scholarly journals MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS

2021 ◽  
Vol 247 ◽  
pp. 10031
Author(s):  
Nicholas P. Luciano ◽  
Brian J. Ade ◽  
Kang Seog Kim ◽  
Andrew J. Conant
Keyword(s):  
Cross Section ◽  
Neutron Transport ◽  
Three Dimensional ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Water Reactors ◽  
Improved Performance ◽  
Burnup Calculation

MPACT is a state-of-the-art core simulator designed to perform high-fidelity analysis using whole-core, three-dimensional, pin-resolved neutron transport calculations on modern parallel computing hardware. MPACT was originally developed to model light water reactors, and its capabilities are being extended to simulate gas-cooled, graphite-moderated cores such as Magnox reactors. To verify MPACT’s performance in this new application, the code is being formally benchmarked using representative problems. Progression problems are a series of example models that increase in complexity designed to test a code’s performance. The progression problems include both beginning-of-cycle and depletion calculations. Reference solutions for each progression problem have been generated using Serpent 2, a continuous-energy Monte Carlo reactor physics burnup calculation code. Using the neutron multiplication eigenvalue ke_ as a metric, MPACT’s performance is assessed on each of the progression problems. Initial results showed that MPACT’s multigroup cross section libraries, originally developed for pressurized water reactor problems, were not sufficient to accurately solve Magnox problems. MPACT’s improved performance on the progression problems is demonstrated using this new optimized cross section library.

scholarly journals DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA

2021 ◽  
Vol 247 ◽  
pp. 10016
Author(s):  
Kang Seog Kim ◽  
Brian J. Ade ◽  
Nicholas P. Luciano
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Group Structure ◽  
Homogenization Method ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Neutron Leakage ◽  
Water Reactors ◽  
Neutronic Characteristics

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnox reactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Neutronic Study of Burnable Poison Materials and Their Alternative Configurations in Fully Ceramic Microencapsulated Loaded Pressurized Water Reactor Fuel Assembly

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Muhammad Qasim Awan ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
Chuanqi Zhao
Keyword(s):  
Life Cycle ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Burnable Poison ◽  
Attractive Option ◽  
Plant Operations ◽  
Water Reactors ◽  
Control Rods ◽  
Double Heterogeneity

Use of FCM fuel in light water reactors is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. This work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and to keep the moderator temperature coefficient of reactivity (MTC) feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configurations including quadruple isotropic (QUADRISO), bi-isotropic (BISO), and Matrix Mix forms. The results obtained from the implicit random treatment of the double heterogeneity of tri-structural isotropic (TRISO), QUADRISO, and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5%. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at end of cycle (EOC); however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At the beginning of cycle (BOC), erbia-loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.

The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes

2002 ◽  
Author(s):  
Jean-Franc¸ois Pignatel
Keyword(s):  
Natural Circulation ◽  
Atomic Energy Commission ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Large Area ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Subject ◽  
Water Reactors ◽  
Predictive Study

Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

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.

Scope

2021 ◽  
pp. 1-2
Author(s):  
Wei Shen ◽  
Benjamin Rouben
Keyword(s):  
Point Of View ◽  
Nuclear Industry ◽  
Light Water Reactors ◽  
Reactor Physics ◽  
Light Water ◽  
Computational Schemes ◽  
The World ◽  
Water Reactors ◽  
And Mathematics

From the educational point of view, there are many textbooks on reactor physics used at various universities in the world. However, most of these textbooks focus either on application to Light Water Reactors (LWRs), or on the theory and mathematics, with a significant number of equations and computational schemes. Or else they were written more than 20, or even more than 60, years ago, and therefore they do not reflect the evolution of reactor concepts and engineering requirements since then. All those categories of books are either difficult to follow for non-physicists working in the nuclear industry, or else are of little value for those who are interested in special features of CANDU reactor physics.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Sign in / Sign up

Export Citation Format

Share Document