Nuclear Safety Analysis Methodology
Best estimate models of physical processes, best estimate plant
states, and most probable system configurations and failure events
provide the most realistic representation of plant behaviour
and consequences during accidents. Deviations from these best
conditions can and will occur, resulting in uncertainty in the
outcome of a best estimate analysis. In order to quantify the
variability and uncertainty in the outcome of an accident, it
to identify and characterize the components contributing to uncertainty
and evaluate their impact on safety consequences. A primary objective
is to define, through the use of an integrated probabilistic
approach, the ranges of key plant parameters that assure that
are met at a prescribed level of probability and confidence.
Research in this area investigates approaches to propagation of
time-dependent variability and uncertainty in plant response during
accidents using dynamic sensitivity analysis and quasilinearization
methods. Of particular interest is the representation of nonlinear
bifurcation behaviour within a framework of quasilinearized sensitivity
analysis since this allows time-dependent solutions for multiple-parameter
variations to be generated from a limited set of detailed computer
simulations obtained from best estimate computer codes.
Experimental Studies in Nuclear Safety Thermalhydraulics
Boiling heat transfer from cylindrical nuclear fuel elements and
fuel channel calandria tubes is an important process in CANDU
reactor accidents. Specifically, changes in heat transfer regimes
from nucleate boiling to film boiling influence both the operational
safety margins and integrity of components during accidents.
Research in this area investigates fundamental aspects of limiting
heat transfer processes in two-phase systems. The major focus is
on investigating the fundamental processes of vapour bubble nucleation
and transport from the walls of horizontal cylindrical heated tubes.
This will be used to improve understanding of heat transfer limitations
during transitions in boiling regimes and vapour formation during
rapid depressurization of hot fluids.
Theoretical Modelling Studies in Nuclear Safety Thermalhydraulics
the integrity of engineered barriers to fission product release
is one primary objective of nuclear safety analysis.
Where heat transfer to fluids occurs across an engineered barrier
to fission product release - such as at the metal cladding (sheath)
of a nuclear fuel element or at the pressure tube/calandria tubes
of a fuel channel - a transition to a more limiting heat transfer
regime (e.g. transition from nucleate boiling to film boiling)
invariably poses a challenge to the integrity of the barrier. Additionally,
transition from the film boiling regime to the nucleate boiling
regime through the quenching process, in which a rapid increase
in heat transfer and concomitant rapid reduction in the temperature
of the wall occurs, is of importance in limiting the time during
which integrity of a barrier is subject to challenge.
Research in this area is focused on mechanistic modelling of near-field
processes including vapour bubble nucleation and transport processes
and associated local heat transfer processes, such as conduction
in metal walls and turbulent boundary layer convection, as they
influence development and spread of vapour film patches (drypatches)
on horizontal heated cylinders. Similarly, fundamental theoretical
mechanistic modelling of the complementary transition from stable
film boiling back to nucleate boiling (often referred to as the “quench” process)
are being investigated.
Selected Recent Publications
J.C.Luxat, “Analytical Evaluation of Probabilistic Safety
Margins”, Proc. 25th CNS Annual Conf., Toronto, Ontario,
J.C.Luxat, “Historical Perspective of the Design Basis in
Canada”, OECD/CSNI/CNRA Workshop on Redefining Large Break
LOCA, Zurich, Switzerland, June 2003.
D.W. Dormuth, P.J. Ingham and J.C.Luxat, “Application of
Heirarchical, Two-Tiered Scaling Method to Channel Coolant Voiding
in CANDU Reactors”, Proc. ANS Winter Meeting, Washington
D.C., November 17-21, 2002.
J.C.Luxat, “Fuel Behaviour During A Power Pulse: A Review
and Assessment of Reactivity Initiated Accident (RIA) Test Data”,
Proc. 23rd CNS Annual Conf., Toronto, Ontario, June 2002.
