Preliminary Neutronics and Thermal-Hydraulics Study on Thorium-Based HTR-PM With Outer Breeding Zone

2018 ◽  
Author(s):  
Qiudong Wang ◽  
Bing Xia ◽  
Jiong Guo ◽  
Ding She ◽  
Lei Shi ◽  
...  
Keyword(s):  
Heavy Metal ◽  
Reactor Core ◽  
Volume Ratio ◽  
Reactor Power ◽  
Normal Operation ◽  
Design Parameters ◽  
Fuel Particle ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Breeding Zone

In this work, a two-zone reactor core, which contains an inner driving zone and an outer ThO2 breeding zone, is designed under the framework of the HTR-PM. The main aim of this work is to investigate the feasibility of thorium utilization in the mature design of the HTR-PM with the inherent safety features. The neutronics and thermal-hydraulics characteristics are investigated to optimize the design parameters by using VSOP. The aim of optimization is to maximize the conversion of thorium to 233U in the breeding zone. The preliminary results indicate that the volume ratio of the breeding zone to the driving zone has significant influence on the power peaking factor and the maximum fuel temperature in normal operation and accidental conditions. On the other hand, the increase of reactor power will lead to increase of maximum fuel temperature after DLOFC accident. More heavy metal loading in the breeding zone will raise 233U yield, while the influence of fuel particle radius on the conversion ratio is negligible. An optimized 200 MWt two-zone reactor design is obtained with volume ratio of the driving zone to the breeding zone of 4:1, and 7 g and 30 g heavy metal per fuel sphere in the driving zone and the breeding zone, respectively.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

Characteristics of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor in Depressurized Loss of Coolant Accidents

2008 ◽  
Author(s):  
Yanhua Zhengy ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Safety Margin ◽  
Reactor Core ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Loss Of Coolant ◽  
High Temperature Gas ◽  
Residual Heat

Depressurized loss of coolant accident (DLOCA) is one of the most important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behavior during DLOCA can provide useful reference to the physics, thermo-hydraulic and structure designs of the reactor core. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), three cases of DLOCA: a instantaneous depressurization along with a flow coastdown and scram at zero time, a main pipe with a diameter of 65mm rupture, and a instrument pipe with a diameter of 10mm broken, are studied by the help of two different kinds of software THERMIX and TINTE. The key parameters of different cases including reactor power, temperature distribution of the core and pressure vessel, and the decay power removal by the passive residual heat remove system (RHRS) are compared in detail. Some uncertainties, such as residual heat calculation, power distribution, heat conductivity of fuel element, etc., are analyzed in order to evaluate the safety margin of the maximum fuel temperature during DLOCA. The calculating results show that, the decay heat in the DLOCA can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed. It also illustrates that the HTR-PM can reach 250MW reactor power per unit and still can keep the inherent safety.

scholarly journals Transient Study on the HTR-PM with TINTE-vPower Coupling Code Package

2020 ◽  
Vol 2020 ◽  
pp. 1-14
Author(s):  
Jun Sun ◽  
Ximing Sun ◽  
Yanhua Zheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Flow ◽  
Nuclear Power ◽  
Data Exchange ◽  
Reactor Power ◽  
Normal Operation ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Operation Conditions

The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.

Reactivity Accident Due to Control Rod Inadvertent Withdrawal of the 250MW High Temperature Gas-Cooled Reactor

2010 ◽  
Author(s):  
Yanhua Zheng ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Fission Products ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Control Rod ◽  
Detail Design ◽  
Fuel Particles ◽  
Reactivity Accident ◽  
High Temperature Gas

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) with single module power of 250MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper, e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not, and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature and Xenon concentration are studied and compared in detail between these different cases. The calculating results show that, the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor

2014 ◽  
Author(s):  
Yanhua Zheng ◽  
Fubing Chen ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Thermal Inertia ◽  
Hydraulic Calculation ◽  
Normal Operation ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Cooling Flow ◽  
Safety Features ◽  
High Temperature Gas

Pebble bed modular high temperature gas-cooled reactors (HTR), due to their characteristics of low power density, slender structure, large thermal inertia of fuel elements and reactor component materials (graphite), have good inherent safety features. However, the reflectors consisting of large piles of graphite blocks will form huge numbers of certain bypass gaps in the radial, axial and circumferential directions, thus affecting the effective cooling flow into the reactor core, which is one of the concerned issues of HTRs. According to the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the thermal-hydraulic calculation model is established in this paper. Based on this model, considering different bypass flow, that is to say, different core cooling flow, fuel element temperature, outlet helium temperature and the core pressure drop in the normal operation, as well as the maximal fuel temperature during the depressurized loss of forced cooling (DLOFC) accident are analyzed. This study on bypass effects on the steady-state and transient phases can further demonstrate the HTR safety features.

