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.

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.

Charts for water hammer in pipelines resulting from valve closure from full opening only

1985 ◽  
Vol 12 (2) ◽  
pp. 241-264 ◽  
Author(s):  
Bryan W. Karney ◽  
Eugen Ruus
Keyword(s):  
Good Accuracy ◽  
Pipe Wall ◽  
Pressure Head ◽  
Wall Friction ◽  
Maximum Pressure ◽  
Valve Closure ◽  
Closure Time ◽  
Basic Parameters ◽  
Valve Opening ◽  
Open Position

Maximum pressure head rises, which result from total closure of the valve from an initially fully open position, are calculated and plotted for the valve end and for the midpoint of a simple pipeline. Uniform, equal-percentage, optimum, and parabolic closure arrangements are analysed. Basic parameters such as pipeline constant, relative closure time, and pipe wall friction are considered with closures from full valve opening only. The results of this paper can be used to draw the maximum hydraulic grade line along the pipe with good accuracy for the closure arrangements considered. It is found that the equal-percentage closure arrangement yields consistently less pressure head rise than does the parabolic closure arrangement. Further, the optimum closure arrangement yields consistently less head rise than the equal-percentage one. Uniform closure produces pressure head rise that usually lies between those produced by the parabolic and the equal-percentage closure arrangements, except for the range of low pressure head rise combined with low or zero friction, where the rise due to uniform closure approaches that produced by optimum closure.

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.

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.

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 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.

Comprehensive Seismic Evaluation of HTTR Against the 2011 Off the Pacific Coast of Tohoku Earthquake

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Masato Ono ◽  
Kazuhiko Iigaki ◽  
Hiroaki Sawahata ◽  
Yosuke Shimazaki ◽  
Atsushi Shimizu ◽  
...  
Keyword(s):  
Seismic Response ◽  
Pacific Coast ◽  
Tohoku Earthquake ◽  
Reactor Power ◽  
Great Earthquake ◽  
Response Analysis ◽  
Seismic Response Analysis ◽  
Control Rod ◽  
The Pacific ◽  
Equipment And Instruments

On Mar. 11, 2011, the 2011 off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the high temperature engineering test reactor (HTTR) had been stopped under the periodic inspection and maintenance of equipment and instruments. A comprehensive integrity evaluation was carried out for the HTTR facility because the maximum seismic acceleration observed at the HTTR exceeded the maximum value of design basis earthquake. The concept of comprehensive integrity evaluation is divided into two parts. One is the “visual inspection of equipment and instruments.” The other is the “seismic response analysis” for the building structure, equipment and instruments using the observed earthquake. All equipment and instruments related to operation were inspected in the basic inspection. The integrity of the facilities was confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the results of inspection of equipment and instruments associated with the seismic response analysis, it was judged that there was no problem for operation of the reactor, because there was no damage and performance deterioration. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013, and 2015. Additionally, the integrity of control rod guide blocks was also confirmed visually when three control rod guide blocks and six replaceable reflector blocks were taken out from reactor core in order to change neutron startup sources in 2015.

Rod Group and Individual Control System

2014 ◽  
pp. 320-351
Author(s):  
Yuri Rozen ◽  
Alexander Siora
Keyword(s):  
Control System ◽  
Remote Control ◽  
Reactor Power ◽  
Nuclear Safety ◽  
International Standards ◽  
Control Rod ◽  
Individual Control ◽  
Actuation System ◽  
The Individual

Chapter 10 considers the Rod Group and Individual Control (RG&IC) system, which is one of the individual I&C systems and a part of the reactor control and protection system. RG&IC is an actuation system, which performs functions initiated by emergency and preventive reactor protection, reactor power control, unloading, limitation and accelerated preventive protection, and remote control rod position commands sent by the power unit personnel. The central part of RG&IC system consists of software-hardware complex SHC RG&IC-R based on the equipment family of the Research and Production Corporation “Radiy” (RADIY PLATFORM – see Chapter 1). The RG&IC system combines functions that belong to A and B categories according to safety impact (IEC, 2009), relates to safety class 2(A) and complies with the fundamental safety principles (IAEA, 1999), requirements that are set forth in international standards (IAEA, 2002, 2012; IEC, 2011), and Ukrainian nuclear safety rules and regulations (NP, 2000, 2008a, 2008b).

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