Influencing Factors of the HTR-10 Core Dynamics in Reactivity Insertion ATWS Tests

2016 ◽  
Author(s):  
Feng Gou ◽  
Fubing Chen ◽  
Yujie Dong
Keyword(s):  
Influencing Factors ◽  
Reactor Power ◽  
Sensitivity Analyses ◽  
Simulation Accuracy ◽  
Control Rod ◽  
Poisoning Effect ◽  
Core Dynamics ◽  
New Energy ◽  
Full Power ◽  
Reactivity Insertion

After the full power operation of the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10), several safety demonstration tests, representing the anticipated transient without scram (ATWS) conditions, were successfully performed on this reactor. Among these tests, two reactivity insertion ATWS tests were conducted by withdrawing a single control rod without reactor scram at 30% rated power. In the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, these two tests have been reanalyzed using the THERMIX code, and the code itself was strictly checked through the test data. According to the previous code benchmark activities utilizing the HTR-10 tests, the temperature coefficient of reactivity (TCR), the residual heat level (RHL) and the xenon poisoning effect (XPE) could be considered the most important influencing factors of the THERMIX simulation accuracy for the core dynamics. In this study, sensitivity analyses are performed on the basis of the assumed variations of TCR, RHL and XPE. The impacts of these concerned parameters on the reactor power transient are qualitatively identified.

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.

Estimation of control rod worth, xenon effect on reactivity and power defect of BAEC TRIGA Mark-II research reactor

2018 ◽  
Vol 33 (39) ◽  
pp. 1850233
Author(s):  
Md. Mehedi Hassan ◽  
K. M. Jalal Uddin Rumi ◽  
Md. Nazrul Islam Khan ◽  
Rajib Goswami
Keyword(s):  
Theoretical Analysis ◽  
Low Power ◽  
Temperature Effects ◽  
Research Reactor ◽  
Power Level ◽  
Fuel Temperature ◽  
Control Rod ◽  
Full Power ◽  
Triga Mark Ii ◽  
Control Rods

In this work, control rod worth, xenon (Xe) effect on reactivity and power defect have been measured by doing experiments in the BAEC TRIGA Mark-II research reactor (BTRR) and through established theoretical analysis. Firstly, to study the xenon-135 effect on reactivity, reactor is critical at 2.4 MW for several hours. Next, experiments have been performed at very low power (50 W) to avoid temperature effects. Moreover, for the power defect experiment, different increasing power level has been tested by withdrawing the control rods. Finally, it is concluded that the total control rods worth of the BAEC TRIGA Mark-II research reactor, as determined through this study, is enough to run the reactor at full power (3 MW) considering the xenon-135 and fuel temperature effects.

RETRAN Application of Turbine Trip and Load Rejection of Startup Test Analysis for Lungmen ABWR

2009 ◽  
Author(s):  
Chen-Lin Li ◽  
Chiung-Wen Tsai ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Su-Chin Chung
Keyword(s):  
Control System ◽  
Water Level ◽  
Control Systems ◽  
Nuclear Power ◽  
Pressure Control ◽  
Reactor Power ◽  
Level Control ◽  
Bypass Flow ◽  
Control Rod ◽  
Recirculation Flow

This study used RETRAN program to analyze the turbine trip and load rejection transients of Taiwan Power Company Lungmen Nuclear Power Plant’s startup test at 100% power and 100% core flow operating condition. This model includes thermal flow control volumes and junctions, control systems, thermal hydraulic models, safety systems, and 1D kinetics model. In Lungmen RETRAN model, four steam lines are simulated as one line. There are four simulated control systems: pressure control system, water level control system, feedwater control system, and speed control system for reactor internal pumps. The turbine trip event, at above 40% power, triggers the fast open of the bypass valves. Upon the turbine trip, the turbine stop valves close. To minimize steam bypassed to the main condenser, recirculation flow is automatically runback and a SCRRI (selected control rod run in) is initiated to reduce the reactor power. The load rejection event causes the fast opening of the bypass valves. Steam bypass will sufficiently control the pressure, because of their 110% bypass capacity. A SCRRI and RIP runback are also initiated to reduce the reactor power. This study also investigated the sensitivity analysis of turbine bypass flow, runback rate of RIPS and SCRRI to observe how they affect fuel surface heat flux, neutron flux and water level, etc. The results show that turbine bypass flow has larger impacts on dome pressure than RIPS runback rate and SCRRI. This study also indicates that test criteria in turbine trip and load rejection transients are met and Lungmen RETRAN model is performing well and applicable for Lungmen startup test predictions and analyses.

TRACE/PARCS Analysis of Full Isolation Startup Test for Lungmen ABWR

2012 ◽  
Author(s):  
Ai-Ling Ho ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Chunkuan Shih
Keyword(s):  
Water Level ◽  
Coupling Model ◽  
Reactor Power ◽  
Signal Delay ◽  
Control Rod ◽  
Pump Speed ◽  
Closure Time ◽  
Delay Times ◽  
Open Position ◽  
Delayed Time

TRACE/PARCS coupling model of Lungmen NPP has been used to analyze the full isolation startup test, defined as a simultaneous full closure of all MSIVs (Main Steamline Isolation Valve). In this analysis, the full closure time of MSIVs was varied by 2 sec, 3sec (base), and 4 sec. As MSIV was closed to 90% full open position, a reactor power scram signal was initiated. For each of MSIV full closure time, the delayed time of scram signal to the start of control rod insertion was varied by base 0.09 sec, 0.04 sec, and 0.14 sec. Because of MSIVs closure, the increasing reactor dome pressure reached the setpoint of relief valves (RVs) and caused RVs to open. The 4 reactor internal pumps (RIPs) not connected to the M/G sets tripped and were runback to minimum pump speed when water level dropped to L3. The 6 RIPs connected to the M/G sets tripped if low water level (L2) or high reactor dome pressure (1140 psia) had been reached. Those important thermal parameters with various MSIV full closure times (2, 3, and 4sec) and various reactor scram signal delay times (0.09, 0.04, and 0.14 sec) are compared and shown in this paper.

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.

Theoretical Analysis of Step-Down of Hydraulic Cylinder Motion for Control Rod Hydraulic Drive System Based on Model II

2017 ◽  
Author(s):  
Qianfeng Liu ◽  
Yuzheng Li ◽  
Benke Qin ◽  
Bo Hanliang
Keyword(s):  
Hydraulic Drive ◽  
Hydraulic Cylinder ◽  
Energy Technology ◽  
Hydraulic Control ◽  
Transient Pressure ◽  
Control Rod ◽  
Tsinghua University ◽  
Motion Time ◽  
New Energy ◽  
Step Down

Hydraulic Control Rod Drive Technology (HCRDT) is a newly invented patent and Institute of Nuclear and New Energy Technology Tsinghua University own HCRDT’s independent intellectual property rights. The hydraulic cylinder is the key part of this technology, so the performance of the hydraulic cylinder directly affects the HCRDT. Firstly, the theoretical model of the cylinder hydraulic has been obtained and verified by the experiment. Second, the step-down process of the cylinder hydraulic is analyzed. The results are shown that the model can analyze the performance of the cylinder, including the motion time of the cylinder, the transient pressure of the cylinder arrival, the transient impact energy of the cylinder arrival. At last, the cylinder and the drive mechanism can be optimized based on the result.

scholarly journals Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

2017 ◽  
Vol 2017 ◽  
pp. 1-25 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Light Water Reactors ◽  
Control Rod ◽  
Light Water ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

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