The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study ...is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium fuel system (Zr–U) is currently used as a standard fuel system in pressurized water reactors around the world. This investigation presents the benefits of utilizing an alternative cladding material such as chromium coating and the effects on the thermal neutron parameters of the way in which the chromium coating is introduced in the reactor fuel. Among these significant benefits is an increase in the reactor fuel’s thermal conductivity, which improves reactor safety. Two types of fuel-cladding systems were investigated: Zircaloy–uranium (Zr–U) and Zircaloy–chromium (Zr–Cr–U) coating fuel systems. Neutronics analysis evaluations were performed for the selected fuel assemblies and a two-dimensional full core based on an APR-1400 reactor design. Neutronics analyses were performed for the application of the new fuel-cladding material systems using the reactor-physics Monte Carlo code Serpent 2.31.
The aim of this study is to investigate the potential improvement of accident-tolerant fuels in pressurized water reactors for replacing existing reference zircaloy (Zr) fuel-cladding systems. Three ...main strategies for improving accident-tolerant fuels are investigated: enhancement of the present state-of-the-art zirconium fuel-cladding system to improve oxidation resistance, replacement of the current referenced fuel-cladding system material with an alternative high-performance oxidation-resistant cladding, and replacement of the current fuel with alternative fuel forms. This study focuses on a preliminary analysis of the neutronic behavior and properties of silicon carbide (SiC)-fuel and FeCrAl cladding systems, which provide a better safety margin as accident-tolerant fuel systems for pressurized water reactors. The typical physical behavior of both cladding systems is investigated to determine their general neutronic performance. The multiplication factor, thermal neutron flux spectrum, 239Pu inventory, pin power distribution, and radial power are analyzed and compared with those of a reference Zr fuel-cladding system. Furthermore, the effects of a burnable poison rod (Gd2O3) in different fuel assemblies are investigated. SiC cladding assemblies present a softer neutron spectrum and a lower linear power distribution compared with the conventional Zr-fuel-cladding system. Additionally, the SiC fuel-cladding system exhibits behaviors that are consistent with the neutronic behavior of conventional Zr fuel-cladding systems, thereby affording greater economic and safety improvements.
Traditional neutronics calculations and simulation behavior of plate type research reactors pose a challenge comparing with the conventional pressurized water reactors. This is due to its complex ...geometry, characteristics, heterogeneity, and neutron leakage that lead to modeling difficulties. The Reactor Monte Carlo code (RMC) is employed in stochastic approach for the neutronic analysis in the plate type research reactor. The goal of this report is to examine the capability and validity of the model built using RMC and compare the results with a well-known Monte Carlo code such as MCNP. The results between RMC and MCNP are in a good agreement for effective multiplications factor (keff), radial power peaking factor, total power peaking factor, and integral and differential control rods worth.
Our aim is to study the neutronic behaviour of potential accident-tolerant fuel (ATF) claddings in a pressurised water reactor under normal operations. This work compares ATF silicon carbide ...composite (SiC/SiC) cladding to conventional ZIRLOTM cladding in APR-1400. Additionally, a “sandwich” cladding design developed by the CEA is used for SiC/SiC. The design structure includes a liner in between two layers of the composite to ensure leak tightness. The two proposed liners are Niobium (Nb) and Tantalum (Ta). Serpent 2, a Monte Carlo reactor physics lattice code, is employed to model both cladding materials in APR-1400 at three different levels: pin cell, fuel assembly, and core. The criticality, neutron spectrum, actinide inventory, and power distribution as a function of burnup are investigated. The simulations show that SiC/SiC with the Nb liner displays a far superior performance than the Ta liner across all examined characteristics. Ta leads to a harder neutron spectrum and increased Pu-239 content throughout the cycle, while Nb presents negligible effects. In fact, SiC/SiC with the Nb liner performs very similarly to ZIRLOTM at all model levels. The results indicate that, in terms of neutronics, the adoption of the SiC/SiC composite would entail little to no changes to current APR-1400 operations.
After the Fukushima-Daiichi event, there has been notable progress in developing accident-tolerant fuel (ATF) cladding designs to improve light-water reactor (LWR) safety features. In this regard, ...the Zirconium alloy (Zircaloy) has significant advantages in enhancing the neutron economy, and its unique beneficial characteristics establish it as a reliable and promising material for ATF claddings in the present era. Depositing protective coatings of different interlayer diffusion metals over Zircaloy-based nuclear fuel claddings is becoming increasingly popular as a near-term solution. This strategy can enhance the ability of cladding to withstand accident scenarios, reduce the oxidation rate at high temperatures, and provide additional benefits during both usual and accidental conditions. Current research efforts have strategically prioritized the development of ATF cladding concepts to improve nuclear fuel safety in normal, transient, and potential accidental situations. Herein, we present a comprehensive review of the current research and development events related to the design, manufacturing techniques, and various performance characteristics of chromium (Cr) coated Zircaloy-cladding ATF materials. These materials can improve both the reactor's economics and protection level. Our discussion focuses on these methods for enhancing safety performance in LWRs. In addition, the report provides a prospective viewpoint on future research accomplishments related to these materials.
•The necessity for surface modification of Zircaloy is discussed in detail.•Various types of manufacturing processes for Cr-coated ATF cladding materials are elaborated.•Various fabrication parameters that may cause coating failure are deeply examined.•Different performance parameters for Cr-coated Zircaloy cladding are reviewed.•A prospective viewpoint on future accomplishments of Cr-coated ATF claddings is discussed.
