ANDREA 2: Improved Version of Code for Reactor Core Analysis

2014 ◽  
Author(s):  
František Havlůj ◽  
Radim Vočka ◽  
Jiří Vysoudil
Keyword(s):  
Experimental Data ◽  
Cross Section ◽  
Reactor Core ◽  
Object Oriented ◽  
Core Analysis ◽  
Modular Architecture ◽  
The Core ◽  
Code Structure ◽  
Ongoing Development ◽  
Data Library

In recent years, the core physics code ANDREA has been significantly improved and its capabilities were vastly extended. The code implementation has been overhauled to more modern, object-oriented and modular architecture. The code structure was adapted to allow use of multiple different neutronics solvers and to tackle various spatial and energy discretization models. The data library formats and processing workflow have been completely generalized, and different transport codes (e.g. HELIOS, SERPENT or SCALE) can be used to prepare several-group cross-section libraries for ANDREA. The new version of the code has gone through extensive validation both on the benchmarks and experimental data. The modular architecture of ANDREA code allow for its ongoing development. Currently we focus on coupling of ANDREA code with TRANSURANUS code.

RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis

2002 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
3D Model ◽  
Reactor Core ◽  
Core Analysis ◽  
Reactor Operation ◽  
The Core ◽  
Calculation Results ◽  
Plant Data ◽  
Comparison Of The Results ◽  
Real Plant ◽  
Good Agreement

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data.

scholarly journals Double Heterogeneity Neutronic Simulation of HTR-10 for First Criticality and Full Power Initial Core with TORT-TD code

2021 ◽  
Vol 927 (1) ◽  
pp. 012037
Author(s):  
Daddy Setyawan
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
The Core ◽  
Full Power ◽  
Double Heterogeneity ◽  
High Temperature Gas

Abstract In order to support the verification and validation of computational methods and codes for the safety assessment of pebble bed High-Temperature Gas-cooled Reactors (HTGRs), the calculation of first criticality and full power initial core of the high-temperature pebble bed reactor 10 MWt (HTR-10) has been defined as one of the problems specified for both code-to-code and code-to-experiment benchmarking with a focus on neutronics. HTR-10 Experimental facility serves as the source of information for the currently designed high-temperature gas-cooled nuclear reactor. It is also desired to verify the existing codes against the data obtained in the facility. In HTR-10, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in the reactor consist of TRISO particles, which are embedded in the graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneity of this type of reactor. In this paper, the first criticality and the power distribution in full power initial core calculations of HTR-10 are used to demonstrate treatment of this double heterogeneity using TORT-TD and Serpent for cross-section generation. HTR-10 has unique characteristics in terms of the randomness in geometry, as in all pebble bed reactors. In this technique, the core structure is modeled by TORT-TD, and Serpent is used to provide the cross-section in a double heterogeneity approach. Results obtained by TORT-TD calculations are compared with available data. It is observed that TORT-TD calculation yield sufficiently accurate results in terms of initial criticality and power distribution in full power initial core of the HTR-10 reactor.

Code Development for the Neutronics/Thermal-Hydraulics Coupling Transient Analysis of Molten Salt Reactors

2017 ◽  
Author(s):  
Tianliang Hu ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
Kun Zhuang
Keyword(s):  
Molten Salt ◽  
Cross Section ◽  
Cross Sections ◽  
Transient Analysis ◽  
Reactor Core ◽  
Dynamics Simulation ◽  
Monte Carlo Code ◽  
Section Data ◽  
Dynamics Simulations ◽  
Data Library

A code system has been developed in this paper for the dynamics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY and TANSY_K which is based on OpenFOAM to perform the steady-state full-core coupled simulations and dynamics simulation. For verification and application of the codes sequence, the simulation of a representative molten salt reactor core MOSART has been performed. For the further study of the characteristics of MSRs, several transients like the code-slug transient, unprotected loss of flow transient and overcooling transient have been analyzed. The numerical results indicated that the TANSY and TANSY_K codes with the cross section library generated by OpenMC has the capability for the dynamics analysis of MSRs.

