘ C compared to a wall-resolved mesh. Thermal-hydraulic behaviors under natural and forced convection conditions were compared within the heated cavity and RCCS. A turbulence model sensitivity analysis indicated that low-Reynolds number k-ɛ, k-ω shear stress transport (SST), and Reynolds stress transport models produce similar wall temperature predictions. A buoyancy modeling sensitivity study revealed that the Boussinesq approximation significantly underpredicted thermal-hydraulic behavior in the RCCS. Based on these findings, modeling recommendations are provided. The validated data set along with identified sensitivities refine the modeling of natural convection in the RCCS. The information produced by this study supports RCCS design, optimization, and safety evaluations, enabling the calibration and verification of reduced-order thermal-hydraulic models.

The purpose of this paper is to evaluate the dosimetric performance of two computed tomography (CT) scanners installed in 2015 and 2020 by comparing measured computed tomography dose index (CTDI1 0 0) values for brain and lung protocols. The study assesses measurement accuracy, angular dose stability, inter-scanner agreement, and compliance with national reference dose (NRD) limits under standardized conditions. Dosimetric measurements were performed using standardized 16-cm and 32-cm polymethyl methacrylate computed tomography dose index (CTDI) phantoms for the brain and lung protocols. For each scanner and protocol, CTDI1 0 0 was measured at five predefined positions using a calibrated pencil ionization chamber. Scanner-reported computed tomography dose index volume (CTDIvol) values were compared with CTDI1 0 0–derived CTDIvol values. Measurement precision was assessed using the standard error of the mean. Paired t-tests and Bland-Altman analysis were applied to evaluate statistical significance and agreement. The older scanner exhibited greater angular dose variability and a larger deviation between the measured and displayed CTDI values for the brain protocol (13.04%, p = .045). The newer scanner showed improved angular stability and closer agreement with the displayed values (1.22%). The lung protocol deviations were similar for both scanners (2.72%) with no significant difference (p = .45). The Bland-Altman analysis demonstrated tighter agreement for lung protocols and poorer agreement for the brain protocol on the older scanner. Lung protocols and brain measurements on the newer scanner were within NRD limits. In conclusion, the newer generation CT scanners provided more stable radiation output and better agreement between the measured and displayed CTDI values, particularly for high-dose protocols. Routine phantom-based quality assurance and independent dosimetric verification remain essential for reliable dose reporting.