scholarly journals A New Numerical Nuclear Reactor Neutronics Code SHARK

2021 ◽  
Vol 9 ◽  
Author(s):  
Chen Zhao ◽  
Xingjie Peng ◽  
Hongbo Zhang ◽  
Wenbo Zhao ◽  
Zhang Chen ◽  
...  

In order to establish the next-generation reactor physics calculation method based on the numerical nuclear reactor technology and realize high-fidelity modeling and calculation, a new numerical nuclear reactor neutronics code SHARK is developed. The code is based on the direct transport method with construct solid geometry (CSG) method, advanced subgroup resonance method, direct transport MOC method in rectangle and hexagonal geometry, large-scale parallel, and CMFD acceleration method. The C5G7, macro BEAVRS and VERA benchmarks are verified to show the accuracy of the code and method. Numerical results show good accuracy and calculation performance of SHARK, and the direct transport method can be adopted on numerical nuclear reactor calculation.

2021 ◽  
Vol 247 ◽  
pp. 06045
Author(s):  
Lu Cao ◽  
Zhouyu Liu ◽  
Qingming He ◽  
Liangzhi Cao

With the development of nuclear reactor, new complicated designs introduce some challenges for the resonance method in the NECP-X code, which is called as the Global-Local method. There are two limitations of the present NECP-X. One is the geometry modeling limitation, and the other one is that the Global-Local resonance method can only deal with the cylindrical fuel rods in the current version. Therefore, some fuels like plate, annular fuels cannot be calculated in the NECP-X code because of these limitations. To overcome above issues, the constructive solid geometry (CSG) method is developed to model the complex problem, and the capability of constructing and solving the equivalent 1D plate and 1D annual fuel pin based on the framework of Global-Local method, is developed. A set of tests are calculated, including multi-annular fuel problems and plate fuel problems. The results show good accuracy of the new developed resonance self-shielding method.


Author(s):  
Wenping Hu ◽  
Shengyao Jiang ◽  
Xingtuan Yang

Pebble-bed nuclear reactor technology, with a reactor core typically composed of spherical pebbles draining very slowly in a continuous refueling process, is currently being revived around the world. But the dense slow pebble flow in the reactor, which has an important impact on reactor physics, is still poorly understood. Under such circumstance, this article studies mathematical models which are potential to research the pebbles motion in the pebble-bed reactor, including void model, spot model and DEM model. The fundamental principles of these models are introduced, the success and deficiency of each model is briefly analyzed. Theoretically, it’s expected that spot model and DEM model may be more practical to apply on studying the pebble dynamics. Though, spot model still needs to be refined based on further experimentation, and more research is necessary to solve the problem of huge computational time in order to make the DEM model simulation technique a really practical notion.


2019 ◽  
Vol 13 (2) ◽  
pp. 161-170
Author(s):  
Mardiyanto Mangun Panitra ◽  
Syahfandi Ahda ◽  
Abu Kahlid Rivai

At present, nuclear reactor technology that is widely used because of its proven reliability is the gen-III + nuclear reactor. Even if it is seen from the aspect of safety and reliability of this generation reactor, it has been proven, but because nuclear energy plays a vital role to meet the growing world energy needs, it is necessary to have a type of nuclear reactor that is tailored to those needs.The next generation of nuclear reactors must meet the requirements of fulfilling safety requirements, be flexible, a longer operating life (more than 60 years), more economical. In order for a reactor to produce higher power, a longer operating life and more economical, reactor structure materials which are capable of being operated at high temperatures are needed. The types of materials that are expected to meet these requirements include various types of ferritic / martensite steel, austenite, alloy steel containing nickel, and metal glass materials and ceramic materials. FeCr metal alloys are alloys that form the metals mentioned above, so it is important to conduct research both in simulation and experiment. Molecular Dynamics simulation of FeCr alloys using Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS) has been done to explore their thermodynamic characteristics such as heat treatment, solubility of Cr, atomic radial distribution function (RDF). The results of the simulation are illustrated using Visual Molecular Dynamics (VMD) code. 


2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Xinyu Zhou ◽  
Zhouyu Liu ◽  
Lu Cao ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
...  

Abstract The constructive solid geometry (CSG) method is widely used in the Monte Carlo community because of its geometry flexibility. As the requirements of designing new reactors and development of deterministic methods, the CSG method is adopted for the geometry modeling in the deterministic numerical nuclear reactor physics codes to construct complex geometries in recent years. In the new developed numerical nuclear reactor physics code NECP-X, the CSG method is also implemented to expand its geometric modeling capability, but it is difficult to efficiently calculate the volume of complex geometries. This work develops a new efficient method for calculating volume of arbitrarily complex geometries for NECP-X. Rather than implementing the sampling method, the new developed method is based on the characteristic ray information, which is used for the method of characteristics (MOC) sweeping in NECP-X. The implementation shows that this method is very convenient to be applied to the MOC codes. A set of cases such as pin-cell and hexagonal rod-bundle assembly problems are tested to show the accuracy and performance.


Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.


Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.


Author(s):  
Robert Zboray ◽  
Domenico Paladino ◽  
Olivier Auban

The present paper discusses experiments carried out to examine mixing of different gases (steam, air) and the evolution their distributions in large-scale, multi compartment geometry imitating nuclear reactor containment compartments. The flow and the mixing process in the experiments are driven by plumes and jets representing source structures with different momentum-to-buoyancy strength. The time evolution of the relevant parameters like gas concentrations, velocities and temperatures are followed using dedicated instrumentation. The data obtained is meant to be used for the validation and development of high-resolution, mainly CFD based, 3D computational tools for nuclear reactor containment safety analysis.


1953 ◽  
Vol 21 (4) ◽  
pp. 300-304
Author(s):  
E. D. Klema ◽  
R. J. Stephenson ◽  
S. Taylor

Author(s):  
Sue Ion

This chapter will cover the nuclear fission option as a future energy supply, and will essentially address the question: can nuclear fission plug the gap until the potential of nuclear fusion is actually realized? (The potential for fusion is considered in detail chapter 7.) To put this question into context, let us first look at some of the key issues associated with nuclear fission, which currently supplies around one fifth of the UK’s electricity. Most large scale power stations produce electricity by generating steam, which is used to power a turbine. In a nuclear power station, the principle is the same, but instead of burning coal, oil, or gas to turn water into steam, the heat energy comes from a nuclear reactor. A reactor contains nuclear fuel, which remains in place for several months at a time, but over that time it generates a huge amount of energy. The fuel is usually made of uranium, often in the form of small pellets of uranium dioxide, a ceramic, stacked inside hollow metal tubes or fuel rods, which can be anything from a metre to four metres in length, depending on the reactor design. Each rod is about the diameter of a pencil, and the rods are assembled into carefully designed bundles, which in turn are fixed in place securely within the reactor. There are two isotopes (or different types) of uranium, and only one of these is a material which is ‘fissionable’—that is to say, if an atom of this uranium isotope is hit by a neutron, then it can split into two smaller atoms, giving off energy in the process and also emitting more neutrons. This, and other pathways, are illustrated in Fig. 6.1 (Source: CEA). Controlling the reaction, so that the energy from the fission of uranium atoms is given out slowly over a period of years, requires two aspects of the process to be carefully balanced. 1. First, there must be enough fissile atoms in the fuel so that—on average— each fission leads to exactly one other. Any fewer, and the reaction will die away.


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