scholarly journals Impact of the cross section library on 93mNb activity in VVER-1000 reactor dosimetry

2020 ◽  
Vol 225 ◽  
pp. 03009
Author(s):  
P. Haroková ◽  
M. Lovecký
Keyword(s):  
Cross Section ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Nuclear Data Library ◽  
The Cross ◽  
Materials Used ◽  
The Impact ◽  
Data Library ◽  
Reactor Dosimetry

One of the objectives of reactor dosimetry is determination of activity of irradiated dosimeters, which are placed on reactor pressure vessel surface, and calculation of neutron flux in their position. The uncertainty of calculation depends mainly on the choice of nuclear data library, especially cross section used for neutron transport and cross section used as the response function for neutron activation. Nowadays, number of libraries already exists and can be still used in some applications. In addition, new nuclear data library was recently released. In this paper, we have investigated the impact of the cross section libraries on activity of niobium, one of the popular materials used as neutron fluence monitor. For this purpose, a MCNP6 model of VVER-1000 was made and we have compared the results between 14 commonly used cross section libraries. A possibility of using IRDFF library in activation calculations was also considered. The results show good agreement between the new libraries, with the exception of the most recent ENDF/B-VIII.0, which should be further validated.

Temperature Dependence of the Westcott g-factor for Neutron Reactions in Activation Analysis

1999 ◽  
Vol 71 (12) ◽  
pp. 2309-2315 ◽  
Author(s):  
N. E. Holden
Keyword(s):  
Cross Section ◽  
Neutron Capture ◽  
Nuclear Reactions ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Nuclear Data Library ◽  
Neutron Reactions ◽  
Using Data ◽  
Data Library ◽  
G Factors

The Westcott g-factors, which allow the user to determine reaction rates for nuclear reactions taking place at various temperatures, have been calculated using data from the Evaluated Neutron Nuclear Data Library, ENDF/B-VI. Nuclides chosen have g-factors which are significantly different from unity and result in different reaction rates compared to nuclides whose neutron capture cross section varies as the reciprocal of the neutron velocity. Values are presented as a function of temperature up to 673.16 K (400 °C).

Calculations of the Dual Functional Lithium Lead-Test Blanket Module for ITER With Three-Dimensional Deterministic Method

2012 ◽  
Author(s):  
GuangChun Zhang ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Youqi Zheng
Keyword(s):  
Ray Tracing ◽  
Cross Sections ◽  
Flux Distribution ◽  
Neutron Transport ◽  
Computing Time ◽  
Three Dimensional ◽  
Nuclear Data ◽  
Dual Functional ◽  
Nuclear Data Library ◽  
Data Library

Neutronics calculations and analysis of ITER test blanket module lay the foundation for the design, construct and experiment of ITER. In this paper, the realistic 3D neutronics calculations of the dual functional lithium lead-test blanket module (DFII-TBM) have been carried out by means of the 3D MOC code and the SPN code, which are both deterministic methods and developed by NECP lab, adopting the multi-group nuclear data library FENDI/MG-2.1. The main features of the TBM nuclear response are assessed, paying a particular attention to the neutron flux and tritium production rate. The 3DMOC code is a coupling a 3D method of characteristics (MOC) to the common geometry module. It could calculate the flux throughout three-dimensional systems by the MOC, which has been proved a very flexible and effective method for the neutron transport calculation in a complex geometry. In this code, a modular ray tracing technique is adopted to reduce the amount of the ray tracing data and the Coarse Mesh Finite Difference (CMFD) acceleration method is employed to save computing time, which could well solve the difficulties when applying MOC in three-dimensional geometries. The SPN code is another three-dimensional Boltzmann transport equation calculation code. The simplified PN method is used to treat the directional variable, and the Nodal method treats the spatial variable. Consequently, this code has an advantage in shorting computing time when applied to big geometry problems. Considering the big geometry of DFII-TBM and the large number of the cross sections of nuclear data library FENDI/MG-2.1, a two-step approach is adopted. Firstly, the DFII-TBM is dissected into some typical independent parts. 3D calculations are performed on these parts respectively with 3D MOC code and FENDL/MG-2.1 library to obtain the detailed heterogeneous flux distribution. Then the homogenization is carried out to calculate the average homogeneous cross sections, followed by the use of homogeneous cross sections to calculate the flux distribution throughout the DFII-TBM with SPN code. The results obtained are herewith presented and critically discussed.

scholarly journals Uncertainty propagation based on correlated sampling technique for nuclear data applications

