IAEA CRP Benchmark of Kalinin VVER-1000 NPP: An Analysis Using EXCEL-TRIHEX-FA Code System

2008 ◽  
Author(s):  
Suhail Ahmad Khan ◽  
V. Jagannathan ◽  
R. P. Jain
Keyword(s):  
Power Distribution ◽  
Boron Concentration ◽  
Transport Theory ◽  
Fuel Cycle ◽  
Diffusion Theory ◽  
Nuclear Data ◽  
Core Level ◽  
Coolant Temperature ◽  
Code System ◽  
Density Values

Two units of VVER-1000Mwe reactors are in an advanced stage of construction at Kudankulam, Tamilnadu, India. With a view to assess the capability of analyzing the physics characteristics of VVER cores, the IAEA CRP benchmark problem of Kalinin VVER-1000 MWe NPP [1] is studied using the indigenous code system EXCEL-TRIHEX-FA [2,3]. The lattice burnup code EXCEL is based on a combination of 1-D multigroup transport theory and 2-D few group diffusion theory. Nuclear data in 172 group WIMS-D format based on JEFF-3.1 [4] has been used in the present analysis. The core level calculations are performed using the code TRIHEX-FA which solves the 3-D multigroup diffusion equation using the finite difference method with fine triangular meshes. Power dependent feedback models for xenon, Doppler, coolant temperature and density values have been incorporated in TRIHEX-FA. Keff for the critical soluble boron concentration, assembly power distribution and axial power distribution are calculated as a function of fuel cycle burnup. In the present paper, lattice level results are compared with the results of other participants reported in Ref. [1]. The results of core level calculations have been compared with the experimental data provided [1].

Boron Transient Analysis by the GRS Coupled Code Systems QUABOX-CUBBOX/ATHLET and TORT-TD/ATHLET

2009 ◽  
Author(s):  
K. Velkov ◽  
A. Seubert ◽  
I. Pasichnyk ◽  
A. Pautz
Keyword(s):  
Transient Analysis ◽  
Transport Theory ◽  
State Of The Art ◽  
Diffusion Theory ◽  
Coarse Mesh ◽  
Code System ◽  
The Core ◽  
Simulation Results ◽  
The Impact ◽  
Transport Approach

The application of modern coupled thermal-hydraulic neutron-kinetic code systems is state of the art for performing safety analyses. In this paper, a radially asymmetric boron dilution transient in a PWR MOX/UO2 core is defined. The transient is analysed using the coupled code system QUABOX-CUBBOX/ATHLET which is based on fuel-assembly coarse mesh diffusion theory, and the results are compared to the solution obtained with the transport theory-based coupled code system TORT-TD/ATHLET with a pin-wise representation of the core. The aim of this study is to investigate the impact of the transport approach and the diffusion approximation on the simulation results.

Comparison of nuclear data uncertainties with other nuclear fuel cycle uncertainty sources

2021 ◽  
Author(s):  
Aris V. Skarbeli ◽  
Rubén Eusebio‐Yebra ◽  
Pablo Romojaro ◽  
Francisco Álvarez‐Velarde ◽  
Daniel Cano‐Ott
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Data ◽  
Nuclear Fuel Cycle ◽  
Uncertainty Sources

scholarly journals Spectroscopy in heterogenous media and applications for bioprocess and environmental monitoring

2002 ◽  
Vol 4 (1) ◽  
pp. 21-26 ◽  
Author(s):  
Frank Schael ◽  
Oliver Reich ◽  
Sonja Engelhard
Keyword(s):  
Diffuse Reflectance ◽  
Near Infrared ◽  
Transport Theory ◽  
Radiation Transport ◽  
Diffusion Theory ◽  
Nile Blue ◽  
Milk Powder ◽  
Photon Migration ◽  
Heterogenous Media ◽  
Reflectance Measurements

Diffuse reflectance measurements and photon migration studies with near infrared (NIR) diode lasers were employed to elucidate experimental methods for determining absorption and scattering coefficients and species concentrations in heterogenous media. Measurements were performed at a number of wavelengths utilizing several laser sources some of which were widely tunable. In order to establish the applicability of simple photon migration models derived from radiation transport theory and to check the experimental boundary conditions of our measurements, simple light scattering solutions (such as suspensions of titanium dioxide, latex particles, and solutions of milk powder) containing dyes (such as nile blue, isosulfan blue) were investigated. The results obtained from diffuse-reflectance studies at different sourcedetector distances were in accordance with predictions from simple photon diffusion theory. Applications of reflectance measurements for monitoring of cell growth during fermentation processes and forin-situinvestigations of soils are presented.

