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

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.

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ANALYSIS OF THE BEAVRS BENCHMARK USING CASMO5/SIMULATE5 WITH JENDL-4.0 AND ENDF/B-VII.1

EPJ Web of Conferences ◽

10.1051/epjconf/202124702014 ◽

2021 ◽

Vol 247 ◽

pp. 02014

Author(s):

Fujita Tatsuya ◽

Sakai Tomohiro

Keyword(s):

Measurement Data ◽

Nuclear Data ◽

Isothermal Temperature ◽

Control Rod ◽

Significant Difference ◽

Nuclear Data Library ◽

Calculation Results ◽

The Difference ◽

Data Library ◽

Good Agreement

The BEAVRS benchmark was analyzed using the CASMO5/SIMULATE5 in order to compare the measurement data and the calculation results based on the JENDL-4.0 and ENDF/B-VII.1 and investigate the difference between those calculation results. For the hot zero power (HZP) physics test, the calculation results showed good agreement with the measurement data for both of cycles 1 and 2. For cycle 1, the calculation results of the isothermal temperature coefficient (ITC) differed from the measurement data by approximately 1 pcm/℉, and the same tendency has been reported in previous studies. For the cycle operation, the calculation results of the boron letdown agreed well with the measurement data. On the other hand, some calculation results of the axial detector signals had a large difference from the measurement data, which is supposedly attributed to the discrepancy of the axial offset (AO) caused by the authors’ approximation for the control rod and shutdown bank positions. In terms of the comparison between the JENDL-4.0 and ENDF/B-VII.1, although approximately 15 ppm difference of the boron letdown in the cycle operation was observed, no significant difference was seen for other core parameters, thus, the influence of the two nuclear data library was small on the present results.

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

EPJ Web of Conferences ◽

10.1051/epjconf/202022503009 ◽

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.

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Uncertainty quantification and sensitivity studies on Thorium-fueled reactors

EPJ Web of Conferences ◽

10.1051/epjconf/202023922008 ◽

2020 ◽

Vol 239 ◽

pp. 22008

Author(s):

Eliot Party ◽

Xavier Doligez ◽

Philippe Dessagne ◽

Maëlle Kerveno ◽

Greg Henning

Keyword(s):

Cross Sections ◽

Nuclear Data ◽

Pressurized Water ◽

Correlation Matrices ◽

Target Uncertainty ◽

Nuclear Data Library ◽

Sensitivity Studies ◽

Water Reactors ◽

Data Library ◽

Good Agreement

This paper shows how Total Monte Carlo (TMC) method and Perturbation Theory (PT) can be applied to quantify uncertainty due to nuclear data on reactor static calculations of integral parameters such as keff and βeff. This work focuses on thorium fueled reactors and it aims to rank different cross sections uncertainty regarding criticality calculations. The consistency of the two methods are first studied. The cross sections set used for the TMC method is computed to build adequate correlation matrices. Those matrices are then multiplied by the sensitivity coefficients obtained thanks to the PT to obtain global uncertainties that are compared to the ones calculated by the TMC method. Results in good agreement allow us to use correlation matrix from the state of the art nuclear data library (JEFF 3-3) that provide insight of uncertainty on keff and βeff for thorium fueled Pressurized Water Reactors. Finally, maximum uncertainties on cross sections are estimated to reach a target uncertainty on integral parameters. It is shown that a strong reduction of the current uncertainty is needed and consequently, new measurements and evaluations have to be performed.

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

Science and Technology of Nuclear Installations ◽

10.1155/2021/6633366 ◽

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.

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APPLICATION OF STUDSVIK’S CMS5 CODE SYSTEM TO ACCIDENT TOLERANT FUEL CORE DESIGN AND ANALYSIS

EPJ Web of Conferences ◽

10.1051/epjconf/202124702020 ◽

2021 ◽

Vol 247 ◽

pp. 02020

Author(s):

Gerardo Grandi ◽

Rodolfo Ferrer ◽

Tamer Bahadir

Keyword(s):

Three Dimensional ◽

Nuclear Data ◽

Light Water Reactors ◽

Code System ◽

Nuclear Data Library ◽

Fuel Pin ◽

Water Reactors ◽

Institute Of Technology ◽

Data Library ◽

Good Agreement

The possible deployment of Accident Tolerant Fuels (ATF) for currently-operating Light Water Reactors (LWR) has prompted interest in the use of Studsvik’s CMS5 code system to support the analysis of such advanced ATF core designs. Various ATF concepts have been proposed; for example, uranium silicide (U3Si2) fuel, together with iron-based (FeCrAl) cladding. The purpose of this work is to showcase the application of the CMS5 code system, which includes the CASMO5 advanced lattice physics code and the SIMULATE5 three-dimensional nodal simulator, to the analysis of a U3Si2/FeCrAl ATF concept. Given that the CMS5 code system was designed from inception to enable the analysis of advanced core designs, only minor changes to the CASMO5 lattice physics code and SIMULATE5 core simulator are necessary. The current CASMO5 586 energy-group nuclear data library provides all the necessary data to support the generation of homogenized data for downstream use by SIMULATE5 for ATF. The SIMULATE5 nodal code, which features a simplified fuel pin model, requires updating various thermophysical properties corresponding to the U3Si2/SiC ATF fuel and the gaseous conductance models. An equilibrium core for the Integral Inherently Safe (I2S) LWR design developed by the Georgia Institute of Technology was selected. The results of the CMS5 simulation were compared with those in the literature and were found to be in good agreement, giving us confidence that the CMS5 package can be used in the modeling of LWR systems with ATF technology.

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