2020 ◽  
Vol 6 ◽  
pp. 8 ◽  
Author(s):  
Axel Laureau ◽  
Vincent Lamirand ◽  
Dimitri Rochman ◽  
Andreas Pautz
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Neutron Transport ◽  
Uncertainty Propagation ◽  
Computation Time ◽  
Sampling Technique ◽  
Nuclear Data ◽  
Correlated Sampling ◽  
Enriched Uranium ◽  
The Impact

A correlated sampling technique has been implemented to estimate the impact of cross section modifications on the neutron transport and in Monte Carlo simulations in one single calculation. This implementation has been coupled to a Total Monte Carlo approach which consists in propagating nuclear data uncertainties with random cross section files. The TMC-CS (Total Monte Carlo with Correlated Sampling) approach offers an interesting speed-up of the associated computation time. This methodology is detailed in this paper, together with two application cases to validate and illustrate the gain provided by this technique: the highly enriched uranium/iron metal core reflected by a stainless-steel reflector HMI-001 benchmark, and the PETALE experimental programme in the CROCUS zero-power light water reactor.

The Effect of Iron Cross-Section in Thermal Region on Neutron Transport in VVER-1000 Mock-Up in LR-0 Reactor

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Martin Schulc ◽  
Michal Košťál ◽  
Davit Harutyunyan ◽  
Marie Švadlenková ◽  
Vojtěch Rypar ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Power Density ◽  
Cross Section ◽  
Thermal Neutron Flux ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Nuclear Data Library ◽  
Thermal Region ◽  
Density Evaluation

The iron cross-section in thermal regions influences the thermal neutron flux prediction in steel structural components of reactors and also in regions adjoining them. The thermal neutron flux level is proportional to pin power density in fuel. This quantity is an important criterion reflected in limits and conditions of reactor operation. The new power density evaluation shows notable, well distinguishable discrepancy between calculations realized using the CENDL-3.1 nuclear data library and experimentally determined pin power density in boundary rows of pins. All experiments were carried out in a water–water energetic reactor (VVER-1000) transport mock-up placed in the LR-0 reactor.

The Testing and Application of Continuous-Energy Neutron Cross Section Library HENDL-ADS/MC

2013 ◽  
Author(s):  
Chong Chen ◽  
Jun Zou ◽  
Dezheng Xu ◽  
Qin Zeng ◽  
Minghuang Wang
Keyword(s):  
Cross Section ◽  
Neutron Energy ◽  
Fission Products ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Neutron Shielding ◽  
Section Data ◽  
Nuclear Data Library ◽  
Data Library ◽  
High Neutron

A point-wise cross-section data library HENDL-ADS/MC (Hybrid Evaluated Nuclear Data Library) has been produced by FDS team to do the nuclear analysis for the ADS system. The HENDL-ADS/MC library contained 408 nuclide cross-section files including actinides, fission products and structural materials for neutron energy up to 150 MeV. The nuclear library also contained several sub-libraries with different temperatures. A series of neutron integral experiments and critical safety benchmarks have been performed to test the availability and reliability of the HENDL-ADS/MC data library. To validate and qualify the reliability of the high neutron energy cross section for HENDL-ADS/MC library further, a series of high neutron shielding experiments have been performed using MCNP. The testing results indicated the accuracy and reliability of HENDL-ADS/MC library.

The Consistency of WIMS-D Library and ACE Format Library

2010 ◽  
Author(s):  
Zhiyan Liu ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Qingjie Liu
Keyword(s):  
Cross Section ◽  
Scaling Function ◽  
Processing System ◽  
Expansion Method ◽  
Nuclear Data ◽  
Calculation Procedure ◽  
Function Expansion ◽  
Nuclear Data Library ◽  
Data Library

Nuclear data library is the cornerstone in the nuclear reactor’s design and calculation. The WIMS-D multi-group library and ACE format library (mainly used in MCNP) is applied frequently in the nuclear calculation. We have developed a new self-shielding calculation procedure based on Wavelets scaling function expansion method. This procedure needs several parts in both WIMS-D and ACE format library. So the consistency of two libraries becomes a very serious problem. This may bring in large errors. In this paper, NJOY cross section processing system is used to produce new WIMS-D and ACE format library from the same ENDF/B data. We compute some homogenous problems using new and old libraries in WIMS-D and ACE format. The results of the two new libraries and the old libraries are compared respectively. It is found that there are consistency problems between the two libraries. The newly produced libraries are more compatible than the old ones.

scholarly journals Verification of CENDL-3.2 Nuclear Data on VENUS-3 Shielding Benchmark by ARES Transport Code

2021 ◽  
Vol 2021 ◽  
pp. 1-13
Author(s):  
Jiaju Hu ◽  
Bin Zhang ◽  
Zhiwei Zong ◽  
Cong Liu ◽  
Yixue Chen
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Transport Code ◽  
Nuclear Data Library ◽  
Transport Calculation ◽  
Calculation Results ◽  
Almost All ◽  
Data Library

