Skip to main content
SpringerLink
Log in

Analysis of Reactor Fuel Rod Behavior

  • Reference work entry
Handbook of Nuclear Engineering

Abstract

The analysis of the behavior of light water reactor (LWR) fuel rods is described. The properties of relevant fuel and cladding materials are discussed and numerical data are given. The basic phenomena taking place in pellet-in-cladding nuclear reactor fuel are described systematically, including neutronic aspect of the fuel, the thermal and mechanical behavior, the fission gas behavior, and radiation effects. Finally typical phenomena and issues in the design and licensing of LWR fuels and their effects on fuel behavior are discussed: the high burnup structure, pellet-cladding interaction, pellet-coolant interaction, loss-of-coolant accidents (LOCA), and reactivity-initiated accidents (RIA).

This is a preview of subscription content, log in via an institution to check access.

Access this chapter

Log in via an institution
Chapter
USD 29.95
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
eBook
USD 1,599.99
Price excludes VAT (USA)
  • Available as EPUB and PDF
  • Read on any device
  • Instant download
  • Own it forever
Hardcover Book
USD 1,999.99
Price excludes VAT (USA)
  • Durable hardcover edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info

Tax calculation will be finalised at checkout

Purchases are for personal use only

Institutional subscriptions

References

  • Abeles B (1963) Lattice thermal conductivity of disordered semiconductor alloys at high temperatures. Phys Rev 131:1906–1911

    Google Scholar 

  • Adamson MG, Aitken EA, Caputi RW (1985) Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels. J Nucl Mater 130:349–365

    Google Scholar 

  • Ainscough JB, Oldfield BW, Ware JO (1973/1974) Isothermal grain growth kinetics in sintered UO2 pellets. J Nucl Mater 49:117–128

    Google Scholar 

  • Aitken EA, Evans SK (1968) A thermodynamic data program involving plutonium and urania. Tech. Rep. USAEC GEAP-5672, General Electric

    Google Scholar 

  • Ambegaokar V (1959) Thermal resistance due to isotopes at high temperatures. Phys Rev 114:488–489

    Google Scholar 

  • Anderson TL (1995) Fracture mechanics – fundamentals and applications, 2nd edn. CRC Press

    Google Scholar 

  • Assmann H (1982) Ãœberblick über zusammenhänge zwischen LWR-brennstoffeigenschaften und verfahrensabläufen bei der Brennstoffproduktion. J Nucl Mater 106:15–34

    Google Scholar 

  • Assmann H, Stehle H (1978) Thermal and in-reactor densification of UO2: mechanisms and experimental results Nucl Eng Des 48:49–67

    Google Scholar 

  • Bailly H, Ménessier D, Prunier C (1999) The nuclear fuel of pressurized water reactors and fast neutron reactors. Lavoisier

    Google Scholar 

  • Baker L, Just LC (1962) Studies of metal-water reactions at high temperatures; III. Experimental and theoretical studies of the zirconium-water reaction. Tech. Rep. ANL-6548, Argonne National Laboratory

    Google Scholar 

  • Barner J, Gunningham MD, Freshley DD, Lanning D (1990) Relationship between microstructure and fission gas release in high burnup UO2 fuel with emphasis on the rim region, International Topical Meeting on LWR Fuel Performance, ANS/ENS, 21-24 April 1991, Avignon, France, Proceedings, pp 538–548

    Google Scholar 

  • Barney W, Wemble B (1958) Metallography of UO2-containing fuel elements. Tech. Rep. KAPL-1836

    Google Scholar 

  • Baron D (1986) Porosity Buildup in the Fuel Periphery at High Burnup. HBEP steering committee meeting

    Google Scholar 

  • Baron D (1998) A fuel thermal conductivity correlation based on the latest experimental results. In: Proceedings of the seminar on thermal performance of high burn-up LWR fuel. Cadarache, France, pp 129–143

    Google Scholar 

  • Baron D, Masson R, Gatt J, Spino J, Laux D (2005) Evolution of the nuclear fuel mechanical properties with burn-up. An extensive European experimental program. In: Proceedings of the international topical meeting on LWR fuel performance, Track 2, Paper 1040. Kyoto, Japan

    Google Scholar 

  • Bates JL (1970) Melting point of irradiated uranium dioxide. J Nucl Mater 36:234–236

    Google Scholar 

  • Beals RJ, Handwerk JH, Wrona BJ (1969) Behavior of urania-rare-earth oxides at high temperatures. J Am Ceram Soc 52:578–581

    Google Scholar 

  • Beauvy M (1992) Nonideality of the solid solution in (U, Pu)O2 nuclear fuels. J Nucl Mater 188:232–238

    Google Scholar 

  • Belle J (1961) Uranium dioxide: properties and nuclear applications. Naval reactor handbooks. US Atomic Energy Commission

    Google Scholar 

  • Berna GA, Beyer CE, Davis KL, Lanning DD (1997) Frapcon-3: a computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup, report NUREG/CR-6534-v2, (PNNL-11513-v2), December 1997

    Google Scholar 

  • Bernaudat C, Pupier P (2005) A new analytical approach to study the rod ejection accident in PWRs. In: Proceedings of the international topical meeting on light water reactor fuel performance. Kyoto, Japan, pp 602–614

    Google Scholar 

  • Bibilashvily Y, Medvedev A, Khostov G, Bogatyr S, Korystine L (2000) Development of the fission gas behavior model in the START-3 code and its experimental support. In: Proceedings of the international seminar on fission gas behaviour in water reactor fuels. Cadarache, France

    Google Scholar 

  • Billaux M (2005) High burnup fuel for LWRs: fuel performance, limits, operational and safety Issues. The 2005 Frédéric Joliot and Otto Hahn Summerschool. Karlsruhe, Germany

    Google Scholar 

  • Billone M et al (2008) Cladding embrittlement during postulated loss-of-coolant accidents. Tech. Rep. NUREG /CR-6967

