Journal of Nuclear Engineering and Radiation Science: Recent submissions
Now showing items 161-180 of 891
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Simulated Environmental Tests for Selected Advanced Technology Fuel Cladding Solutions
(The American Society of Mechanical Engineers (ASME), 2022)Current development on advanced technology fuel (ATF) claddings is aiming at improved high temperature integrity of new candidate materials designed on existing zirconium cladding materials. To assess their performance in ... -
Weight Gain and Hydrogen Absorption in Supercritical Water At 500 °C of Chromium-Coated Zirconium-Based Alloys: Transverse Versus Longitudinal Direction
(The American Society of Mechanical Engineers (ASME), 2022)Canadian Nuclear Laboratories has an on-going Research & Development program to support the development of a scaled–down 300 MWe version of the Canadian Super-Critical Water Reactor concept. The 300 MWe and 170–channel ... -
Testing the Modified Subchannel tempa-sc Code in Comparison With Experiments and Other Computer Codes
(The American Society of Mechanical Engineers (ASME), 2022)The paper describes a modified version of the tempa-sc computer program designed to calculate temperature fields in bundles of rods cooled by a supercritical pressure (SCP) fluid. This version of the program is based on ... -
Dr. Romney B. Duffey on His 80th Birthday
(The American Society of Mechanical Engineers (ASME), 2022)Romney B. DuffeyDr. Romney Duffey is an internationally recognized multi-disciplinary scientist, consultant, manager, speaker, author, and poet. Born on June 26, 1942, and educated in England, Dr. Duffey has over 50 years ... -
Validating the serpent-ants Calculation Chain Using BEAVRS Fresh Core Hot Zero Power Data
(The American Society of Mechanical Engineers (ASME), 2022)The serpent Monte Carlo code and the serpent-ants two-step calculation chain are used to model the hot zero power physics tests described in the BEAVRS benchmark. The predicted critical boron concentrations, control rod ... -
The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term
(The American Society of Mechanical Engineers (ASME), 2022)The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory, etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, and choice ... -
Experimental Investigation on the Retention of Soluble Particles by Pool Scrubbing
(The American Society of Mechanical Engineers (ASME), 2022)In the late stage of a severe loss-of-coolant accident, the pressure in the containment building of a nuclear power plant could rise beyond the design limits and thus endanger its structural integrity. Therefore, to avoid ... -
Nuclear Power and Nuclear Technologies Can Benefit From Regional Implementation of Multinational Approaches
(The American Society of Mechanical Engineers (ASME), 2022)Nuclear energy is a proven low-carbon technology that can provide the dispatchable electricity needed to stabilize national grids that have increasing shares of renewables. Other nuclear technologies are applied in medicine, ... -
Deposition of ITER Vacuum Vessel Dust Inside the Pressure Suppression System During a Loss of Coolant Accident: Experimental and Numerical Analyses
(The American Society of Mechanical Engineers (ASME), 2022)This paper deals with an experimental and numerical analysis of the deposition of International thermonuclear experimental reactor (ITER) dust simulant inside a reduced scale vacuum vessel pressure suppression system (VVPSS) ... -
serpent 2 Validation for Radiation Shielding Applications
(The American Society of Mechanical Engineers (ASME), 2022)This paper contributes to the validation of serpent's photon transport and coupled neutron–photon transport routines. Two benchmarks presenting measurements of neutron and photon flux through different sized iron and lead ... -
Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One
(The American Society of Mechanical Engineers (ASME), 2022)In this paper, shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW (Holtec International Transfer Cask Variable Weight) loaded with spent fuel assemblies (SFA) ... -
Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2
(The American Society of Mechanical Engineers (ASME), 2022)In this work, the effect of averaging operating history parameters such as power history, boron concentration and coolant density, and temperature on spent nuclear fuel properties was investigated. The examined properties ... -
Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years
(The American Society of Mechanical Engineers (ASME), 2022)This article gives an impact analysis of utilization of nuclear power plant full scope simulator on operation parameters, training and education in nuclear power plant Krško (NEK). The Slovenian Nuclear Safety Administration ... -
Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code
(The American Society of Mechanical Engineers (ASME), 2022)The scope of the current research in the field of fuel performance is primarily aimed at an improvement of the operating reliability, safety, and cost-effectiveness of the reactors in operation. The current requirement of ... -
Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor
(The American Society of Mechanical Engineers (ASME), 2022)The core configuration of the Training Research Isotopes General Atomics (TRIGA) MARK I IPR-R1 nuclear research reactor at Nuclear Technology Development Centre (CDTN), Belo Horizonte, Brazil, has been modified six times ... -
On the Self-Shielding in the Unresolved Resonance Range
(The American Society of Mechanical Engineers (ASME), 2022)Resonance behavior is a feature of nuclear reaction cross sections. Resonance density increases with increasing incident particle energy and they begin to overlap, until they can no longer be resolved experimentally, but ... -
Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation
(The American Society of Mechanical Engineers (ASME), 2022)The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise ... -
Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System
(The American Society of Mechanical Engineers (ASME), 2022)Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have ... -
Flow and Heat Transfer Simulation in a Complete Pressurized Water Reactor Fuel Assembly Using Wall-Modeled RANS
(The American Society of Mechanical Engineers (ASME), 2022)Reproduction of turbulent flow and heat transfer inside a pressurized water reactor (PWR) fuel assembly is a challenging task due to the complex geometry and the huge computational domain. Capability of a wall-modeled ... -
Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code
(The American Society of Mechanical Engineers (ASME), 2022)The paper investigates the influence of the used thermal-hydraulic approximations on the coupled calculations of gas-cooled fast reactor design (hereby GFR 2400). The NESTLE code is used as coupled simulation tool and ...