Accident tolerant fuel cladding development: Promise, status, and challenges

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

Modelling and analysis of the MSFR transient behaviour

In nuclear reactor physics, deterministic and hybrid calculation methods dominate the field of transient analysis. This implies that important safety assessments are subject to many approximations, which are needed by these methods. This paper proposes the Dynamic Monte Carlo method (Dynamic MC), which solves the coupled Boltzmann and kinetic equations with exact geometry and continuous energy, using only Monte Carlo techniques.For Dynamic MC a number of new techniques are developed, e.g., precursor tracking, forced decay for precursors, and the branchless method. Also, the particle source of the simulation has to be determined differently from what is current standard Monte Carlo practice, and the simulation scheme is adapted.A few example cases are simulated, demonstrating the effectiveness of Dynamic MC. The sample cases vary from simple homogeneous systems to full fuel assemblies with an asymmetric flux profile during the transient. Since Dynamic MC is implemented in the general-purpose Monte ...

Dynamic Monte Carlo Method for Nuclear Reactor Kinetics Calculations

(1987). Introductory Nuclear Reactor Dynamics. Nuclear Technology: Vol. 76, No. 2, pp. 311-312.

Introductory Nuclear Reactor Dynamics

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

https://catatanstudi.files.wordpress.com/2009/11/2006-the-pbmr-steady-state-and-coupled-kinetics-core-thermal-hydraulics-benchmark-test-problems_reitsma.pdf

The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems

The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5. The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature. For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor-10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system. It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.

Validation of the dalton-thermix code system with transient analyses of the htr-10 and application to the pbmr

/pdf/advanced-demonstration-and-test-reactor-options-study-2kq7ltbve8.pdf

Gerhard Strydom

Papers

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems

Validation of the dalton-thermix code system with transient analyses of the htr-10 and application to the pbmr

Advanced Demonstration and Test Reactor Options Study

The US Department of Energy’s high temperature reactor research and development program – Progress as of 2019