HPLWR fine mesh core analysis
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E. Temesvári
Abstract
The European version of Supercritical Water Reactors (SCWR), the High Performance Light Water Reactor (HPLWR) operates in the thermodynamically supercritical region of water. Our basic objective was to elaborate a stationary coupled neutronic-thermohydraulic code capable for the calculation of the actual 3-pass core design with fuel assembly clusters. The calculations covered the neutronic transport calculations of HPLWR fuel assemblies, the coupled neutronic-thermohydraulic global calculations and the pin-wise analysis. Applying conservative assumptions, the relation to the linear heat rate and maximum cladding temperature limits was checked for the equilibrium cycle of HPLWR with this new code system.
Kurzfassung
Die europäische Version des superkritischen Wasserreaktors, der sog. High Performance Light Water Reactor (HPLWR), arbeitet im superkritischen Bereich des Wassers. In diesem Beitrag wird die Entwicklung eines stationären gekoppelten Codes zur Berechnung des aktuellen 3-Wege Kerndesigns mit HPLWR Brennelementen beschrieben. Dabei berücksichtigen die Berechnungen die Neutronentransportgleichungen der Brennelemente, die gekoppelten Neutronen-Thermohydraulikrechnungen und die Einzelstabanalyse. Mit konservativen Annahmen wurde die Beziehung der linearen Wärmerate und der maximalen Hüllrohrtemperaturgrenzen für den Gleichgewichtszyklus eines HPLWR überprüft.
References
1 Schulenberg, T.; Maraczy, Cs.; Heinecke, J.; Bernnat, W.: Design and Analysis of a Thermal Core for a High Performance Light Water Reactor. Nuclear Engineering and Design241 (2011) 442010.1016/j.nucengdes.2010.09.025Search in Google Scholar
2 Schulenberg, T.; Starflinger, J.; Heinecke, J.: Three pass core design proposal for a high performance light water reactor. Progress in Nuclear Energy50 (2008) 52610.1016/j.pnucene.2007.11.038Search in Google Scholar
3 Maráczy, Cs.; Hegyi, Gy.; Hordósy, G.; Temesvári, E.: HPLWR equilibrium core design with the KARATE code system. Progress in Nuclear Energy53 (2011) 26710.1016/j.pnucene.2010.12.002Search in Google Scholar
4 Derstine, K. L.: DIF3D: A Code to Solve One-, Two, and Three-Dimensional Finite-Difference Diffusion Theory Problems. ANL-82-64, Argonne National Laboratory, USA (1984)10.2172/7157044Search in Google Scholar
5 Keresztúri, A.; Hegyi, Gy.; Korpás, L.; Maráczy, Cs.; Makai, M.; Telbisz, M.: General features and validation of the recent KARATE-440 code system. Int. J. Nuclear Energy Science and Technology5 (2010) 20710.1504/IJNEST.2010.033476Search in Google Scholar
6 J.Heinecke: Personal communication (2009)Search in Google Scholar
7 Watts, M. J.; Chou, C. T.: Mixed convection heat transfer to supercritical pressure water. Proceedings of the 7th IHTC, Munchen, Germany, p. 495 (1982)10.1615/IHTC7.2970Search in Google Scholar
© 2014, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Editorial
- Technical Contributions/Fachbeiträge
- Highly enriched alternatives of VVER-440 fuel assembly
- “FULL-CORE” VVER-440 calculation benchmark
- Development of approximation method to evaluate isotopic composition of burnt fuel
- Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
- Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code
- Development and verification of new nodal methods in the KIKO3DMG code
- HPLWR fine mesh core analysis
- Assessment of reactor scram effectiveness based on measured worth of separate CR groups
- Engineering factors of the macrocode MOBY-DICK
- CFD investigation of flow in the MATIS-H test facility
- Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
- Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
- Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
- Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Editorial
- Technical Contributions/Fachbeiträge
- Highly enriched alternatives of VVER-440 fuel assembly
- “FULL-CORE” VVER-440 calculation benchmark
- Development of approximation method to evaluate isotopic composition of burnt fuel
- Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
- Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code
- Development and verification of new nodal methods in the KIKO3DMG code
- HPLWR fine mesh core analysis
- Assessment of reactor scram effectiveness based on measured worth of separate CR groups
- Engineering factors of the macrocode MOBY-DICK
- CFD investigation of flow in the MATIS-H test facility
- Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
- Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
- Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
- Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients
