How to Simulate the Microstructure Induced by a Nuclear Reactor with an Ion Beam Facility : DART

2009 ◽  
Vol 1215 ◽  
Author(s):  
Laurence Luneville ◽  
David Simeone ◽  
Gianguido Baldinozzi ◽  
Dominique Gosset ◽  
yves serruys
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Ion Beam ◽  
Fast Neutrons ◽  
Relaxation Processes ◽  
Pressurized Water ◽  
Water Reactors ◽  
Displacement Per Atom ◽  
The Impact

AbstractEven if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and ion beam facilities on a defined material. In this paper, a formalism is presented to evaluate the displacement cross-sections pointing out the effect of the anisotropy of nuclear reactions. From this formalism, the impact of fast neutrons (with a kinetic energy En superior to 1 MeV) is accurately described. This point allows calculating accurately the displacement per atom rates as well as primary and weighted recoil spectra. Such spectra provide useful information to select masses and energies of ions to perform realistic experiments in ion beam facilities.

A Systematic Review of PCTRAN-Based Pressurized Water Reactor Transient Analysis

2021 ◽  
Author(s):  
Suubi Racheal ◽  
Yongkuo Liu ◽  
Miyombo Ernest ◽  
Abiodun Ayodeji
Keyword(s):  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Research Education ◽  
User Expertise ◽  
Scale Down ◽  
Water Reactors ◽  
The Impact ◽  
And Training

Abstract The impact of nuclear accidents has been a topic of debate since the construction of the first nuclear reactor, and still stands as a key issue of public concern. Several codes and simulators have been used to study the transient progression in pressurized water reactors, and to evaluate the technical measures adopted to scale down the risk of accidents. However, some of these codes are not suitable for multipurpose research and training as they require significant user expertise, leading to analysis uncertainties largely from the code user effect. This paper presents a bird-eye view of one of the most widely used nuclear reactor transient analyzer — the Personal Computer Transient Analyzer (PCTRAN). This paper discusses the comparative advantages of the simulator from the users’ perspective, with specific attention to its utilization both for research and training. The paper also demonstrates the ease of usage by simulating common transient in a pressurized water reactor. Finally, observations and possible improvements to the code to increase its usability in research, education and training are discussed. This work aims to evaluate the robustness of the simulator towards better utilization for research and training, especially in nuclear newcomer countries.

Proposed Solutions to Achieve Nuclear Fusion

Engevista ◽  
2017 ◽  
Vol 19 (5) ◽  
pp. 1496
Author(s):  
Relly Victoria Virgil Petrescu ◽  
Raffaella Aversa ◽  
Antonio Apicella ◽  
Florian Ion Petrescu
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Nuclear Fusion ◽  
Nuclear Fission ◽  
Light Nuclei ◽  
Fast Neutrons ◽  
Fusion Reactions ◽  
Nuclear Fusion Reactions ◽  
The Masses ◽  
Fusion Products

Despite research carried out around the world since the 1950s, no industrial application of fusion to energy production has yet succeeded, apart from nuclear weapons with the H-bomb, since this application does not aims at containing and controlling the reaction produced. There are, however, some other less mediated uses, such as neutron generators. The fusion of light nuclei releases enormous amounts of energy from the attraction between the nucleons due to the strong interaction (nuclear binding energy). Fusion it is with nuclear fission one of the two main types of nuclear reactions applied. The mass of the new atom obtained by the fusion is less than the sum of the masses of the two light atoms. In the process of fusion, part of the mass is transformed into energy in its simplest form: heat. This loss is explained by the Einstein known formula E=mc2. Unlike nuclear fission, the fusion products themselves (mainly helium 4) are not radioactive, but when the reaction is used to emit fast neutrons, they can transform the nuclei that capture them into isotopes that some of them can be radioactive. In order to be able to start and to be maintained with the success the nuclear fusion reactions, it is first necessary to know all this reactions very well. This means that it is necessary to know both the main reactions that may take place in a nuclear reactor and their sense and effects. The main aim is to choose and coupling the most convenient reactions, forcing by technical means for their production in the reactor. Taking into account that there are a multitude of possible variants, it is necessary to consider in advance the solutions that we consider them optimal. The paper takes into account both variants of nuclear fusion, and cold and hot. For each variant will be mentioned the minimum necessary specifications.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Crack-Opening Displacement and Leak-Rate Calculations for Full Structural Weld Overlays