    Google Scholar 

  • Blank H, Matzke H (1973) The effect of fission spikes on fission gas re-solution. Radiat Eff 17:57–64

    Google Scholar 

  • Booth AH (1957) A method of calculating fission gas diffusion from UO2 fuel and its application to the x-2-f loop test. Tech. Rep. CRDC-721, AECL, Chalk River, Ontario, Canada

    Google Scholar 

  • Booth AH (1957) A suggested method for calculating the diffusion of radioactive rare gas fission products from UO2 elements and a discussion of proposed in-reactor experiments that may be used to test its validity. Tech. Rep. DCI-27, AECL, Chalk River, Ontario, Canada

    Google Scholar 

  • Bossis P, Pêcheur D, Hanifi K, Thomazet J, Blat M (2006) Comparison of the high burn-up corrosion on M5 and low tin Zircaloy-4. ASTM Int 3, DOI:10.1520/JAI12,404

    Google Scholar 

  • Carbajo JJ, Yoder GL, Popov SG, Ivanov VK (2001) A review of the thermophysical properties of MOX and UO2 fuels. J Nucl Mater 299:181–198

    Google Scholar 

  • Cathcart JV, Powel RE et al (1977) Zirconium metal-water oxidation kinetics, w. reaction rate studies. Tech. Rep. ORNLINUREG-17, Oak Ridge National Laboratory

    Google Scholar 

  • Cazalis B, Bernaudat C, Yvon P, Desquines J, Poussard C, Averty X (2005) The PROMETRA program: a reliable material database for highly irradiated Zircaloy-4, ZIRLOTM and M5TM fuel cladding. In: Proceedings of the 18th international conference on structural mechanics in reactor technology (SMiRT 18), no. C02-1 in SMiRT

    Google Scholar 

  • Chadwick MB, Oblozinsky P, Herman M et al (2006) ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology. Nucl Data Sheets 107: 2931–3060

    Google Scholar 

  • Christensen JA (1962) Irradiation effects on uranium dioxide melting Tech. Rep. Report HW-69234, Hanford, California

    Google Scholar 

  • Christensen JA, Allio RJ, Biancheria A (1964) Uranium dioxide thermal conductivity. Trans Am Nucl Soc 7:390–391

    Google Scholar 

  • Chung H (2005) Fuel behavior under loss-of-coolant accident situations. Nucl Eng Technol 37:327–362

    Google Scholar 

  • Chung H, Kassner T (1998) Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup. Nucl Eng Des 186: 411–427

    Google Scholar 

  • Cox B (1990) Pellet-clad interaction (PCI) failures of zirconium alloy fuel cladding a review. J Nucl Mater 172:249–292

    Google Scholar 

  • Cozzo C, Staicu D, Pagliosa G, Papaioannou D, Rondinella V, Konings RJM, Walker C, Barker M, Herve PJ (2010) Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM. J Nucl Mater, doi:10.1016/j.jnucmat.2010.03.006

    Google Scholar 

  • Croff AG, Bjerke MA, Morrison GW, Petrie LM (1978) Revised uranium-plutonium cycle pwr and bwr models for the ORIGEN computer code. Tech. Rep. ORNL/TM-6051, Oak Ridge National Laboratory

    Google Scholar 

  • Cunningham M, Feshley M, Lanning D (1992) Development and characteristics of the rim region in high burnup UO2 fuel pellets. J Nucl Mater 188:19–27

    Google Scholar 

  • Cunningham ME, Beyer CE, Medvedev PG, Berna GA, Scott H (2001) Fraptan: a computer code for the transient analysis of oxide fuel rods, NUREG/CR-6739, vol 1, PNNL-13576

    Google Scholar 

  • Delafoy C, Dewes P, Miles T (2007) AREVA NP Cr2O3-doped fuel development for BWRs. In: Proceedings of the 2007 international LWR fuel performance meeting, Paper 1071. San Francisco, California

    Google Scholar 

  • Desquines J, Cazalis B, Bernaudat C, Poussard C, Averty X, Yvon P (2005) Mechanical properties of Zircaloy-4 PWR fuel cladding with burnup 54–64 MWd/kgU and implications for RIA behavior. In: Proceedings of the 14th international symposium on zirconium in the nuclear industry, ASTM STP 1467, pp 851–872

    Google Scholar 

  • Dupin N, Ansara I, Servant C, Toffolond C, Lemaignanc C, Brachet JC (1999) A thermodynamic database for zirconium alloys. J Nucl Mater 275: 287–295

    Google Scholar 

  • Duriez C, Alessandri JP, Gervais T, Philipponneau Y (2000) Thermal conductivity of hypostoichiometric low Pu content (U,Pu)O2 − x mixed oxide. J Nucl Mater 277:143–158

    Google Scholar 

  • Federici E, Lamare F, Bessiron V, Papin J (2001) The SCANAIR code version 3.2: main features and status of qualification. In: Proceedings of the technical committee meeting on fuel behaviour under transient and LOCA conditions, TECDOC-1320. IAEA, pp 88–101

    Google Scholar 

  • Fink JK (2000) Thermophysical properties of uranium dioxide. J Nucl Mater 279:1–18

    Google Scholar 

  • Fischer U, Wiese HW (1983) Verbesserte konsistente berechnung des nuklearen inventars abge brannter dwr-brennstoffe auf der basis des zell-abbrand-verfahrens mit korigen. Tech. Rep. KfK-3014, Kernforschungszentrum Karlsruhe

    Google Scholar 

  • Forgeron T, Brachet JC, Barcelo F, Castaing A, Hivroz J, Mardon J, Bernaudat C (2000) Experiment and modeling of advanced fuel rod cladding behavior under LOCA conditions: alpha-beta phase transformation kinetics and EDGAR methodology ASTM STP 1354, pp 256–279

