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.

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.

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 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.

Behavior of solid fission products in coated fuel particles of a high-temperature gas-cooled reactor

1989 ◽  
Vol 67 (2) ◽  
pp. 594-599 ◽  
Author(s):  
A. N. Gudkov ◽  
V. A. Kashparov ◽  
A. A. Kotlyarov ◽  
N. N. Ponomarev-Stepnoi ◽  
I. G. Prikhod'ko ◽  
...  
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Fuel Particles ◽  
High Temperature Gas

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10

2008 ◽  
Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Power Level ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Code System ◽  
Safety Features ◽  
Forced Cooling ◽  
High Temperature Gas

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

scholarly journals Bypass flow in small absorber sphere channels of the high-temperature gas-cooled reactor pebble-bed module

2017 ◽  
Vol 10 (3) ◽  
pp. 128-139 ◽  
Author(s):  
Ziping Liu ◽  
Zeguang Li ◽  
Jun Sun
Keyword(s):  
High Temperature ◽  
Network Model ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Bypass Flow ◽  
Pebble Bed ◽  
Flow Network ◽  
Control Rod ◽  
High Temperature Gas

In the high-temperature gas-cooled reactor pebble-bed module, the helium bypass flow among graphite blocks cannot be ignored due to its effect on the temperature distribution as well as the maximum temperature in the reactor core. Bypass flow was previously analyzed in the discharging tube, in vertical gaps between graphite reflectors, and in control rod channels. The focus of this study is on the bypass flow that connects the small absorber sphere channels. Different from bypass flow connecting the control rod channels, there was no evident inlet or outlet flow paths into or out of the small absorber sphere channels at the top or bottom of the reactor core. Therefore, the bypass flow connecting the pebble bed with the small absorber sphere channels was mainly caused by the horizontal gaps, in which those gaps would also be irregular due to installation, thermal expansion, or irradiation of the graphite reflectors. After clarifying the resistant coefficients of those gaps by computational fluid dynamic tools, the bypass flow distribution was calculated by the flow network model including the flow in the reactor core, small absorber sphere channels, as well as horizontal gaps. Cases with various size combinations of gaps were adopted into the flow network model to test the sensitivity of bypass flow distribution to those parameters. Finally, the bypass flow in the small absorber sphere channels was concluded to be not significant in the reactor core.

Study on the Rod Drop Performance of High-Temperature Gas-Cooled Reactor

2014 ◽  
Author(s):  
He Yan ◽  
Xingzhong Diao
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Insertion Time ◽  
Operational Mode ◽  
Control Rod ◽  
Rod System ◽  
Key Factor ◽  
Control Rods ◽  
High Temperature Gas

In this paper, the theoretical study and experimental investigation on the rod drop performance of high-temperature gas-cooled reactor (HTGR) pebble-bed module have been presented. The control rod drive mechanisms (CRDMs), serving as the first shutdown system of the reactor, are positioned above the reactor pressure vessel. When the reactor is operated at the power regulation mode, the control rods are pulled up-and-down in their channels around the reactor core. The CRDM provides a fail-safe operational mode for the control rod system. If the reactor emergency shutdown is required the control rods could drop into their channels by gravity. Thus the key factor, emergency insertion time of the whole control rod stroke, which represents the inherent safety of the CRDM, is crucially important and should be measured precisely. In the final objective of ensuring reliability of the CRDM, a full size drive line had been built and tested to obtain the overall performance function of the CRDM. Every component of the CRDM test line was simulated at the scale 1:1, including a 15 meters high test bench that was used as the substitution of the pressure vessel. At current stage, the rod drop performance had been experimental investigated at ambient temperature and pressure. The emergency insertion time of an 8 meters stroke was measured to be less than 50 seconds. A mathematical model of CRDM also had been developed. The rod motion characteristic equations show that the rod dropping speed approaches to a constant during the emergency insertion. The theoretical results are in agreement with the test results.

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 of the Gas Sampling and Analyzing System of HTR-PM

2017 ◽  
Author(s):  
Jiang Zhu ◽  
Feng Xie
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Fission Products ◽  
Structural Materials ◽  
Primary Coolant ◽  
Primary Loop ◽  
Gas Sampling ◽  
Purification System ◽  
Safety Features ◽  
High Temperature Gas

The high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is a modular high temperature gas-cooled reactor demonstration power plant, is characterized by inherent safety features and high generating efficiency. It adopts numerous graphite for structural materials in the reactor core, helium as primary coolant, and tristructural isotropic (TRISO) coated particles embedded in the graphite matrix as fuel elements. However, at high temperature the impurities in the helium can react with the graphite to cause corrosion of structural materials. Therefore, it is very necessary to monitor and control the composition and content of gaseous impurities in the primary coolant. In HTR-PM, the gas sampling and analyzing system has been designed to sample the primary helium at different positions in the helium purification system which is used to reduce the quantity of chemical impurities and remove the radioactive dust and gaseous fission products in the primary loop, and monitor the gas composition and individual concentration online. In the current paper, the composition of the gaseous impurities which need to be monitored in the primary loop of HTR-PM is presented, the design of the gas sampling positions in the helium purification system is discussed, and the main gas analyzing instruments are introduced.

scholarly journals Temperature Response of the HTR-10 during the Power Ascension Test

2015 ◽  
Vol 2015 ◽  
pp. 1-13 ◽  
Author(s):  
Fubing Chen ◽  
Yujie Dong ◽  
Zuoyi Zhang
Keyword(s):  
High Temperature ◽  
Temperature Response ◽  
Temperature Limit ◽  
Reactor Power ◽  
Helium Pressure ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Comparison Results ◽  
Safety Features ◽  
High Temperature Gas

The 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-Cooled Reactor in China. With the objective of raising the reactor power from 30% to 100% rated power, the power ascension test was planned and performed in January 2003. The test results verified the practicability and validity of the HTR-10 power regulation methods. In this study, the power ascension process is preliminarily simulated using the THERMIX code. The code satisfactorily reproduces the reactor transient parameters, including the reactor power, the primary helium pressure, and the primary helium outlet temperature. Reactor internals temperatures are also calculated and compared with the test values recorded by a number of thermocouples. THERMIX correctly simulates the temperature variation tendency for different measuring points, with good to fair agreement between the calculated temperatures and the measured ones. Based on the comparison results, the THERMIX simulation capability for the HTR-10 dynamic characteristics during the power ascension process can be demonstrated. With respect to the reactor safety features, it is of utmost importance that the maximum fuel center temperature during the test process is always much lower than the fuel temperature limit of 1620°C.

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