scholarly journals GCFR Coupled Neutronic and Thermal-Fluid-Dynamics Analyses for a Core Containing Minor Actinides

2009 ◽  
Vol 2009 ◽  
pp. 1-8 ◽  
Author(s):  
Diego Castelliti ◽  
Eleonora Bomboni ◽  
Nicola Cerullo ◽  
Guglielmo Lomonaco ◽  
Carlo Parisi
Keyword(s):  
Fluid Dynamics ◽  
Energy Production ◽  
Nuclear Power ◽  
Climate Changes ◽  
Fuel Cycle ◽  
Production Of Hydrogen ◽  
Gen Iv ◽  
Safety Systems ◽  
Dynamics Analyses ◽  
Thermal Hydraulic Analysis

Problems about future energy availability, climate changes, and air quality seem to play an important role in energy production. While current reactor generations provide a guaranteed and economical energy production, new nuclear power plant generation would increase the ways and purposes in which nuclear energy can be used. To explore these new technological applications, several governments, industries, and research communities decided to contribute to the next reactor generation, called “Generation IV.” Among the six Gen-IV reactor designs, the Gas Cooled Fast Reactor (GCFR) uses a direct-cycle helium turbine for electricity generation and for a -free thermochemical production of hydrogen. Additionally, the use of a fast spectrum allows actinides transmutation, minimizing the production of long-lived radioactive waste in an integrated fuel cycle. This paper presents an analysis of GCFR fuel cycle optimization and of a thermal-hydraulic of a GCFR-prototype under steady-state and transient conditions. The fuel cycle optimization was performed to assess the capability of the GCFR to transmute MAs, while the thermal-hydraulic analysis was performed to investigate the reactor and the safety systems behavior during a LOFA. Preliminary results show that limited quantities of MA are not affecting significantly the thermal-fluid-dynamics behavior of a GCFR core.

Steady-State Thermal-Hydraulic Analysis of a Particle-Bedded Boiling Water Reactor

2009 ◽  
Author(s):  
Xiaoyu Cai ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Changyou Zhao
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Superheated Steam ◽  
Nuclear Reactor ◽  
Hydraulic Analysis ◽  
Light Water ◽  
Safety Systems ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis

The current Light Water Reactors both BWR and PWR have extensive nuclear reactor safety systems, which provide safe and economical operation of Nuclear Power Plants. During about forty years of operation history the safety systems of Nuclear Power Plants have been upgraded in an evolutionary manner. The cost of safety systems, including large containments, is really high due to a capital cost and a long construction period. These conditions together with a low efficiency of steam cycle for LWR create problems to build new power plants in the USA and in the Europe. An advanced Boiling Water Reactor concept with micro-fuel elements (MFE) and superheated steam promises a radical enhancement of safety and improvement of economy of Nuclear Power Plants. In this paper, a new type of nuclear reactor is presented that consists of a steel-walled tube filled with millions of TRISO-coated fuel particles (Micro-Fuel Elements, MFE) directly cooled by a light-water coolant-moderator. Water is used as coolant that flows from bottom to top through the tube, thereby fluidizing the particle bed, and the moderator water flows in the reverse direction out of the tube. The fuel consists of spheres of about 2.5 mm diameter of UO2 with several coatings of different carbonaceous materials. The external coating of steam cycle the particles is silicon carbide (SiC), manufactured with chemical vapor deposit (CVD) technology. Steady-State Thermal-Hydraulic Analysis aims at providing heat transport capability which can match with the heat generated by the core, so as to provide a set of thermal hydraulic parameters of the primary loop. So the temperature distribution and the pressure losses along the direction of flow are calculated for equilibrium core in this paper. The calculation not only includes the liquid region, but the two phase region and the superheated steam region. The temperature distribution includes both the temperature parameters of micro-fuel elements and the coolant. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.

The Fast Breeder Reactor — Energy without Depletion of Natural Resources

1976 ◽  
Vol 190 (1) ◽  
pp. 163-175
Author(s):  
R. D. Vaughan ◽  
A. A. Farmer
Keyword(s):  
Energy Production ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
Primary Energy ◽  
Western World ◽  
Fast Reactors ◽  
Generating Capacity ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Fuel Reprocessing

SYNOPSIS. Nuclear power should account for 20% of primary energy production in the Western world by the end of the century, but only if growth of generating capacity can be freed of the constraint of uranium supply. It is shown that, providing fast breeder reactors and their associated fuel reprocessing facilities are developed quickly, a substantial increase in nuclear capacity could be provided by fast reactors. The relative importance of various fuel cycle parameters is spelt out and brief accounts are given of the alternative fast reactors being developed to meet the requirements.