    Google Scholar 

  • Forsberg K, Massih AR (2001) Theory of fission gas release in a growing grain 16th International Conference on Structural Mechanics in Reactor Technology (SMIRT-16), Washington DC, USA, 12–17 August 2001

    Google Scholar 

  • Fuketa T, Sasajima H, Mori Y, Ishijima K (1997) Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions. J Nucl Mater 248:249–256

    Google Scholar 

  • Fukushima S, Ohmichi T, Handa M (1986) The effect of rare earths on thermal conductivity of uranium, plutonium and their mived ovide fuels. J Less-Common Met 121:631–639

    Google Scholar 

  • Garcia P, Struzik C, Agard M, Louche V (2002) Mono-dimensional mechanical modelling of fuel rods under normal and off-normal operating conditions. Nucl Eng Des 216:183–201

    Google Scholar 

  • Garner N, Rentmeister T, Lippert HJ, Mollard P (2007) Upgraded fuel assemblies for BWRs. In: Proceedings of the 2007 international LWR fuel performance meeting, Paper 1091. San Francisco, California

    Google Scholar 

  • Geelhood KJ, Beyer CE, Cunningham ME (2004) Modifications to Fraptran to predict fuel rod failures due to PCMI during RIA-type accidents. In: Proceedings of the 2004 international meeting on LWR fuel performance, Paper 1097. Orlando, Florida, pp 585–595

    Google Scholar 

  • Gibby RL (1971) The effect of plutonium content on the thermal conductivity of (U, Pu)O2 solid solutions J Nucl Mater 38:163–177

    Google Scholar 

  • Grimes RW, Catlow CRA (1991) The stability of fission products in uranium dioxide. Phil Trans R Soc London A 335:609–634

    Google Scholar 

  • Gyori C, Hózer Z, Lassmann K, Schubert A, van de Laar J, Cvan M, Hatala B (2003) In:EU Research in Reactor Safety, Conclusion Symposium on Shared-Cost and Concerted Actions (FISA-2003), Proceedings EUR 21026, Luxembourg, November 10–13, 2003, pp 584–589

    Google Scholar 

  • Hagrman DL, Reymann GA (1979) MATPRO version 11-A, Handbook of materials properties for use in the analysis of light water reactor fuel rod behavior, 3rd edn. TREENUREC-1280, Advanced Inorganic Chemistry

    Google Scholar 

  • Hanley HJM (1973) The viscosity and thermal conductivity coefficients of dilute argon, krypton, and xenon. J Phys Chem Ref Data 2:619–642

    Google Scholar 

  • Harbottle JE, Kennard MW, Sunderland DJ, Strasser AA (1994) The behaviour of defective BWR barrier and non-barrier fuel. In: Proceedings of the international topical meeting on LWR fuel performance. West Palm Beach, Florida, April 17–21

    Google Scholar 

  • Harding J, Martin D (1989) A recommendation for the thermal conductivity of UO2. J Nucl Mater 166:223–226

    Google Scholar 

  • Haüssinger P, Glatthaar R, Rhode W, Kick H, Benkmann C, Weber J, Wunschel HJ, Stenke V, Leicht E, Stenger H, Groll G (1991) Noble gases. In: Ullmann’s encyclopedia of industrial chemistry. VCH Verlagsgesellschaft AG, Weinheim, pp 485–540

    Google Scholar 

  • Heins L, Groß H, Nissen K, Wunderlich F (1991) Statistical analysis of QC data and estimation of fuel rod behavior. J Nucl Mater 178:287–295

    Google Scholar 

  • Hermann A, Yagnik S, Gavillet D (2007) Effect of local hydride accumulations on Zircaloy cladding mechanical properties. In: 15th international symposium on zirconium in the nulcear industry. Sunriver, Oregon

    Google Scholar 

  • Hiernaut JP, Hyland GJ, Ronchi C (1993) Premelting transition in uranium dioxide. Int J Thermophys 14:259–283

    Google Scholar 

  • Higgs JD, Lewis BJ, Thompson WT, He Z (2007) A conceptual model for the fuel oxidation of defective fuel. J Nucl Mater 366:99–128

    Google Scholar 

  • Hoffmann PB, Dewes P (2004) Post-irradiation examination and ramp testing of fuel rods with Fe-enhanced Zr liner cladding at high burnup. In: Proceedings of the 2004 international meeting on LWR fuel performance, Paper 1059. Orlando, Florida

    Google Scholar 

  • Hollasky N, Valtonen K, Hache G, Gross H, Bakker K, Recio M, Bart G, Zimmermann M, van Duisboerg W, Killeen J, Meyer R (2000) Fuel safety criteria technical review. Tech. Rep. NEA/CSNI/R(99)25, Nuclear Energy Agency, Committee on the safety of nuclear installations

    Google Scholar 

  • Hoppe N (1980) Improvements to Comethe III-j fuel rod modelling code. Nucl Eng Des 56:123–133

    Google Scholar 

  • IAEA (1995) Characteristics and use of urania-gadolinia fuels. Tech. Rep. IAEA-TECDOC-844, International Atomic Energy Agency

    Google Scholar 

  • IAEA (1997) Thermophysical properties of materials for light water reactors. Tech. Rep. IAEA-TECDOC-949

    Google Scholar 

  • IAEA (2006) Thermophysical properties database of materials for light water reactors and heavy water reactors. Tech. Rep. IAEA-TECDOC-1496

    Google Scholar 

  • Ishimoto S, Hirai M, Ito K, Korei Y (1994) Effects of soluble fission products on thermal conductivities of nuclear fuel pellets. J Nucl Sci Technol 31:796–802

    Google Scholar 

  • Jackson PA, Turnbull JA, White RJ (1990) Enigma fuel performance code. Nucl Energy 29:107–114

    Google Scholar 

  • Jankus VZ, Weeks RW (1972) LIFE-II — A computer analysis of fast-reactor fuel-element behavior as a function of reactor operating history. Nucl Eng Des 18:83–96

