The Deep-Coupling and Preprocessed Photon Transport Based on RMC Codes

2018 ◽  
Author(s):  
Pan Qingquan ◽  
Wang Kan
Keyword(s):  
Computational Complexity ◽  
Conventional Method ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Photoelectric Effect ◽  
Photon Transport ◽  
Transport Calculation ◽  
Calculation Results ◽  
Transport Method ◽  
Photon Coupling

The conventional method for neutron-photon coupling transport calculation lacks of clear physical meanings, where the process of neutron transport and photon transport are independent, and only ensures the numbers of photons to be coupling with the neutrons. At the same time, when dealing with photoelectric effect, the nuclear data will be processed frequently, increasing the amount of calculation. By modifying the RMC codes, the deep-coupling and preprocessed photon transport is achieved. This new coupling method can satisfy the physical requirements and reduce the computational complexity while ensuring the accuracy of the calculation. At the same time, the preprocessing of the photoelectric effect nuclear data can accelerate the calculation without changing the calculation results. Through the deep-coupling and preprocessed photon transport method, the RMC codes can finished the high-precision shielding calculation. A typical LWR component is calculated with the new method, and the results prove the effectiveness.

scholarly journals The Application of the Combined Fission Matrix Theory in Fast Reactors

2021 ◽  
Vol 9 ◽  
Author(s):  
Donghao He ◽  
Tengfei Zhang ◽  
Xiaojing Liu
Keyword(s):  
High Efficiency ◽  
Matrix Theory ◽  
Neutron Transport ◽  
Thermal Reactor ◽  
Fast Reactors ◽  
Sodium Fast Reactor ◽  
The Monte Carlo Method ◽  
Transport Calculation ◽  
Rms Error ◽  
Transport Method

The combined fission matrix theory is a recently-developed hybrid neutron transport method. It features high efficiency, fidelity, and resolution whole-core transport calculation. The theory is based on the assumption that the fission matrix element ai,j is dominated by the property of the destination cell i. This assumption can be well explained in thermal reactors, and the combined fission matrix method has been validated in a series of thermal neutron system benchmarks. This work examines the feasibility of the combined fission matrix theory in fast reactors. The European Sodium Fast Reactor is used as the numerical benchmark. Compared to the Monte Carlo method, the combined fission matrix theory reports a 64 pcm keff difference and 8.3% 2D RMS error. The error is much larger than that in thermal reactors, and the correction ratio cannot significantly reduce the material discontinuity error in fast reactors. Overall, the combined fission matrix theory is more suited for thermal reactor transport calculations. Its application in fast reactors needs further developments.

scholarly journals Activation of high flux test module sample holder after IFMIF-DONES operation

2020 ◽  
Vol 60 (1) ◽  
Author(s):  
Simona Breidokaitė ◽  
Gediminas Stankūnas ◽  
Andrius Tidikas
Keyword(s):  
Neutron Transport ◽  
Nuclear Data ◽  
Total Activity ◽  
High Flux ◽  
Sample Holder ◽  
Dose Rates ◽  
The Monte Carlo Method ◽  
Nuclear Data Library ◽  
Calculation Results ◽  
Test Module

Nuclear safety assessment in nuclear fusion devices relies on the Monte Carlo method based neutron transport calculations. This paper presents information about the calculation results of the activities and dose rates caused by neuron irradiation for the structural materials of the high flux test module sample holder of IFMIF-DONES. The neutron induced activities and dose rates at shutdown were calculated by means of the FISPACT-2010 code with data from the EAF-2010 nuclear data library. Neutron fluxes and spectra were obtained with MCNP neutron transport calculations. The activities and dose rates were calculated at the end of irradiation of the assumed device operation scenario for cooling times of 0 s – 1000 year. In addition, radionuclides with contribution of at least 0.5% to the total value of activation characteristics at the previously mentioned cooling times were identified. After the operation, the most active radionuclide is 55Fe, with an activity share ranging from 30% (M200) to 63% (M8), and at the end of the prediction it accounts for 86% of the total activity. The highest dose rates at the end of irradiation are attributed to 56Mn radionuclide. 54Mn and 60Co are the most dominant radionuclides during intermediate and long cool-down periods.

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 Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

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.

Nuclear Data and the Effect of Gadolinium in the Moderator

2012 ◽  
Vol 1 (1) ◽  
pp. 21-25 ◽  
Author(s):  
J.C. Chow ◽  
F.P. Adams ◽  
D. Roubstov ◽  
R.D. Singh ◽  
M.B. Zeller
Keyword(s):  
Cross Section ◽  
Neutron Capture ◽  
Capture Cross Section ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Calibration Data ◽  
Capture Cross ◽  
Neutron Capture Cross Section ◽  
Reactivity Effect ◽  
Section Data

Recent cross-section measurements on gadolinium have raised concerns over the accuracy of moderator poison reactivity coefficient calculations. Measurements have been made at the ZED-2 (Zero Energy Deuterium) critical facility, Chalk River Laboratories, AECL, to study the reactivity effect of gadolinium in the moderator. Since the neutron capture cross-section of boron is well known, measurements were also made with boron to provide calibration data for measurements with gadolinium. The measurements have been used to quantify the bias of the reactivity effect in full-core simulations of ZED-2 using MCNP, a neutron transport code used extensively for simulations of nuclear systems, along with the ENDF/B-VII.0 cross-section data. The results showed a bias of -0.41 ± 0.07 mk/ ppm, or -2.1% ± 0.3%, given a reactivity worth of -20.1 mk/ppm for gadolinium. Additional simulations also show that the gadolinium neutron capture cross-section has been over-corrected, relative to previous evaluations, in a beta version of ENDF/B VII.1, which incorporates the Leinweber data.

Sign in / Sign up

Export Citation Format

Share Document