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Licensed Unlicensed Requires Authentication Published by De Gruyter April 12, 2019

Determination of radiological source term of CHASHMA-1 NPP during LOCA

Abschätzung der bei einem Kühlmittelverluststörfalls aus dem KKW CHASNUPP-1 freigesetzten Radioaktivität
  • K. Mehboob and M. S. Aljohani
From the journal Kerntechnik

Abstract

The CHASHMA Nuclear Power Plant unit 1 is known as CHASNUPP-1. The CHASNUPP-unit 1 is a 996 MWth intermediate type pressurised water reactor that began commercial operation in June 2000 in Pakistan. The CHASNUPP unit 1 is a conventional two loop PWR operated by the Pakistan Atomic Energy Commission (PAEC). The radiological source term of CHASNUPP-1 has been evaluated and compared with the advanced modular reactor (SMART) and KORI-1 reactor. For this purpose, modelling and simulation has been carried out in MATLAB. a kinetic model has been developed to carry out the simulation of the release of radionuclides. The core and coolant activity of CHASNUPP-1 is compared with the similar type reactor KORI-1. The developed model uses the ORIGEN 2.2 core inventory as a subroutine. The coolant inventory has been evaluated with 0.25% fuel damage and compared with SMART and KORI-1 reactor.

Kurzfassung

Der CHASHMA-Kernkraftwerksblock 1 ist international als CHASNUPP-1 bekannt. Der CHASNUPP-Block 1 ist ein mittelgroßer Druckwasserreaktor mit einer thermischen Leistung von 996 MWth, der im Juni 2000 in Pakistan den kommerziellen Betrieb aufgenommen hat. Der CHASNUPP-Block 1 ist ein konventioneller Zweikreis-DWR, der von der Pakistan Atomic Energy Commission (PAEC) betrieben wird. Berechnungen zur erwartenden Strahlenbelastung bei Kühlmittelverluststörfällen in CHASNUPP-1 werden in diesem Beitrag vorgestellt. Dazu wurde mit dem Programm MATLAB eine Modellierung des KKW erstellt und die Szenarien berechnet. Für die Simulation der Freisetzung von Radionukliden wurde ein kinetisches Modell entwickelt. Die Kern- und Kühlmittelaktivität von CHASNUPP-1 wird mit bestehenden Analysen zum ähnlichen Reaktor KORI-1 sowie mit dem fortschrittlichen modularen Reaktorkonzept SMART verglichen. Das entwickelte Modell verwendet das ORIGEN 2.2 Kernprogramm als Unterprogramm.


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References

1 European Nuclear Society: Nuclear power plants, world-wide. URL: https://www.Euronuclear.Org/Info/Encyclopedia/N/Nuclear-Power-Plant-World-Wide.Htm. (2018) (Accessed on 2018-01-30).Search in Google Scholar

2 Mehboob, K.; Xinron, C.; Ali, M.: Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR. Research Journal of Applied Sciences Engineering and Technology16 (2012) 2895290Search in Google Scholar

3 Mehboob, K.; Xinrong, C.: Source term evaluation of two loop PWR under hypothetical severe accidents. Annals of Nuclear Energy50 (2012) 27128410.1016/j.anucene.2012.07.011Search in Google Scholar

4 Mehboob, K; Xinrong, C.; Ahmed, R.; Ali, M.: Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents. Nuclear Engineering and Design262 (2013) 4354510.1016/j.nucengdes.2013.05.021Search in Google Scholar

5 Mehboob, K.; Park, K.; Khan, R.: Quantification of in-containment fission products source term for 1000 MWe PWR under loss of coolant accident. Annals of Nuclear Energy75 (2015) 36537610.1016/j.anucene.2014.08.056Search in Google Scholar

6 Mehboob, K.; Aljohani, S. M.: Modelling and Simulation of Radio-Iodine Released inside the Containment as Result of an Accident. Progress in Nuclear Energy88 (2016) 758710.1016/j.pnucene.2015.11.013Search in Google Scholar

