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.

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.

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 Practical, visible wavelength nuclear-pumped laser

1993 ◽  
Vol 11 (3) ◽  
pp. 559-566 ◽  
Author(s):  
Y.R. Shaban ◽  
G.H. Miley
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Visible Region ◽  
Laser Cavity ◽  
Pumping Power ◽  
University Of Illinois ◽  
Triga Reactor ◽  
Visible Wavelength ◽  
The University

A practical, visible nuclear-pumped laser (NPL) has been sought at the University of Illinois and other laboratories for a number of years. Yet, the results from successful visible NPLs to date have not been fully satisfactory, e.g., the threshold pumping power is too high for conventional applications. Progress in recent studies of 3He-Ne-H2 as a candidate NPL operating in the visible region at 585.3 nm on the 2P1,-1S2 Ne transition is described. We obtained lasing on the above transition for 3He-Ne-H2 concentrations of 1,140, 588, and 412 torr, respectively, with the laser cavity placed in the beamport of the University of Illinois TRIGA reactor. The threshold thermal neutron flux is 1014 n/cm2-s, corresponding to a threshold pumping power of 5 W/cm3.

Thermal neutron flux characterization at aircraft altitudes with the TinMan detector

2017 ◽  
Author(s):  
Suzanne Nowicki ◽  
Stephen Wender ◽  
Aaron Couture ◽  
Laura Dominik ◽  
Adam Warniment ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Neutron Flux Characterization

Determination of thermal neutron flux distribution at the rotary rack served for elemental concentration analysis using the k0-INAA method

2021 ◽  
Vol 10 (1) ◽  
pp. 11-20
Author(s):  
Tho Nguyen Thi ◽  
Anh Tran Tuan ◽  
Cuong Trinh Van ◽  
Doanh Ho Van ◽  
Duong Tran Quoc ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Flux Distribution ◽  
Thermal Neutron Flux ◽  
Epithermal Neutron ◽  
Epithermal Neutron Flux ◽  
Irradiation Position ◽  
Neutron Flux Distribution ◽  
Sample Irradiation

The accuracy of elements concentration determination using the k0-standardization method directly depends on irradiation and measurement parameters including Non-1/E epithermal neutron flux distribution shape α (ϕ epi ≈1/E1+α ) , thermal-to-epithermal neutron flux ratio f, efficiency ε, peak area… In the case of the irradiation position at the rotary rack of the Dalat Nuclear Research Reactor (DNRR), the difference of thermal neutron flux between the bottom (3.54x1012 n.cm-2.s-1) and the top (1.93x1012 n.cm-2.s-1) of the 15 cm aluminum container is up to 45%. Therefore, it is necessary to accurately determine above-mentioned parameters in the sample irradiation position. The present paper deals with the determination of the distribution of thermal neutron flux along the sample irradiation container by using 0.1% Au–Al wire activation technique. The thermal neutron flux was then used to calculate the concentration of elements in the Standard Reference Material 2711a and SMELS type III using k0-INAA method at different positions in the container. The obtained results with the neutron flux correction were found to be in good agreement with the certified values. In conclusion, the proposed technique can be applied for activation analyses without sandwiching flux monitors between samples during irradiations.

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