Nuclear Data for the Thorium Fuel Cycle and the Transmutation of Nuclear Waste

2016 ◽  
pp. 207-214
Author(s):  
F. Gunsing ◽  
S. Altstadt ◽  
J. Andrzejewski ◽  
L. Audouin ◽  
M. Barbagallo ◽  
...  
Keyword(s):  
Nuclear Waste ◽  
Fuel Cycle ◽  
Nuclear Data

Nuclear Data Development Related to the Th–U Fuel Cycle in China

2016 ◽  
pp. 195-198
Author(s):  
Haicheng Wu ◽  
Zhigang Ge ◽  
Weixiang Yu ◽  
Xiaolong Huang ◽  
Nengchuan Shu ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Nuclear Data

A New Method on Flux Mapping by Function Expansion

2010 ◽  
Author(s):  
Bin Zhong ◽  
Kan Wang ◽  
Ganglin Yu
Keyword(s):  
Power Distribution ◽  
Synthesis Method ◽  
Expansion Method ◽  
Mapping Method ◽  
Least Square ◽  
New Method ◽  
Coolant Temperature ◽  
Fuel Temperature ◽  
Function Expansion ◽  
Flux Mapping

The core flux (power) distribution is very important to safe and economical operation of nuclear reactor. It can be obtained by many methods depending on the desired accuracy and execution time. For on-line core surveillance and regulation, we need to get the real-time flux distribution. If the true local parameters such as fuel temperature, coolant temperature and material density were known, the solution of the diffusion equation with instantaneous parameters could, in principle, provide the necessary spatial details. However, in reality, it is impossible to obtain the operational “readings” of these parameters for each fuel cell. The detector results at certain locations can be applied to improve the results of the only diffusion calculations by Flux Mapping methods. Function expansion method is employed to express the approximate real distribution by the combination of several Flux Mapping method results as the expansion basis functions. The Harmonics Synthesis Method (HSM) and Least-Square method are combined to get a new Flux Mapping method in this paper. The simulation results show that the new method can be used for Flux Mapping and get better results.

Design Study of Small Lead-Cooled Fast Reactors Using SiC Cladding and Structure

2006 ◽  
Author(s):  
Abu Khalid Rivai ◽  
Minoru Takahashi
Keyword(s):  
Power Distribution ◽  
Inner Core ◽  
Fuel Cycle ◽  
Outer Core ◽  
Structure Type ◽  
Natural Uranium ◽  
Neutron Energy Spectrum ◽  
Fast Reactors ◽  
Steel Type ◽  
Type Reactor

Effects of SiC cladding and structure on neutronics of reactor core for small lead-cooled fast reactors have been investigated analytically. The fuel of this reactor was uranium nitride with 235U enrichment of 11% in inner core and 13% in outer core. The reactors were designed by optimizing the use of natural uranium blanket and nitride fuel to prolong the fuel cycle. The fuels can be used without reshuffling for 15 years. The coolant of this reactor was lead. A calculation was also conducted for steel cladding and structure type as comparison with SiC cladding and structure type. The results of calculation indicated that the neutron energy spectrum of the core using SiC was slightly softer than that using steel. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have critical condition with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The result of the design analysis showed that neutron flux distributions and power distribution was made flatter because the outer core enrichment was higher than inner core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 13% for 125 MWt reactor and 25% for 375 MWt reactor at EOC.

Application of Fine Depletion Calculation for LLFPs In-Core Transmutation Evaluation in PWRs

2012 ◽  
Author(s):  
Kun Liu ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Youqi Zheng ◽  
Changhui Wang
Keyword(s):  
Power Distribution ◽  
Critical Concentration ◽  
Verification And Validation ◽  
Code System ◽  
Fuel Management ◽  
Numerical Accuracy ◽  
Analysis And Evaluation ◽  
Evaluation Code ◽  
Flux Calculation ◽  
Numerical Performance

An in-core transmutation analysis and evaluation code, named CATE, considering in-core fine flux calculation and fine depletion process, is verified and validated in the present paper. Verification and Validation of implementations for the OECD/NEA PWR cell benchmark for actinides transmutation, IAEA PWR benchmark and infinite homogenized plate problem to confirm reliability and numerical accuracy for the code have been performed in presented paper. The numerical performance of the code system is demonstrated in the analyses of the in-core fuel management calculation. It is found that the present code system gives stability in prediction of critical concentration of boric solution and radial power distribution. Based on the verifications and validations, a preliminary LLFP transmutation pattern is calculated. Numerical results indicate that CATE can be used not only for the fuel management calculation, but also for in-core transmutation evaluation of PWR.

scholarly journals CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

2018 ◽  
Vol 4 ◽  
pp. 10 ◽  
Author(s):  
Guillaume Ritter ◽  
Romain Eschbach ◽  
Richard Girieud ◽  
Maxime Soulard
Keyword(s):  
Nuclear Fuel ◽  
Cross Section ◽  
Cross Sections ◽  
Power Plants ◽  
Fuel Cycle ◽  
Nuclear Data ◽  
Computer Code ◽  
Test Case ◽  
Industrial Facilities ◽  
Transport Code

CESAR stands in French for “simplified depletion applied to reprocessing”. The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ∼400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with “industrial nuclear” constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR’s). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA − Cadarache.

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