J.C.Luxat, “Mechanistic Modeling of Heat Transfer Processes
Governing Pressure Tube-To-Calandria Tube Contact and Fuel Channel
Failure”, Proc. 23rd CNS Annual Conf., Toronto, Ontario,
A. Popescu, J. Pascoe and J.C. Luxat , “Modifications of
ACAP for Use in Accuracy Assessment of Safety Analysis Software
at OPG”, Proc. 23rd CNS Annual Conf., Toronto, Ontario, June
J.C.Luxat and R. Huget, “Prototype Application Of Best Estimate
And Uncertainty Safety Analysis Methodology to Large LOCA Analysis”,
Proc. 22nd CNS Annual Conf., Toronto, Ontario, June 2001.
R. Huget, J.C.Luxat and D.K. Lau, “Application Of Best Estimate
And Uncertainty Safety Analysis Methodology to Loss of Flow Events
at Ontario Power Generation’s Darlington Nuclear Generating
Station”, Proc. 22nd CNS Annual Conf., Toronto, Ontario,
J.C.Luxat, “Review and Assessment Of Heavy Water Transport
Properties For Accident Analysis At High Temperature Conditions”,
Proc. 22nd CNS Annual Conf., Toronto, Ontario, June 2001.
J.C. Luxat, R.G Huget and F. Tran, “Development and Application
of Ontario Power Generation’s Best Estimate Nuclear Safety
Analysis Methodology”, Proc. ANS Topical Meeting on Best
Estimate Methods (BE-2000), Washington DC, November, 2000.
J.C.Luxat, “Safety Analysis Technology: Evolution, Revolution
and the Drive to Regain Margin”, Proc. 21st CNS Annual Conf.,
Toronto, Ontario, June 2000 (also CNS Bulletin, Vol. 21, No. 2,
J.C. Luxat, W. Kupferschmidt, P.D. Thompson and M-A Petrilli, “The
Industry Standard Toolset (IST) of Codes for CANDU Safety Analysis”,
Proc. OECD/CSNI Workshop: Advanced Thermal-Hydraulic And Neutronic
Codes: Current And Future Applications, Barcelona, Spain, April
J. Pascoe, B.H. McDonald, J.C. Luxat, D.J. Richards and B. Rouben, “Qualification
Programs for Computer Codes Used for Safety & Transient Analysis
in Canada”, Proc. OECD/CSNI Workshop: Advanced Thermal-Hydraulic
And Neutronic Codes: Current And Future Applications, Barcelona,
Spain, April 10-13, 2000.
J.C. Luxat, V. Snell, M-A Petrilli and P.D. Thompson, “Implementation
of Common Industry Safety Analysis Codes”, Proc. 20th CNS
Annual Conf., Montreal, Quebec, May30 – June 2, 1999.
P. Boczar, A. Dastur, K. Dormuth, A. Lee, D. Meneley, D. Pendergast,
J.C. Luxat, “Global Warming and Sustainable Energy Supply
with CANDU Nuclear Power Systems”, Progress in Nuclear Energy,
Vol. 32, No. 3/4, pp 297-304, 1998.
D.J. Richards, V.S. Krishnan and J..C. Luxat, “Best-Estimate
Methods in CANDU Reactor LOCA Analysis”, Proc. OECD/NEA/CSNAI
Seminar on Utilization of Best Estimate Methodology in Safety Analysis,
Ankara, Turkey, 1998.
P.J. Ingham, J.C. Luxat, A.J. Melnyk and T.V. Sanderson, “Natural
Circulation in an Integral CANDU Test Facility”, IAEA T/C
Mtg. On Experimental Tests and Qualification of Analytical Methods
to Address Thermohydraulic Phenomena in Advanced Water Cooled Reactors,
Villigen, Switzerland, September 14-17, 1998.
A.Gloaguen, J.C. Luxat, W. Thomas, S. Izumi, A.S. Polyakov, R.
Dodds, M. Bunn, L. Koch and N. Zarimpas, “Present Status
and Environmental Implications of the Different Fuel Cycles – Key
Issue Paper No. 2”, Proc. IAEA Int. Symp. on Nuclear Fuel
Cycle and Reactor Strategy: Adjusting to New Realities, Vienna,
Austria, June 3-6, 1997.
Boczar, P.G. , J.R. Hopkins, H. Feinroth and J.C. Luxat, “Plutonium
Disposition in CANDU”, Proc. IAEA Technical Meeting on Recycling
of Plutonium and Uranium in Water Reactor Fuels, Windermere, U.K.