Development of Security and Safety Fuel for Pu-Burner HTGR: Part 3 — Preliminary Reactor Safety Analyses Under Depressurization Accidents

2017 ◽  
Author(s):  
Masaaki Nakano ◽  
Kazutaka Ohashi ◽  
Koji Okamoto
Keyword(s):  
Thermal Power ◽  
Reactor Core ◽  
Reactor Safety ◽  
Analytical Results ◽  
Uranium Fuel ◽  
Core Width ◽  
Accident Management ◽  
Reactivity Coefficient ◽  
Plutonium Fuel ◽  
Is Design

Preliminary reactor safety analyses were performed for the Pu-burner HTGR (High Temperature Gas-cooled Reactor) to investigate plutonium fuel influence on reactor thermal and kinetic behavior for the depressurization accident which is design base accident, and the depressurization accident without scram, which is beyond design base accident. The analytical results of depressurization accident show that the reactor thermal power is restricted lower than the reference reactor with uranium fuel, because the core width is wider for better Pu-burning performance. The analytical results of depressurization accident without scram show that the re-criticality is delayed compare to the reference reactor with uranium-fueled core, due to its temperature reactivity coefficient. The delay of re-criticality brings that operators have more margins for accident management, for example, manual insertion of shutdown devices to the reactor core.

scholarly journals Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4610
Author(s):  
Ahmed Amin E. Abdelhameed ◽  
Chihyung Kim ◽  
Yonghee Kim
Keyword(s):  
Operating Temperature ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Inherent Safety ◽  
Oxide Fuel ◽  
Fast Reactors ◽  
Power Oscillation ◽  
Negative Reactivity ◽  
Oscillation Problem ◽  
Reactivity Coefficient

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

Heat Exchanger Design for Ocean Thermal Difference Power Plants

1975 ◽  
Vol 97 (3) ◽  
pp. 1035-1045 ◽  
Author(s):  
J. G. McGowan ◽  
J. W. Connell
Keyword(s):  
Heat Exchanger ◽  
Power Plants ◽  
Thermal Power ◽  
Thermal Power Plants ◽  
Optimization Program ◽  
Preliminary Results ◽  
Heat Exchanger Design ◽  
Analytical Results ◽  
Thermal Difference

This paper discusses variations in heat exchanger design and configuration for a class of ocean thermal power plants. Details of the heat exchanger models are summarized and analytical results for component and cycle variations are presented. A heat exchanger optimization program is discussed in detail and preliminary results for this study are given.

Effects of Thickness of Zirconium Liner on Stress-Strain Characteristics of U10Mo Monolithic Plates

2013 ◽  
Author(s):  
Hakan Ozaltun ◽  
Robert M. Allen ◽  
You Sung Han
Keyword(s):  
Diffusion Barrier ◽  
Reactor Core ◽  
Stress Strain ◽  
Strain Behavior ◽  
Uranium Fuel ◽  
Plate Design ◽  
Enriched Uranium ◽  
Test Reactor ◽  
Irradiation Performance ◽  
Fuel Elements

The effects of the thickness of Zirconium liner on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation processes were studied. Monolithic plate-type fuel elements is a new fuel form being developed for research and test reactors to achieve higher uranium densities which allows the use of low-enriched uranium fuel in reactor core. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. Early RERTR experiments indicated that the presence of an interaction layer between the fuel and cladding materials causes mechanical problems. To minimize the fuel/cladding interaction, employing a diffusion barrier between the cladding and the fuel materials was proposed. Current monolithic plate design employs a 0.025 mm thick, 99.8% pure annealed Zirconium diffusion barrier between the fuel foil (U10Mo) and the cladding materials (AL6061-O). To benchmark the irradiation performance, a number of plates were irradiated in the Advanced Test Reactor (ATR) with promising irradiation performance. To understand the effects of the thickness of the Zirconium diffusion barrier on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate from RERTR-12 experiments (Plate L1P7A0) was selected and simulated. Both fabrication and irradiation stages were considered. Simulations were repeated for various Zirconium thicknesses to understand the effects of the thickness of the diffusion barrier. Results of fabrication simulations indicated that Zirconium thickness has noticeable effects on foil’s stresses. Irradiation simulations revealed that the fabrication stresses of the foil would be relieved rapidly in the reactor. Results also showed that Zirconium thickness has little or no effects on irradiation and shutdown stresses.

Neutronics Optimization of UN Fuel Pin Assemblies for a Sodium-Cooled, Small Modular Reactor

2011 ◽  
Author(s):  
Timothy M. Schriener ◽  
Mohamed S. El-Genk
Keyword(s):  
Thermal Power ◽  
Reactor Core ◽  
Central Cavity ◽  
Preliminary Results ◽  
The Core ◽  
Small Modular Reactor ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Reference Design ◽  
Centerline Temperature

This paper presents preliminary results of neutronics and thermal-hydraulics design analysis of a sodium cooled, small modular reactor (SMR). The reactor’s nominal thermal power is 150 MWth at sodium inlet and exit temperatures of 630 and 780 K. The reactor core is comprised of three rings of shrouded hexagonal assemblies of 19.8% enriched UN fuel pins and a hexagonal assembly of enriched B4C pins in the central cavity for a coarse reactivity control. The objectives are to provide enough excess reactivity for achieving a refueling cycle > 5 year, maintaining a more even coolant flow in the core assemblies and keeping the peak centerline temperature of UN fuel pins < 1300 K. Fuel assemblies with scalloped shroud walls, 4 rings and 1.942 cm diameter fuel pins with p/d = 1.098 are selected for the reference design of the present SMR. In this design, peak fuel centerline temperature is only 1240 K and the beginning-of-life, cold-clean excess reactivity is $26.67.

