scholarly journals NEUTRONIC MODELING OF A FAST CORE WITH MODERATING MATERIALS USING APOLLO3R CODE

2021 ◽  
Vol 247 ◽  
pp. 06009
Author(s):  
M. Zajaczkowski ◽  
J.-M. Palau ◽  
V. Pascal ◽  
C.de Jean Saint
Keyword(s):  
Atomic Energy Commission ◽  
Calculation Scheme ◽  
Zirconium Hydride ◽  
Control Rod ◽  
Fuel Assemblies ◽  
Development And Validation ◽  
And Control ◽  
Important Design ◽  
Alternative Energies ◽  
Average Neutron Energy

In order to improve passive safety of Sodium-cooled Fast Reactors The French Alternative Energies and Atomic Energy Commission (CEA) has proposed a new core design called CADOR - an SFR core with enhanced Doppler reactivity feedback. One of its most important design features is the introduction of solid moderating materials inside each fuel assemblies to slightly decrease the average neutron energy. The article focuses on development and validation of a neutronics calculation scheme able to produce accurate results in case of CADOR and other fast cores with moderating materials. The study uses two different fuel assembly models moderated by metallic beryllium and zirconium hydride (ZrH2) respectively The study includes discussion of neutron scattering treatment and different ways of spatial homogenization and energy condensations. The results indicate that the accurate scattering treatment leads to much better estimation of Doppler constant, especially in case of ZrH2 moderated core. By using combined deterministic-Monte Carlo calculation scheme we are able to quantify the biases on global reactivity, reactivity feedbacks and control rod worth. We demonstrate that spatial homogenization plays a more important role in case of moderated CADOR assemblies and thus preserving certain level of heterogeneity within fuel assemblies can lower the calculation bias significantly.

Analysis of a Loss of Feedwater Case at the Ringhals-3 NPP Using RELAP5/PARCS Coupled Codes

2008 ◽  
Author(s):  
Jo´zsef Ba´na´ti ◽  
Mathias Sta˚lek ◽  
Christophe Demazie`re ◽  
Magnus Holmgren
Keyword(s):  
Temperature Fields ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
History Effects ◽  
Coupled Codes ◽  
Fuel Assemblies ◽  
Development And Validation ◽  
Secondary Side ◽  
Relap5 Code ◽  
Good Agreement

This paper deals with the development and validation of a coupled RELAP5/PARCS model of the Swedish Ringhals-3 pressurized water reactor against a Loss of Feedwater transient, which occurred on August 16, 2005. At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modeled in individually in both codes. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, because of the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were more challenged in this transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analysis. The calculated values of the parameters show good agreement with the measured data.

scholarly journals Study of Hydrogen Generation of Aluminum-Containing Compositions with Boric Acid

2018 ◽  
Vol 3 (3) ◽  
pp. 230
Author(s):  
Salakhova A.A. ◽  
Suvorov V.A. ◽  
Firsova A. I. ◽  
Belozerov V.I. ◽  
Milinchuk V.K.
Keyword(s):  
Boric Acid ◽  
Hydrogen Generation ◽  
Sodium Metasilicate ◽  
Chemical Processes ◽  
System Management ◽  
Control Rod ◽  
Radiation Fields ◽  
Fuel Elements ◽  
And Control ◽  
Kinetics Of

The results of investigations of the kinetics of hydrogen generation compositions with aluminum, chemical activators (hydrated sodium metasilicate, oxide and calcium hydroxide) boric acid. Aluminium and its alloys used for the manufacture of protective sheaths of fuel elements and control rod protection system management, pipelines, tanks, and various support structures in the active zone of atomic reactors RBMK, research water-cooled reactors. The aluminum is protected from direct contact with water and steam surface layer of metal oxide having a high corrosion resistance at high temperatures in powerful radiation fields. However, after removal or when the discontinuity of the oxide layer of activated metal efficiently decompose water to hydrogen. It is established that the hydrogen aluminum-containing compositions is dependent on the concentration of boric acid. The discovery of the involvement of boric acid in these reactions expands the ideas about regularities of chemical processes of formation of hydrogen flowing in the water coolant of VVER reactors with the participation of the corrective additives and impurities.

scholarly journals Development and Validation of a MicroRNA Panel to Differentiate Between Patients with Rheumatoid Arthritis or Systemic Lupus Erythematosus and Controls

