Study of (n,2n) reaction cross section of fission product based on neural network and decision tree models
Xiaodong SunZihao WeiDuan WangRuirui XuYuan TianXi TaoYingxun ZhangYue ZhangZhi ZhangZhigang GeJimin WangHouqiong XiaNengchuan Shu
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Abstract:
The neutron induced nuclear reaction cross sections of fission products are related with the neutron fiux and the reactor burnup, which are important for the accurate of nuclear engineering design. To predict the (n,2n) reaction cross section, especially those lack of experimental measurements, we analyzed the relevant features and establish the experimental data set on the basis of sorting out the experimental data recorded in EXFOR library. The back propagation artificial neural network (ANN) and decision tree (DT) models are built to learn the experimental data set, respectively, adopting PyTorch and XGBOOST toolboxes. we report that machine learning models are applied to analysis and predicate (n,2n) reaction cross section.Keywords:
Nuclear data
Data set
Burnup
Experimental data
Tree (set theory)
Nuclear fission product
Section (typography)
Burnup
Nuclear fission product
Plutonium-240
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A spectrometric system was developed for spent fuel burnup evaluations at the LVR-15 research reactor, which employed highly enriched (36%) IRT-2M-type fuel. Such a system allows the measurement of fission product axial distribution by measuring certain nuclides, such as Cs137, Cs134, and their ratios, respectively. Within the paper, a comparison between experimental data provided by the spectrometric system and calculations in operational code called NODER is provided.
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Nuclear fission product
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A method is developed to estimate spent-fuel burnup using gamma-ray spectrometry of the short-lived fission product 140La. The 140La activity was established by reirradiating the spent fuel in a reactor core. Based on the measured 140La activity, burnup values can be deduced by iterative calculations. In this method, the fuel irradiation history is not needed. To verify its validity, burnup values deduced from 140La activities were compared with those deduced from the conventional long-lived I37Cs activities and 134Cs/137Cs activity ratios; good agreement was obtained. This method is applicable to reactors loaded with highly enriched, thin plate-type fuels.
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The objective of this document is to describe a method for non-destructive burnup evaluation of the AGR-1 TRISO fuel compacts using gamma spectrometry. Initially, gamma-ray spectra taken from the AGR-1 experiment TRISO fuel compacts using the Precision Gamma Scanner (PGS) system were analyzed. The spectrometry analysis included a characterization of the PGS system and the measured fission product activities for the compacts. The measured activities were compared to detailed production/depletion fuel inventory simulations of the AGR-1 experiment. Two methods for relating burnup to measured fission product activity were developed from the results of the fuel inventory simulations and applied to the measured fission product activities. Burnup derived from the PGS measurements was compared to the predicted burnup from simulation.
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Point isotopic depletion methods are used to develop spatially dependent fission product and heavy metal inventories for the TMI-2 core. Burnup data from 1239 fuel nodes (177 elements, 7 axial nodes per element) are utilized to preserve the core axial and radial power distributions. A full-core inventory is calculated utilizing 12 fuel groups (four burnup ranges for each of three initial enrichments). Calculated isotopic ratios are also presented as a function of burnup for selected nuclides. Specific applications of the isotopic ratio data include correlation of fuel debris samples with core location and estimates of fission product release fractions. 24 figs., 25 tabs.
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Pressurized water reactor
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Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.
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The accuracy of power distribution solution and burnup algorithms in standalone fuel behaviour codes is limited. For improved accuracy, fuel behaviour codes can be coupled with external reactor physics solvers. A coupled calculation system between the Monte Carlo neutronics and burnup calculation solver Serpent and the fuel behaviour solver TRANSURANUS was developed in an earlier work. The data exchanged between the solvers included fuel pin axial and radial power and temperature distributions, the axial fast flux distribution in the cladding and the changes in the pin radii. In this work the coupling was further improved by developing the capability to transfer and utilize Serpent calculated nuclide compositions in TRANSURANUS within the coupled calculations. Additionally, support for corrector type burnup algorithms in the coupled solution was now implemented for increased accuracy of the nuclide solution. The new capabilities were demonstrated with a coupled single rod burnup calculation utilizing power and coolant history data from the Loviisa nuclear power plant. The focus at this stage was on fission product nuclides as their accumulation adversely affects the fuel performance, making their accurate solution important to ensure safe yet economical use of fuel. The demonstration showed small but visible differences when compared to simulations that were run without using fission products calculated by Serpent. The developed capabilities will now facilitate import of further fission product nuclides to be taken into account in TRANSURANUS.
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