scholarly journals 3-D COUPLED SIMULATION OF A VVER 1000 WITH PARCS/ATHLET

2021 ◽  
Vol 247 ◽  
pp. 06015
Author(s):  
Romain Henry ◽  
Yann Périn ◽  
Kiril Velkov ◽  
Sergei Pavlovich Nikonov
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Simulation Code ◽  
Control Rod ◽  
Lattice Codes ◽  
Physics Simulation ◽  
Two Phases ◽  
Effective Multiplication

A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems. Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations, neither on the geometry nor in the nuclear data. Since the whole core SERPENT models for production of such cross-section libraries would be computationally costly (and the standard 2-D approach may introduce unnecessary large approximations), 3-D models of each assembly type in infinite radial lattice configurations have been created. These cross-sections are then used to evaluate effective multiplication factors for different core configurations with the diffusion code PARCS. The results are compared with the reference SERPENT calculations. In the next step, a thermal-hydraulic model with the system code ATHLET applying an assembly-wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transient test calculations. This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, a simple exercise with coupled code system PARCS/ATHLET is presented and analysed.

scholarly journals Trends on Major Actinides from an Integral Data Assimilation

2019 ◽  
Vol 211 ◽  
pp. 03001
Author(s):  
Gerald Rimpault ◽  
Virginie Huy ◽  
Gilles Noguère
Keyword(s):  
Data Assimilation ◽  
Energy Range ◽  
Cross Section ◽  
Cross Sections ◽  
Capture Cross Section ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Capture Cross ◽  
Integral Data ◽  
Recent Evaluation

Nuclear data evaluations on major actinides can be improved by integral data assimilation. Appropriate integral measurements with reliable experimental techniques have been selected such as ICSBEP, IRPhE and MASURCA critical masses, PROFIL irradiation experiments and the FCAIX experimental programme (critical masses and spectral indices). Highly reliable analyses are possible with the use of as-built geometries calculated with the TRIPOLI4 Monte Carlo code. The C/E values have been used in an integral data assimilation solving the Bayes equation. The trends on the JEFF3.1.1 235U capture cross section are quite consistent with recent differential measurements. Assimilation results suggest up to a 2.5% decrease for 238U capture from 3 keV to 60 keV, and a 4-5% decrease for 238U inelastic in the plateau region. For this energy range, uncertainties are respectively reduced from 3-4 to 1-2% and from 6-9% to 2-2.5% for 238U capture and 238U inelastic. Results on 239Pu fission cross sections are included in posterior uncertainties. The increase trends on 239Pu capture cross-section is of around 3% in the [2 keV-100 keV] energy range. For 240Pu capture cross section, the increase is of around 4% in the [3 keV-100 keV] energy range and goes in the same direction as a recent evaluation.

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.

scholarly journals Analysis of cross sections on iron and oxygen using Cf-252 neutron source

2020 ◽  
Vol 239 ◽  
pp. 18005
Author(s):  
Bohumil Jansky ◽  
Jiri Rejchrt ◽  
Evzen Novak ◽  
Anatoly Blokhin
Keyword(s):  
Cross Section ◽  
Neutron Source ◽  
Neutron Spectrum ◽  
Heavy Water ◽  
Cross Sections ◽  
Energy Region ◽  
Nuclear Data ◽  
Data Section ◽  
Neutron Spectra ◽  
Artificial Changes

The leakage neutron spectra measurements have been done on benchmark spherical assemblies with Cf-252 source in center of 1) heavy water sphere with diameter of 30 cm (with Cd cover) and of 2) iron spheres with diameter of 100 cm and 50 cm. It has been stated for years that transport calculations by iron overestimate measured spectra in energy region around 300 keV by about 20-40 % (calculation to measurement ratio C/E = 1.2-1.4). The influence of an artificial changes in cross-section XS-Fe-56 (n,elastic)designed by IAEA, Nuclear Data Section, has been studied on the iron spheres. Influence of those XS-corrections to calculated neutron spectrum is presented.

