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.

ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents

2006 ◽  
Author(s):  
N. Reinke ◽  
K. Neu ◽  
H.-J. Allelein
Keyword(s):  
Fission Products ◽  
Feasibility Studies ◽  
Current Status ◽  
Light Water ◽  
Fast Running ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Water Reactors ◽  
Reactor Types

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.

Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break1

2015 ◽  
Vol 137 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Pipe ◽  
Short Period ◽  
Main Steam Line ◽  
Main Steam ◽  
Transient Thermal ◽  
Secondary Side

This paper presents a multidimensional numerical analysis of the transient thermal-hydraulic response of a steam generator (SG) secondary side to a double-ended guillotine break of the main steam line attached to the SG at a pressurized water reactor (PWR) plant. A simplified analysis model is designed to include both the SG upper space, which the steam occupies and a part of the main steam line between the SG outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport (SST) turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break (MSLB) accident, a constant amount of steam is assumed to be generated from the bottom of the SG upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present computational fluid dynamics (CFD) model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2–8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.

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.

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.

scholarly journals Anthropogenic Radiocarbon: Past, Present, and Future

Radiocarbon ◽  
1986 ◽  
Vol 28 (2A) ◽  
pp. 668-672 ◽  
Author(s):  
Pavel Povinec ◽  
Martin Chudý ◽  
Alexander Šivo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Pressurized Water Reactors ◽  
Anthropogenic Radionuclides ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Weapon Testing ◽  
Water Reactors

14C is one of the most important anthropogenic radionuclides released to the environment by human activities. Weapon testing raised the 14C concentration in the atmosphere and biosphere to +100% above the natural level. This excess of atmospheric C at present decreases with a half-life of ca 7 years. Recently, a new source of artificially produced 14C in nuclear reactors has become important. Since 1967, the Bratislava 14C laboratory has been measuring 14C in atmospheric 14CO2 and in a variety of biospheric samples in densely populated areas and in areas close to nuclear power plants. We have been able to identify a heavy-water reactor and the pressurized water reactors as sources of anthropogenic 14C. 14C concentrations show typical seasonal variations. These data are supported by measurements of 3H and 85Kr in the same locations. Results of calculations of future levels of anthropogenic 14C in the environment due to increasing nuclear reactor installations are presented.

Evaluation of a Numerical Analysis Model for the Transient Response of Nuclear Steam Generator Secondary Side to a Sudden Steam Line Break

2016 ◽  
Vol 139 (3) ◽  
Author(s):  
Jong Chull Jo ◽  
Bok Ki Min ◽  
Jae Jun Jeong
Keyword(s):  
Numerical Analysis ◽  
Steam Generator ◽  
Steam Line ◽  
Pressurized Water Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Transient Thermal ◽  
Phase Change Heat Transfer ◽  
Secondary Side ◽  
Good Agreement

This paper presents an evaluation of the applicability of a numerical analysis model to the transient thermal-hydraulic response of steam generator (SG) secondary side to blowdown following a steam line break (SLB) at a pressurized water reactor (PWR). To do this, the numerical analysis model was applied to simulate the same blowdown situation as in an available experiment which was conducted for a simplified SG blowdown model, and the numerical results were compared with the measurements. As a result, both are in reasonably good agreement with each other. Consequently, the present numerical analysis model is evaluated to have the applicability for numerical simulations of the transient phase change heat transfer and flow situations in PWR SGs during blowdown following a SLB.

scholarly journals An Overview of the Pressurized Thermal Shock Issue in the Context of the NURESIM Project

2009 ◽  
Vol 2009 ◽  
pp. 1-13 ◽  
Author(s):  
D. Lucas ◽  
D. Bestion ◽  
E. Bodèle ◽  
P. Coste ◽  
M. Scheuerer ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Cfd Simulation ◽  
Two Phase ◽  
Pressurized Water ◽  
Simulation Tools ◽  
Pressurized Thermal Shock ◽  
Qualitative Level ◽  
Water Reactors ◽  
Near Term ◽  
Accident Scenarios

Within the European Integrated Project NURESIM, the simulation of PTS is investigated. Some accident scenarios for Pressurized Water Reactors may cause Emergency Core Coolant injection into the cold leg leading to PTS situations. They imply the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. This paper focuses on two-phase conditions that are potentially at the origin of PTS. It summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor,that is, the cold leg and the downcomer, where the “PTS fluid-dynamics" is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two-phase flow subphenomena are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance is still limited. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet well controlled, leading to slightly conservative predictions.

scholarly journals Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

2012 ◽  
Vol 66 (3) ◽  
pp. 291-299 ◽  
Author(s):  
Grégory Lefèvre ◽  
Ljiljana Zivkovic ◽  
Anne Jaubertie
Keyword(s):  
Nuclear Reactor ◽  
Primary Circuit ◽  
Particle Removal ◽  
Mechanical Seals ◽  
Solution Ph ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Key Factor ◽  
Water Reactors ◽  
Chemical Conditions

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition) in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek) theory and used as such to interpret this industrial phenomenon.

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.

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