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

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

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

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 Resonance Region Nuclear Data Analysis to Support Advanced Fuel Cycle Development

2011 ◽  
Vol 59 (2(3)) ◽  
pp. 1207-1212
Author(s):  
M. E. Dunn ◽  
H. Derrien ◽  
L. C. Leal ◽  
C. Gil ◽  
D. Kim
Keyword(s):  
Data Analysis ◽  
Fuel Cycle ◽  
Resonance Region ◽  
Nuclear Data ◽  
Advanced Fuel

scholarly journals Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

Nukleonika ◽  
2019 ◽  
Vol 64 (3) ◽  
pp. 87-96 ◽  
Author(s):  
Piotr Darnowski ◽  
Michał Pawluczyk
Keyword(s):  
Fuel Cycle ◽  
Neutron Transport ◽  
Measurement Data ◽  
Nuclear Data ◽  
Data File ◽  
Power Measurement ◽  
Future Research ◽  
Pressurized Water ◽  
Control Rod ◽  
Nodal Diffusion

Abstract This paper presents an analysis of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) performed using SCALE 6.1.2 and PARCS 3.2 computer codes. The benchmark specification contains a detailed design, operational data and measurements for a real 4-loop Westinghouse pressurized water reactor (PWR). The lattice physics simulations were prepared using TRITON depletion sequence and NEWT neutron transport solver (SCALE package). The 238-neutron group library based on evaluated nuclear data file – ENDF/B-VII nuclear data libraries was applied. A set of branch and burnup calculations was prepared, and group constants in the form of PMAXS files were generated with GenPMAXS. The full-core models were prepared using the PARCS nodal-diffusion core simulator. The PMAXS libraries were used with PARCS to investigate the core operation. The hot zero power measurement data, including control rod worths and critical boron concentrations, were compared using simulations, and satisfactory results were achieved. The first fuel cycle was simulated, and acceptable agreement with boron letdown curve and measurements were obtained. Finally, conclusions and recommendations for future research were presented.

scholarly journals A NOVEL COMPUTATIONAL PLATFORM FOR THE PROPAGATION OF NUCLEAR DATA UNCERTAINTIES THROUGH THE FUEL CYCLE CODE ANICCA

2021 ◽  
Vol 247 ◽  
pp. 13007
Author(s):  
Augusto Hernandez Solis ◽  
Ivan Merino Rodriguez ◽  
Luca Fiorito ◽  
Gert Van den Eynde
Keyword(s):  
Fuel Cycle ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Uncertainty Assessment ◽  
Time Dependent ◽  
Relative Standard ◽  
Neutron Reaction ◽  
The Creation ◽  
Data Libraries ◽  
Computational Platform

This paper presents the first results of a computational platform dedicated to the propagation of nuclear data covariances, all the way to fuel cycle scenario observables. Such platform, based on in-house codes developed at SCK•CEN in Belgium, both for the creation of the many-randomized nuclear data libraries based on ENDF format and for fuel cycle scenario-studies (known as SANDY and ANICCA, respectively), was employed for the uncertainty assessment of the time-dependent inventory computed from a mono-recycling of Plutonium scenario based on a PWR fleet. An essential part of the procedure that deals with the creation of input data libraries to ANICCA, has been carried out this time by the SERPENT2 code. Due to the fact that its neutron transport and depletion parallelized calculation in 72 cores for up to 1640 days and 60 MWd/kg-HM takes almost one hour, it is feasible to finish a total of 100 ANICCA runs based on randomized input libraries created from ENDF/B-VII.1 neutron-reaction covariances in about one week. Therefore, it is consider that the computation of the output population statistics can be inferred from 100 observables representing time-dependent mass inventories. To mention a few results from the aforementioned NEA/OECD benchmark scenario, it was found out that the relative standard deviation of the accumulated plutonium in the final disposal after 120 years was of 7%, while for curium it corresponded to 8%. Thus, sources of uncertainty arising from neutron-reaction covariances do have an impact in the final quantitative analysis of the fuel cycle output uncertainties.

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