Best estimate plus uncertainty analysis of LBLOCA for Indian PHWR
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A. Srivastava
Abstract
Deterministic safety analysis is an important tool for confirming the adequacy of provisions within the defense-in-depth concept for the safety of nuclear power plants. One of the important design basis events is considered to be a complete double-ended guillotine rupture i.e. Loss of Coolant Accident (LOCA) of largest and coldest pipe in the primary coolant circuit. The present work deals with this scenario for an Indian PHWR. It highlights the identification of critical break size leading to the maximum clad temperature using the best estimate code RELAP5. Further, important parameters affecting the clad temperature are described along with results of initial sensitivity studies to select dominant uncertain parameters. For the uncertainty propagation, Latin Hypercube Sampling (LHS) is used instead of simple random sampling for Monte-Carlo simulation. The inherent characteristic of LHS is to reduce the required runs for Monte-Carlo simulation to manageable order for the current computing capability. The 95th percentile value of peak cladding temperature (PCT) is obtained by the method and compared with acceptance criteria.
Kurzfassung
Deterministische Sicherheitsanalysen sind ein wichtiges Instrument der Sicherheitsuntersuchungen von Kernkraftwerken. Einer der wichtigsten zu untersuchenden Auslegungsstörfälle ist ein kompletter 2F Bruch der größten und kältesten Leistung im Primärkreislauf – auch LBLOCA genannt. Im aktuellen Beitrag werden Untersuchungen eines LBLOCA für einen schwerwassermoderierten Druckwasserreaktor (PHWR) in Indien vorgestellt. Dabei werden die Einflussfaktoren, wie z.B. die kritische Leckgröße, die zu maximalen Brennstabtemperaturen führen, bestimmt und untersucht. Die Berechnungen mit RELAP5 werden durch eine Unsicherheitsanalyse basierend auf der Stichprobenmethode Latin Hypercube Sampling (LHS) ergänzt. Die so berechnete 95%-Perzentile der maximalen Brennstabtemperatur (PCT) wird mit den Akzeptanzkriterien verglichen.
References
1 Bajaj, S. S.;Gore, A. R.: The Indian PHWR. Nuclear Engineering and Design236 (2006) 701–72210.1016/j.nucengdes.2005.09.028Search in Google Scholar
2 Emergency Core Cooling Systems; Revisions to Acceptance Criteria. Federal Register53180 (1988) 35996–36005Search in Google Scholar
3 Boyack, B. E.; et al.: Quantifying Reactor Safety Margins, Part 1: An Overview Of The Code Scaling, Applicability, and Uncertainty Evaluation Methodology. Nuclear Engineering and Design119 (1990) 1–1510.1016/0029-5493(90)90071-5Search in Google Scholar
4 Prosek, A.: Optimal Statistical Estimator for Efficient Generation of the Response Surface. International Meeting on Best Estimate Methods in Nuclear Installation Safety Analysis, Washington DC, November, 2000Search in Google Scholar
5 Glaeser, H.: GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications. Science and Technology of Nuclear Installations2008 (2008) Article ID 798901, doi:10.1155/2008/79890110.1155/2008/798901Search in Google Scholar
6 Wilks, S. S.: Determination of Sample Sizes for Setting Tolerance Limits. The Annals of Mathematical Statistics12 (1941) 91–9610.1214/aoms/1177731788Search in Google Scholar
7 Lee, J. H.: Comparison of Latin Hypercube Sampling and Simple Random Sampling applied to neural network modeling of HfO2 thin film fabrication. Transactions on Electronic and Electric Materials7 (2006) 210–21410.4313/TEEM.2006.7.4.210Search in Google Scholar
8 Helton, J. C.;Davis, F. J.: Latin Hypercube Sampling and the propagation of uncertainty in analyses of complex systems. Reliability Engineering and System Safety81 (2003) 23–6910.1016/S0951-8320(03)00058-9Search in Google Scholar
9 Fletcher, G. D.;Schultz, R. R.: RELAP5/MOD3.2 Code Manual. Idaho National Engineering Laboratory, Idaho, 1995Search in Google Scholar
10 Ingham, P. J.;McGee, G. R.;Krishnan, V. S.: LOCA assessment experiments in a full-elevation, CANDU-typical test facility. Nuclear Engineering and Design122 (1990) 401–41210.1016/0029-5493(90)90223-KSearch in Google Scholar
11 Cho, Y. J.;Jeun, G. D.: Assessments of RELAP5/mod3.2 and RELAP5/CANDU in a Reactor Inlet Header Break Experiment B9401 of RD-14M. Journal of the Korean Nuclear Society35 (2003) 426–441Search in Google Scholar
12 Majumdar, P.;Srivastava, A.;Gupta, S. K.: Nodalisation Study on RD14M Test Facility. 1st National Conference on Nuclear Reactor Safety, BARC, Mumbai, November 25–27 (2002)Search in Google Scholar
13 Iman, R. L.;Helton, J. C.: A comparison of uncertainty and sensitivity analysis techniques for computer models. NUREG/CR-3904, SAND84-1461, Sandia National Laboratories (1985)Search in Google Scholar
© 2012, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Simulation of HALDEN IFA-650 loss-of-coolant accidents tests with TRACE
- On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures
- Best estimate plus uncertainty analysis of LBLOCA for Indian PHWR
- Remarks on boiling water reactor stability analysis – part 1: theory and application of bifurcation analysis
- Upgrading (V)HTR fuel elements for generationIV goals by SiC encapsulation
- Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor
- PCA-based ANN approach to leak classification in the main pipes of VVER-1000
- Analytical investigation of the properties of the neutron noise induced by vibrating absorber and fuel rods
- Diffusion approximation for certain scattering parameters of the Anli-Güngör phase function
- Calculation of beta induced Bremsstrahlung exposure from therapeutic radionuclide 198Au in tissues, DNA and RNA
- Dosimetric aspects of 103Pd radioactive stent source
- Radiological significance of coal, slag and fly ash samples from the Eastern Black Sea region
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Simulation of HALDEN IFA-650 loss-of-coolant accidents tests with TRACE
- On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures
- Best estimate plus uncertainty analysis of LBLOCA for Indian PHWR
- Remarks on boiling water reactor stability analysis – part 1: theory and application of bifurcation analysis
- Upgrading (V)HTR fuel elements for generationIV goals by SiC encapsulation
- Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor
- PCA-based ANN approach to leak classification in the main pipes of VVER-1000
- Analytical investigation of the properties of the neutron noise induced by vibrating absorber and fuel rods
- Diffusion approximation for certain scattering parameters of the Anli-Güngör phase function
- Calculation of beta induced Bremsstrahlung exposure from therapeutic radionuclide 198Au in tissues, DNA and RNA
- Dosimetric aspects of 103Pd radioactive stent source
- Radiological significance of coal, slag and fly ash samples from the Eastern Black Sea region
