scholarly journals Neutronics of Lead and Bismuth

2021 ◽  
pp. 177-213
Author(s):  
Cheol Ho Pyeon
Keyword(s):  
Numerical Simulations ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Design Parameters ◽  
Integral Parameter ◽  
Considerable Impact ◽  
Test Materials ◽  
Design Calculations ◽  
Data Libraries

AbstractCross-section uncertainties of Pb and Bi isotopes could consequently affect the precision of nuclear design calculations of preliminary analyses, before the actual operation of upcoming ADS, since Pb and Bi are composed partly of coolant material (lead-bismuth eutectic: LBE) in ADS facilities. The main characteristics of LBE in ADS are recognized as follows: chemically inactive; high boiling point mechanically; excellent neutron economy caused by large scattering cross sections. From the viewpoint of neutronics, LBE exerts considerable impact on nuclear design parameters for numerical simulations of neutron interactions of Pb and Bi isotopes. As a suitable way of investigating cross-section uncertainties, sample reactivity worth measurements in critical states are considered effective with the use of reference and test materials in a zero-power state, such as a critical assembly, because integral parameter information on cross sections of test materials can be acquired experimentally. For the required experimental study on Pb and Bi nuclear data uncertainties, the sample reactivity worth experiments are carried out at the KUCA core by the substitution of reference (aluminum) for test (Pb or Bi) materials, and numerical simulations are performed with stochastic and deterministic calculation codes together with major nuclear data libraries.

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 Libraries for the Nuclear Design Calculations of the NPP Krško with the CORD-2 System

2021 ◽  
Author(s):  
Marjan Kromar ◽  
Bojan Kurincic
Keyword(s):  
Cross Sections ◽  
Fission Products ◽  
Nuclear Data ◽  
Isothermal Temperature ◽  
Nuclear Systems ◽  
Full Power ◽  
Design Calculations ◽  
Data Libraries ◽  
Control Rods ◽  
Operating Cycles

Abstract Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the NPP Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 libraries are considered. In the first part of the paper the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few hundred pcm in multiplication factor for a fresh fuel due to differences in 235U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1 %. Some differences in isothermal temperature coefficient were observed, however they only marginally affect core power defect going from zero to full power.

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 Investigating the Deformation and Failure Mechanism of a Submarine Tunnel with Flexible Joints Subjected to Strike-Slip Faults

2021 ◽  
Vol 9 (12) ◽  
pp. 1412
Author(s):  
Guangxin Zhou ◽  
Qian Sheng ◽  
Zhen Cui ◽  
Tianqiang Wang ◽  
Yalina Ma
Keyword(s):  
Numerical Simulations ◽  
Fault Zone ◽  
Cross Section ◽  
Strike Slip ◽  
Strike Slip Fault ◽  
Design Parameters ◽  
Flexible Joints ◽  
Flexible Joint ◽  
Deformation And Failure ◽  
Tunnel Cross Section

Knowledge from historical earthquake events indicates that a submarine tunnel crossing active strike-slip faults is prone to be damaged in an earthquake. Previous studies have demonstrated that the flexible joints are an effective measure for a submarine tunnel crossing a strike-slip fault. The background project of this paper is the second submarine tunnel of Jiaozhou bay. In this work, model tests and numerical simulations are conducted to investigate the deformation and failure mechanism of a submarine tunnel with flexible joints under a strike-slip fault dislocation. The influence of strike-slip faults on a tunnel with flexible joints has been investigated by examining the deformation of rock mass surface, analyzing lining stains, and crack propagation from model tests. Numerical simulations are conducted to study the effects of the design parameters of a tunnel with flexible joints on the mechanical response of the lining. The results showed that the ‘articulated design’ measure can improve the ability of the tunnel to resist the strike-slip faults. In terms of the mechanism of design parameters of a tunnel with flexible joints, this paper finds that increasing the lining thickness, decreasing the lining segment length, and decreasing the tunnel diameter to a reasonable extent could effectively improve the performance of this faulting resistance measure for a tunnel under the strike-slip fault zone dislocation. Compared with the horseshoe tunnel cross-section, the circular tunnel cross-section can improve the ability of the faulting resistance of a tunnel with flexible joints, while the optimal angle of the tunnel crossing the fault zone is 90º. It is concluded that the wider fault zone, smaller flexible joint width, and less stiffness of the flexible joint could make lining safer under a strike-slip fault dislocation. The above research results can serve as a necessary theoretical reference and technical support for the design of reinforcement measures for a submarine tunnel with flexible joints under strike-slip fault dislocation.

