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Abstract Background The optimal treatment approach for low‐risk prostate cancer (LRPC) remains controversial. While active surveillance is an increasingly popular option, definitive local treatments, including radical prostatectomy (RP), external beam radiotherapy (EBRT), and prostate seed implantation (PSI), are also commonly used. This study aimed to evaluate the survival outcomes of patients with LRPC using a large patient population from the National Cancer Database (NCDB). Methods We analyzed data from 195,452 patients diagnosed with LRPC between 2004 and 2015 using the NCDB. Patients were classified based on their treatment modalities, including RP, EBRT, PSI, or no local treatment (NLT). Only patients with Charlson–Deyo comorbidity scores of 0 or 1 were included to ensure comparability. Propensity score analysis was used to balance the treatment groups, and the accelerated failure time model was used to analyze the survival rates of the treatment groups. Results After a median follow‐up of 70.8 months, 24,545 deaths occurred, resulting in an all‐cause mortality rate of 13%. RP demonstrated a survival benefit compared with NLT, particularly in patients younger than 74 years of age. In contrast, radiation treatments (EBRT and PSI) did not improve survival in the younger age groups, except for patients older than 70 years for EBRT and older than 65 years for PSI. Notably, EBRT in patients younger than 65 years was associated with inferior outcomes. Conclusion This study highlights the differences in survival outcomes among LRPC treatment modalities. RP was associated with improved survival compared to NLT, especially in younger patients. In contrast, EBRT and PSI showed survival benefits primarily in the older age groups. NLT is a reasonable choice, particularly in younger patients when RP is not chosen. These findings emphasize the importance of individualized treatment decisions for LRPC management.

Abstract We report an improved calculation of the electron backscattering coefficients (BSCs) for beryllium, molybdenum and tungsten at electron energies of 0.1–100 keV based on an up-to-date Monte Carlo simulation method with different input of energy loss function (ELF) data. The electron inelastic cross-section is derived from the relativistic dielectric functional formalism, where the full Penn’s algorithm is applied for the extension of the ELF from the optical limit of $$q \to 0$$ q → 0 into the $$\left( {q,\omega } \right)$$ q , ω -plane. We have found that the accuracy of energy loss function may affect largely the calculated BSC. We also show that this has close relationship with the f- and ps-sum rules.

ZrC dispersion-strengthened W exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. In this study, surface modification of 0.5 wt.% ZrC dispersion-strengthened W (WZrC) under low energy and high fluence He plasma irradiation at high temperature was presented. Under the energy of 90 eV and fluence ranging from 6 × 10 ^24 He·m ^−2 –2 × 10 ^26 He·m ^−2 He irradiation at 920 °C, a typical fuzz nanostructure appeared on the W matrix of WZrC. The thickness of fuzz layer is proportional to the square root of He irradiation fluence. The fuzz showed comparable thickness and structural features to pure W, indicating limited effects of the particle’s addition on resistance to high fluence He irradiation at high temperatures. Under continuous He injection, the fuzz would grow extending onto the particle area, making the particle obscured. Besides, the erosion behavior of particles under He plasma irradiation has been investigated, which is thought to be dominated by a sputtering process. Under the He influence of 6 × 10 ^24 He·m ^−2 , only nanopores were observed in the surface region. With fluence increasing to 5 × 10 ^25 He·m ^−2 , the surface became relatively uneven with larger holes. W aggregated in spots and distributed on the surface of the particle, which might be the result of subthreshold sputtering and deposition. When fluence further increased to 2 × 10 ^26 He·m ^−2 , the particles were eroded completely and covered by the extended fuzz, forming cavities. In addition, distinctive layered nanotendrils were observed above the cavities, which were characterized to consist of inner W-riched skeletons and outer Zr-riched layers. It indicates that the layered nanotendrils should be the result of fuzz extension combined with particle sputtering/deposition.