    Google Scholar 

  • Jerlerud Pérez R, Massih AR (2007) Thermodynamic evaluation of the Nb-O-Zr system. J Nucl Mater 360:242–254

    Google Scholar 

  • Jernkvist L (2006) Computational assessment of burnup-dependent fuel failure thresholds for reactivity initiated accidents. J Nucl Sci Technol 43:546–561

    Google Scholar 

  • Kaji Y, Tsuru T (2008) Investigation of model for stress corrosion cracking of cladding materials. In: Proceedings of the 7th international workshop materials modelling and simulations or nuclear fuel MMSNF 7, EUR 23556 EN. Karlsruhe, Germany

    Google Scholar 

  • Kamemaya T, Matsumura T, Kinoshita M (1994) Numerical analysis for microstructure change of a light water reactor fuel pellet at high burnup. Nucl Technol 106:334–341

    Google Scholar 

  • Kang KW, Yang JH, Kim JH, Rhee YW, Kim DJ, Kim KS, Song KW (2007) The solidus and liquidus temperatures of UO2-Gd2O3 and UO2-Er2O3 fuels. Thermochim Acta 455:134–137

    Google Scholar 

  • Karlsson J, Lysell G, Pettersson H, Rönnberg G (2004) In-pile testing of liner-cladding and pellet performance in failed fuel rods. In: Proceedings of the international topical meeting on LWR fuel performance, Paper 1046. Orlando, Florida

    Google Scholar 

  • Kato M, Morimoto K, Sugata H, Konashi K, Kashimura M, Abe T (2008) Solidus and liquidus temperatures in the UO2-PuO2 system. J Nucl Mater 373:237–245

    Google Scholar 

  • Killeen J, Turnbull J, Sartori E (2007) Fuel modelling at extended burnup: IAEA coordinated research project FUMEX-II. In: Proceedings of the 2007 international LWR fuel performance meeting, paper 1102. San Francisco, California, pp 261–273

    Google Scholar 

  • Kinoshita M, Sonoda T, Kitajima S, Kolstad E, Matzke H, Rondinella V, Stalios A, Walker C, Ray I, Sheindlin M, Halton D, Ronchi C (2000) High burn-up rim project, (ii) irradiation and examination to investigate rim-structured fuel. In: Proceedings of the international topical meeting on LWR fuel performance. Park City, Utah

    Google Scholar 

  • Kogai T (1997) Modelling of fission gas release and gaseous swelling of light water reactor fuels. J Nucl Mater 244:131–140

    Google Scholar 

  • Koo YH, Sohn DS, Volkov B (1997) A comparative analysis of UO2 and MOX fuel behaviour under reactivity initiated accident conditions. Ann Nucl Energy 24:859–870

    Google Scholar 

  • Koo YH, Sohn DS, Yoon YK (1994) An analysis method for the fuel rod gap inventory of unstable fission products during steady-state operation. J Nucl Mater 209:62–78

    Google Scholar 

  • Kreyns P, Spahr G, McCauley J (1976) An analysis of iodine stress corrosion cracking of Zircaloy-4 tubing. J Nucl Mater 61:203–212

    Google Scholar 

  • Kurchatov SY, Likhanskii VV, Sorokin AA, Khoruzhii OV (2002) RTOP-code simulation of the radial distribution of heat release and plutonium isotope accumulation in high burnup oxide fuel. Atom Energy 92:349–356

    Google Scholar 

  • Kuroda M, Yamanaka S, Nagase F, Uetsuka H (2001) Analysis of the fracture behavior of hydrided fuel cladding by fracture mechanics. Nucl Eng Des 203:185–194

    Google Scholar 

  • Lassmann K (1980) The structure of fuel element codes. Nucl Eng Des 57:17–39

    Google Scholar 

  • Lassmann K (1992) TRANSURANUS: a fuel rod analysis code ready for use. J Nucl Mater 188:295–302

    Google Scholar 

  • Lassmann K, Benk H (2000) Numerical algorithms for intragranular fission gas release. J Nucl Mater 280: 127–135

    Google Scholar 

  • Lassmann K, Hohlefeld F (1987) The revised URGAP model to describe the gap conductance between fuel and cladding. Nucl Eng Des 103:215–221

    Google Scholar 

  • Lassmann K, O’Carroll C, van de Laar J, Walker CT (1994) The radial distribution of plutonium in high burnup UO2 fuels. J Nucl Mater 208:223–231

    Google Scholar 

  • Lassmann K, Schubert A, van de Laar J, Vennix C (2000) Recent developments of the TRANSURANUS code with emphasis on high burnup phenomena. In: Proceedings of the IAEA technical committee meeting on nuclear fuel behaviour modelling at high burnup, IAEA-TECDOC-1233. Windermere, UK, pp 387–406

    Google Scholar 

  • Lassmann K, Schubert A, Van Uffelen P, van de Laar J (2005) TRANSURANUS Handbook v1m1j09 (2009), European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany

    Google Scholar 

  • Lassmann K, Van Uffelen P (2004) The structure of fuel rod codes. Tech. Rep. EUR 21400 EN, JRC Publications, European Commission

    Google Scholar 

  • Lassmann K, Walker CT, van de Laar J (1998) Extension of the TRANSURANUS burnup model to heavy water reactor conditions. J Nucl Mater 255: 222–233

    Google Scholar 

  • Le Saux M (2008) Comportement et rupture de gaines en Zircaloy-4 d’etendu vierges, hydrurées ou irradiées en situation accidentelle de type RIA. Ph.D. thesis, Ecole National Superieure des Mines, MINES Paris Tech

    Google Scholar 

  • Le Saux M, Besson J, Carassour S, Poussard C, Averty X (2008) A model to describe the anisotropic viscoplastic mechanical behviour of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions. J Nucl Mater 378:60–69

