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Volumn 46, Issue 2, 2005, Pages 127-141

A review of nuclear fuel performance codes

Author keywords

Fuel performance; Nuclear fuel

Indexed keywords

ACCIDENT PREVENTION; COMPUTER SIMULATION; NUCLEAR ENERGY; NUCLEAR FUEL CLADDING; NUCLEAR FUEL PELLETS; NUCLEAR POWER PLANTS; NUCLEAR REACTOR ACCIDENTS; NUCLEAR REACTORS; POWER GENERATION; SAFETY FACTOR; STRUCTURAL DESIGN;

EID: 17644404728     PISSN: 01491970     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.pnucene.2005.01.004     Document Type: Article
Times cited : (37)

References (44)
  • 4
    • 17644382404 scopus 로고
    • FRAPCON-2 Changes in Order to Create FRAPCON-2/V1M4
    • (February 7 W.N. Rausch, and D.D. Lanning FRAPCON-2 Changes in Order to Create FRAPCON-2/V1M4 Batelle Memorandum 1983
    • (1983) Batelle Memorandum
    • Rausch, W.N.1    Lanning, D.D.2
  • 12
    • 17644395044 scopus 로고
    • Thermal Contact Conductance of Reactor Fuel Elements
    • N. Todreas, and G. Jacobs Thermal Contact Conductance of Reactor Fuel Elements Nuclear Science and Engineering 50 1973 283
    • (1973) Nuclear Science and Engineering , vol.50 , pp. 283
    • Todreas, N.1    Jacobs, G.2
  • 14
    • 17644398310 scopus 로고    scopus 로고
    • BNWL-1875 (Nov. C.E. Beyer, and C.R. Hann 1974
    • BNWL-1875 (Nov. C.E. Beyer, and C.R. Hann 1974
  • 15
    • 0014587119 scopus 로고
    • Grain Boundary Gas Release and Swelling in High Burnup UO2
    • R.G. Bellamy, and J.B. Rich Grain Boundary Gas Release and Swelling in High Burnup UO2 J. Nucl. Materials 33 1969 64
    • (1969) J. Nucl. Materials , vol.33 , pp. 64
    • Bellamy, R.G.1    Rich, J.B.2
  • 22
    • 17644422881 scopus 로고
    • Semiempirical Model for Radioactive Fission Gas Release from UO2
    • C.E. Beyer, and R.O. Meyer Semiempirical Model for Radioactive Fission Gas Release from UO2 Trans. Am. Nucl. Soc. 23 1976 172
    • (1976) Trans. Am. Nucl. Soc. , vol.23 , pp. 172
    • Beyer, C.E.1    Meyer, R.O.2
  • 23
    • 17644422530 scopus 로고
    • User's Guide for PIN: A Computer Program for the Calculation of the Thermal Behaviour of an Oxide Fuel Rod
    • plc. UJV-6124T F. Pazdera, and M. Valach User's Guide for PIN: A Computer Program for the Calculation of the Thermal Behaviour of an Oxide Fuel Rod Nuclear Research Institute Rez 1982
    • (1982) Nuclear Research Institute Rez
    • Pazdera, F.1    Valach, M.2
  • 25
    • 17644388921 scopus 로고    scopus 로고
    • Description of the PIN micro Innovation to the PIN99W code
    • plc. UJV-11321-T,M (November F. Strizhov, and M. Valach Description of the PIN micro Innovation to the PIN99W code Nuclear Research Institute Rez 1999
    • (1999) Nuclear Research Institute Rez
    • Strizhov, F.1    Valach, M.2
  • 26
    • 17644424729 scopus 로고    scopus 로고
    • The IAEA CRP FUMEX Influence on the Fuel Rod Performance Modelling Quality in the Czech Republic
    • Portland, OR, USA
    • (2-6 March R. Svoboda The IAEA CRP FUMEX Influence on the Fuel Rod Performance Modelling Quality in the Czech Republic ANS International Topical Meeting on Light Water Reactor Fuel Performance Portland, OR, USA 1997
    • (1997) ANS International Topical Meeting on Light Water Reactor Fuel Performance
    • Svoboda, R.1
  • 27
    • 0026870578 scopus 로고
    • Development of Characteristics of the Rim Region in High Burnup UO2 Fuel Pellets
    • M.E. Cunningham, M.D. Freshley, and D.D. Lanning Development of Characteristics of the Rim Region in High Burnup UO2 Fuel Pellets J. Nucl. Mater. 188 1992 19 27
    • (1992) J. Nucl. Mater. , vol.188 , pp. 19-27
    • Cunningham, M.E.1    Freshley, M.D.2    Lanning, D.D.3
  • 28
    • 17644406132 scopus 로고    scopus 로고
    • Modelling Fission Gas Release at High Burnup
    • (May P. Losonen Modelling Fission Gas Release at High Burnup OECD Halden Reactor Project No. F3.5 1999
    • (1999) OECD Halden Reactor Project , Issue.35
    • Losonen, P.1
  • 31
    • 0022113610 scopus 로고
    • FEMAXI-IV: A Computer Code for the Analysis of Fuel Rod Behaviour under Transient Conditions
    • T. Nakajima FEMAXI-IV: A Computer Code for the Analysis of Fuel Rod Behaviour under Transient Conditions Nucl. Eng. Design 88 1985 69 84
    • (1985) Nucl. Eng. Design , vol.88 , pp. 69-84
    • Nakajima, T.1
  • 32
    • 0022062765 scopus 로고
    • Comparison between Fission Gas Release Data and FEMAXI-IV Code Calculations
    • T. Nakajima, and H. Saito Comparison between Fission Gas Release Data and FEMAXI-IV Code Calculations Nucl. Eng. Design 101 1987 267 279
    • (1987) Nucl. Eng. Design , vol.101 , pp. 267-279
    • Nakajima, T.1    Saito, H.2
  • 33
    • 0028445844 scopus 로고
    • FEMAXI-IV: A Computer Code for the Analysis of Fuel Rod Behaviour under Transient Conditions
    • T. Nakajima, H. Saito, and T. Osaka FEMAXI-IV: A Computer Code for the Analysis of Fuel Rod Behaviour under Transient Conditions Nucl. Eng. Design 148 1994 41 52
    • (1994) Nucl. Eng. Design , vol.148 , pp. 41-52
    • Nakajima, T.1    Saito, H.2    Osaka, T.3
  • 40
    • 0017928103 scopus 로고
    • The Mechanistic Prediction of Transient Fission-Gas Release from LWR Fuel
    • J. Rest The Mechanistic Prediction of Transient Fission-Gas Release from LWR Fuel Nucl. Eng. Design 56 1980 233 256
    • (1980) Nucl. Eng. Design , vol.56 , pp. 233-256
    • Rest, J.1
  • 43
    • 0020832690 scopus 로고
    • Cracking and Relocation Behaviour of Nuclear Fuel Pellets during Rise to Power
    • M. Oguma Cracking and Relocation Behaviour of Nuclear Fuel Pellets During Rise to Power Nucl. Eng. Design 76 1983 35 45
    • (1983) Nucl. Eng. Design , vol.76 , pp. 35-45
    • Oguma, M.1


* 이 정보는 Elsevier사의 SCOPUS DB에서 KISTI가 분석하여 추출한 것입니다.