Research on scaling design and applicability evaluation of integral thermal-hydraulic test facilities: A review - ScienceDirect
FHR and MSR Thermal Hydraulics Research at UC Berkeley
PDF) The three-level scaling approach with application to the Purdue University MultiDimensional Integral Test Assembly (PUMA | Himanshu Pokharna - Academia.edu
18 PUMA component heat loss estimation | Download Table
A historical perspective of nuclear thermal-hydraulics - ScienceDirect
19 Containment heat loss in PUMA with addition of fibrous-based... | Download Table
PDF) Design and Thermo-Hydraulic Analysis of Upgraded PUMA System for the Development of a Test Facility of Superconducting CICCs
1 PUMA scalmg Methodology Flow Chart (Continued) | Download Table
Euratom FP6 Research Projects and Training Activities Volume III
Review on neutronic/thermal-hydraulic coupling simulation methods for nuclear reactor analysis - ScienceDirect
18 PUMA component heat loss estimation | Download Table
Nea Csni R (2016) 14 | PDF | Scalability | Nuclear Power
Overview on PANDA and ISP-42, T18 / / Aksan, 1 Laboratory for Thermal-Hydraulics Nuclear Energy and Safety Department IAEA-ICTP Natural Circulation. - ppt download
10 PUMA Drywell Design 7-17 | Download Table
Full article: Scaling Methodology for Integral Effects Tests in Support of Fluoride Salt–Cooled High-Temperature Reactor Technology
Energies | Free Full-Text | Hybrid 1D + 2D Modelling for the Assessment of the Heat Transfer in the EU DEMO Water-Cooled Lithium-Lead Manifolds | HTML
PUMA : Purdue University Multi-Dimensional Integral Test Assembly Scientific Design of the Scaled Facility for GE SBWR Design SBWR : Simplified Boiling. - ppt download
C1_HTGR_design_safety_JCK
13 PUMA PCCS/ICS Condenser Design Parameters | Download Table
1 PUMA Scaling Methodology Flow Chart PUMA SCALING | Download Table
14 PUMA PCCS/ICS Pool Design Parameters | Download Table
Perspective on Thermal Hydraulic Researches in the Nuclear Field - ppt download
Energies | Free Full-Text | The H2020 McSAFER Project: Main Goals, Technical Work Program, and Status | HTML
Overview on PANDA and ISP-42, T18 / / Aksan, 1 Laboratory for Thermal-Hydraulics Nuclear Energy and Safety Department IAEA-ICTP Natural Circulation. - ppt download
PDF) Design and Thermo-Hydraulic Analysis of Upgraded PUMA System for the Development of a Test Facility of Superconducting CICCs
Full - NKS - Nordic Nuclear Safety Research
C1_HTGR_design_safety_JCK
