Uncertainty studies on hydrogen source term with MAAP5 code
-
T.-C. Wang
und
M. Lee
Abstract
In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.
Abstract
In der vorliegenden Studie wird eine Methodik zur Quantifizierung der Unsicherheiten von speziellen Modellparametern des integralen schweren Störfallanalyse-Codes MAAP5 entwickelt. Hier wird die Wasserstoffproduktion im Reaktorbehälter während eines Kernschmelzunfalls für das Kernkraftwerk Lungmen der Taiwan Power Company, einem fortgeschrittenen Siedewasserreaktor, analysiert. Es werden Sensitivitätsstudien durchgeführt, um die Parameter zu identifizieren, die einen Einfluss auf die Ausgabeparameter haben. Dazu werden Mehrfachberechnungen von MAAP5 mit Eingabekombinationen durchgeführt, die mit Latin Hypercube Sampling (LHS) generiert wurden. Die Ergebnisse werden analysiert, um das 95. Perzentil mit einem Konfidenzniveau von 95% für die Menge der Wasserstoffproduktion im Behälter zu bestimmen. Die Berechnungen zeigen, dass die Standard-Modelloptionen für IOXIDE und FGBYPA empfohlen werden können. Der Pearson-Korrelationskoeffizient (PCC) wurde verwendet, um den Einfluss der Modellparameter auf die Ziel-Ausgangsparameter zu bestimmen und zeigt, dass die drei Parameter TCLMAX, FCO, FOXBJ einen hohen Einfluss auf die Wasserstofferzeugung im Behälter haben. Es werden Vorschläge für die Auswahl dieser drei Parameter gegeben.
Nomenclature
- RCS
reactor coolant system
- SAMGs
severe accident management guidelines
- SNL
Sandia national laboratories
- NRC U.S.
nuclear regulatory commission
- MAAP5
module accident analysis program)
- FAI
Fauske & associated inc
- EPRI
electric power research institute
- ABWR
advanced boiling water reactor
- MUG
MAAP users’ group
- Zr
zirconium
- BWR
boiling water reactors
- LHS
latin hypercube sampling
Acknowledgements
We thank the Taiwan Power Company (TPC) and Lungmen Vuclear Power Plant for supporting this work.
References
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© 2021 Walter de Gruyter GmbH, Berlin/Boston
Artikel in diesem Heft
- Frontmatter
- Experimental investigation on pool boiling for downward facing heating with different concentrations of Al2O3 nanofluids
- 244Cm contributions to the alpha source term of CANDU reactors
- Methodology for integral analysis of ATWS for Kuosheng nuclear power station
- LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
- Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events
- Uncertainty studies on hydrogen source term with MAAP5 code
- Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor
- Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition
- Temperature adjusted cross section libraries used for criticality calculations
Artikel in diesem Heft
- Frontmatter
- Experimental investigation on pool boiling for downward facing heating with different concentrations of Al2O3 nanofluids
- 244Cm contributions to the alpha source term of CANDU reactors
- Methodology for integral analysis of ATWS for Kuosheng nuclear power station
- LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
- Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events
- Uncertainty studies on hydrogen source term with MAAP5 code
- Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor
- Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition
- Temperature adjusted cross section libraries used for criticality calculations
