Jordi Freixa Terradas

Lecturer
Campus Sud (UPC), Edifici H-PavC, Despatx 31.06
Avda. Diagonal, 647
08028 Barcelona

Telèfon: +34 934017454
E-Mail: jordi.freixa-terradas

upc.edu
Web del grup de recerca: https://ant.upc.edu
Web personal: https://futur.upc.edu/181605

Short CV

Jordi Freixa Graduated in Physics by the University of Barcelona in 2001 just before joining the PhD Nuclear Engineering Program at the Technical University of Catalonia (UPC). He finished his PhD on “SBLOCA transients with boron dilution” in 2007. Jordi worked as a post-doc for 4 years at the Paul Scherrer Institut (PSI, Switzerland) participating in several projects including safety analyses of LOCA and SBLOCA scenarios for the EPR Olkiluoto III nuclear power plant. He continued at PSI as a senior scientist for 2 additional years and lectured at the Nuclear Engineering Master program of the Technical University of Zurich (ETH). Jordi joined ANT in January 2013 under a UPC/PQS contract. His main areas of research are the validation and application of best estimate thermal hydraulic codes for LWRs which he combines with teaching activities in the European Master of Nuclear Engineering.

1- Core-wide multi-physics simulation of fuel behaviour during large-break LOCA using NEXUS
Peakman, A.; Gregg, R.; Martínez, V.; Freixa, J.
Annals of nuclear energy 215, article 111228 (2025)

2- Doppler effect evaluation of the TRU-based Moroccan TRIGA Mark II reactor: applying novel methodologies
Mira, M.; el Hajjaji, O.; Freixa, J.; el Bardouni, T.; Boukhal, H.; Boulaich, Y.; el Ghalbzouri, T.
Annals of nuclear energy 217, 111353 (2025)

3- Scaling in nuclear thermal hydraulics: advances, applications and future perspectives in Spain
Freixa, J.; Berna, C.; Escrivá, F.A.; Gallardo, S.; Lorduy, M.; Martinez, V.; Martin, K.; Mendizabal, R.; Muñoz-Cobo, J.L.; Reventos, F.; Sánchez Perea, Miguel
Nuclear engineering and design 426, article 113394 (2024)

4- Application of the ANT assessment methodology for validating LOCUST 1.2 thermal-hydraulic code
Martin, K.; Casamor, M.; Martinez, V.; Perez, M.; Zhongyun, J.; Freixa, J.
Nuclear engineering and design 427, article 113410 (2024)

5- Assessment of the choked flow model of RELAP5 for application of inverse quantification methods
Osés, P.; Freixa, J.; Mendizabal, R.; Sánchez Perea, Miguel
Nuclear technology , (2024)

6- Spanish contribution to the development and application of best estimate plus uncertainty methodologies: past, present and future
Freixa, J.; Barrachina, T.; Berna, C.; Bocanegra, R.; Carlos, S.; Castro, E.; Cuervo, D.; Durán, L.; Escrivá, F.A.; Feria, F.; Fernández, K.; Martinez, V.; Perez, M.; Pericas, R.; Reventos, F.
Nuclear engineering and design 417, 112837 (2024)

7- Spanish research related to SMRs projects
Queral Salazar, José César; Redondo, E.; Sanchez, J.; Jimenez, G.; Larriba, S.; Cuervo, D.; Cabellos, O.; Durán, L.; Herranz, L.; García, M.; Martinez, V.; Freixa, J.
Nuclear engineering and design 417, 112818 (2024)

8- Full scope scaling analysis and recommendations for PWR SGTR scenarios
Martin, K.; Martinez, V.; Freixa, J.
Nuclear engineering and design 418, 112864 (2024)

9- Development and application in multiscale and multiphysics methodologies in Spain: present and future trends
Gallardo, S.; Alvarez-Velarde, F.; Barrachina, T.; Cabellos, O.; Castro, E.; Casamor, M.; Cuervo, D.; Escrivá, F.A.; Freixa, J.; García, N.; Martinez, V.; Miro Herrero, Rafael; Queral, C.; Rivera, Y.
Nuclear engineering and design 421, 113096 (2024)

10- Activities in Spain on the integration of probabilistic and deterministic safety analysis methods. Discussion on their applicability to DEC-A scenarios
Mendizabal, R.; Sánchez, M.; Queral Salazar, José César; Hortal, F.; Martorell, S.; Pelayo, F.; Freixa, J.
Nuclear engineering and design 419, 112944 (2024)

11- A systematic approach for the adequacy analysis of a set of experimental databases: application in the framework of the ATRIUM activity
Baccou, J.; Glantz, T.; Ghione, A.; Sargentini, L.; Fillion, P.; Damblin, G.; Sueur, R.; Iooss, B.; Fang, J.; Freixa, J.
Nuclear engineering and design 421, 113035 (2024)

12- Kv-scaling in thermal hydraulics: Background, applications and forthcoming uses
Martinez, V.; Freixa, J.; Reventos, F.
Nuclear engineering and design 404, 112141 (2023)

13- Multi-physics framework for whole-core analysis of transient fuel performance after load following in a pressurised water reactor
Peakman, A.; Gregg, R.; Bennett, T.; Casamor, M.; Martinez, V.; Freixa, J.; Pericas, R.; Rossiter, G.
ANNALS OF NUCLEAR ENERGY 173, 109086:1-109086:19 (2022)

14- Off-line vs. semi-implicit TH-TH coupling schemes: A BEPU comparison
Casamor, M.; Avramova, M.; Reventos, F.; Freixa, J.
Annals of nuclear energy 178, 109344:1-109344:14 (2022)

