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 After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased.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.

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

2006 ◽  
Author(s):  
Mingang Lang ◽  
Yujie Dong
Keyword(s):  
Gas Turbine ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
Positive Reactivity ◽  
New Energy ◽  
Test Reactor ◽  
Intrinsic Safety

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, INET has planned 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 basic 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 accident on 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 lager than 800°C, the reactor was scrammed. During the accident sequence the maximum fuel temperature was 1200.9°C. It was lower than the fuel temperature limitation of 1230°C. The paper 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 when the fuel safety limits are the same.

A Study on an Evaluation of Indirect Cycle Plant Systems for HTGR

2008 ◽  
Author(s):  
Isao Minatsuki ◽  
Sunao Oyama ◽  
Yorikata Mizokami ◽  
Bernard Ballot
Keyword(s):  
Gas Turbine ◽  
Reactor Core ◽  
System Optimization ◽  
Reactor Power ◽  
Combined Cycle ◽  
Point Of View ◽  
Outlet Temperature ◽  
Optimization Study ◽  
Gas Turbine Cycle ◽  
Plant Efficiency

In the world now, several types of indirect system concept have been investigated for the High Temperature Gas cooled Reactor power plant (HTGR). From a point of optimization of HTGR, it is important to investigate and to compare their power conversion systems from a technical and an economical view point. In the first step of this study, an indirect steam cycle (ID-SC), an indirect gas turbine cycle (ID-GT), an indirect gas turbine combined cycle (ID-CCGT) and a direct gas turbine cycle (D-GT) has been chosen as the systems to be compared. The followings are chosen items for comparison analysis: a) Plant efficiency; b) Amount of commodities (which can estimate capital cost); c) Flexibility of reactor core design; d) Technical issues to be developed; e) Compatibility with hydrogen production system, etc. And for the second step, as the system optimization study among the selected system, sensitiveness to plant efficiency by changing the inlet and the outlet temperature of reactor core has been investigated from an economical and plant efficiency point of view.

Safety Evaluation of the HTTR-IS Nuclear Hydrogen Production System

2010 ◽  
Author(s):  
Hiroyuki Sato ◽  
Hirofumi Ohashi ◽  
Yujiro Tazawa ◽  
Nariaki Sakaba ◽  
Yukio Tachibana
Keyword(s):  
Hydrogen Production ◽  
Production System ◽  
Evaluation Method ◽  
Cooling System ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Fuel Temperature ◽  
System Pressure ◽  
Production Of Hydrogen ◽  
The Government

Establishment of a safety evaluation method is one of the key issues for the nuclear hydrogen production demonstration since fundamental differences in the safety philosophy between nuclear plants and chemical plants exist. In the present study, a practical safety evaluation method, which enables to design, construct and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification for the HTTR-IS nuclear hydrogen production system is conducted in order to select abnormal events which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the indentified five Anticipated Operational Occurrences (AOOs) and three ACciDents (ACDs) show that evaluation items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case, and therefore the reactor core was not damaged and cooled sufficiently. These results will contribute to the safety review from the government and demonstration of the nuclear production of hydrogen.

Research on High Temperature Combustor for Advanced Reheat Gas Turbine

1986 ◽  
Author(s):  
K. Mori ◽  
J. Kitajima ◽  
T. Kimura ◽  
T. Nakamura
Keyword(s):  
High Temperature ◽  
Gas Turbine ◽  
Thermal Barrier Coating ◽  
Ceramic Tile ◽  
Cooling System ◽  
Thermal Barrier ◽  
Outlet Temperature ◽  
National Project ◽  
Barrier Coating ◽  
Double Wall

Preliminary studies on the high temperature combustors (combustor outlet temperature = 1670–1770K) for the prototype advanced reheat gas turbine were carried out under the national project of Japan - Moonlight Project. This paper describes an outline of the studies; (i) application of an advanced ceramic thermal barrier coating and ODS superalloy, (ii) development of the advanced double-wall cooling system, (iii) research on the ceramic tile combustor.

Thermal Evaluation for a Ultra High Temperature Reactor With Pebble Type Fuels

2009 ◽  
Author(s):  
Motoo Fumizawa ◽  
Yuta Kosuge ◽  
Hidenori Horiuchi
Keyword(s):  
High Temperature ◽  
Power Density ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Small Scale ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
System Pressure ◽  
High Temperature Reactor ◽  
Ultra High Temperature

This study presents a predictive thermal-hydraulic analysis with packed spheres in a nuclear gas-cooled reactor core. The predictive analysis considering the effects of high power density and the some porosity value were applied as a design condition for an Ultra High Temperature Reactor (UHTR). The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 °C was achieved in the case of an UHTREX at LASL, which was a small scale UHTR using hollow-rod as a fuel element. In the present study, the fuel was changed to a pebble type, a porous media. Several calculation based on HTGR-GT300 through GT600 were 4.8 w/cm3 through 9.6 w/cm3 respectively. As a result, the relation between the fuel temperature and the power density was obtained under the different system pressure and coolant outlet temperature. Finally, available design conditions are selected.

scholarly journals Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes

2020 ◽  
Vol 2020 ◽  
pp. 1-7
Author(s):  
Van Khanh Hoang
Keyword(s):  
Safety Margin ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Control Rod ◽  
The Core ◽  
Small Modular Reactor ◽  
Active Core ◽  
And Performance

