High Temperature Intrinsically Safe Nuclear Reactor

2015 ◽  
Author(s):  
Robin J. McDaniel
Keyword(s):  
High Temperature ◽  
Alternative Energy ◽  
Nuclear Reactor ◽  
Cooling System ◽  
Clean Energy ◽  
Normal Operation ◽  
The Core ◽  
Small Modular Reactor ◽  
Loss Of Coolant ◽  
Minimum Number

Small Modular Reactor (SMR) technologies have been recently included by the DOE as clean energy, a low carbondioxide emitting “alternative energy” source. The objective of this paper is to further the discussion of intrinsically safe nuclear reactors in the context of passive safety design principles and introduction of a novel conceptual reactor design. After a multiple year research study of past fast neutron reactor designs and recent reactor proposals, the following design is the result of analysis of the best concepts discovered. An improved fast reactor of the liquid metal cooled type including a core configuration allowing for only two operational states, “Power” or “Rest”. The flow of the primary cooling fluid suspends the core in the “Power” state, with sufficient flow to remove the heat to an intermediate heat exchanger during normal operation. This design utilizes the force of gravity to shut down the reactor after any loss of coolant flow, either a controlled reactor shut down or a Loss of Coolant Accident (LOCA) event, as the core is controlled via dispersion of fuel elements. Electromagnetic pumps incorporating automatic safety electrical cut-offs are employed to shutdown the primary cooling system to disassemble the core to the “Rest” configuration due to a loss of secondary coolant or loss of ultimate heat sink. This design is a hybrid pool-loop pressurized high-temperature reactor unique in its use of a minimum number of components, utilizing no moving mechanical parts, no rotating seals, and no control rods. This defines an elegantly simple Gen IV intrinsically safe nuclear reactor. [Advanced Small Modular Reactor (aSMR)]

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

CFD Analysis on the Influence of Coolant Distribution Blocks on the Core Hot Spot Temperature in a Prismatic Very High Temperature Reactor

2011 ◽  
Author(s):  
Min-Hwan Kim ◽  
Nam-il Tak ◽  
Jae Man Noh ◽  
Goon-Cherl Park
Keyword(s):  
High Temperature ◽  
Hot Spot ◽  
Maximum Velocity ◽  
Unit Cell ◽  
Normal Operation ◽  
The Core ◽  
Very High Temperature Reactor ◽  
Velocity Deviation ◽  
High Temperature Reactor ◽  
Very High

Two design options of core distribution block (CDB) for a cooled-vessel design in the Very High Temperature Reactor (VHTR) were developed and the influence on the core hot spot was investigated by the commercial computational fluid dynamics (CFD) code, CFX-11. Isothermal CFD analyses were performed to estimate the coolant flow variation at the inlet of the coolant channel. The results predicted about 5% of the maximum velocity deviation when applying the core pressure drop of NHDD PMR200. A unit-cell CFD model was used to access the effect of the velocity deviation on the core hot spot. The unit-cell analyses were carried out for the velocity deviation of 0%, 5%, and 10%. Not only a constant power but also a local maximum power profile was considered. According to the results, the maximum fuel temperature was increased by about 30°C for the velocity deviation of 10% but still below the normal operation limit of 1250°C.

HTR-PM Technology in Energy Supply of Petrol Chemical Industry in China: An Economic Feasibility View

2016 ◽  
Author(s):  
Gang Zhao ◽  
Ping Ye ◽  
Jie Wang ◽  
Xiaoyong Yang
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Alternative Energy ◽  
Fossil Fuel ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Chemical Plant ◽  
Economic Feasibility ◽  
Cogeneration Plant ◽  
High Temperature Steam