scholarly journals EN Analysis of modern approaches to improve the efficiency of blackout accident management at nuclear power plants

2019 ◽  
Vol 55 (4) ◽  
pp. 227-234
Author(s):  
A. Denysova ◽  
V. Skalozubov ◽  
V. Spinov ◽  
D. Spinov ◽  
D. Pirkovskiy ◽  
...  
Keyword(s):  
Steam Generator ◽  
Power Plants ◽  
Real Life ◽  
Reactor Core ◽  
Heat Removal ◽  
Integrated Approach ◽  
Necessary Condition ◽  
Multiple Failures ◽  
Feedwater Pump ◽  
Accident Management

The paper analyzes the approaches to improve the efficiency of blackout accident management taking into account the lessons of the great accident at Fukushima Daiichi NPP in 2011. It is found that the afterheat removal passive systems by natural circulation through steam generators cannot provide conditions for adequate safety functions to remove heat from the reactor and maintain the required feedwater level in the steam generator during blackout accidents and multiple failures of safety-related systems. The application of alternative approaches using auxiliary feedwater steam generator driven pumps requires additional experiment-calculated operability / reliability qualification for a blackout accident and multiple failures of NPP safety-related systems. However, implementation of alternative SDEFP system requires in-depth qualification for the conditions of blackout accidents. Safety systems of passive heat removal from the steam generator (adequately to active safety electrical systems) cannot ensure safety functions on control of required feedwater level in the steam generator and heat removal from the reactor core during blackout accidents (at least 72 hours) and multifailure accidents. The system of the steam generator driven emergency feedwater pump can be the alternative solution to ensure safety functions on heat removal through the steam generator during blackout accidents. Additional study of efficiency of steam driven pumps at the experimental facilities that meet real-life criteria of hydrodynamic similarity is a necessary condition for implementation of system of the steam driven emergency feedwater pump. Application of an integrated approach to manage blackout accidents is reasonable. At the initial stage of accident with relatively high steam pressure in the steam generator it is required supply of feedwater by the steam driven emergency pump

scholarly journals Numerical modelling of modular high-temperature gas-cooled reactors with thorium fuel

Nukleonika ◽  
2021 ◽  
Vol 66 (4) ◽  
pp. 133-138
Author(s):  
Mikołaj Oettingen ◽  
Jerzy Cetnar
Keyword(s):  
High Temperature ◽  
Neutron Transport ◽  
Reactor Core ◽  
Computation Time ◽  
Homogenization Method ◽  
Uranium Fuel ◽  
Neutron Multiplication ◽  
3D Numerical Model ◽  
Effective Neutron ◽  
High Temperature Gas

Abstract The volumetric homogenization method for the simplified modelling of modular high-temperature gas-cooled reactor core with thorium-uranium fuel is presented in the paper. The method significantly reduces the complexity of the 3D numerical model. Hence, the computation time associated with the time-consuming Monte Carlo modelling of neutron transport is considerably reduced. Example results comprise the time evolutions of the effective neutron multiplication factor and fissionable isotopes (233U, 235U, 239Pu, 241Pu) for a few configurations of the initial reactor core.

scholarly journals Implications for accident management of adding water to a degrading reactor core

1994 ◽  
Author(s):  
P. Kuan ◽  
D.J. Hanson ◽  
D.J. Pafford ◽  
K.S. Quick ◽  
R.J. Witt
Keyword(s):  
Reactor Core ◽  
Accident Management

Using Artificial Neural Networks in the Reactor Safety Research

2008 ◽  
Author(s):  
Alexander Kratzsch ◽  
Wolfgang Ka¨stner ◽  
Rainer Hampel
Keyword(s):  
Neural Networks ◽  
Artificial Neural Networks ◽  
Heat Dissipation ◽  
Model Development ◽  
Reactor Core ◽  
Differential Pressure ◽  
Reactor Safety ◽  
Safety Research ◽  
Artificial Neural ◽  
Residual Heat

The paper deals with the creation of a differential pressure model with artificial neural networks (ANN). Particular, model development and verification tests are considered. One of the main features in reactor safety research is the safe heat dissipation from the reactor core and the reactor containment of light-water reactors. In the case of loss of coolant accident (LOCA) the possibility of the entry of isolation material into the reactor containment and the building sump of the reactor containment and into the associated systems to the residual heat exhaust is a serious problem. This can lead to a handicap of the system functions. To ensure the residual heat exhaust it is necessary the emergency cooling systems to put in operation which transport the water from the sump to the condensation chamber and directly to the reactor pressure vessel. A high allocation of the sieves with fractionated isolation material, in the sump can lead to a blockage of the strainers, inadmissibly increase of differential pressure, build-up at the sieves and to malfunctioning pumps. Hence, the scaling and retention of fractionated isolation material in the building sump of the reactor containment must be estimated. This allows the potential plant status in case of incidents to be assessed. The differential pressure is the essential parameter for the assessment of allocation of the strainers. For modelling we use artificial neural networks. To build up the ANN, the available experimental data are used to train the ANN.

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