Mechanical Aspects of the Canadian Generation IV Supercritical Water-Cooled Pressure Tube Reactor

2012 ◽  
Author(s):  
M. Yetisir ◽  
M. Gaudet ◽  
D. Martin
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
High Efficiency ◽  
Fuel Cycle ◽  
Pressure Tube ◽  
Steam Power ◽  
Power Cycle ◽  
Design Activities ◽  
Gen Iv ◽  
Safety Features

Canada is developing a next generation (Gen IV) reactor, the Canadian Super-Critical Water-cooled Reactor (SCWR), which has the potential of meeting all design goals of the Gen IV International Forum (GIF). These design goals include enhanced safety features (inherent safe operation and deploying passive safety features), improved economic and resource utilization (∼40% more efficient than current nuclear power generating stations, sustainable fuel cycle (will burn thorium, which is three times more abundant than uranium), enhanced sustainability, and greater proliferation resistance compared to the current fleets (i.e., Gen III and Gen III+) of reactors. The Canadian SCWR concept is a pressure-tube type reactor that uses supercritical water as a coolant, a separate low-pressure heavy water moderator, a high efficiency fuel channel that includes a zirconium alloy pressure and a ceramic zirconia insulator, and a direct steam power cycle. This paper presents the evolution of the Canadian SCWR core design, simplifications introduced in the design to deal with design challenges, and ongoing design activities.

Small Break LOCA Analysis of ACR-700 NPP

2006 ◽  
Author(s):  
Limin Zheng ◽  
Sen Shen ◽  
David Wright
Keyword(s):  
Nuclear Power ◽  
Sensitivity Study ◽  
Hydraulic Analysis ◽  
Reactor Outlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Subsequent Loss ◽  
Safety Systems ◽  
Limiting Case ◽  
Thermal Hydraulic Analysis

A small break loss of coolant accident (SB-LOCA) analysis to assess a preliminary conceptual design of the ACR-700 PHWR nuclear power plant (NPP) developed by AECL has been performed with CATHENA MOD 3.5d, a PHWR system thermal-hydraulic analysis code. The limiting break size has been found by performing a sensitivity study for three different break locations [i.e. reactor inlet header (RIH), HTS pump suction (PS) pipe and reactor outlet head (ROH)] under the limiting case (i.e. SB-LOCA with subsequent loss of class IV power with all safety systems available). The analysis results indicate that the SB-LOCA acceptance criteria are satisfied.

scholarly journals Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

2018 ◽  
Vol 2018 ◽  
pp. 1-17
Author(s):  
Frederic Salaun ◽  
David R. Novog
Keyword(s):  
Fuel Cycle ◽  
Control Rod ◽  
Neutron Kinetics ◽  
3 Dimensional ◽  
Thermodynamic Critical Point ◽  
Sensitivity Studies ◽  
Gen Iv ◽  
Safety Systems ◽  
Supercritical Water Cooled Reactor ◽  
Control Rods

The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

Discrete element method–computational fluid dynamics analyses of flexible fibre fluidization

2021 ◽  
Vol 910 ◽  
Author(s):  
Yiyang Jiang ◽  
Yu Guo ◽  
Zhaosheng Yu ◽  
Xia Hua ◽  
Jianzhong Lin ◽  
...  
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Discrete Element Method ◽  
Discrete Element ◽  
Dynamics Analyses ◽  
Element Method

Abstract

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Nuclear Power in Latin America: An Overview of Its Present Status

1983 ◽  
Vol 25 (3) ◽  
pp. 377-415 ◽  
Author(s):  
Margarete K. Luddemann
Keyword(s):  
Developing Countries ◽  
Latin America ◽  
Nuclear Power ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Field Studies ◽  
Energy Agency ◽  
Self Sufficiency ◽  
Policy Concern ◽  
Alternative Sources

The pivotal role energy plays in national economics not only converts the access to sources of supply into a vivid issue of foreign policy concern, but also causes an understandable preoccupation with investment capabilities and self-sufficiency. A report prepared by the International Atomic Energy Agency (IAEA) in 1974 predicted a bright future for nuclear energy in the i developing countries and encouraged use of this form of energy after numerous field studies.A nation that commits itself to nuclear energy by purchasing nuclear power-generating technology but not fuel cycle facilities incurs the risk of becoming dependent upon the supplier country because a quick switch to alternative sources of supply is difficult in cases of curtailment of fuel.

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