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Using Mpi Portable Parallel Programming With The Message Passing Interface

This is cne of the r-are text books written in the discipline of Nuclear Reactor Analysis, where the author has considered the interests of both scientists and practising engineers, in addition to serving the needs of the academia. The most attractive feature of this book is a balanced treatment of theory and practice of the subject matter. The theoretical foundations of the reactor design methods are explained with simplified definitions and relevant practical illustrations. The author scans through quickly the traditional aspects of the so-called reactor physics and takes the reader through the details of the analytical aspects in a conventional manner. Hcwever, there is a definite departure from the classical method of approach in order to avoid lengthy and repetitive discussions, that are available in many standard text books on reactor physics. The chief departure fran tradition is the priority accorded to the treatment of the energy part of the problems as opposed to the spatial Dart normally devoted to by other authors . A similar unorthodox approach has been applied while dealing with the solution of the various equations by giving priority to computer oriented mrethods as opposed to the classical solutions.

Nuclear reactor analysis

Abstract Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.

Physics-based multiscale coupling for full core nuclear reactor simulation

Neutron Physics By K H Beckurts and K Wirtz Pp x + 444 (Berlin: Springer-Verlag, 1964) 68 DM Introduction to Neutron Distribution Theory By L C Woods (Methuen's Monographs on Physical Subjects) Pp xii + 132 (London: Methuen and Co, Ltd; New York: John Wiley and Sons, Inc, 1964) 28s net

Nuclear Reactor Physics

A new high-fidelity neutronics code NECP-X

A new three-dimensional method of characteristics for the neutron transport calculation

The pseudo-resonant-nuclide subgroup method (PRNSM) based global–local self-shielding calculation scheme is proposed to simultaneously resolve the local self-shielding effects (including spatial self-shielding effect and the resonance interference effect) for large-scale problems in reactor physics calculations. This method splits self-shielding calculation into global calculations and local calculations. The global calculations obtain the Dancoff correction factor for each pin cell by neutron current method. Then an equivalent one-dimensional (1D) cylindrical problem for each pin cell is isolated from the lattice system by preserving Dancoff correction factor. The local calculation is to perform self-shielding calculations of the equivalent 1D cylindrical problem by the PRNSM. The numerical results show that PRNSM obtains accurate spatial dependent self-shielded cross sections and improves the accuracy of dealing with the resonance interference over the conventional Bondarenko iteration method an...

Hongchun Wu

Papers

A new high-fidelity neutronics code NECP-X

A new three-dimensional method of characteristics for the neutron transport calculation

The pseudo-resonant-nuclide subgroup method based global–local self-shielding calculation scheme

NECP-Atlas: A new nuclear data processing code

NECP-MCX: A hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems