scholarly journals An improved control rod selection algorithm for core power control at TRIGA PUSPATI Reactor

2020 ◽  
Vol 14 (1) ◽  
pp. 6362-6379 ◽  
Author(s):  
Mohd Sabri Minhat ◽  
Nurul Adilla Mohd Subha ◽  
Fazilah Hassan ◽  
Norjulia Mohamad Nordin
Keyword(s):  
Power Control ◽  
Reactor Core ◽  
Reactor Power ◽  
Fine Tuning ◽  
Selection Algorithm ◽  
Acceptable Error ◽  
Control Rod ◽  
The Core ◽  
Significant Difference ◽  
Point Kinetics

The 1 MWth TRIGA PUSPATI Reactor known as RTP undergoes more than 37 years of operation in Malaysia. The current core power control utilized Feedback Control Algorithm (FCA) and a conventional Control Rod Selection Algorithm (CRSA). However, the current power tracking performance suffers and increase the workload on Control Rod Drive Mechanism (CRDM) if the range between minimum and maximum rod worth value for each control rod has a significant difference. Thus, it is requiring much time to keep the core power stable at the power demand value within the acceptable error bands for the safety requirement of the RTP. In conventional CRSA, regardless of the rod worth value, the lowest position of the control rod is selected for up-movement to regulate the reactor power with 2% chattering error. To improve this method, a new CRSA is introduced named Single Control Absorbing Rod (SCAR). In SCAR, only one rod with highest reactivity worth value will be selected for coast tuning during transient and the lowest reactivity worth value will be selected for fine-tuning rod movement during steady-state. The simulation model of the reactor core is represented based on point kinetics model, thermal-hydraulic models and reactivity model. The conventional CRSA model included with control rod position dynamic model and actual reactivity worth curve data from RTP. The FCA controller is designed based on Proportional-Integral (PI) controller using MATLAB Simulink simulation. The core power control system is represented by the integration of a reactor core model, CRSA model and FCA controller. To manifest the effectiveness of the proposed SCAR algorithm, the results are compared to the conventional CRSA in both simulation and experimentation. Overall, the results shows that the SCAR algorithm offers generally better results than the conventional CRSA with the reduction in rising time up to 44%, workload up to 35%, settling time up to 26% and chattering error up to 18% of the nominal value.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs

2012 ◽  
Author(s):  
Kuniyoshi Takamatsu ◽  
Kazuhiro Sawa
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Control ◽  
Reactor Power ◽  
Reactor Kinetics ◽  
The Core ◽  
Power Control System ◽  
Stop Test ◽  
High Temperature Reactor ◽  
Forced Cooling

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results. We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.

The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core

2010 ◽  
Author(s):  
Minggang Lang ◽  
Yujie Dong
Keyword(s):  
Gas Turbine ◽  
Cooling System ◽  
Negative Temperature ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
System Pressure ◽  
Temperature Limitation

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1242.4°C. It was a little higher than 1230°C–the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.

The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core Under the Loss of the System Pressure

2014 ◽  
Author(s):  
Minggang Lang
Keyword(s):  
Gas Turbine ◽  
Cooling System ◽  
Negative Temperature ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
Design Basis ◽  
System Pressure

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and to use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basis accidents and beyond design basis accidents of HTR10-GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake under the loss of the system pressure. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted. At the same time, the system pressure was supposed to lose by some reason. Thus the reactor power increased and the temperature of the core increased. And the protection system warns with two scram signal: too high of the negative varying rate of the system pressure and too high of the reactor power, which should induce the reactor to scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure continued to increase but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1203.4°C. It was a little bit lower than 1230°C — the fuel temperature limitation of HTR-10 and there is no release of any radioactivity. So the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10.

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

scholarly journals Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Energies ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 7377
Author(s):  
Michał Górkiewicz ◽  
Jerzy Cetnar
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Reactor Performance ◽  
Fission Rate ◽  
Control Rod ◽  
Structured Design ◽  
The Core ◽  
Stable Core ◽  
Control Rods ◽  
Burnup Calculation

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

scholarly journals Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling

2019 ◽  
Vol 5 (4) ◽  
pp. 297-303
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Temperature Field ◽  
Mass Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Cfd Modeling ◽  
Problem Solution ◽  
Conductivity Equation ◽  
The Core ◽  
Significant Difference

Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

PWR MOX/UO2 Transient Benchmark Calculation Using Monte Carlo Serpent 2 Code and Open Nodal Core Simulator ADPRES

2020 ◽  
Author(s):  
Muhammad Imron ◽  
Donny Hartanto
Keyword(s):  
Monte Carlo ◽  
Diffusion Coefficients ◽  
Power Distribution ◽  
Reactor Core ◽  
Reactor Power ◽  
Monte Carlo Code ◽  
Step Method ◽  
Control Rod ◽  
Benchmark Calculation ◽  
Transient Solutions

Abstract This paper presents static and transient solutions for the PWR MOX/UO2 transient benchmark by Serpent 2 Monte Carlo code and open nodal core simulator called ADPRES. The presences of MOX fuels and burn-up variation in the benchmark’s reactor core pose challenges for reactor simulators due to severe flux gradient across fuel assemblies. In this work, the two-step method was used, in which the assembly level two-group constants were generated from single assembly calculations with zero net current boundary conditions using Serpent 2 Monte Carlo code, and later the core calculation was performed using ADPRES open nodal core simulator. Two types of diffusion coefficients were generated: the conventional B1 leakage corrected and Cumulative Migration Method (CMM). Finally, the solutions of Serpent 2/ADPRESS, including multiplication factor, power distribution, control rod worth, and critical boron concentration using both diffusion coefficients were compared against solutions from heterogeneous Serpent 2 calculations where the fuel and cladding are explicitly modeled. The reactor power during transients were also compared qualitatively against other nodal core simulators. The results showed that Serpent 2/ADPRES were able to predict the heterogeneous Monte Carlo solutions very well with reasonable differences. The transient solutions were also quite accurate compared to other nodal core simulators. As for the diffusion coefficients comparison, it was found that the CMM diffusion coefficient provide more accurate solutions for the benchmark compared to the B1 leakage corrected diffusion coefficients.

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