The massive use of fossil fuel has caused huge carbon emission and serious air pollution in China. Now all kinds of alternative energy technology are developing rapidly to solve such problem in China. Electricity produced by non-fossil fuel energy is continued to increase sharply in China. But it’s hard for regular alternative energy, such as wind power, solar power, hydroelectricity power, nuclear power and so on, to easily provide process heat for industry, especially high temperature steam. High temperature Gas-cooled Reactor (HTGR, sometimes also called HTR) is a kind of nuclear reactor, which are demonstrated very high efficiencies, safety and availability features by American and German power plant. HTR differs from water nuclear reactors by offering a high thermal efficiency for electricity generation and a high level of passive safety features. Now HTR-PM project is built in Shidao Bay of China. Moreover, HTR is the only nuclear reactor, which can provide high temperature steam comparing with other water nuclear reactors. So HTR can provide a versatile cogeneration solution for industry. In this paper, a case was studied, how to provide heat for a refinery and petro-chemical plant with HTR. Firstly, the energy need of a typical large chemical plant in china was investigated. Steam supply diagram of an oil refinery plant, which produced 10 million tons oil products and 1 million tons ethylene in China, was calculated. Secondly, technical feasibility of energy providing by HTR cogeneration plant was discussed. Extraction steam from HTR system was designed for the chemical plant. It would meet the requirement of steam supply for chemical plant and would replace the captive power plant, where coal was burning. The balance of steam, enthalpy and temperature was calculated. At last, economic evaluation for such cogeneration plants was carried out. The steam supply cost from captive coal power plant and HTR cogeneration plant was compared. Some economical conclusion was made from the discussion.

Numerical Simulation and Experiment of the Fire Protecting System of the Manipulator in High Temperature Environment

2019 ◽  
Author(s):  
Byeong Cheon Kim ◽  
Kyoungsik Chang
Keyword(s):  
Numerical Simulation ◽  
High Temperature ◽  
Fire Protection ◽  
Cooling System ◽  
Normal Operation ◽  
Hydraulic Pressure ◽  
Flexible Pipe ◽  
Ansys Fluent ◽  
Temperature Environment ◽  
High Temperature Environment

Abstract In the present work, the strategy for cooling the manipulator in high temperature environment is studied using both numerical and experimental methods. Since the manipulator is designed to operate in the environment with the maximum 250 °C temperature, fire protection system and the cooling system should be installed for normal operation of the manipulator. The para-aramid-filament with the thickness of 0.5 mm and Graphite felt with the thickness of 5.5mm is considered for fire protection suit and air blowing technique is applied for cooling the electronic circuit and hydraulic pressure cylinders. For numerical simulation, ANSYS Fluent V18.2 is adopted to simulate the convective heat transfer flows and the radiation with the model, S2S (Surface to surface). Two types of blowing techniques are considered, global blowing and local one. Even though the global blowing at the inlet is most effective for cooling system, so much amount of compressed air is required, which means that extra big compression system should be added in the system. The local blowing is applied to the component with small holes of the flexible pipe and the magnitude of the local blowing mass flow rate is 0.0166kg/s. The technique of local blowing is more effective than the global blowing for cooling the system. To validate numerical simulation, the model is tested within the hot temperature chamber whose mean temperature is approximately 250 °C.

Study on the Core Cooling Scheme After Accident Shutdown of the Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2012 ◽  
Author(s):  
Yanhua Zheng ◽  
Lei Shi ◽  
Fubing Chen
Keyword(s):  
High Temperature ◽  
Temperature Distribution ◽  
Normal Operation ◽  
Helium Temperature ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Decay Heat ◽  
The Core ◽  
Core Cooling ◽  
High Temperature Gas