Design Optimization of PERCS in RELAP5 Using Parallel Processing and a Multi-Objective Non-Dominated Sorting Genetic Algorithm

2018 ◽  
Author(s):  
Paul R. Wilding ◽  
Nathan R. Murray ◽  
Matthew J. Memmott
Keyword(s):  
Genetic Algorithm ◽  
Parallel Processing ◽  
Pareto Front ◽  
Operating Time ◽  
Reactor Core ◽  
Normal Operation ◽  
Design Parameters ◽  
Outlet Temperature ◽  
Multi Objective Optimization ◽  
Multi Objective

Multi-objective optimization is a powerful tool that has been successfully applied to many fields but has seen minimal use in the design and development of nuclear power plant systems. When applied to design, multi-objective optimization involves the manipulation of key design parameters in order to develop optimal designs. These design parameters include continuous and/or discrete variables and represent the physical design specifications. They are modified across a specific design space to accomplish a number of set objective functions, representing the goals for both system design and performance, which conflict and cannot be combined into a single objective function. In this paper, a non-dominated sorting genetic algorithm (NSGA) and parallel processing in Python 3 were used to optimize the design of the passive endothermic reaction cooling system (PERCS) model developed in RELAP5/MOD 3.3. This system has been proposed as a retrofit to currently-operating light water reactors (LWR) and is designed to remove decay heat from the reactor core via the endothermic decomposition of magnesium carbonate (MgCO3) and natural circulation of the reactor coolant. The PERCS design is currently a shell-and-tube heat exchanger, with the coolant flowing through the tube side and MgCO3 on the shell side. During a station blackout (SBO), the PERCS initially keeps the reactor core outlet temperature from exceeding 635 K and then reduces it to below 620 K for 30 days. The optimization of the PERCS was performed with three different objectives: (1) minimization of equipment costs, (2) minimization of deviation of the core outlet temperature during a SBO from its normal operation steady-state value, and (3) minimization of fractional consumption of MgCO3, a metric that is measurable and directly related to the operating time of the PERCS. The manipulated parameters of the optimization include the radius of the PERCS shell, the pitch, hydraulic diameter, thickness and length of the PERCS tubes, and the elevation of the PERCS with respect to the reactor core. The NSGA methodology works by creating a population of PERCS options with varying design parameters. Using the evolutionary concepts of selection, reproduction, mutation, and survival of the fittest, the NSGA method repeatedly generates new PERCS options and gets rid of less fit ones. In the end, the result was a Pareto front of PERCS designs, each thermodynamically viable and optimal with respect to the three objectives. The Pareto front of options as a whole represents the optimized trade-off between the objectives.

Two-Way Coupling Between the Reactor Dynamics Code DYN3D and the Fuel Performance Code TRANSURANUS at Assembly Level

2014 ◽  
Author(s):  
L. Holt ◽  
U. Rohde ◽  
M. Seidl ◽  
A. Schubert ◽  
P. Van Uffelen ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Computational Cost ◽  
Normal Operation ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Fuel Performance ◽  
Control Rod ◽  
Reactor Dynamics ◽  
Coupling Approach ◽  
Fuel Behavior

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel thermal hydraulics code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in greater detail. Still these code systems lack a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. To our knowledge a two-way coupling to a fuel performance code hasn’t so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models. A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switching from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and the possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states. Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up. The numerical convergence for DYN3D-TRANSURANUS is quick and stable. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.

Reactivity Accident in a High Temperature Gas-Cooled Reactor Due to Inadvertent Withdrawal of Control Rod

2010 ◽  
Vol 133 (5) ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Fission Products ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Control Rod ◽  
Detail Design ◽  
Fuel Particles ◽  
Reactivity Accident ◽  
High Temperature Gas

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese pebble-bed modular high temperature gas-cooled reactor (HTR-PM) with single module power of 250 MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper (e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not) and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature, and xenon concentration are studied and compared in detail between these different cases. The calculating results show that the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way so that the temperature limits of the fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Dynamic Testing in Nuclear Power Plants for Model Validation

1976 ◽  
Vol 98 (3) ◽  
pp. 340-347 ◽  
Author(s):  
T. W. Kerlin ◽  
E. M. Katz ◽  
A. T. Chen ◽  
J. G. Thakkar ◽  
S. I. Chang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Dynamic Testing ◽  
Reactor Power ◽  
Normal Operation ◽  
Fuel Temperature ◽  
Variable Model ◽  
Pressurized Water ◽  
Frequency Responses

Dynamics tests were preformed at the Oconee pressurized water reactor to obtain information for checking a theoretical plant model. Low level, periodic reactivity perturbations were introduced and several system responses (reactor power, temperatures, pressures) were monitored. The data were processed off-line to give frequency responses. A linear state-variable model for the plant was formulated and used to compute theoretical frequency responses. A computerized, model-reference identification procedure was used to identify the fuel temperature coefficient of reactivity and the overall fuel-to-coolant heat transfer coefficient. The study showed that dynamic tests can be performed in operating nuclear power plants with insignificant interference to normal operation. Also, the use of automatic parameter identification procedures was demonstrated.

Integrated Multi-Modular Fast Reactor Concept Design

2012 ◽  
Author(s):  
Vishal Patel ◽  
Pavel Tsvetkov
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Negative Temperature ◽  
Thermal Hydraulics ◽  
Concept Design ◽  
Fuel Temperature ◽  
Feasibility Assessment ◽  
Balance Of Plant ◽  
Critical Reactor ◽  
Economics Analysis

The Integrated Multi-Modular Fast reactor is a pre-conceptual small modular fast reactor design consisting of 7 self-consist subcritical modules, each utilizing a BeO-MOX concept fuel with complete supercritical CO2 brayton cycle turbo-machinery. The subcritical modules, when brought into proximity of one another, form a complete critical reactor core. The feasibility of the reactor is assessed on a full core level, which includes a neutronics, thermal hydraulics, balance of plant, economics, and economics analysis. It has been shown that a critical configuration lasting for 14 years at 10 MWth can be achieved. A hot channel thermal hydraulics and safety analysis shows that the reactor can operate within safety limits with negative temperature coefficients of reactivity as well as stay within fuel temperature limits. A plant thermal efficiency of 36% was achieved and there is room for optimization to achieve higher efficiencies. An economical feasibility assessment shows that the reactor can be economical based on an economy of serial production argument. The analysis also leads to a licensing discussion.

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