scholarly journals generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data

2017 ◽  
Vol 57 (1) ◽  
pp. 42-53 ◽  
Author(s):  
Vytenis Barkauskas ◽  
Rita Plukienė ◽  
Artūras Plukis ◽  
Vidmantas Remeikis
Keyword(s):  
Experimental Data ◽  
Nuclear Fuel ◽  
Cross Section ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Transport Code ◽  
Burnable Absorber ◽  
Acceptable Agreement ◽  
Calculation Code ◽  
Nuclide Composition

Depletion of RBMK-1500 spent nuclear fuel (SNF) with and without an erbium burnable absorber was modelled, and one-group burn-up dependent cross-section libraries for Origen-ARP were created. Depletion calculations for the generation of cross-section libraries were performed using the SCALE 6.1 code package with the TRITON control module, which employs the NEWT deterministic 2D transport code with the 238-group energy library based on the ENDF-B VII library and the ORIGEN-S nuclide composition calculation code. Concentrations of the most important actinides for criticality safety were calculated using the created libraries and were compared with the available experimental data and the newest modelling results. Available experimental data of fission products (Nd and Cs isotopes) were also compared to the modelling results. Composition differences were evaluated for several fuel enrichments and water densities. The comparison shows an acceptable agreement between the values obtained using new one-group cross-section libraries and experimental data except for238Pu and241Am, as well as the causes of discrepancy are discussed. It has been found that the enrichment and presence of the burnable absorber play an important role in the SNF composition. At the highest evaluated burn-up (29 GWd/tU) isotopic composition differences between 2% enrichment fuel and 2.8% burn-up for actinides important to burn-up credit (BUC) applications varied from 11 to 52%.

Conceptual Study of Neutron Physics of Nuclear Fuel Cycle for Ceramic Fast Reactor

2021 ◽  
Author(s):  
Xuesong Yan ◽  
Yaling Zhang ◽  
Yucui Gao ◽  
Lei Yang
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Fast Reactor ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
Nuclear Fuel Cycle ◽  
Purification Method ◽  
Neutron Physics ◽  
Conceptual Study

Abstract To make the nuclear fuel cycle more economical and convenient, as well as prevent nuclear proliferation, the conceptual study of a simple high-temperature dry reprocessing of spent nuclear fuel (SNF) for a ceramic fast reactor is proposed in this paper. This simple high-temperature dry (HT-dry) reprocessing includes the Atomics International Reduction Oxidation (AIROX) process and purification method for rare-earth elements. After removing the part of fission products from SNF by a HT-dry reprocessing without fine separation, the remaining nuclides and some uranium are fabricated into fresh fuel which can be used back to the ceramic fast reactor. Based on the ceramic coolant fast reactor, we studied neutron physics of nuclear fuel cycle which consists operation of ceramic reactor, removing part of fission products from SNF and preparation of fresh fuels for many time. The parameters of the study include effective multiplication factor (Keff), beam density, and nuclide mass for different ways to remove the fission products from SNF. With the increase in burnup time, the trend of increasing 239Pu gradually slows down, and the trend of 235U gradually decreases and become balanced. For multiple removal of part of fission products in the nuclear fuel cycle, the higher the removal, the larger the initial Keff.

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 Depleted Uranium Oxides and Silicates as Spent Nuclear Fuel Waste Package Fill Materials

1996 ◽  
Vol 465 ◽  
Author(s):  
C. W. Forsberg
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Chemical Species ◽  
Fission Products ◽  
Depleted Uranium ◽  
Improve Performance ◽  
Waste Package ◽  
Reducing Conditions ◽  
Nuclear Fuel Waste

ABSTRACTA new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated. The WP uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be loaded with SNF. Void spaces would then be filled with DU (∼0.2 wt % 235U) dioxide (UO2) or DU silicate-glass beads.Fission products and actinides can not escape the SNF UO2 crystals until the UO2 dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of WP groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion minimizes water flow in the degraded WP. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

A Strategic Recovery of Rare-Metal Fission Products in Spent Nuclear Fuel

2004 ◽  
Vol 148 (3) ◽  
pp. 348-357 ◽  
Author(s):  
Yuichi Sano ◽  
Yoshihiko Shinoda ◽  
Masaki Ozawa
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Rare Metal

scholarly journals New trends in the reprocessing of spent nuclear fuel. Separation of minor actinides by solvent extraction

2016 ◽  
Vol 71 (1) ◽  
pp. 123
Author(s):  
Jerzy Narbutt
Keyword(s):  
Solvent Extraction ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Natural Uranium ◽  
Heterocyclic Ligands ◽  
High Level Nuclear Waste ◽  
Extraction Processes ◽  
High Level

<p>Recycling of actinides from spent nuclear fuel by their selective separation followed by transmutation in fast reactors will optimize the use of natural uranium resources and minimize the long-term hazard from high-level nuclear waste. This paper describes solvent extraction processes recently developed, aimed at the separation of americium from lanthanide fission products as well as from curium present in the waste. Depicted are novel poly-N-heterocyclic ligands used as selective extractants of actinide ions from nitric acid solutions or as actinide-selective hydrophilic stripping agents.</p>

scholarly journals The new results of experimental studies of fission products release from spent nuclear fuel of the icebreaker “Lenin” reactor dumped in the Kara Sea

2011 ◽  
Vol 46 (6) ◽  
pp. S363-S369 ◽  
Author(s):  
V. Soyfer ◽  
V. Goryachev ◽  
A. Salyuk ◽  
O. Dudarev ◽  
D. Andreev ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Experimental Studies ◽  
Fission Products ◽  
Kara Sea
Sign in / Sign up

Export Citation Format

Share Document