05 Nuclear fuels (scientific, technical)

06/01614 Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core Adorni, M. et al. Annals of Nuclear Energy, 2005, 32, (15), 1679-1692. The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA- TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.

06/01615 Application of two preconditioned generalized conjugate gradient methods to three-dimensional neutron and photon transport equations Chen, G. S. and Sheu, R. D. Progress' in Nuclear Energy, 2004, 45, (1), 11-23. In this paper the preconditioned generalized conjugate gradient methods are applied to solve the linear system of equations that arise in three-dimensional neutron and photon transport equations. These generalized conjugate gradient methods are the CGS (conjugate gradient square) algorithm and the Bi-CGSTAB (N-conjugate gradient stabilized) algorithm. Several subroutines are developed from these preconditioned generalized conjugate gradient methods for use in fime-independent multi-group three-dimensional neutron and photon transport equations. These subroutines are connected to the computer program TORT. The reason for choosing the preconditioned gener- alized conjugate gradient methods is that these methods have good residual error control procedures during computation and have good convergence rates. A problem was tested having an eight set of matrix equations with 24255 unknowns each, in a personal computer with an AMD Athlon-XP1600+ central processing unit (CPU) and using Mandrake Linux 6.3 as the operating system. The point-wise incomplete LU factorization (ILU) and modified point-wise incom- plete LU factorization (MILU) are the preconditioning techniques used in the test problem. It was found that the preconditioned CGS and Bi-CGSTAB methods with the preconditioner ILU are more efficient than with the preconditioner MILU in the test problem. The numerical solution of flux by the preconditioned Bi-CGSTAB and CGS methods produces the same results as those obtained by the successive over relaxation method (SOR) in the TORT program which is usually used for the calculation of radiation shielding in nuclear power plant and nuclear spent fuel storage.

06101616 Controlling corrosion in the back end of fuel cycle using nitric acid grade stainless steels Kain, V. and De, P. K. hzternational Journal of Nuclear Energy Science and Technology, 2005, 1, (2-3), 220-231, The materials for use in nitric acid service have to be resistant to intergranular corrosion and have low general corrosion rates. End grain corrosion resulting from inclusions and segregation along dislocation bands needs to be minimised. The nitric acid grade (NAG) of stainless steel resistant to these corrosion problems has been indigenously produced. It is emphasised that proper fabrication practices have to be employed for NAG grade to take full advantage of its cleanliness, controlled chemistry and microstructure. The problems associated with fabrication of large diameter pipes, oxide embedment and welding are described with the authors" experience in its production and quality assessment.

06101617 CRONOS2 and APOLL02 results for the NEA C5G7 MOX benchmark Moreau, F. et al. Progress in Nuclear Energy, 2004, 45, (2 4), 17%200. Results for the NEA C5G7 MOX benchmark are presented. The results for the 2D benchmark were obtained with CRONOS2 diffusion and $8 finite elements and with the characteristic method of APOLL02, and for the 3D benchmark with CRONOS2 $4 finite elements. Detailed results show the reliability and accuracy of the methods available in CRONOS2 and APOLL02. A recent calculation with the method of characteristics for structured heterogeneous cells is also discussed.

06/01618 DYNID-MSR dynamics code for molten salt reactors Krepel, J. et al. Annals o f Nuclear Energy, 2005, 32, (17), 1799-1824. This paper reports about the DYNID-MSR code development and dynamics studies of the molten salt reactors (MSR) - one of the 'Generation IV International Forum' concepts. In this forum the

graphite-moderated channel type MSR based on the previous Oak Ridge National Laboratory research is considered. The liquid molten salt serves as a fuel and coolant, simultaneously and causes two physical peculiarities: the fission energy is released predominantly directly into the coolant and the delayed neutrons precursors are drifted by the fuel flow. The drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit and it can lead to a reactivity loss or gain in the case of fuel flow acceleration or deceleration, respectively. Therefore, specific 3D tool based on in house code DYN3D was developed in FZR. The code DYN3D-MSR is based on the solution of two-group neutron diffusion equation by the help of a nodal expansion method and it includes models of delayed neutrons drift and specific MSR heat release distribution. In this paper the development and verification of 1D version DYN1D-MSR of the code is described. The code has been validated with the experimental data gained from the molten salt reactor experiment performed in the Oak Ridge and after the validation it was applied to several typical transients (overcooling of fuel at the core inlet, reactivity insertion, and the fuel pump trip).

06/01619 Global physical and numerical stability of a nuclear reactor core Morales-Sandoval, J. and Hernfindez-Solis, A. Annals of Nuclear Energy, 2005, 32, (15), 1666 1678. Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modelled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented.

06101620 HELIOS, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark Ivanov, B. D. et al. Progress in Nuclear Energy, 2004, 45, (2-4), 119- 124. As part of an effort to test the ability of current transport codes to treat reactor core problems without spatial homogenization, the lattice code HELIOS was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 MOX fuel assembly problem specified in an earlier study. This problem, known as C5G7 MOX benchmark, comprises two cases - two and three-dimensional geometry. There are four fuel assemblies - two with MOX fuel, the other two with UO2 fuel. Each fuel assembly is made up of a 17x17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the Benchmark Specification, which also provides seven-group trans- port-corrected isotropic scattering cross-sections for UO2, the three MOX enrichments, the guide tubes, the fission chamber and the moderator. This paper presents the methodology employed in solving the C5G7 MOX fuel assembly problem using the transport code HELIOS.

06•01621 Influence of the external neutron sources in the criticality prediction using I lM curve Pereira, V. et al. Annals of Nuclear Energy, 2005, 32, (17), 1875 1888. The influence of external neutron sources in the process to obtain the criticality condition is estimated. To reach this objective, the three- dimensional neutron diffusion equation in two groups of energy is solved, for a subcritical PWR reactor core with external neutron sources. The results are compared with the solution of the correspond- ing problem without external neutron sources, that is an eigenvalue problem. The method developed for this purposes it makes use of both the nodal method (for calculation of the neutron flux) and the finite differences method (for calculation of the adjoint flux). A coarse mesh finite difference method was developed for the adjoint flux calculation, which uses the output of the nodal expansion method. The results regarding the influence of the external neutron source presence for attaining criticality have shown that far from criticality it is necessary to calculate the reactivity values of the system.

250 Fuel and Energy Abstracts July 2006