7 Yangmo, Z.; Jianghua, G.; Chu, N.; Youhua, Z.: Simulation and dose analysis of a hypothetical accident in Sanmen nuclear power plant. Annals of Nuclear Energy65 (2014) 20721310.1016/j.anucene.2013.11.016Search in Google Scholar

8 Pappas, C.; Ikonomopoulos, A.; Sfetsos, A.; Andronopoulos, S.; Varvayanni, M.; Catsaros, N.: Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1. Nuclear Engineering and Design274 (2014) 10011710.1016/j.nucengdes.2014.04.008Search in Google Scholar

9 Zhao, Y.; Zhang, L.; Tong, J.: Development of rapid atmospheric source term estimation system for AP1000 nuclear power plan. Progress in Nuclear Energy81 (2015) 26427510.1016/j.pnucene.2015.02.008Search in Google Scholar

10 Ammirabile, L.; Bujan, A.; Sangiorgi, M.: PWR Medium Break LOCA source term analysis using ASTEC code Plan. Progress in Nuclear Energy81 (2015) 304210.1016/j.pnucene.2015.01.001Search in Google Scholar

11 Poggi, L. A.; Malizia, A.; Ciparisse, J. F.; Gaudio, P.: A novel integrated approach for the hazardous radioactive dust source terms estimation in future nuclear fusion power plants. Heliyon2 (2016) e00184 PMid:27812553; 10.1016/j.heliyon.2016.e00184Search in Google Scholar PubMed PubMed Central

12 Tong, L. L.; Huang, G. F.; Cao, X. W.: Simulation of fission products behavior in severe accidents for advanced passive PWR. Annals of Nuclear Energy, Vol. 76 (2015) 49350310.1016/j.anucene.2014.10.031Search in Google Scholar

13 FSAR: Chashma nuclear power plant unit-1. Final Safety Analysis Report, Chashma, Punjab, Pakistan (1989)Search in Google Scholar

14 Soffer, L.; Burson, S. B.; Ferrell, C. M.; Lee, R. Y.; Ridgely, J. N.: Accident source term for Light-Water Nuclear power plants. NUREG-1465. United States Nuclear Regulatory Commission (1995) 10.2172/29438Search in Google Scholar

15 Croft, A. G.: A User's Manual for the ORIGEN2 Computer Code. Rep. ORNL/TM-7175. Oak Ridge National Laboratory, Oak Ridge, TN (1980)Search in Google Scholar

16 USNRC: Reactor safety study. An assessment of accident risk in U.S. commercial nuclear power plants. WASH-1400, United States Nuclear Regulatory Commission (1975)Search in Google Scholar

17 USNRC: Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. Regulatory Guide1. 183 (2000)Search in Google Scholar

18 El-Jaby, A.; Lewis, B. J.; Thompson, W. T.; Iglesias, F.; Ip, M.: A general model for predicting coolant activity behavior for fuel failure monitoring analysis. Journal of Nuclear Material399 (2010) 8710010.1016/j.jnucmat.2010.01.006Search in Google Scholar

19 NRC: Severe accident risk: An assessment of five U.S. Commercial plants. NUREG- 1150, united states nuclear regulatory commission (1990)Search in Google Scholar

20 Sümer, S., Asim, S.: Experimental evaluation of surveillance capsule assemblies for life assessment of CHASNUPP unit-1 reactor pressure vessel. Annals of Nuclear Energy89 (2016) 909810.1016/j.anucene.2015.11.019Search in Google Scholar

21 Tahir, M.; Chughtai, I. R.; Aslam, M.: Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1. Annals of Nuclear Energy35 (2008) 1986199310.1016/j.anucene.2008.06.006Search in Google Scholar

22 Pakistan Nuclear Regulatory Authority (PNRA): Fourth national report for the convention on nuclear safety. Fourth National Report of Pakistan, PNRA, Islamabad, Pakistan, (2007)Search in Google Scholar

23 Kim, S. U.; Kang, C. S.: Evaluation of Radioactive Source term in the System integrated Modular Advanced Reactor. Journal of the Korean Nuclear Society31 (1999) 916Search in Google Scholar

Received: 2018-01-29
Published Online: 2019-04-12
Published in Print: 2019-04-15

© 2019, Carl Hanser Verlag, München

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