Fusion and Breakup Reactions of 17S + 208Pb and 15C + 232ThHalo Nuclei Systems

2015 ◽  
Vol 11 (2) ◽  
pp. 2972-2978
Author(s):  
Fouad A. Majeed ◽  
Yousif A. Abdul-Hussien
Keyword(s):  
Experimental Data ◽  
Cross Section ◽  
Fusion Reaction ◽  
Fusion Cross Section ◽  
Reaction Cross ◽  
Recent Experimental Data ◽  
Barrier Distribution ◽  
Coupled Channel ◽  
Total Fusion ◽  
Full Quantum

In this study the calculations of the total fusion reaction cross section have been performed for fusion reaction systems 17F + 208Pb and 15C + 232Th which involving halo nuclei by using a semiclassical approach.The semiclassical treatment is comprising the WKB approximation to describe the relative motion between target and projectile nuclei, and Continuum Discretized Coupled Channel (CDCC) method to describe the intrinsic motion for both target and projectile nuclei. For the same of comparsion a full quantum mechanical clacualtions have been preforemd using the (CCFULL) code. Our theorticalrestuls are compared with the full quantum mechaincialcalcuations and with the recent experimental data for the total fusion reaction  checking the stability of the distancesThe coupled channel calculations of the total fusion cross section σfus, and the fusion barrier distribution Dfus. The comparsion with experiment proves that the semiclassiacl approach adopted in the present work reproduce the experimental data better that the full quantal mechanical calcautions. 

SETTING A DIAGRAM APPROACH TO CALCULATING VIBRATED, CENTRIFUGED AND VIBROCENTRIFUGED REINFORCED CONCRETE COLUMNS WITH A VARIATROPIC STRUCTURE

2020 ◽  
pp. 22-34
Author(s):  
Л. Р. Маилян ◽  
С. А. Стельмах ◽  
Е. М. Щербань ◽  
М. П. Нажуев
Keyword(s):  
Experimental Data ◽  
Reinforced Concrete ◽  
Bearing Capacity ◽  
Cross Section ◽  
Concrete Columns ◽  
Reinforced Concrete Columns ◽  
The Cross ◽  
Concrete Elements ◽  
Diagram Approach

Состояние проблемы. Железобетонные элементы изготавливаются, как правило, по трем основным технологиям - вибрированием, центрифугированием и виброцентрифугированием. Однако все основные расчетные зависимости для определения их несущей способности выведены, исходя из основного постулата - постоянства и равенства характеристик бетона по сечению, что реализуется лишь в вибрированных колоннах. Результаты. В рамках диаграммного подхода предложены итерационный, приближенный и упрощенный способы расчета несущей способности железобетонных вибрированных, центрифугированных и виброцентрифугированных колонн. Выводы. Расчет по диаграммному подходу показал существенно более подходящую сходимость с опытными данными, чем расчет по методике норм, а также дал лучшие результаты при использовании дифференциальных характеристик бетона, чем при использовании интегральных и, тем более, нормативных характеристик бетона. Statement of the problem. Reinforced concrete elements are typically manufactured according to three basic technologies - vibration, centrifugation and vibrocentrifugation. However, all the basic calculated dependencies for determining their bearing capacity were derived using the main postulate, i.e., the constancy and equality of the characteristics of concrete over the cross section, which is implemented only in vibrated columns. Results. Within the framework of the diagrammatic approach, iterative, approximate and simplified methods of calculating the bearing capacity of reinforced concrete vibrated, centrifuged and vibrocentrifuged columns are proposed. Conclusions. The calculation according to the diagrammatic approach showed a significantly better convergence with the experimental data than that using the method of norms, and also performs better when using differential characteristics of concrete than when employing integral and particularly standard characteristics of concrete.