    Google Scholar 

  • Leclercq S, Parrot A, Leroy M (2008) Failure characteristics of cladding tubes under RIA conditions. Nucl Eng Des 238: 2206–2218

    Google Scholar 

  • Lee BH, Koo YH, Oh JY, Cheon JS, Sohn DS (2007) Improvement of fuel performance code COSMOS with recent in-pile data for MOX and UO2. Fuels Nucl Technol 157:53–64

    Google Scholar 

  • Lee CB, Kim BG, Song JS, Bang JG, Jung YH (2000) RAPID model to predict radial burnup distribution in LWR UO2 fuel. J Nucl Mater 282:196–204

    Google Scholar 

  • Lemehov S, Nakamura J, Suzuki M (2001) PLUTON: A three-group model for the radial distribution of plutonium, burnup, and power profiles in highly irradiated LWR fuel rods. Nucl Technol 133:153–168

    Google Scholar 

  • Lewis BJ, Thompson WT, Akbari F, Thompson DM, Thurgood C, Higgs J (2004) Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel. J Nucl Mater 328:180–196

    Google Scholar 

  • Link T, Koss D, Motta A (1998) Failure of Zircaloy cladding under transverse plane-strain deformation. Nucl Eng Des 186:379–394

    Google Scholar 

  • Lösönen P (2000) On the behaviour of intragranular fission gas in UO2 fuel. J Nucl Mater 280:56–72

    Google Scholar 

  • Lyon WL, Bailey WE (1967) The solid-liquid phase diagram for the UO2-PuO2 system. J Nucl Mater 22:332–339

    Google Scholar 

  • MacDonald P, Seiffert S, Martinson Z, McCardell R, Owen D, Fukuda S (1980) Assessment of light-water-reactor fuel damage during a reactivity-initiated-accident. Nucl Safety 21:582–602

    Google Scholar 

  • Manzel R, Walker CT (2002) EPMA and SEM of fuel samples from PWR rods with an average burn-up of around 100 MWd/kgHM. J Nucl Mater 301:170–182

    Google Scholar 

  • Marchal N, Campos C, Garnier C (2009) Finite element simulation of pellet-cladding interaction PCI. Comput Mat Sci 821–826

    Google Scholar 

  • Martin DG (1988) The thermal expansion of solid UO2 and (U, Pu) mixed oxides - a review and recommendations. J Nucl Mater 152:94–101

    Google Scholar 

  • MATPRO (1981) Version 11 (Revision 2), A handbook of material properties for use in the analysis of light water reactor fuel rod behaviour

    Google Scholar 

  • MATPRO (1990) Report NUREG/CR-5273, EGG-2555

    Google Scholar 

  • MATPRO (1993) NUREG/CR-6150, EGG-2720, SCDAP/RELAP5/MOD3.1 Code Manual volume IV. In: Hagrman DT (ed) MATPRO – A library of materials properties for Light-Water Reactor accident analysis. November 1993, Idaho National Engineering Laboratory

    Google Scholar 

  • Mattas R, Yaggee F, Neimark L (1979) Iodine stress-corrosion cracking in irradiated Zircaloy cladding. In: Proceedings topical meeting LWR fuel. Portland, Oregon

    Google Scholar 

  • Mattys HM (1968) Plutonium oxide as nuclear fuel. Actinides Rev 1:165–182

    Google Scholar 

  • Matzke H (1980) Gas release mechanisms in UO2—a critical review Radiat Eff 53:219–242

    Google Scholar 

  • Meyer R, McCardell R, Chung H, Diamond H, Scott H (1996) A regulatory assessment of test data for reactivity initiated accidents. Nucl Safety 37:271–288

    Google Scholar 

  • Michel B, Sercombe J, Nonon C, Fandeur O (2010) Modelling of pellet cladding interaction. In: Konings RJM, Allen TR, Stoller RE, Yamanaka S (eds) Comprehensive Nuclear Materials. Elsevier, Oxford

    Google Scholar 

  • Michel B, Sercombe J, Thouvenin G (2008) A new phenomenological criterion for pellet cladding interaction. Nucl Eng Des 238:1612–1628

    Google Scholar 

  • Miller A, Ocken H, Tasooji A (1981) Iodine stress corrision cracking of Zircaloy: laboratory data, a phenomenological mode, and predictions of in-reactor behavior. J Nucl Mater 254–268

    Google Scholar 

  • Misfeldt I (1977) A stress corrosion failure model. In: Proceedings of IAEA specialists’ meeting on fuel element performance computer modelling blackpool. Report SRE-14–77. Risoe National Laboratory

    Google Scholar 

  • Mogensen M, Pearce J, Walker C (1999) Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from XRF investigation. J Nucl Mater 264:99–112

    Google Scholar 

  • Musienko A, Cailletaud G (2009) Simulation of inter- and intragranular crack propagation in polycrystlline aggregates due to stress corrosion cracking. Acta Mater 57:3840–3855

    Google Scholar 

  • Nagase F, Fuketa T (2004) Effects of pre-hydriding on thermal resistance of Zirealoy-4 cladding under simulated loss-of-coolant accident conditions. J Nucl Sci Tech 41(7):723–730

    Google Scholar 

  • Newton TD, Hutton JL (2002) The next generation WIMS lattice code: WIMS9 In: PHYSOR 2002 – international conference on the new frontiers of nuclear technology: reactor physics, safety and high-performance computing, 14A-04, American Nuclear Society, Seoul, Korea, October 7–10, 2002, ISBN 0-89448-672-1

    Google Scholar 

  • Nichols AL, Aldama DL, Verpelli M (2008) Handbook of nuclear data for safeguards. Tech. Rep. INDC(NDS)-0502, IAEA – International Nuclear Data Committee, Vienna