One of the most important properties of the modular high temperature gas-cooled reactor is that the decay heat in the core can be carried out solely by means of passive physical mechanism after shutdown due to accidents. The maximum fuel temperature is guaranteed not to exceed the design limitation, so as to the integrity of the fuel particles and the ability of retaining fission product will keep well. Nonetheless, the auxiliary active core cooling should be design to help removing the decay heat and keeping the reactor in an appropriate condition effectively and quickly in case of reactor scram due to any transient and the main helium blower or steam generator unusable. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor, assuming that the core cooling will be started up 1 hour after the scram, different core cooling schemes are studied in this paper. After the reactor shutdown, a certain degree of natural convection will come into being in the core due to the non-uniform temperature distribution, which will accordingly change the core temperature distribution and in turn influence the outlet hot helium temperature. Different cooling flow rates are also analyzed, and the important parameters, such as the fuel temperature, outlet hot helium temperature and the pressure vessel temperature, are studied in detail. A feasible core cooling scheme, as well as the reasonable design parameters could be determined based on the analysis. It is suggested that, considering the temperature limitation of the structure material, the coolant flow direction should be same as that of the normal operation, and the flow rate could not be too large.

Development of Inherently Safe Technologies for Large Scale BWRs: (4) Hydrogen Explosion Prevention System Using SiC Fuel Claddings

2014 ◽  
Author(s):  
Ryo Ishibashi ◽  
Tomohiko Ikegawa ◽  
Kenji Noshita ◽  
Kazuaki Kitou ◽  
Mamoru Kamoshida
Keyword(s):  
High Temperature ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Hydrogen Gas ◽  
Metallic Layer ◽  
Normal Operation ◽  
Air Cooling ◽  
Composite Substrate ◽  
Prevention System ◽  
High Temperature Steam

In the aftermath of the lessons learned from the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), including a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents. One of inherently safe technologies currently under development is a system to prevent hydrogen explosion during severe accidents (SAs). This hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding of silicon carbide (SiC), and a passive autocatalytic recombiner (PAR). Replacing the zircaloy (Zry) claddings currently used in LWRs with the SiC claddings decreases the hydrogen generation and thus decreases the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B) such as an operation floor. The PAR recombines the leaked hydrogen gas so as to maintain the hydrogen concentration at less than the explosion limit of 4 % in the R/B. The advantages of using SiC claddings in the system were examined through experiments and SA analysis. Results of steam oxidation tests confirmed that SiC was estimated to show 2 to 3 orders of magnitude lower hydrogen generation rates during oxidation in a high temperature steam environment than Zry. Results of SA analysis showed that the total amount of hydrogen generation from fuels was reduced to one fifth or less. Calculation also showed that the lower heat of the oxidation reaction of SiC moderated the steep generation with the temperature increase. We expected this moderated steep generation to reduce the pressure increase in the PCV as well as prevent excess amounts of leaked hydrogen from hydrogen disposal rate capacity using PARs. The SiC cladding under consideration consists of an inner metallic layer, a SiC/SiC composite substrate, and an outer environment barrier coating (EBC). A thin inner metallic layer in combination with a SiC/SiC composite substrate functions as a barrier for fission products. EBC is introduced to have both corrosion resistance in high temperature water environments during normal operation and oxidation resistance in high temperature steam environments during SA. Further reduction of the hydrogen generation rate in high temperature steam by improving the EBC is expected to decrease the total amount of hydrogen generation even more.

Sensitivity Analysis of a PWR Response During a Loss of Coolant Accident Under a Hypothetical Core Blockage Scenario Using RELAP5-3D

2014 ◽  
Author(s):  
Timothy Crook ◽  
Rodolfo Vaghetto ◽  
Alessandro Vanni ◽  
Yassin A. Hassan
Keyword(s):  
Cooling System ◽  
Safety Concern ◽  
System Response ◽  
Flow Paths ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pass Through ◽  
Core Cooling