Instrumental Photon Activation and Applications in a Nuclear-Waste Inspection Purpose

2003 ◽  
Author(s):  
F. Jeanneau ◽  
M. Gmar ◽  
N. Huot ◽  
F. Laine´ ◽  
A. Lyoussi ◽  
...  
Keyword(s):  
Waste Management ◽  
Nuclear Waste ◽  
Cross Sections ◽  
Detection Limits ◽  
Current Method ◽  
High Energy ◽  
Monte Carlo Code ◽  
Photon Activation ◽  
Simulation Code ◽  
Passive Noise

The development of non-destructive methods to inspect nuclear-waste containers is important for radioactive-waste management and non-proliferation purposes. This paper will present studies and results carried out by a method based on photon interrogation (photofission) which allows the determination of the actinide quantity contained in the waste. High-energy photons (produced by an electron accelerator associated with a Bremsstrahlung tungsten target) will induce photofission reactions on the actinides. Then the flux of delayed neutrons, which is directly proportional to the amount of actinides, is measured with 3He detectors. Since the beginning of 1990’s, our team in CEA has been working on the development of this method and the improvement of the existing simulation code. The two main tools will be introduced: OPERA (tool for the simulation of photonuclear reactions) which includes photonuclear cross sections in a Monte-Carlo code based on MCNP4C, and SAPHIR (Irradiation and Photon-Activation System), a device allowing experimentations for research and development programs. The applications of these tools will be illustrated mainly with two examples: 1) The feasibility study of an inspection device for old concrete containers will be reported. Two campaigns of measurements have been performed in order to determine the sensitivity and the detection limits in the case of four different types of concrete containers, in terms of nature and geometry. 2) Nuclear-waste producers and managers have been interested by the active photon interrogation possibilities to measure actinide quantity in wastes of high activity, vitrified or compacted, with constraints like a dose rate around 400 Gy/h at 27 cm from the container. The simulation-code improvement has allowed some calculations, based on the SAPHIR facility, which have shown a good linearity between the actinide mass and the number of detected neutrons, in spite of a very high passive noise and the presence of a lead protection. Several R&D programs will be also presented. On one hand, measurements are performed on real wastes, chosen for parameter which could define a limitation of the measurements, in order to improve the method and to evaluate the detection limits. For instance, tomography can be performed with this experimental device: quantity and position of actinides in the waste can be calculated. On the other hand, a new method is studied, using the delayed-gamma flux in order to quantify and to identify the different actinide isotopes contained in the waste. These methods and device offer a large panel of results in terms of measurements and simulations. Our team is now involved in several prospecting and R&D programs in order to improve the current method and to find some new applications for nuclear-waste management.

New empirical formula for calculating (n,p) reaction cross-sections at 14.5MeV neutrons

2020 ◽  
Vol 29 (08) ◽  
pp. 2050052
Author(s):  
Dashty T. Akrawy ◽  
Ali H. Ahmed ◽  
E. Tel ◽  
A. Aydin ◽  
L. Sihver
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Empirical Formula ◽  
Nuclear Data ◽  
Reaction Cross ◽  
Section Data ◽  
Nuclear Data Library ◽  
Number Formula ◽  
New Formula ◽  
Reaction Cross Sections

An empirical formula to calculate the ([Formula: see text], [Formula: see text] reaction cross-sections for 14.5[Formula: see text]MeV neutrons for 183 target nuclei in the range [Formula: see text] is presented. Evaluated cross-section data from TENDL nuclear data library were used to test and benchmark the formula. In this new formula, the nonelastic cross-section term is replaced by the atomic number [Formula: see text], while the asymmetry parameter-dependent exponential term has been retained. The calculated results are presented in comparison with the seven previously published formulae. We show that the new formula is significantly in better agreement with the measured values compared to previously published formulae.

scholarly journals Validating nuclear data uncertainties obtained from a statistical analysis of experimental data with the “Physical Uncertainty Bounds” method