    Google Scholar 

  • Noirot L (2005) MARGARET: an advanced mechanistic model of fission gas behavior in nuclear fuel. In: International topical meeting on light water reactor fuel performance. ANS, Kyoto, Japan, October 2–5, 2005. OECD paper No. 1067

    Google Scholar 

  • Nonon C, Menard J, Lansiart S, Noirot J, Martin S, Decroix G, Rabouille O, Delafoy C, Petitprez B (2004) PCI behaviour of chromium oxide doped fuel. In: Proceedings of the international seminar on pellet-clad interaction in water reactor fuels. OECD-NEA, Aix-en Provence, France

    Google Scholar 

  • Oguma M (1983) Cracking and relocation behavior of nuclear fuel pellets during rise to power. Nucl Eng Des 76:35–45

    Google Scholar 

  • Olander D (1976) Fundamental aspects of nuclear reactor fuel elements. Tech. Rep. TID-26711-P1, Technical Information Center, Office of Public Affairs Energy Research and Development Administration

    Google Scholar 

  • Olander DR, Kim YS, Wang WE, Yagnik SK (1999) Steam oxidation of fuel in defective LWR rods. J Nucl Mater 270:11–20

    Google Scholar 

  • Olander DR, Mubayi V (1999) Review of the materials-chemistry models in the VICTORIA code. J Nucl Mater 270:1–10

    Google Scholar 

  • Olander DR, Uffelen PV (2001) On the role of grain boundary diffusion in fission gas release. J Nucl Mater 288:137–147

    Google Scholar 

  • Palmer ID, Hesketh KW, Jackson PA (1983) Water reactor fuel element performance computer modelling. Applied Science Barking, UK, p 321

    Google Scholar 

  • Park S, Kim J, Lee MH, Jeon YH (2008) Stress-corrosion crack initiation and propagation behavior of Zircaloy-4 cladding under iodine environment. J Nucl Mater 372:293–303

    Google Scholar 

  • Peck SO (1980) Automated uncertainty analysis methods in the FRAP computer codes, IAEA Specialist’s Meeting on Fuel Element Performance Computer Modeling, Blackpool, UK, March 17–21, 1980, pp 260–266, IAEA Summary report IWGFPT/7

    Google Scholar 

  • Powers DA (2000) Technical issues associated with air ingress during core degradation. Tech. Rep. SAND2000, 1935C, SANDIA National Laboratory

    Google Scholar 

  • Rashid J, Montgomery R, Lyon W, Yang R (2001)A cladding failure model for fuel rods subjected to operational transients and accident transients. In: Nuclear fuel behaviour modeling at high burnup and its experimental support. Proceedings of a Technical Committee meeting, June 19–23, 2000. Windermere, UK, IAEA-TECDOC-1233. IAEA, pp 187–199

    Google Scholar 

  • Rest J, Zawadski SA (1992) Fastgrass: a mechanistic model for the prediction of xe, i, cs, te, ba and sr release from nuclear fuel under normal and sever-accident conditions. Tech. Rep. NUREG/CR-5840, ANL-92/3, Argonne National Laboratory

    Google Scholar 

  • Ronchi C, Elton PT (1986) Radiation re-solution of fission gas in uranium dioxide and carbide. J Nucl Mater 140:228–244

    Google Scholar 

  • Ronchi C, Sheindlin M, Staicu D, Kinoshita M (2004) Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t − 1. J Nucl Mater 327:58–76

    Google Scholar 

  • Ronchi C, Shiendlin M, Musella M, Hyland GJ (1999) Thermal conductivity of uranium dioxide up to 2900 K from simultaneous measurement of the heat capacity and thermal diffusivity. J Appl Phys 85:776–789

    Google Scholar 

  • Ross AM, Stoute RL (1962) Heat transfer coefficient between UO2 and Zircaloy-2. Tech. Rep. AECL-1552, AECL

    Google Scholar 

  • Rousselier G, Leclercq S, Diard O (2003) Scenario for the damage of PWR fuel cladding in situations of pellet-cladding interaction. In: Transactions of the 17th international conference on structural mechanics in reactor technology (SMiRT 17). Prague, Czech Republic, pp C03–3

    Google Scholar 

  • Sartori E, Killeen J, Turnbull J (2007) Interna-tional fuel performance experiments (IFPE) database (January 2010). http://www.nea.fr/html/science/fuel/ifpelst.html 2009.

  • Sasajima H, Fuketa T, Nakamura T, Nakamura J, Kikuchi K (2000) Behavior of irradiated atr/mox fuel under reactivity initiated accident conditions. J Nucl Sci Technol 37:455

    Google Scholar 

  • Schire D, Grapengiesser B, Hallstadius L, Lundholm L, Lysell G, Frenning G, Ronnberg G, Jonsson A (1994) Secondary defect behaviour in sc abb bwr fuel. In: Proceedngs international topical meeting on LWR fuel performance. West Palm Beach, Florida, pp 398–409

    Google Scholar 

  • Schmidt HE (1970) Commission of the European Communities. European Institute for Transuranium Elements, Karlsruhe, Tech. Rep. Progress Report No. 9, July-December 1969, No. 2576, p. 29

    Google Scholar 

  • Schmidt HE (1971) Die Warmeleitfahigkeit von Uran- und Uran-Plutonium dioxyd. High Temp-High Press 3:345–353

    Google Scholar 

  • Schubert A, Van Uffelen P, van de Laar J, Walker CT, Haeck W (2008) Extension of the TRANSURANUS burn-up model. J Nucl Mater 376:1–10

    Google Scholar 

  • Schuster I, Lemaignan C (1992) Influence of texture on iodine-induced stress corrosion cracking of Zircaloy-4 cladding tubes. J Nucl Mater 189:157–166

    Google Scholar 

  • Shimada S, Etoh E, Hayashi H, Tukuta Y (2004) A metallographic and fractographic study of outside-in cracking caused by power ramp tests. J Nucl Mater 327:97–113

