Fuel Cladding Integrity – Postulated Accidents

Nuclear Fuel - TemperaturesFuel cladding is the outer layer of the fuel rods, standing between the reactor coolant and the nuclear fuel (i.e. fuel pellets). It is made of a corrosion-resistant material with low absorption cross section for thermal neutrons (~ 0.18 × 10–24 cm2), usually zirconium alloy. Cladding prevents radioactive fission products from escaping the fuel matrix into the reactor coolant and contaminating it. Cladding constitute one of barriers in ‘defence-in-depth‘ approach, therefore its coolability is one of key safety aspects.

Fuel Cladding Integrity – Postulated Accidents

Fuel rod cladding is the first barrier for retention of fission products, and the structural integrity of the cladding ensures coolable core geometry. Integrity and coolability of the fuel cladding are the two major issues to be considered under all the reactor operating conditions, especially during postulated Design-basis Accidents (DBAs). Note that, a design-basis accident is a postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structures, and components necessary to ensure public health and safety.
There are two main categories of postulated accidents, for which fuel cladding integrity and core coolability must be ensured:
  1. LOCA conditions. LOCA (loss of coolant accident) accidents are postulated accidents that result in a loss of reactor coolant at a rate in excess of the capability of the reactor makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system. LOCA conditions are associated with rapid decrease in system pressure, cladding balloning, rupture and high temperature steam oxidation. In January 1974, the USNRC published 10 CFR 50.46 establishing acceptance criteria for the ECCSs for LWRs, addressing safety limits that must be assured under LOCA conditions:
    1. Maximum zircaloy cladding temperature (Peak Cladding Temperature),
    2. Maximum oxidation of cladding,
    3. Maximum amount of hydrogen generated by chemical reaction of the zirconium alloy with water and/or steam,
    4. Coolable core geometry,
    5. Long term cooling.
  2. Non-LOCA conditions. Non-LOCA conditions that affect fuel cladding integrity typically include accidents as a main steamline break (MSLB) and reactivity initiated accidents (RIAs). MSLBs are associated with power excursion and with the risk of departure from nucleate boiling, which results in high temperature steam oxidation. RIAs consist of postulated accidents which involve a sudden and rapid insertion of positive reactivity. In these accidents, the large and rapid deposition of energy in the fuel can result in melting, fragmentation, and dispersal of fuel. The mechanical action associated with fuel dispersal can be sufficient to destroy the cladding and the rod-bundle geometry of the fuel and produce pressure pulses in the primary system.

See also: Enthalpy of Nuclear Fuel

 

References:
Nuclear and Reactor Physics:
      1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
      2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
      3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
      4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
      5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
      6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
      7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
      8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

      1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
      2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
      3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2. 
      4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See above:

Fuel Cladding