Articles in Scholarly Journals [Incomplete List]

1. The individual channel monitoring (ICM) proposal to improve the safety performance of RBMK
   Nuclear Engineering and Design, 2007
2. The multiple pressure tube rupture (MPTR) issue in RBMK safety technology
   Nuclear Engineering and Design, 2007
3. Thermal–hydraulic performance of confinement system of RBMK in case of accidents
   Nuclear Engineering and Design, 2007
4. A procedure to optimize the timing of operator actions of accident management procedures
   Nuclear Engineering and Design, vol. 237, no. 22, pp. 2151–2156, 2007
5. Use of coupled code technique for Best Estimate safety analysis of nuclear power plants
   Progress in Nuclear Energy, vol. 49, no. 1, pp. 1–13, 2007
6. A model for the analysis of pump start-up transients in Tehran Research Reactor
   Progress in Nuclear Energy, vol. 49, no. 7, pp. 499–510, 2007
7. Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code
   Progress in Nuclear Energy, vol. 48, no. 8, pp. 806–819, 2006
8. Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code
   Annals of Nuclear Energy, vol. 33, no. 7, pp. 646–652, 2006
9. Quantitative assessment of thermal–hydraulic codes used for heavy water reactor calculations
   Nuclear Engineering and Design, vol. 236, no. 3, pp. 295–308, 2006
10. Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled thermalâ??hydraulic 3D kinetics code
    Nuclear Engineering and Design, vol. 236, no. 12, pp. 1240–1255, 2006
11. Assessment study of the coupled code RELAP5/PARCS against the Peach Bottom BWR turbine trip test
    Nuclear Engineering and Design, vol. 235, no. 16, pp. 1727–1736, 2005
12. Methodology for the reliability evaluation of a passive system and its integration into a Probabilistic Safety Assessment
    Nuclear Engineering and Design, vol. 235, no. 24, pp. 2612–2631, 2005
13. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core
    Annals of Nuclear Energy, vol. 32, no. 15, pp. 1679–1692, 2005
14. Analysis of natural circulation phenomena in VVER-1000
    Nuclear Engineering and Design, vol. 229, no. 1, pp. 25–46, 2004
15. Transient and stability analysis in single-phase natural circulation
    Annals of Nuclear Energy, vol. 31, no. 10, pp. 1177–1198, 2004
16. Dynamic calculations of the IAEA safety MTR research reactor Benchmark problem using RELAP5/3.2 code
    Annals of Nuclear Energy, vol. 31, no. 12, pp. 1385–1402, 2004
17. Reliability evaluation of a natural circulation system
    Nuclear Engineering and Design, vol. 224, no. 1, pp. 79–104, 2003
18. Editorial
    Nuclear Engineering and Design, vol. 215, no. 1-2, pp. vii–ix, 2002
19. Use of a natural circulation map for assessing PWR performance
    Nuclear Engineering and Design, vol. 215, no. 1-2, pp. 111–126, 2002
20. Review of quantitative accuracy assessments with fast Fourier transform based method (FFTBM)
    Nuclear Engineering and Design, vol. 217, no. 1-2, pp. 179–206, 2002
21. Analyses of PACTEL passive safety injection experiments with APROS, CATHARE and RELAP5 codes
    Nuclear Engineering and Design, vol. 198, no. 3, pp. 261–286, 2000
22. Author's reply to comments by P. Ingham et al. on code validation and uncertainties in system thermalhydraulics
    Progress in Nuclear Energy, vol. 36, no. 2, p. 235, 2000
23. Circadian changes of the arterial distensibility in the elderly
    American Journal of Hypertension, vol. 13, no. 6, p. S229, 2000
24. Analysis of single-phase natural circulation experiments by system codes
    International Journal of Thermal Sciences, vol. 38, no. 11, pp. 977–983, 1999
25. Code validation and uncertainties in system thermalhydraulics
    Progress in Nuclear Energy, vol. 33, no. 1-2, pp. 175–216, 1998
26. Hetereogeneity of vascular involvement in patients with essential hypertension.
    American Journal of Hypertension, vol. 9, no. 4, p. 60A, 1996
27. Halden reactors IFA-511.2 and IFA-54x: Experimental series under adverse core cooling conditions
    Experimental Thermal and Fluid Science, vol. 11, no. 1, pp. 77–100, 1995
28. Foreword
    Nuclear Engineering and Design, vol. 145, no. 1-2, pp. xi–xiv, 1993
29. The SSN: an emergency system based on intentional coolant depressurization for PWRs
    Nuclear Engineering and Design, vol. 143, no. 1, pp. 25–54, 1993
30. Characterization of instabilities during two-phase natural circulation in typical PWR conditions
    Experimental Thermal and Fluid Science, vol. 3, no. 6, pp. 641–650, 1990
31. Flowrate and density oscillations during two-phase natural circulation in PWR typical conditions
    Nuclear Engineering and Design, vol. 122, no. 1-3, pp. 209–218, 1990
32. Assessment of scaling principles for the simulation of small break LOCA experiments in PWRs
    Nuclear Engineering and Design, vol. 102, no. 2, pp. 129–141, 1987
33. Assessment of scaling criteria adopted in designing nuclear power plants experimental simulators
    Journal of Nuclear Materials, vol. 130, pp. 51–63, 1985