During a Loss of Coolant Accident (LOCA) a substantial amount of debris may be generated in containment during the blowdown phase. This debris can become a major safety concern since it can potentially impact the Emergency Core Cooling System (ECCS). Debris, produced by the LOCA break flow and transported to the sump, could pass through the filtering systems (debris bed and sump strainer) in the long term cooling phase. If the debris were to sufficiently accumulate at the core inlet region, the core flow could theoretically decrease, affecting the core coolability. Under such conditions, the removal of decay heat would only be possible by coolant flow reaching the core through alternative flow paths, such as the core bypass (baffle). There are certain plant specific features that can play a major role in core cooling from this bypass flow. One of these of key interest is the pressure relief holes. A typical 4-loop Pressurized Water Reactor (PWR) was modeled using RELAP5-3D to simulate the reactor system response during the phases of a large break LOCA and the effectiveness of core cooling under full core blockage was analyzed. The simulation results showed that the presence of alternative flow paths may significantly increase core coolability and prevent cladding temperatures from reaching safety limits, while the lack of LOCA holes may lead to a conservative over-prediction of the cladding temperature.

Operating Experience With Mechanical Seals in Reactor Shutdown Cooling Pumps

2009 ◽  
Author(s):  
G. Staniewski ◽  
A. Keshavarz ◽  
W. J. Ferguson
Keyword(s):  
High Pressure ◽  
Nuclear Reactor ◽  
Cooling System ◽  
Operating Experience ◽  
Normal Operation ◽  
Mechanical Seals ◽  
Reactor Inlet ◽  
Start Up ◽  
Service Water ◽  
Reactor Shutdown

The CANDU nuclear reactor has a dedicated shutdown cooling system (SDC) to remove heat from the primary heat transport system and transfer it to the high pressure service water open system. There are four individual SDC circuits, each connected to one of the four quadrants of the reactor. The four circuits are independent and each of them consists of one SDC pump and one heat exchanger. When the reactor is within normal operation, all four SDC loops are isolated but kept pressurized to the reactor inlet header of 9.7 MPa (1410 psi) and warmed up to 93°C. During reactor shutdown condition and at reactor start-up operation, the SDC pumps are in-service. They operate in several different configurations and are exposed to a number of different pressure conditions. This paper will present typical failures of the seal components.

Modeling Passive Safety in the Westinghouse Small Modular Reactor: Assessment of Wall Condensation and CMT Natural Circulation

2013 ◽  
Author(s):  
Jun Liao ◽  
Vefa N. Kucukboyaci
Keyword(s):  
Natural Circulation ◽  
Passive Safety ◽  
Reactor Safety ◽  
Heat Transfer Mechanism ◽  
The Core ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Small Modular Reactor ◽  
Loss Of Coolant ◽  
Safety Design

Passive safety design that utilizes gravity, natural circulation, heat sink and stored potential energy for reactor safety functions is being increasingly adopted in advanced reactors, especially in the small modular reactor (SMR) designs. The passive safety design of the Westinghouse SMR is described in details and compared with the AP1000® passive safety design. The natural circulation loops and heat transfer mechanism in a postulated Westinghouse SMR loss of coolant accident (LOCA) are discussed. The key thermal hydraulic phenomena pertinent to the passive safety design of the Westinghouse SMR have been identified in the small break LOCA Phenomena Identification and Rank Table (PIRT). Among the identified phenomena, condensation on the containment wall and natural circulation in core makeup tank (CMT) loop are highly ranked. Those passive safety phenomena are expected to be assessed using the WCOBRA/TRAC-TF2 LOCA thermal hydraulic code, which will provide the design basis LOCA analysis in the SMR design control documentation. In this paper, the progress on the assessing two key phenomena in passive safety of Westinghouse SMR is reported. The preliminary assessments against UCB tube condensation tests and Westinghouse core makeup tank tests reveals the capability of WCOBRA/TRAC-TF2 code to reasonably predict the condensation on the containment wall and natural circulation in the core makeup tank (CMT) loop.

scholarly journals Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

2019 ◽  
Vol 5 (4) ◽  
pp. 289-295 ◽  
Author(s):  
Olga I. Bulakh ◽  
Oleg K. Kostylev ◽  
Vladimir N. Nesterov ◽  
Eldar K. Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Alternative Energy ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Calculated Data ◽  
Fuel Burnup ◽  
The Core ◽  
High Temperature Reactors ◽  
High Temperature Gas

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

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