This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

A Neutronic Evaluation of the HTR-10 Using Scale, MCNPX and MCNP5 Nuclear Codes

2014 ◽  
Author(s):  
Rômulo V. Sousa ◽  
Clarysson A. M. Silva ◽  
Ângela Fortini ◽  
Cláubia Pereira ◽  
Maria Auxiliadora F. Veloso ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Heat Removal ◽  
Fission Products ◽  
Negative Temperature ◽  
Electricity Production ◽  
Three Dimension ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Coated Particles ◽  
The Core

The HTR-10 (High Temperature Gas-cooled Test Reactor) is a 10 MW modular pebble bed type reactor, built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China. As an advanced reactor, it has good passive safety characteristics: capacity of retaining all fission products inside the coated particles (up to 1,600° C), passive decay heat removal, large heat capacity of the core to mitigate temperature transition, large fuel temperature margin and negative temperature reactivity coefficient sufficient to accommodate reactivity insertion and small amount of excess reactivity in the core. This reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles, is used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and cogeneration of heat, and provide experience in PBR design, operation and construction. Using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation), the MCNPX 2.6.0 (Monte Carlo N-Particle eXtended) and the MCNP 5 (Monte Carlo N-Particle) nuclear codes, the HTR-10 first critical core described in the Evaluation of The Initial Critical Configuration of The HTR-10 Pebble-Bed Reactor was modeled and analyzed. A three-dimension model was simulated and the keff was obtained and compared with the reference. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel.

scholarly journals ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM

2016 ◽  
Vol 19 (2) ◽  
pp. 75
Author(s):  
Syarip, Khoirul Anam, Dwi Priyantoro
Keyword(s):  
High Temperature ◽  
Power Reactor ◽  
Operation Time ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Reactor Operation ◽  
Control Rod ◽  
Negative Reactivity ◽  
Control Rods ◽  
High Temperature Gas

ANALISISPENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. ABSTRAK ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM. Telah dilakukan analisis simulasi pengaturan posisi batang-batang kendali untuk melanjutkan operasi reaktor daya eksperimental (RDE) paska scram setelah beroperasi pada periode waktu tertentu. Pengendalian reaktivitas pada reaktor RDE yang akan dibangun di Indonesia dengan rujukan high temperature gas reactor (HTR) 10 MWt, dilakukan dengan 10  pasang batang-batang kendali atau control rod (CR). Apabila terrjadi kondisi abnormal maka CR secara otomatis akan jatuh tersisip ke dalam reflektor  reaktor sehingga reaktor scram dan berada pada kondisi subkritis. Untuk melanjutkan operasi reaktor pasca scram diperlukan analisis terkait pengaruh reaktivitas negatif dari Xenon dan suhu. Pada makalah ini disajikan hasil simulasi yang dilakukan untuk penentuan posisi CR paling optimum untuk melanjutkan operasi reaktor, menggunakan simulator PCTRAN-HTR. Simulasi dilakukan pada variasi 70%, 85% dan 100% dari tingkat daya penuh dan dengan variasi waktu operasi 50 s, 10.000 s, dan 20.000 s di mana setelah reaktor beroperasi pada tingkat-tingkat daya dan waktu operasi tersebut reaktor mengalami scram. Untuk melanjutkan operasi lagi maka CR harus dinaikkan lagi dan diatur ke posisi tertentu sampai   reaktor mencapai kondisi kritis lagi pada tingkat daya nominal tersebut. Hasil yang telah diperoleh menunjukkan bahwa dengan posisi CR naik 52 % sudah bisa menghasilkan kondisi kritis dan mampu mengatasi reaktivitas negatif peracunan xenon maupun suhu. Kata kunci: RDE, HTR, operasi reaktor, batang kendali, reaktivitas, scram ABSTRACT POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. Analytical study using PC-based simulator has been carried out on control rods position adjustment of the Indonesian experimental power reactor concept or reaktor daya ekperimental (RDE) in a post reactor scram to continue operation after a certain operation period. Reactivity control of the RDE uses 10 pairs of control rods (CRs), which is based on that applied in the high temperature gas reactor (HTR) 10 MW(t). If an abnormal operating condition occurs, these control rods automatically dropped to the reflector that bring the reactor into a scram and subcritical condition. To continue reactor operation after a period of time, the CRs should be withdrawn to achieve recriticality. Prior to any CRs withdrawal, an analysis of negative reactivity effects of Xenon (poissoning) and fuel temperature coefficient should be done. Simulations using PCTRAN-HTR simulator to determine the optimum CRs positions in achieving reactor criticality for continuation of reactor operation is presented in this paper. The simulations were conducted by varying the reactor power levels at 70%, 85% and 100% of full power, respectively. The reactor operation time was varied at 50s, 10000s, and 20000 s prior to the reactor scram. Adjustment of CRs position should be done to continue reactor operation at those nominal power levels by withdrawing the CRs to the proper positions. The simulation results show that recriticality can be achieverd by whitdrawing the CRs 52% of farther and the negative reactivity from xenon poisoning and temperature could be overcome. Keywords : RDE, HTR, reactor operation, control rod, reactivity, scram.

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.

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