This study explores a hypothetical scenario where low-activation refractory multi-component alloys (RMCAs) are considered for use as divertor target materials in fusion reactors. To investigate the surface modifications under divertor service conditions, a multi-phase W–Ta–V–Cr RMCA is irradiated with low-energy helium (He) plasma at varying temperatures to approximately 5.0 × 10 ^25 He·m ^−2 . The W-rich and Cr-rich phases in the multi-phase RMCA mimic the mono-phase W–Ta–V–Cr RMCA and segregation at grain boundaries, respectively. Following irradiation, fuzz layer formation is observed for all phases at temperatures lower than pure W requires. Additionally, nano He bubbles are identified in the fuzz tendrils at 920 °C. The modified layers exhibit reduced V and Cr content with increasing temperature, transitioning into W-Ta binary alloys at 920 °C. Notably, the fuzz layer on the W-rich phase is thinner than that on pure W at 920 °C. While a thinner fuzz layer suggests the alloy’s potential as a divertor target material, challenges include a lower fuzz formation temperature and potential high V and Cr sputtering yield, raising concerns for plasma contamination during fusion reactor operation. This dual perspective highlights both the promise and challenges of utilizing RMCAs as divertor target materials in severe fusion reactor environments.

The heat flux width ( λ _q ) is a key parameter determining the heat load at divertor targets. In recent years, drifts have been found to play a remarkable role in the edge plasma transport, while its influence on λ _q has not been well understood. Investigations of the influence of drifts on λ _q , systematic simulations using the SOLPS-ITER code are performed in this work. The statistics of the simulation results show that the drift under favorable/unfavorable B _T tends to increase the λ _q in the outer/inner side and decrease the λ _q in the other side, which is consistent with the experiment observations. At the upstream and the target, the mechanisms of the influence of the drifts on λ _q are different. The upstream λ _q ( λ _q _,u ) is directly affected by the drift-induced convective heat flux, while λ _q at the target ( λ _q _,t ) is indirectly influenced through heat conduction (in the high-recycling regime) and the sheath (in the detached regime) due to the change of plasma parameters there. Furthermore, the synergetic effect of geometry and drift under favorable B _T leads to an anomalously large λ _q _,t in the inner side at high density.

Purpose: Surgery is the standard-of-care treatment in patients with localized renal cell carcinoma (RCC), offering excellent chance of cure. However, there is a subset of patients who are ineligible for surgery and instead manage with ablative therapies, such as stereotactic ablative body radiation therapy (SABR). We used the National Cancer Database to examine trends in the use of SABR for inoperable RCC and identify any predictors of outcome. Methods and Materials: We queried the National Cancer Database for patients with unresected RCC between 2004 and 2016 who were treated with SABR. Kaplan–Meier analyses were used to determine overall survival. Multivariable Cox regression was used to identify predictors of survival. Results: We identified 347 patients meeting eligibility criteria. Median age was 74, and the majority of patients were clinical stage T1-2 (80%) and N0 (97%). The median tumor size was 3.8 cm (interquartile range [IQR], 2.8-5.2 cm). Six percent of patients received systemic therapy. The median dose of SABR was 45 Gy (IQR, 35-54 Gy) in 3 fractions (IQR, 1-5 fractions). The median follow-up was 36 months (IQR, 1-156 months). Predictors of decreased survival were age >74, larger tumors, and N1 or M1 disease. Median survival across the entire cohort was 58 months. Median survival was 92 months, 88 months, 44 months, and 26 months for primary tumors ≤2.5 cm, 2.6-3.5 cm, 3.5-5.0 cm, and >5.0 cm, respectively (P

Tungsten (W) is used as the plasma-facing material in the divertor region of future fusion reactors, such as ITER; however, its concentration in the core plasma must be maintained at an extremely low level. W transport in the scrape-off layer (SOL), which is related to the source of core W contamination, has been extensively studied. In this study, the influence of E × B drift on the transport of W impurity in the SOL is studied via numerical simulations of a model case based on EAST upper single-null configuration with high recycling divertor plasma. W transport is simulated using DIVIMP on the background plasma obtained from scape-off layer plasma simulation-ITER simulation including drifts. The E × B drift of W ions is introduced based on the background electric field. Therefore, both the direct E × B drift effect of W ion and the indirect effect via background plasma on W transport in the SOL are studied. The influence on the flux of W impurities entering confined plasma across the last closed flux surface Γ _enter is focused on, which is expected to be proportional to the core W concentration. Results reveal that Γ _enter is mainly from the outer (inner) target under a favorable (unfavorable) toroidal field B _T and can be increased by more than one order of magnitude compared with the case without drifts; this reflects the significant effect of E × B drift. The effects due to the background plasma and the poloidal and radial E × B drift of W ion, as well as the related mechanisms, are analyzed in detail for three stages of W transport in the SOL: effective sputtering from the target, leakage from the divertor, and entry into the confined plasma.