2009 ◽  
Author(s):  
D.-J. Shim ◽  
E. Kurth ◽  
F. Brust ◽  
G. Wilkowski ◽  
A. Csontos ◽  
...  
Keyword(s):  
Crack Opening ◽  
Nuclear Industry ◽  
Leak Rate ◽  
Opening Displacement ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
The Impact ◽  
Weld Overlays

Full structural weld overlays have been used in the U.S. nuclear power industry for over twenty years in boiling water reactors (BWRs). Primary water stress corrosion cracking (PWSCC) in nickel-based dissimilar metal welds (DMWs) has been experienced in pressurized water reactors (PWRs) since the early 1990s. As a result, the nuclear industry is implementing full structural weld overlays (FSWOL) as a PWSCC mitigation technique that may be used on primary coolant lines previously approved for Leak-Before-Break (LBB). This work investigates the effect of the FSWOL on the leakage behavior of these lines with postulated defects. In this paper, finite element (FE) based crack-opening displacements (CODs) were developed for pipes with a FSWOL with postulated complex cracks. The COD solutions were then employed in standard leak-rate calculations, where equivalent crack morphology parameters were developed to consider a flow through two different crack morphologies, i.e., PWSCC through the DMW and corrosion fatigue through the weld overlay. The results of the sensitivity study and a discussion on the impact of the weld overlay on the leakage behavior concludes this paper.

Small Modular Reactor and Large Nuclear Reactor Fuel Cost Comparison

2014 ◽  
Author(s):  
Christopher P. Pannier ◽  
Radek Škoda
Keyword(s):  
Nuclear Reactor ◽  
Cost Comparison ◽  
Design Parameters ◽  
Fuel Cost ◽  
Uranium Enrichment ◽  
Pressurized Water ◽  
Nuclear Reactor Fuel ◽  
Water Reactors ◽  
Near Term ◽  
Fuel Costs

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.

scholarly journals Fast neutron spectrum unfolding of a TRIGA Mark II reactor and measurement of spectrum-averaged cross sections: integral tests of differential cross sections of neutron threshold reactions

2013 ◽  
Vol 101 (10) ◽  
pp. 613-620 ◽  
Author(s):  
M. S. Uddin ◽  
S. Sudár ◽  
S. M. Hossain ◽  
R. Khan ◽  
M. A. Zulquarnain ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Cross Sections ◽  
Nuclear Reactions ◽  
Neutron Field ◽  
Fast Neutrons ◽  
Metal Foils ◽  
Spectrum Unfolding ◽  
Fission Spectrum ◽  
Data Files ◽  
Triga Mark Ii

Summary The spectrum of fast neutrons having energies from 0.5 to 20 MeV in the core of the 3MW TRIGA Mark II reactor at Savar, Dhaka, Bangladesh, was unfolded by activating several metal foils to induce threshold nuclear reactions covering the whole spectrum, and then doing necessary iterative calculations utilizing the activation results and the code SULSA. The analysed shape of the spectrum in the TRIGA core was found to be similar to that of the pure 235U-fission spectrum, except for the energies between 0.5 and 1.5 MeV, where it was slightly higher than the fission spectrum. Spectrum-averaged cross sections were determined by integral measurements. The integral values measured in this work were compared with the recommended values for a pure fission spectrum as well as with the integrated data deduced from measured and evaluated excitation functions of a few reactions given in some data files. The good agreement between integral measurements and integrated data in case of well-investigated reactions shows that the fast neutron field at the TRIGA Mark II reactor can be used for validation of evaluated data of neutron threshold reactions.

Analysis of Major Group Structures Used for Nuclear Reactor Simulations

2018 ◽  
Author(s):  
Marco Di Filippo ◽  
Jiri Krepel ◽  
Konstantin Mikityuk ◽  
Horst-Michael Prasser
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Group Structure ◽  
Major Group ◽  
Computational Time ◽  
Sensitivity Threshold ◽  
Energy Structure ◽  
Reactor Spectrum ◽  
Group Structures ◽  
The Impact

Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation. The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated. The aim was to identify a group structure with very low dependency on the actual reactor spectrum.