    Google Scholar 

  • Sihver L, Hallstadius L, Wikmark G (1997) Recent ABB BWR failure experience. In: Proceedings international topical meeting on LWR fuel performance. Portland, Oregon, pp. 356–364

    Google Scholar 

  • Solyany VI, Bibilashvily YK, Tonkov VY (1983) In: Proceeding OECD-NEA-CSNI/IAEA Specialists meeting on water reactor fuel safety and fission product release in off-normal and accident conditions, May 16–20, 1983. Risoe, Denmark, p 163

    Google Scholar 

  • Sonoda T, Kinoshita M, Ray I, Wiss T, Thiele H, Pellottiero D, Matzke H (2002) Transmission electron microscopy observation on irradiation-induced microstructural evolution in high burnup UO2 disk fuel. Nucl Instr Methods Phys Res B191:622–628

    Google Scholar 

  • Sonoda T, Matzke H, Kinoshita M (1999) High burnup rim project: (iv) threshold burnup of rim structure formation. In: Project OHR (ed) Proceedings of the enlarged Halden program group meeting. Loen, Norway

    Google Scholar 

  • Sontheimer F, Landskron H, Billaux M (1998) A fuel thermal conductivity correlation based on the latest experimental results. In: Proceedings of the seminar on thermal performance of high burn-up LWR fuel. Cadarache, France, p 119

    Google Scholar 

  • Speight MV (1969) A calculation on the migration of fission gas in material exhibiting precipitation and re-solution of gas atoms under irradiation. Nucl Sci Eng 37:180–185

    Google Scholar 

  • Spino J, Stalios A, Santa Cruz H, Baron D (2006) Stereological evolution of the rim structure in PWR-fuels at prolonged irradiation. J Nucl Mater 354:66–84

    Google Scholar 

  • Stan, M. (2009) Discovery and Design of Nuclear Fuels. Materials Today 12, 36-44.

    Google Scholar 

  • Stammler JJ, Boerresen S, Casal JJ, Forslund P (1996) Helios - Verification against Kritz and other critical experiments. In: PHYSOR 1996 – international conference on physics of reactors, September 16–20, 1996, American Nuclear Society, Mito, Ibaraki, Japan

    Google Scholar 

  • Suk HC, Wang W, Kim BG, Kim KS, Heo YH (1992) In: Technical committee meeting on fission gas release and fuel chemistry related to extended burnup, April 1992, Pembroke, Ontario, Canada, pp 193–201

    Google Scholar 

  • Suzuki M (2000) Analysis of high burnup fuel behavior in Halden reactor by FEMAXI-V code. Nucl Eng Des 201:99–106

    Google Scholar 

  • Suzuki M (2000) Light water reactor fuel analysis code Femaxi-v (ver.1)

    Google Scholar 

  • Trkov A, Molnár GL, Révay Z, Mughabghab SF, Firestone RB, Pronyaev V, Nichols AL, Moxon MC (2005) Revisiting the 238U thermal capture cross section and gamma-ray emission probabilities from 239Np Decay. Nucl Sci Eng 150:336–348

    Google Scholar 

  • Tsederberg NV, Popov VN, Morozova NA (1971) Thermodyanmic and thermophysical properties of Helium. Israel Program for Scientific Translations, Jerusalem

    Google Scholar 

  • Turnbull JA (1980) A review of irradiation induced re-solution in oxide fuels. Rad Effects 53:243–250

    Google Scholar 

  • Turnbull JA, Friskney C, Findlay J, Johnson F, Walter AJ (1982) The diffusion coefficients of gaseous and volatile species during the irradiation of uranium dioxide. J Nucl Mater 107:168– 184

    Google Scholar 

  • Uetsuka H, Furuta T, Kawasaki S (1983) failure-bearing capability of oxidized zircaloy-4 cladding under simulated loss-of-coolant Condition J Nucl Sci Techn 20:941–950

    Google Scholar 

  • Une K, Amaya M, Imamura M, Korei Y (1995) Fission gas release from defective BWR fuels. J Nucl Mater 226:323–326

    Google Scholar 

  • Une K, Imamura M, Amaya M, Korei Y (1995) Fuel oxidation and irradiation behaviors of defective BWR fuel rods. J Nucl Mater 223:40–50

    Google Scholar 

  • Urbanic VF, Heidrick TR (1978) High-temperature oxidation of Zircaloy-2 and Zircaloy-4 in steam. J Nucl Mater 75:251–261

    Google Scholar 

  • US-NRC (1981) Standard review plan for the review of safety analysis reports for nuclear power plants, LWR edition.

    Google Scholar 

  • van der Linde A (1965) Calculation of the safe life time expectancy of zirconium alloy canning in the fuel Tech. Rep. RCN-41, Reactor Centrum Nederland

    Google Scholar 

  • Van Swam L, Strasser A, Cook J, Burger J (1997) Behaviour of Zircaloy-4 and zirconium liner Zircaloy-4 cladding at high burnup. In: Proceedings international topical meeting on LWR fuel performance. Portland, Oregon, pp 421–431

    Google Scholar 

  • Van Uffelen P (2002) Contribution to the modelling of fission gas release in light water reactor fuel. Ph.D. thesis, University of Liege, Nuclear Engineering

    Google Scholar 

  • Van Uffelen P, Gyori C, Schubert A, van de Laar J, Hózer Z, Spykman G (2008) Extending the application range of a fuel performance code from normal operating to design basis accident conditions. J Nucl Mater 383:137-143

    Google Scholar 

  • Van Uffelen P, Jonnet J, Ronchi C (2004) Open questions related to the high burnup structure in nuclear fuels. In: Proceedings of the workshop on materials modelling and simulations for nuclear fuels. Washington, DC