The recently released CENDL-3.2 nuclear data library is deemed as an important achievement in the field of nuclear data research in China. To verify the applicability of the library to the shielding calculation of PWR and analyze the influence of multigroup cross-section parameters on the shielding calculation, ARES-MACXS module is used to process the MATXS format multigroup library based on CENDL-3.2 to generate multigroup working cross sections for PWR shielding calculation. VENUS-3 experimental facility has a clear and complete geometry. It is often used to test the ability of the advanced transport calculation method of calculating RPV fast neutron flux and to evaluate the accuracy of cross-section library. Different cross-section parameters are chosen for ARES to calculate VENUS-3 benchmark, and equivalent neutron flux of 58Ni(n,p)58Co, 115In(n,n′)115mIn and 27Al(n,α)24Na detectors is calculated according to the data provided by the benchmark report. The numerical results demonstrate that almost all the relative deviations between the calculated results and the experimental results are within 20%, which satisfies the requirement of shielding calculation. CENDL-3.2 is suitable for PWR shielding calculation. The comparison of various cross-section parameters results indicates that multigroup cross-section parameters have large effects on the transport calculation results.

scholarly journals Comparison of Deterministic and Stochastic Depletions in Graphite-Filled MOX Fuel Assembly Design

2020 ◽  
Vol 36 (4) ◽  
Author(s):  
Thanh Mai Vu ◽  
Donny Hartanto
Keyword(s):  
Fuel Assembly ◽  
Nuclear Data ◽  
Absolute Difference ◽  
Assembly Design ◽  
Mox Fuel ◽  
Nuclear Data Library ◽  
Calculation Results ◽  
The Impact ◽  
Data Library ◽  
Good Agreement

A comparison between stochastic and deterministic depletion calculations based on a graphite-filled MOX fuel assembly configuration is presented in this paper. The infinite multiplication factors and isotope inventory changes as a function of burnup obtained by Monte Carlo method module SCALE/KENO and deterministic method module SCALE/NEWT are compared with those obtained by deterministic code HELIOS. The impact in calculation results by using different nuclear data library is also investigated. The SCALE/KENO results show a good agreement with SCALE/NEWT results in the eigenvalue as a function of burnup (less than 0.1%). However, the absolute difference in the initial k¥ between SCALE/KENO and NEWT modules and HELIOS results is quite large (around 1.1%) and the isotope inventory changes show quite differently at the end of cycle. The uranium and plutonium depletion rates calculated by SCALE/KENO and SCALE/NEWT have quite good agreement. By using the same data library, the good agreement between stochastic and deterministic code’s results were confirmed.

scholarly journals BEAVRS BENCHMARK EVALUATION WITH CASMO5 AND SIMULATE5

2021 ◽  
Vol 247 ◽  
pp. 06022 ◽  
Author(s):  
Tamer Bahadir
Keyword(s):  
Reaction Rate ◽  
Boron Concentration ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Control Rod ◽  
Significant Difference ◽  
Nuclear Data Library ◽  
Full Power ◽  
The Impact ◽  
Benchmark Evaluation

The MIT BEAVRS benchmark problem, which was primarily setup for the verification and validation of high-fidelity tools that have coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion models, has also found extensive usage in the reactor physics community for validating core analysis tools. The primary purpose of this paper is to provide an accurate, comprehensive evaluation of the BEAVRS benchmark with CASMO5 and SIMULATE5 codes. The CMS5 calculated results for low-power physics tests (hot zero power critical boron, control rod worth and isothermal temperature coefficients) and full power operation (boron let-down and flux map reaction rate distributions) are compared to plant measured data provided in the benchmark specification. The CMS5 model, using ENDF/BVII.1 nuclear data library, predicts HZP critical boron concentration for all-rods-out conditions within 10 ppm for Cycle-1, and 25 ppm in Cycle-2; the control rod worth is predicted with a difference of 0.7% ± 3.8%, where the maximum difference is less than 10%. For the core follow calculations at the hot full power condition, the average difference in predicting the critical boron concentration is less than 20 ppm. In addition, the radial and nodal reaction rate distributions are predicted with a mean difference of about 1.6% and 3.8%, respectively. The CMS5 calculations are repeated using the most recent ENDF/B-VIII.0 library. No significant difference is observed in predicting measured plant parameters with different nuclear data libraries. Additionally, the impact of various modeling options, which are typically employed with nodal diffusion codes, on the predictions of important core parameters are presented as part of the benchmark evaluation.

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