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.

scholarly journals IMPACT OF NUCLEAR DATA LIBRARIES ON PREDICTED FAST REACTOR PERFORMANCE

2021 ◽  
Vol 247 ◽  
pp. 09026
Author(s):  
A.G. Nelson ◽  
K.M. Ramey ◽  
F. Heidet
Keyword(s):  
Cross Sections ◽  
Fast Reactor ◽  
Power Distribution ◽  
Evaluation Process ◽  
Nuclear Data ◽  
Reactor Performance ◽  
Data Set ◽  
Experimental Database ◽  
Data Libraries ◽  
Nuclear Data Libraries

The nuclear data evaluation process inherently yields a nuclear data set designed to produce accurate results for the neutron energy spectra corresponding to a specific benchmark suite of experiments. When studying reactors with spectral conditions outside of, or not well represented by, the experimental database used to evaluate the nuclear data, care should be given to the relevance of the nuclear data used. In such cases, larger biases or uncertainties may be present than in a reactor with well-represented spectra. The motivation of this work is to understand the magnitude of differences between recent nuclear data libraries to provide estimates for expected variability in criticality and power distribution results for sodiumcooled, steel-reflected, metal-fueled fast reactor designs. This work was specifically performed by creating a 3D OpenMC model of a sodium-cooled, steel-reflected, metal-fueled fast reactor similar to the FASTER design but without a thermal test region. This OpenMC model was used to compare the differences in eigenvalues, reactivity coefficients, and the spatial and energetic effects on flux and power distributions between the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 nuclear data libraries. These investigations have revealed that reactivity differences between the above libraries can vary by nearly 900 pcm and the fine-group fluxes can vary by up to 18% in individual groups. Results also show a strong variation in the flux and power distributions near the fuel/reflector interface due to the high variability in the 56Fe cross sections in the libraries examined. This indicates that core design efforts of a sodium-cooled, steel-reflected, metalfueled reactor will require the application of relatively large nuclear data uncertainties and/or the development of a representative benchmark-quality experiment.

scholarly journals MEASUREMENT, EVALUATION AND VALIDATION OF MOLYBDENUM CROSS SECTIONS

2021 ◽  
Vol 247 ◽  
pp. 09007
Author(s):  
Isabelle Duhamel ◽  
Nicolas Leclaire ◽  
Luiz Leal ◽  
Atsushi Kimura ◽  
Shoji Nakamura
Keyword(s):  
Cross Sections ◽  
Path Length ◽  
Measurement Instrument ◽  
Reduction Process ◽  
Nuclear Data ◽  
Flight Path ◽  
Proton Accelerator ◽  
Reactor Physics ◽  
Safety Issues ◽  
Data Libraries

Available nuclear data for molybdenum included in the nuclear data libraries are not of sufficient quality for reactor physics or criticality safety issues and indeed information about uncertainties and covariance is either missing or leaves much to be desired. Therefore, IRSN and JAEA performed experimental measurements on molybdenum at the J-PARC (Japan Proton Accelerator Research Complex) facility in Japan. The aim was to measure capture cross section and transmission of natural molybdenum at the ANNRI (Accurate Neutron-Nucleus Reaction measurement Instrument) in the MLF (Material Life and science Facility) of J-PARC. The measurements were performed on metallic natural molybdenum samples with various thicknesses. A NaI detector, placed at a flight-path length of about 28 m, was used for capture measurements and a Li-glass detector (flight-path length of about 28.7 m) for transmission measurements. Following the data reduction process, the measured data are being analyzed and evaluated to produce more accurate cross sections and associated uncertainties.

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.

Cross-section measurements of the (n,2n) and (n,p) reactions on 124,126,128,130,131,132Xe in the 14 MeV region and theoretical calculations of their excitation functions

2021 ◽  
Author(s):  
Junhua Luo ◽  
Li Jiang ◽  
junchen liang ◽  
Fei Tuo ◽  
Long He ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Theoretical Calculations ◽  
Germanium Detector ◽  
Reaction Model ◽  
Reaction Cross ◽  
Model Parameters ◽  
China Academy ◽  
The Cross ◽  
Data Libraries

Abstract The reaction cross-sections of 124Xe(n, 2n)123Xe, 126Xe(n, 2n)125Xe, 128Xe(n, 2n)127Xe, 130Xe(n, 2n)129mXe, 132Xe(n, 2n)131mXe, 130Xe(n, p)130I, 131Xe(n, p)131I, and 132Xe(n, p)132I were measured at the 13.5, 13.8, 14.1, 14.4, and 14.8 MeV neutron energies. The monoenergetic neutrons were generated through the 3H(d,n)4He reaction at the China Academy of Engineering Physics using the K-400 Neutron Generator with a solid 3H-Ti target. A high-purity germanium detector was used to measure the activities of the product. The reactions 93Nb(n, 2n)92mNb and 27Al(n, α)24Na served for neutron flux calibration. The cross sections of the (n,2n) and (n,p) reactions of the xenon isotopes were acquired within the 13–15 MeV neutron energy range. These cross-sections were then compared with the IAEA-exchange format (EXFOR) database-derived experimental data together with the evaluation results of the CENDL-3, ENDF/B-VIII.0, JENDL-4.0, RUSFOND, and JEFF-3.3 data libraries as well as the theoretical excitation function obtained using the TALYS-1.95 code. The cross-sections of the reactions (except for the 124Xe(n, 2n)123Xe and 132Xe(n, p)132I) at 13.5, 13.8, and 14.1 MeV are reported for the first time in this work. The present results are helpful to provide better cross-section constraints for these reactions in the 13–15 MeV region, thus improving the quality of the corresponding database. Meanwhile, these data can also be used for the verification of relevant nuclear reaction model parameters.

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