### PREDICTION OF AP1000’S NUCLEAR REACTOR PRESSURE VESSEL TEMPERATURE DURING NORMAL OPERATION

DOI: http://dx.doi.org/10.17146/tdm.2022.24.3.6684

#### Abstract

Modeling of thermal-hydraulic calculations for the AP1000 core to predict the reactor pressure vessel (RPV) temperature has been carried out. The reactor’s primary coolant system transfers the heat produced in the reactor fuel during reactor operation to the steam generator. Part of the heat will also be transferred from the coolant to the reactor vessel and the pipe. This paper presents the calculation result of the RPV temperature prediction during AP1000 normal operation. Calculations were performed using COBRA-EN code for analyzing the core thermal hydraulics and using analytics for predicting the RPV temperature. These methods were carried out with the aim to predict the RPV temperature as well as at steady state nominal power conditions, at the function of flow, and at power fluctuation conditions. The calculation results at nominal power 3400 MWt (100% heat generated in fuel was assumed) and thermal design flow with 10% tube plugging (TDF2) of 48,443.7 ton/hr, for the minimum system pressure of 15.1 MPa, nominal system pressure of 15.513 MPa, and design system pressure of 17.133 MPa, show that the core outlet coolant temperature is 326.96°C, 327.01°C, and 327.22°C, and the RPV temperature is 303.65°C, 303.87°C, and 306.67°C, and the minimum departure from nucleate boiling ratio (MDNBR) is 3.21, 3.29, and 3.01, respectively. During reactor operation at a fixed nominal power of 3400 MWt, nominal system pressure, and under the condition of flow fluctuation, the maximum RPV temperature is shown to be 303.87°C.

#### Full Text:

PDF#### References

- AP1000 European Design Control Document, EPS-GW-GL-700 Revision 1: Westinghouse; 2009. Chapter 4 Reactor and Chapter 5 Reactor Coolant System.
- Raghavaiah N.V., Overview of Pressure Vessel Design using ASME Boiler and Pressure Vessel Code Section VIII Division-1 and Division-2. Intl. Journal of Research in Engineering, Science and Management. 2019.
**2(6):**pp. 525 – 526. - Sri Sudadiyo, Taryo T., Setiadipura T., Nugroho A., Krismawan, Preliminary Design of Reactor Pressure Vessel for RDE. Int. Journal of Mechanical Engineering and Technology. 2018
**. 9(6):**pp. 889-898. - Sri Sudadiyo, Cylindrical Shell Analysis of Reactor Pressure Vessel for RDE. Ganendra. 2021.
**24(1):**pp.1-10. - Frith R., Stone M., A Proposed New Pressure Vessel Design Class. Int. Journal of Pressure Vessels and Piping. 2016. 13, pp. 4-11.
- Khattak M.A., Mukhtar A., Rafique A.F., Zareen N., Reactor Pressure Vessel Design and Fabrication: Literature Review. Journal of Advanced Research in Applied Mechanics. 2016.
**22(1):**pp.1-12. - Mairing M.P., A Conceptual Design of Reactor Pressure Vessel for NPP PWR Type IPR1000 Model. J Perangkat Nuklir. 2012.
**6(1):**pp.41-50. (Indonesian) - Kedoh P.W., Budiarsa N., Subagia I.G.G.A., Studi Penentuan Titik Kritis Bejana Tekan Reaktor PWR terhadap Kombinasi Temperatur dan Tekanan. Jurnal Ilmiah Teknik Desain Mekanik. 2017.
**6(1):**pp.108-112. (Indonesian) - Elfrida S., Himawan R., Thermal Stress Analysis on the Wall of PWR Pressure Vessel. Sigma Epsilon. 2017.
**21(1):**40-47. (Indonesian) - Isnaini M.D., Widodo S., Subekti M., Thermal-Hydraulics Analysis on Radial and Axial Power Fluctuation for AP1000 Reactor, Tri Dasa Mega. 2015.
**17(2):**79-86. - ISNAINI M.D., SUBEKTI M., Validation of SIMBAT-PWR Using Standard code of COBRA-EN on Reactor Transient Condition. Tri Dasa Mega. 2016.
**18(1):**pp.41-50. - ISNAINI M.D., MUTIARA E., A Comparison in Thermal-Hydraulics Analysis of PWR-1000 Using Fixed and Temperature Function of Thermal Conductivity. Jurnal Pengembangan Energi Nuklir. 2016. 18(1): pp.31-38.
- AGHAIE M., ZOLFAGHARI A., MINUCHEHR M., NOROUZI A., Enhancement of COBRA-EN Capability for VVER Reactor Calculations. Annals of Nuclear Energy. 2012.
**46:**p.236-243. - RAHIMI M.H., JAHANFARIA G. Thermal-Hydraulic Core Analysis of the VVER-1000 Reactor Using Porous Media Approach. Journal of Fluids and Structures. 2014.
**51:**pp. 85-96. - TODREAS N.E., KAZIMI M.S., Nuclear Systems I: Thermal Hydraulics Fundamentals. Taylor & Francis. 1993, pp. 442-444.

### Refbacks

- There are currently no refbacks.