We present an up-to-date Monte Carlo simulation of electron backscattering coefficient of beryllium, which is an important material in fusion reactor, at an impact energy range of electrons between 100 eV and 100 keV. The Mott’s cross section calculated with more accurate scattering potential and a relativistic dielectric functional formalism with full Penn algorithm and experimental optical data are used in the modelling of electron elastic and inelastic scatterings, respectively. This Monte Carlo simulation modelling enabled us to derive comprehensive theoretical values of backscattering coefficient, which are found significantly smaller than the previous published experimental data particularly at energies below 10 keV. Simulation of electron backscattering spectra was then carried out with and without Auger electron emission, which confirmed that Auger electrons have negligible contribution to backscattering coefficient. To validate our simulation results, we have also performed calculation for amorphous boron and carbon, while the comparison with the available experimental data shows reasonable agreement. Further simulation for carbon and water covered Be sample has revealed that the surface contamination in low vacuum conditions with several atomic/molecular layers can drastically alter the measurement values for beryllium at low energies. The low backscattering coefficient values of beryllium are partly attributed to the extremely strong forward elastic scattering for low atomic number element. Therefore, we recommend to use the here reported backscattering coefficient data of beryllium for applications.

An integration design platform is under development for the design of the China Fusion Engineering Test Reactor (CFETR). It mainly includes the integration physical design platform and the integration engineering design platform. The integration engineering design platform aims at performing detailed engineering design for each tokamak component (e.g., breeding blanket, divertor, and vacuum vessel). The vacuum vessel design and analysis module is a part of the integration engineering design platform. The main idea of this module is to integrate the popular CAD/CAE software to form a consistent development environment. Specifically, the software OPTIMUS provides the approach to integrate the CAD/CAE software such as CATIA and ANSYS and form a design/analysis workflow for the vacuum vessel module. This design/analysis workflow could automate the process of modeling and finite element (FE) analysis for vacuum vessel. Functions such as sensitivity analysis and optimization of geometric parameters have been provided based on the design/analysis workflow. In addition, data from the model and FE analysis could be easily exchanged among different modules by providing a unifying data structure to maintain the consistency of the global design. This paper describes the strategy and methodology of the workflow in the vacuum vessel module. An example is given as a test of the workflow and functions of the vacuum vessel module. The results indicate that the module is a feasible framework for future application.

Theoretical analysis of dimensional measurements made using a scanning electron microscope (SEM) requires a raw image to be corrected for the probe shape of the incident electron beam. In this paper, an effective electron beam shape (EEBS), which differs from the Gaussian profile conventionally assumed in SEM image analysis, was studied using a Monte Carlo simulation method. A model of a focusing electron beam with finite width due to aberration was used in simulating the SEM image of gold particles on a carbon substrate. It was found that the EEBS is local to the electron beam landing position and deviates significantly from the Gaussian profile; it is strongly dependent on both the sample topography and the electron beam focusing condition. [ABSTRACT FROM AUTHOR]