An investigation of effects of level density models and gamma ray strength functions on cross-section calculations for the production of 90Y, 153Sm, 169Er, 177Lu and 186Re therapeutic radioisotopes via (n,γ) reactions

2019 ◽  
Vol 108 (1) ◽  
pp. 11-17
Author(s):  
Mert Şekerci ◽  
Hasan Özdoğan ◽  
Abdullah Kaplan
Keyword(s):  
Experimental Data ◽  
Cross Section ◽  
Production Cross Section ◽  
Gamma Ray ◽  
Production Cross ◽  
Level Density ◽  
Experimental Studies ◽  
Level Density Models ◽  
Strength Functions ◽  
Cancer Diseases

Abstract One of the methods used to treat different cancer diseases is the employment of therapeutic radioisotopes. Therefore, many clinical, theoretical and experimental studies are being carried out on those radioisotopes. In this study, the effects of level density models and gamma ray strength functions on the theoretical production cross-section calculations for the therapeutic radioisotopes 90Y, 153Sm, 169Er, 177Lu and 186Re in the (n,γ) route have been investigated. TALYS 1.9 code has been used by employing different level density models and gamma ray strength functions. The theoretically obtained data were compared with the experimental data taken from the literature. The results are presented graphically for better interpretation.

SEMIEMPIRICAL SCF–LCAO–MO TREATMENT OF THIOPHENE, FURAN, AND PYRROLE

1965 ◽  
Vol 43 (5) ◽  
pp. 1569-1576 ◽  
Author(s):  
N. Solony ◽  
F. W. Birss ◽  
John B. Greenshields
Keyword(s):  
Experimental Data ◽  
Satisfactory Agreement ◽  
Electronic Structures ◽  
Spectroscopic Data ◽  
Variable Parameter ◽  
Coulomb Repulsion ◽  
Resonance Integral ◽  
Electronic Transitions ◽  
The Core ◽  
Lcao Mo

The semiempirical SCF–LCAO–MO method of Pariser–Parr–Pople is utilized in the study of the π-electronic structures of thiophene, furan, and pyrrole. The core Hamiltonian expansion contains a Uz++ term, the potential due to the ionized hetero-atom contributing two electrons to the π-system. The γzz, one-center coulomb repulsion integral for the hetero-atom is evaluated from the experimental spectroscopic data only. With the resonance integral βczc as the only variable parameter, the calculated π*–π electronic transitions are in a satisfactory agreement with the experimental data.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

The probe particle scattering cross-section off the many-fermion systems in the impulse approximation

2021 ◽  
Author(s):  
Yahya Younesizadeh ◽  
Fayzollah Younesizadeh
Keyword(s):  
Experimental Data ◽  
Distribution Function ◽  
Spectral Function ◽  
Cross Section ◽  
Response Function ◽  
Momentum Distribution ◽  
Impulse Approximation ◽  
System Response ◽  
Probe Particle ◽  
Momentum Distribution Function

In this work, we study the differential scattering cross-section (DSCS) in the first-order Born approximation. It is not difficult to show that the DSCS can be simplified in terms of the system response function. Also, the system response function has this property to be written in terms of the spectral function and the momentum distribution function in the impulse approximation (IA) scheme. Therefore, the DSCS in the IA scheme can be formulated in terms of the spectral function and the momentum distribution function. On the other hand, the DSCS for an electron off the [Formula: see text] and [Formula: see text] nuclei is calculated in the harmonic oscillator shell model. The obtained results are compared with the experimental data, too. The most important result derived from this study is that the calculated DSCS in terms of the spectral function has a high agreement with the experimental data at the low-energy transfer, while the obtained DSCS in terms of the momentum distribution function does not. Therefore, we conclude that the response of a many-fermion system to a probe particle in IA must be written in terms of the spectral function for getting accurate theoretical results in the field of collision. This is another important result of our study.

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