    Google Scholar 

  • Van Uffelen P, Sheindlin M, Rondinella V, Ronchi C (2004) On the relations between the fission gas behaviour and the pellet-cladding mechanical interaction in LSR fuel rods In: Seminar on Pellet-clad interaction in water reactor fuels (PCI-2004), CEA Cadarache/DEN/DEC, Aix en Provence, France, March 9–11 2004, OECD, Paper 14

    Google Scholar 

  • Vitanza C (2006) RIA failure threshold and LOCA limit at high burn-up. J Nucl Sci Technol 43:1074–1079

    Google Scholar 

  • Vitanza C (2007) A review and interpretation of RIA experiments. Nucl Eng Technol 39:591–602

    Google Scholar 

  • Vitanza C, Conde Lopez JM (2005) PCMI implicatons for high burn-up light water reactor fuel in reactivity-initiated accidents. In: Seminar on pellet-clad interaction in water reactor fuels (PCI-2004) Aix-en-Provence, March 9–11, 2004, OECD, Paper 5

    Google Scholar 

  • Vitanza C, Graziani U, Fordestrommen NT, Vilpponen KO (1978) Fission gas release from in-pile measurements. Tech. Rep. HPR-221.10, OECD Halden Reactor Project

    Google Scholar 

  • Volkov BY, Viktorov VF, Platonov PA, Rjazantzeva A (1989) Library of subprograms on physical and mechanical properties of the n1-alloy fuel rod cladding material. Tech. Rep. KIAE-4941/11, Institute of Atomic Energy I.V. Kurchatov

    Google Scholar 

  • Walker C, Kamemaya T, Kitajama S, Kinoshita M (1992) Concerning the microstructure changes that occur at the surface of UO2 pellets on irradiation to high burnup. J Nucl Mater 188: 73–79

    Google Scholar 

  • Walker CT, Bremier S, Portier S, Hasnaoui R, Goll W (2009) SIMS analysis of an UO2 fuel irradiatedat low temperature to 65 MWd/kgHM. J Nucl Mater 393:212–223

    Google Scholar 

  • Walker CT, Staicu D, Sheindlin M, Papaioannou D, Goll W, Sontheimer F (2006) On the thermal conductivity of UO2 nuclear fuel at burnup of around 100 MWd/kgHM. J Nucl Mater 350:19–39

    Google Scholar 

  • White RJ (2004) The development of grain-face porosity in irradiated oxide fuel. J Nucl Mater 325:61–77

    Google Scholar 

  • White RJ, Tucker MO (1983) A new fission-gas release model. J Nucl Mater 118: 1–38

    Google Scholar 

  • Wiesenack W (1997) In: Proceedings of the international topical meeting on light water reactor fuel performance, March 2–6, 1997, Portland, Oregon. Assessment of UO2 conductivity degradation based on in-pile temperature data

    Google Scholar 

  • Wood J (1972) Factors affecting stress corrosion cracking of Zircaloy in iodine vapour. J Nucl Mater 45:105–122

    Google Scholar 

  • Xin X, Lai W, Liu B (2009) Point defect properties in hcp and bcc Zr with trace solute Nb revealed by ab initio calculations. J Nucl Mater 393: 197–202

    Google Scholar 

  • Yamanouchi S, Tachibana T, Tsukui K, Oguma M (1988) Melting Temperature of Irradiated UO2 and UO2-2wt%Gd2O3 Fuel Pellets up to Burnup of about 30 GWd/tU. J Nucl Sci Techol 25:528–533

    Google Scholar 

  • Yamnikov VM, Malanchenko LL (1977) Variation of thermal conductivity of a gas mixture under the fuel-element jacket during burn-up. Atomic Energy 42:358–360

    Google Scholar 

  • Zhou G, LindbÃ¥ck JE, Schutte HC, Jernkvist LO, Massih AR (2005) Modelling pellet clad interaction during power ramps,. In: Proceedings international seminar on pellet-clad interaction in water reactor fuels. OECD-NEA, Aix-en Provence, France

    Google Scholar 

  • Zimmerer W (1978) Darstellung der neu integrierten stoffdaten-funktionen im system maplib in tabellarischer und graphischer form. Tech. Rep. KfK-Ext.8/78-3, Kernforschungszentrum Karlsruhe

    Google Scholar 

Download references

Acknowledgment

The authors wish to thank D. Staicu (Institute for Transuranium Elements) for fruitful discussions, W. Goll (areva np) and Nathalie Dupin (Calcul Thermo) for providing some of the figures and J. van der Laar for help for the preparation of some of the figures.

Author information

Authors and Affiliations

  1. European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany

    Paul Van Uffelen

  2. European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany

    Rudy J. M. Konings

  3. OECD Halden Reactor Project, Halden, Norway

    Carlo Vitanza

  4. 202 Nuclear Science Center, University of Florida, Gainesville, FL, USA

    James Tulenko

Authors
  1. Paul Van Uffelen
    View author publications

    You can also search for this author in PubMed Google Scholar

  2. Rudy J. M. Konings
    View author publications

    You can also search for this author in PubMed Google Scholar

  3. Carlo Vitanza
    View author publications

    You can also search for this author in PubMed Google Scholar

  4. James Tulenko
    View author publications

    You can also search for this author in PubMed Google Scholar

Editor information

Editors and Affiliations

  1. Institute for Nuclear Technology and Reactor Safety, Karlsruher Institut fur Technologie (KIT), Gotthard-Franz-Str. 9, 76131, Karlsruhe, Germany

    Dan Gabriel Cacuci

Rights and permissions

Reprints and permissions

Copyright information

© 2010 Springer Science+Business Media, LLC

About this entry

Cite this entry

Uffelen, P.V., Konings, R.J.M., Vitanza, C., Tulenko, J. (2010). Analysis of Reactor Fuel Rod Behavior. In: Cacuci, D.G. (eds) Handbook of Nuclear Engineering. Springer, Boston, MA. https://doi.org/10.1007/978-0-387-98149-9_13

Download citation

Publish with us

Policies and ethics

Search

Navigation