One of the key issues to achieve the phase I mission of China fusion engineering test reactor (CFETR) is steady-state operation with a modest fusion power up to 200 MW. To study this feasibility, time-dependent modeling of CFETR baseline scenario is performed. The time-dependent scenario design examines actual discharge features, including auxiliary heating and current drive scheme, equilibrium, transport, PF coil systems and plasma control system during the plasma evolution. During plasma current ramp-up, 10 MW of EC or LH heating is used to effectively save volt–second consumption to a reasonable level within magnet coil capability and to reduce li for better plasma control. Three reference scenarios are designed to target full non-inductive operation with different external heating and current drive (H&CD), which are NB  +  EC, NB  +  LH, and LH  +  EC, respectively. Two scenarios with neutral beams reach the required fusion power of CFETR phase I target, while for LH  +  EC scenario the fusion power and confinement is somewhat lower. A reversed magnetic shear with is sustained with off-axis current drive and bootstrap current for all cases examined. Ideal MHD analysis shows that both large toroidal mode number n ballooning mode and low n kink mode are stable in the three reference cases independent of whether the current is totally relaxed or not. The L mode and H–L back transition ramp-down are designed to successfully reduce the plasma current to 2 MA and keep li below 2 with ramp-down rate of 80 kA s−1. The H–L back transition ramp-down yields improvements in the power transfer and coil current evolution as well as the li evolution. The heating and current drive characteristics of three auxiliary heating sources are indicated by scanning studies. A 2D scan of NB energy and tangency radius indicates the features of NB power deposition, current drive, torque and shine through issue. Both ramp-up and flattop are considered in the scanning of radio frequency parameters. For EC waves, the ramp-up phase constrains the parameter space more. The 2D scan of EC launch angle shows that EC waves launched above the midplane yield a better CD efficiency than using a midplane launcher. Specifically, 10 MW of EC waves launched from LFS above the midplane yield a maximum on axis driven current of about 0.55 MA, while waves launched from top yield a maximum near edge driven current of about 0.45 MA in the region of . Scanning of in LH spectrum shows that should be set around 1.7 to get enough penetration (inside ) for current drive and avoid strong accessibility problem. It is found that LH waves launched from HFS drives more current than that launched from LFS when is small. The maximum LH driven current reaches 1.1 MA with 10 MW of LH power launched from HFS. [ABSTRACT FROM AUTHOR]

Impurity effects on fusion performance of China fusion engineering test reactor (CFETR) due to extrinsic seeding are investigated. An integrated 1.5D modeling workflow evolves plasma equilibrium and all transport channels to steady state. The one modeling framework for integrated tasks framework is used to couple the transport solver, MHD equilibrium solver, and source and sink calculations. A self-consistent impurity profile constructed using a steady-state background plasma, which satisfies quasi-neutrality and true steady state, is presented for the first time. Studies are performed based on an optimized fully non-inductive scenario with varying concentrations of Argon (Ar) seeding. It is found that fusion performance improves before dropping off with increasing , while the confinement remains at high level. Further analysis of transport for these plasmas shows that low-k ion temperature gradient modes dominate the turbulence. The decrease in linear growth rate and resultant fluxes of all channels with increasing can be traced to impurity profile change by transport. The improvement in confinement levels off at higher . Over the regime of study there is a competition between the suppressed transport and increasing radiation that leads to a peak in the fusion performance at (~2.78 for CFETR). Extrinsic impurity seeding to control divertor heat load will need to be optimized around this value for best fusion performance. [ABSTRACT FROM AUTHOR]

The confidence in ramp up scenario design of the China fusion engineering test reactor (CFETR) can be significantly enhanced using validated transport models to predict the current profile and temperature profile. In the tokamak simulation code (TSC), two semi-empirical energy transport models (the Coppi-Tang (CT) and BGB model) and three theory-based models (the GLF23, MMM95 and CDBM model) are investigated on the CFETR relevant ramp up discharges, including three DIII-D ITER-like ramp up discharges and one EAST ohmic discharge. For the DIII-D discharges, all the transport models yield dynamic within 0.15 deviations except for some time points where the experimental fluctuation is very strong. All the models agree with the experimental except that the CT model strongly overestimates in the first half of ramp up phase. When applying the CT, CDBM and GLF23 model to estimate the internal flux, they show maximum deviations of more than 10% because of inaccuracies in the temperature profile predictions, while the BGB model performs best on the internal flux. Although all the models fall short in reproducing the dynamic evolution for the EAST tokamak, the result of the BGB model is the closest to the experimental . Based on these comparisons, we conclude that the BGB model is the most consistent among these models for simulating CFETR ohmic ramp-up. The CT model with improvement for better simulation of the temperature profiles in the first half of ramp up phase will also be attractive. For the MMM95, GLF23 and CDBM model, better prediction of the edge temperature will improve the confidence for CFETR L-mode simulation. Conclusive validation of any transport model will require extensive future investigation covering a larger variety discharges. [ABSTRACT FROM AUTHOR]