Numerical study on the effect of the PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code
-
Cheng-Lun Yu
and
Chen-Hsien Chen
Abstract
In this work, the boiling water reactor (BWR) was adopted as a physical model to study the influence of PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code. The designed reactor pressure can be simulated through the open area of a valve modified by PI controller. The proportional gain (kp) is equal to one in all cases. For the integral gain (ki) smaller than or equal to 100 (i.e., ki = 0.1, 1, 10 and 100), a scrammed reactor incurred by the high reactor water level (Level-8) that is due to the large reactor pressure drop does not occur in the simulation. Compared to ki = 1, 10 and 100, for ki = 0.1, the reactor pressure modified by the PI controller is more close to the designed reactor pressure; however, the time to meet the designed reactor pressure is longer. The reason is that ki = 1, 10 and 100 incur a larger overshoot in the modified reactor pressure through the feedback system; although the time to meet the designed reactor pressure can be shortened, the amplitude of the reactor pressure varying with time is obvious. To exactly simulate the reactor pressure of a nuclear power plant under the normal operation, our results can offer MAAP code users an important reference to design the PI controller.
-
Research ethics: Not applicable.
-
Author contributions: The authors have accepted responsibility for the entire content of this manuscript and approved its submission.
-
Competing interests: The authors state no conflict of interest.
-
Research funding: None declared.
-
Data availability: The raw data can be obtained on request from the corresponding author.
Appendix A
IF TIM > 0.0
Error = (Pset − PPS)/Pset
SumP = kp*Error
Sum = SumP + SumI
ASRV(1) = 0.01*Sum
REPEAT
END
in which TIM represents time (s); Pset and PPS represent the expected and actual reactor pressure (pa), respectively; TD is the time step (s); kp and ki represent the proportional and integral gain, respectively; ASRV is the open area of a valve (m2). Note that TIM, PPS, and ASRV are the default parameters in MAAP 5.04 code.
References
Anagnost, J.J. and Desoer, C.A. (1991). An elementary proof of the Routh–Hurwitz stability criterion. Circuits Syst. Signal Process 10: 101, https://doi.org/10.1007/BF01183243.Search in Google Scholar
Chen, Y.S. and Huang, L.Y. (2021). Evaluation of the reactor building environmental conditions under a LOCA for the decommissioning Chinshan plant. Ann. Nucl. Energy 156: 108205, https://doi.org/10.1016/j.anucene.2021.108205.Search in Google Scholar
Golnaraghi, F., and Kuo, B.C. (2017). Automatic control systems. McGraw-Hill, New York.Search in Google Scholar
Shih, Y.H., Wang, S.J., Chuang, K.C., and Huang, T.E. (2014). A study of the containment venting strategy for the Fukushima Daiichi accident. Nucl. Technol. 186: 340, https://doi.org/10.13182/NT12-145.Search in Google Scholar
Taiwan Power Company (2007). Final safety analysis Report for Kuosheng nuclear power station units 1 and 2.Search in Google Scholar
USA Electric Power Research Institute (EPRI) (2008). User’s manual volume II (code structure and theory).Search in Google Scholar
Wang, T.C., Wang, S.J., and Teng, J.T. (2005). Analysis of severe accident management strategy for a BWR-4 nuclear power plant. Nucl. Technol. 152: 253, https://doi.org/10.13182/NT05-A3674.Search in Google Scholar
Wang, Y.C., Wang, S.J., Chien, C.S., and Chiang, S.C. (2010). Analysis of pressurizer level control system using MAAP4 code. Nucl. Eng. Des. 240: 160, https://doi.org/10.1016/j.nucengdes.2009.10.001.Search in Google Scholar
Xia, Z. and Zhang, D. (2012). The analysis of PWR SBO accident with RELAP5 based on linux. Phys. Procedia 24: 1329–1334, https://doi.org/10.1016/j.phpro.2012.02.198.Search in Google Scholar
Yu, C.L. and Wang, T.C. (2019). Numerical study on cut-off diameter of aerosol particle for filtered containment venting system in nuclear power plant. Kerntechnik 84: 48, https://doi.org/10.3139/124.110971.Search in Google Scholar
Yu, C.L., Wang, T.C., and Liu, C.Y. (2019a). Numerical investigation on the influence of the type of PI controller combined with MAAP code on the steady-state performance of a nuclear power plant. Nucl. Eng. Des. 343: 248, https://doi.org/10.1016/j.nucengdes.2019.01.004.Search in Google Scholar
Yu, C.L., Wang, T.C., Liu, C.Y., and Wang, W.S. (2019b). Numerical study on PI controller combined with MAAP code for simulating the steady-state core flow of a nuclear power plant. In: 27th international conference on nuclear engineering, May 19–24, 2019, Ibaraki.10.1299/jsmeicone.2019.27.1325Search in Google Scholar
© 2024 Walter de Gruyter GmbH, Berlin/Boston
Articles in the same Issue
- Frontmatter
- Numerical study on the effect of the PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code
- Analyses of the unavailability dynamics of emergency core cooling system
- Study on spent fuel heatup during spent fuel pool complete loss of coolant accident
- Numerical simulation analysis of high-temperature bent sodium heat pipes
- Influence of the twisting and nano fluids on performance of a triangular double tube heat exchanger
- Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method
- Gain scheduled internal model control based on the dynamic sliding mode method for the water level of nuclear steam generators
- Verification and validation optimization method for signal quality bits in digital control system application software of nuclear power plant
- Investigation of Li–Be and B halides as blanket in future fusion molten salt reactor
- A study on porosity investigation of compacted bentonite in various densities by using micro-computed tomography images analysis
- CTAB modification bentonite for enhanced Re adsorption and diffusion suppression
- Study on advection–dispersion behavior for simulation of 3H, 99Tc, and 90Sr transport in crushed sandstone of column experiments
- Investigating advection–dispersion behavior for simulation of HTO and 238Pu transport in argillaceous shale with different varying degrees of weathering
- Study on analysing the potential benefits of utilizing nuclear waste for biodiesel production
- Calendar of events
Articles in the same Issue
- Frontmatter
- Numerical study on the effect of the PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code
- Analyses of the unavailability dynamics of emergency core cooling system
- Study on spent fuel heatup during spent fuel pool complete loss of coolant accident
- Numerical simulation analysis of high-temperature bent sodium heat pipes
- Influence of the twisting and nano fluids on performance of a triangular double tube heat exchanger
- Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method
- Gain scheduled internal model control based on the dynamic sliding mode method for the water level of nuclear steam generators
- Verification and validation optimization method for signal quality bits in digital control system application software of nuclear power plant
- Investigation of Li–Be and B halides as blanket in future fusion molten salt reactor
- A study on porosity investigation of compacted bentonite in various densities by using micro-computed tomography images analysis
- CTAB modification bentonite for enhanced Re adsorption and diffusion suppression
- Study on advection–dispersion behavior for simulation of 3H, 99Tc, and 90Sr transport in crushed sandstone of column experiments
- Investigating advection–dispersion behavior for simulation of HTO and 238Pu transport in argillaceous shale with different varying degrees of weathering
- Study on analysing the potential benefits of utilizing nuclear waste for biodiesel production
- Calendar of events