15- OECD/NEA PKL-4 benchmark activity. Code assessment of the relevant phenomena associated to a blind IBLOCA experiment
Martinez, V.; Szogradi, M.; Schollenberger, S.; Sánchez Perea, Miguel; Sandberg, N.; Zhongyun, J.; Freydier, P.; Freixa, J.
Nuclear engineering and design 389, 111632 (2022)

16- Methodology for phenomenological code assessment with integral test data
Perez, M.; Martinez, V.; Casamor, M.; Freixa, J.
Nuclear engineering and design 387, 111608 (2022)

17- Application of a BEPU-based code assessment to the ATLAS upper head SB-LOCA test
Al-Awad; Freixa, J.; Perez, M.
Annals of nuclear energy 164, 1-13 (2021)

18- Effectiveness of the ASVAD valve in a reactor vessel bottom leak scenario
Freixa, J.; Laborda, A.; Martinez, V.
Annals of nuclear energy 160, 108387: 1-108387: 10 (2021)

19- On the validation of BEPU methodologies through the simulation of integral experiments: Application to the PKL test facility
Freixa, J.; Martinez, V.; Casamor, M.; Reventos, F.; Mendizabal, R.; Sánchez Perea, Miguel
Nuclear engineering and design 379, 111238: 1-111238: 11 (2021)

20- Development of flow regime maps for lead lithium eutectic–helium flows
Mas de les Valls, E.; Cegielski, A.; Jaros, M.; Pérez, M.; Batet, L.; Freixa, J.
Fusion engineering and design 158, 111691:1-111691:8 (2020)

21- Modelling guidelines for safety analysis of Station Black Out sequences based on experiments at the PKL test facility
Freixa, J.; Martinez, V.; Reventos, F.
Annals of nuclear energy 138, 107179:1-107179:11 (2020)

22- On the scaling of uncertainties in thermal hydraulic system codes
Casamor, M.; Martinez, V.; Reventos, F.; Mendizabal, R.; Freixa, J.
Annals of nuclear energy 136, (2020)

23- Perfecting the use of hybrid models in scaling analysis
Reventos, F.; Freixa, J.; Martinez, V.
Nuclear engineering and design 354, 110187:1-110187:7 (2019)

24- Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes – PREMIUM benchmark
Skorek, T.; de Crécy, A.; Kovtonyuk, A.; de Alfonso, E.; Reventos, F.; Freixa, J.
Nuclear engineering and design 354, 110199:1-110199:23 (2019)

25- PVST, a tool to assess the power to volume scaling distortions associated to code simulations
Martinez, V.; Freixa, J.; Reventos, F.
Nuclear engineering and design 332, 173-185 (2018)

26- Testing methodologies for quantifying physical models uncertainties. A comparative exercise using CIRCE and IPREM (FFTBM)
Freixa, J.; De Alfonso, E.; Reventos, F.
Nuclear engineering and design 305, 653-665 (2016)

27- Qualification of a full plant nodalization for the prediction of the core exit temperature through a scaling methodology
Freixa, J.; Martinez, V.; Reventos, F.
Nuclear engineering and design 308, 115-132 (2016)

28- Modelling guidelines for core exit temperature simulations with system codes
Freixa, J.; Martinez, V.; Zerkak, O.; Reventos, F.
Nuclear engineering and design 286, 116-129 (2015)

29- Applying UPC scaling-up methodology to the LSTF-PKL counterpart test
Martinez, V.; Reventos, F.; Freixa, J.
Science and technology of nuclear installations 2014, 1-18 (2014)

30- Simulation of condensation in a closed, slightly inclined horizontal pipe with a modified RELAP5 code
Szijártó, R.; Freixa, J.; Prasser, H.
Nuclear engineering and design 273, 288-297 (2014)

31- "Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation"
Freixa, J.; Manera, A.; Kim, T-W.
Nuclear engineering and design 264, 153-160 (2013)

32- Remarks on Consistent Development of Plant Nodalizations: An Example of Application to the ROSA Integral Test Facility
Freixa, J.; manera, A.
Science and technology of nuclear installations 2012, Article ID 158617 (2012)

33- "Analysis of the ISP-50 direct vessel injection sbloca in the atlas facility with the relap5/mod3.3 code"
Sherabi, M.; Freixa, J.
Nuclear engineering and technology 44, 709-718 (2012)

34- "Thermal-hydraulic analysis of an intermediate LOCA test at the ROSA facility including uncertainty evaluation"
Freixa, J.; Kim, T-W.; Manera, A.
Nuclear engineering and design 249, 97-103 (2012)

35- Verification of a TRACE EPR™ model on the basis of a scaling calculation of an SBLOCA ROSA test
Freixa, J.; manera, A.
Nuclear engineering and design 241, 888-896 (2011)

36- "Analysis of an RPV upper head SBLOCA at the ROSA facility using TRACE"
Freixa, J.; Manera, A.
Nuclear engineering and design 240, 1779-1788 (2010)

37- "SBLOCA with boron dilution in pressurized water reactors. Impact on operation and safety"
Freixa, J.; Reventos, F.; Pretel, C.; Batet, L.; Sol, I.
Nuclear engineering and design 293, 749-760 (2009)

38- "An analytical comparative exercise on the OECD-SETH PKL E2.2 experiment"
Reventos, F.; Freixa, J.; Batet, L.; Pretel, C.; Luebbesmeyer, D.; Spaziani, D.; Jiri, M.; Lahovsky, F.; Kasahara, F.; Umminger, K.; Wegner, R.
Nuclear engineering and design 238, 1146-1154 (2008)

39- "Boron transport model with physical diffusion for RELAP5"
Freixa, J.; Reventos, F.; Pretel, C.; Batet, L.
Nuclear technology 160, 205-215 (2007)