Background
Patients with aggressive lymphomas are at higher risk for venous thromboembolism (VTE). ThroLy is a risk assessment model (RAM) derived to predict the occurrence of VTE in various types of ...lymphomas. In this study, we assess the clinical application of ThroLy RAM in a unified group of patients with diffuse large B-cell lymphoma (DLBCL).
Methods
Hospital databases were searched for patients with DLBCL and radiologically-confirmed VTE. Items in the ThroLy RAM, including prior VTE, reduced mobility, obesity, extranodal disease, mediastinal involvement, neutropenia and hemoglobin < 10.0 g/dL, were retrospectively reviewed.
Results
A total of 524 patients, median age 49 (range: 18-90) years were included. Patients had high disease burden; 57.3% with stage III/IV and 34.0% with bulky disease. All were treated on unified guidelines; 63 (12.0%) had primary refractory disease. Venous thromboembolic events were reported in 71 (13.5%) patients. Among 121 patients with high (> 3) ThroLy score, 22.3% developed VTE compared to 8.4% and 12.4% in those with low and intermediate risk scores, respectively (P = .014). Simplifying the ThroLy model into two risk groups; high-risk (score ≥ 3) and low risk (score < 3) can still segregate patients; VTE developed in 44 (17.2%) high-risk patients (n = 256) compared to 27 (10.1%) in the low-risk group (n = 268), P = .038. Neutropenia, a component of the ThroLy, was encountered in only 14 (2.7%) patients.
Conclusions
ThroLy RAM can identify patients with DLBCL at high risk for VTE. Model can be modified by dividing patients into two, rather than three risk groups, and further simplified by omitting neutropenia.
In 2010, the Department of Nuclear Engineering at Khalifa University of Science and Technology (NUCE) established the MSc degree in Nuclear Engineering. This program was established to support the ...United Arab Emirates (UAE) emerging civil nuclear energy program and produce graduate engineers capable of supporting the safe construction, operation, maintenance and eventual decommissioning of the UAE nuclear reactors. In order to enhance and consolidate classroom teaching, an overseas “Field Trip” module was integrated into the MSc curriculum in order to provide practical exercises, demonstrations and site visits. With respect to the consolidation of reactor physics teaching in particular, the field trips over the past 9 years have utilized mainly the AGN Zero Power reactor at Kyung Hee University (2012-2015) and the Breazeale TRIGA reactor at Penn State University (2017-2018). In both cases students gained “hands-on” supervised experience in classical reactor experiments, shielding measurements, health physics and radiological protection control, and radioanalytical measurement methods, such as neutron activation analysis. At Penn State University, the experiments were extended to surveillance measurements, taking advantage of the facilities specialized modules in nuclear security.
This paper provides details of the arrangements made for the field trips and the learning outcomes these practical arrangements helped to satisfy meeting the overall degree outcomes. The case study will also illustrate where a nuclear energy newcomer State can take advantage of international collaborations without the need to invest in a national research reactor.
Covariance data evaluation for 233U Alrwashdeh, Mohammad
Applied radiation and isotopes,
March 2018, 2018-03-00, 20180301, Volume:
133
Journal Article
Peer reviewed
The FITWR program was developed and applied to experimental nuclear data by using the weighted least square method with nonlinear regression for high-order polynomials to meet the growing demands of ...nuclear data. It can deal with provided variance and covariance data along with experimental data to obtain evaluation data. The aim of this study was to investigate the validity and accuracy of the cross-sectional data for 233U to perform the calculations of the critical system by performing the pre-calculations using the SAMMY code as follows. First, by using the Bayes method, which is available in the SAMMY code, and then by using the FITWR code, which consists of facilitated weighted least square method with nonlinear regression. To validate the FITWR fitting method, a comparison was made with the Bayes method for several benchmarks, and it was tested in both thermal and fast-energy regions. All the computational work done for this paper was carried out at IMAN1 Center – Jordan’s National Supercomputing Center.
239Pu evaluation comparison study Alrwashdeh, Mohammad
Annals of nuclear energy,
August 2018, 2018-08-00, Volume:
118
Journal Article
Peer reviewed
•The analysis of 239Pu was performed using the SAMMY Code for Neutron Transmission, Capture, and Fission cross-sections in the resolved resonance region up to 2.5 keV.•Due to the difficulties of ...processing large covariance matrices, the evaluations were performed in the following distinct uncorrelated energy regions: 0.0–1.0 keV, 1.0 keV–2.0 keV, and 2.0 keV–2.5 keV.
The analysis of 239Pu was performed using the SAMMY Code for Neutron Transmission, Capture, and Fission cross-sections in the resolved resonance region up to 2.5 keV. Due to the difficulties of processing large covariance matrices, the evaluations were performed in the following distinct uncorrelated energy regions: 0.0–1.0 keV, 1.0 keV–2.0 keV, and 2.0 keV–2.5 keV. By using version 8.0.0 of the SAMMY code, the treatment of large resonance parameter covariance matrices was improved significantly. The aim of this work is to analyze the experimental data using a self-developed experimental nuclear data fitting code FITWR and analyze it with the SAMMY code, then compare the results with Bayes method. Finally, the output results from these methods is processed by the SAMMY code in order to obtain covariance matrix resonance parameters in the energy range from 0 to 2.5 keV. All the computational experiments done for this paper was carried out at IMAN1 Center – Jordan’s National Supercomputing Center (IMAN1).
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