2019 ◽  
Vol 47 (2) ◽  
pp. 188-196 ◽  
Author(s):  
Michelle J. Ormseth ◽  
Joseph F. Solus ◽  
Quanhu Sheng ◽  
Fei Ye ◽  
Qiong Wu ◽  
...  
Keyword(s):  
Rheumatoid Arthritis ◽  
Systemic Lupus Erythematosus ◽  
Lupus Erythematosus ◽  
Noncoding Rna ◽  
Characteristic Curve ◽  
Systemic Lupus ◽  
Control Subjects ◽  
Development And Validation ◽  
And Control ◽  
Bioinformatic Approach

Objective.MicroRNA (miRNA) are short noncoding RNA that regulate genes and are both biomarkers and mediators of disease. We used small RNA (sRNA) sequencing and machine learning methodology to develop an miRNA panel to reliably differentiate between rheumatoid arthritis (RA) or systemic lupus erythematosus (SLE) and control subjects.Methods.Plasma samples from 167 RA and 91 control subjects who frequency-matched for age, race, and sex were used for sRNA sequencing. TIGER was used to analyze miRNA. DESeq2 and random forest analyses were used to identify a prioritized list of miRNA differentially expressed in patients with RA. Prioritized miRNA were validated by quantitative PCR, and lasso and logistic regression were used to select the final panel of 6 miRNA that best differentiated RA from controls. The panel was validated in a separate cohort of 12 SLE, 32 RA, and 32 control subjects. Panel efficacy was assessed by area under the receiver operative characteristic curve (AUC) analyses.Results.The final panel included miR-22-3p, miR-24-3p, miR-96-5p, miR-134-5p, miR-140-3p, and miR-627-5p. The panel differentiated RA from control subjects in discovery (AUC = 0.81) and validation cohorts (AUC = 0.71), seronegative RA (AUC = 0.84), RA remission (AUC = 0.85), and patients with SLE (AUC = 0.80) versus controls. Pathway analysis showed upstream regulators and targets of panel miRNA are associated with pathways implicated in RA pathogenesis.Conclusion.An miRNA panel identified by a bioinformatic approach differentiated between RA or SLE patients and control subjects. The panel may represent an autoimmunity signature, perhaps related to inflammatory arthritis, which is not dependent on active disease or seropositivity.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Control Strategy of Soluble Boron-Free Operation for a Small PWR

2017 ◽  
Author(s):  
Guangwen Bi ◽  
Chuntao Tang ◽  
Bo Yang
Keyword(s):  
Control Strategy ◽  
Power Distribution ◽  
Short Term ◽  
Control Rod ◽  
Burnable Poison ◽  
Power Operation ◽  
And Control ◽  
Power Distribution Control ◽  
Selection Of

Elimination of soluble boron will be a challenge to reactor operation for PWR. This paper is to promote a control strategy of soluble boron-free operation for a small PWR, through selection of burnable poison (BP), BP loading and control rod loading, based on the reactivity balance and manage requirement. The analysis for on-power operation and shutdown condition indicated that this strategy could be suitable for long-term and short-term reactivity and power distribution control for soluble boron-free operation.

Coupling Dependence of Multiple Operating Parameters on Burnup Credit Calculations for BWR Spent Fuel Assemblies

2018 ◽  
Author(s):  
Shang-Chien Wu ◽  
Der-Sheng Chao ◽  
Jenq-Horng Liang
Keyword(s):  
Computer Code ◽  
Operating Parameters ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Fuel Assemblies ◽  
Moderator Density ◽  
And Control ◽  
Data Library ◽  
Precise Assessment ◽  
Effective Multiplication

This study aims to investigate the coupling dependence resulting from three and four operating parameters for burnup credit calculations in boiling water reactor (BWR) spent fuel assemblies. Four operating parameters are under investigation, including fuel temperature, axial burnup profile, axial moderator density profile and control blade usage. In this study, the effects of variation on the curve of effective multiplication factor (keff) versus burnup (B) resulting from one and multiple operating parameters were defined as the single and compound effects, respectively. Particularly, the compound effects adopt more practical operating parameters than single effects does and thus affect the precise assessment to some extent. In our previous study, the compound effects resulting from two operating parameters were investigated in depth. However, the influence of compound effects resulting from three and four operating parameters on burnup credit calculation is still unknown. Therefore, this constitutes the purpose of this study. All the calculations were performed using SCALE 6.1 computer code together with the ENDF/B-VII 238 energy group neutron data library. Two geometrical models were established to represent the typical GE14 10 × 10 BWR fuel assembly and the GBC-68 storage cask. The results revealed that the reactivity deviation (or changes of keff, Δk) resulting from the compound effects was not a summation of the Δk’s resulting from the associated single effects. Moreover, such Δk discrepancies increase as B increases. In this study, the curves of keff versus B due to single and compound effects were approximated by a second degree polynomial of B. A general formula was thus proposed to express these curves.

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