Regulatory Perspective on Advanced Finite Element Flaw Growth Analysis of Alloy 82/182 Butt Welds

2008 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Aladar A. Csontos ◽  
Timothy R. Lupold ◽  
Chia-Fu Sheng
Keyword(s):  
Finite Element ◽  
Nuclear Power ◽  
Regulatory Review ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Primary Water ◽  
Water Reactors ◽  
The Impact ◽  
Growth Analyses

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed preweld overlay inspections using manual ultrasonic testing (UT) techniques on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal (DM) and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe end DM butt welds that the licensee attributed to primary water stress corrosion cracking (PWSCC). These indications were significantly larger and more extensive than previously seen for the case of circumferential indications in commercial pressurized water reactors. As a result of the NRC staff’s initial flaw growth analyses, the NRC staff obtained commitments from the nuclear power industry licensees to complete pressurizer nozzle DM butt weld inspections on an accelerated basis. In addition, the industry informed the NRC staff that it would undertake a task to refine the crack growth analyses using more realistic assumptions to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds. These new analyses are referred to as advanced finite element (AFE) analyses. This paper will discuss the regulatory review of the industry’s AFE analyses. This discussion will include the NRC staff’s approach to the review, the differences between the industry’s AFE analyses and the NRC staff’s confirmatory analyses, and the NRC staff’s acceptance criteria. The paper will discuss the impact of the AFE analyses on the regulatory process. Finally, the paper will discuss possible future regulatory and research applications for AFE analyses as well as additional NRC research projects intended to address some of the uncertainties in this type of analysis.

scholarly journals Qualification and characterization of electronics of the fast neutron Hodoscope detectors using neutrons from CABRI core

2018 ◽  
Vol 170 ◽  
pp. 04016 ◽  
Author(s):  
S. Mirotta ◽  
J. Guillot ◽  
V. Chevalier ◽  
B. Biard
Keyword(s):  
Fast Neutrons ◽  
Pressurized Water ◽  
Distinctive Features ◽  
Power Pulse ◽  
Integral Effect ◽  
Proton Recoil ◽  
Complete Set ◽  
Water Reactors ◽  
Threshold Tuning ◽  
Fission Chambers

The study of Reactivity Initiated Accidents (RIA) is important to determine up to which limits nuclear fuels can withstand such accidents without clad failure. The CABRI International Program (CIP), conducted by IRSN under an OECD/NEA agreement, has been launched to perform representative RIA Integral Effect Tests (IET) on real irradiated fuel rods in prototypical Pressurized Water Reactors (PWR) conditions. For this purpose, the CABRI experimental pulse reactor, operated by CEA in Cadarache, France, has been strongly renovated, and equipped with a pressurized water loop. The behavior of the test rod, located in that loop in the center of the driver core, is followed in real time during the power transients thanks to the hodoscope, a unique online fuel motion monitoring system, and one of the major distinctive features of CABRI. The hodoscope measures the fast neutrons emitted by the tested rod during the power pulse with a complete set of 153 Fission Chambers and 153 Proton Recoil Counters. During the CABRI facility renovation, the electronic chain of these detectors has been upgraded. In this paper, the performance of the new system is presented describing gain calibration methodology in order to get maximal Signal/Noise ratio for amplification modules, threshold tuning methodology for the discrimination modules (old and new ones), and linear detectors response limit versus different reactor powers for the whole electronic chain.

Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Lena Andriolo ◽  
Clément Meriot ◽  
Nikolai Bakouta
Keyword(s):  
Heat Dissipation ◽  
Analytical Calculation ◽  
Heat Fluxes ◽  
Reference Case ◽  
Pressurized Water ◽  
Vessel Steel ◽  
Small Modular Reactor ◽  
Lower Head ◽  
Water Reactors ◽  
The Impact

The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EDF) Research and Development (R&D) Center, in the Pericles department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMR) with the modular accident analysis program version 5 in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a stepwise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of the lower head geometry, vessel steel ablation, and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. The results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identify modeling improvements.

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