2020 ◽  
Vol 6 ◽  
pp. 19
Author(s):  
Denise Neudecker ◽  
Morgan Curtis White ◽  
Diane Elizabeth Vaughan ◽  
Gowri Srinivasan
Keyword(s):  
Experimental Data ◽  
Statistical Analysis ◽  
Information Content ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Current Version ◽  
Neutron Data ◽  
Data Standards ◽  
Actual Evaluation

Concerns within the nuclear data community led to substantial increases of Neutron Data Standards (NDS) uncertainties from its previous to the current version. For example, those associated with the NDS reference cross section 239Pu(n,f) increased from 0.6–1.6% to 1.3–1.7% from 0.1–20 MeV. These cross sections, among others, were adopted, e.g., by ENDF/B-VII.1 (previous NDS) and ENDF/B-VIII.0 (current NDS). There has been a strong desire to be able to validate these increases based on objective criteria given their impact on our understanding of various application uncertainties. Here, the “Physical Uncertainty Bounds” method (PUBs) by Vaughan et al. is applied to validate evaluated uncertainties obtained by a statistical analysis of experimental data. We investigate with PUBs whether ENDF/B-VII.1 or ENDF/B-VIII.0 239Pu(n,f) cross-section uncertainties are more realistic given the information content used for the actual evaluation. It is shown that the associated conservative (1.5–1.8%) and minimal realistic (1.1–1.3%) uncertainty bounds obtained by PUBs enclose ENDF/B-VIII.0 uncertainties and indicate that ENDF/B-VII.1 uncertainties are underestimated.

scholarly journals Processing and application of nuclear data for low temperature criticality assessment

2020 ◽  
Vol 239 ◽  
pp. 14006
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Tara Hanlon ◽  
Richard Hiles ◽  
Malcolm Lingard ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Low Temperature ◽  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Monte Carlo Code ◽  
Temperature Range ◽  
Nuclear Data Library ◽  
Criticality Assessment ◽  
Good Agreement

Until recently, criticality safety assessment codes had a minimum temperature at which calculations can be performed. Where criticality assessment has been required for lower temperatures, indirect methods, including reasoned argument or extrapolation, have been required to assess reactivity changes associated with these temperatures. The ANSWERS Software Service MONK® version 10B Monte Carlo criticality code, is capable of performing criticality calculations at any temperature, within the temperature limits of the underlying nuclear data in the BINGO continuous energy library. The temperature range of the nuclear data has been extended below the traditional lower limit of 293.6 K to 193 K in a prototype BINGO library, primarily based on JEFF-3.1.2 data. The temperature range of the thermal bound scattering data of the key moderator materials was extended by reprocessing the NJOY LEAPR inputs used to produce bound data for JEFF-3.1.2 and ENDF/B-VIII.0. To give confidence in the low temperature nuclear data, a series of MONK and MCBEND calculations have been performed and results compared against external data sources. MCBEND is a Monte Carlo code for shielding and dosimetry and shares commonalities to its sister code MONK including the BINGO nuclear data library. Good agreement has been achieved between calculated and experimental cross sections for ice, k-effective results for low temperature criticality benchmarks and calculated and experimentally determined eigenvalues for thermal neutron diffusion in ice. To quantify the differences between ice and water bound scattering data a number of MONK criticality calculations were performed for nuclear fuel transport flask configurations. The results obtained demonstrate good agreement with extrapolation methods. There is a discernible difference in the use of ice and water data.

scholarly journals Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

2008 ◽  
Vol 2008 ◽  
pp. 1-5
Author(s):  
I. Kodeli
Keyword(s):  
Uncertainty Analysis ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Maximum Information ◽  
Tritium Production ◽  
Section Data ◽  
Combined Use ◽  
Deterministic Transport ◽  
The Cross

An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL) concept will be performed in 2008 in the Frascati Neutron Generator (FNG) in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n) and (n,3n) reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

scholarly journals Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
A. Rais ◽  
D. Siefman ◽  
G. Girardin ◽  
M. Hursin ◽  
A. Pautz
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Transient Behavior ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Nuclear Data Library ◽  
And Control

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

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