Benchmark experiment on copper with D-T neutrons for verification of neutron transport and related nuclear data of ?-3.1

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.

Download Full-text

Nuclear Data and the Effect of Gadolinium in the Moderator

AECL Nuclear Review ◽

10.12943/anr.2012.00004 ◽

2012 ◽

Vol 1 (1) ◽

pp. 21-25 ◽

Cited By ~ 2

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.

Download Full-text

ORNL benchmark experiment for neutron transport in iron and stainless steel

10.2172/4492432 ◽

1973 ◽

Author(s):

R. E. Maerker ◽

F. J. Muckenthaler

Keyword(s):

Stainless Steel ◽

Neutron Transport ◽

Benchmark Experiment

Download Full-text

Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

Nukleonika ◽

10.2478/nuka-2019-0011 ◽

2019 ◽

Vol 64 (3) ◽

pp. 87-96 ◽

Cited By ~ 2

Author(s):

Piotr Darnowski ◽

Michał Pawluczyk

Keyword(s):

Fuel Cycle ◽

Neutron Transport ◽

Measurement Data ◽

Nuclear Data ◽

Data File ◽

Power Measurement ◽

Future Research ◽

Pressurized Water ◽

Control Rod ◽

Nodal Diffusion

Abstract This paper presents an analysis of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) performed using SCALE 6.1.2 and PARCS 3.2 computer codes. The benchmark specification contains a detailed design, operational data and measurements for a real 4-loop Westinghouse pressurized water reactor (PWR). The lattice physics simulations were prepared using TRITON depletion sequence and NEWT neutron transport solver (SCALE package). The 238-neutron group library based on evaluated nuclear data file – ENDF/B-VII nuclear data libraries was applied. A set of branch and burnup calculations was prepared, and group constants in the form of PMAXS files were generated with GenPMAXS. The full-core models were prepared using the PARCS nodal-diffusion core simulator. The PMAXS libraries were used with PARCS to investigate the core operation. The hot zero power measurement data, including control rod worths and critical boron concentrations, were compared using simulations, and satisfactory results were achieved. The first fuel cycle was simulated, and acceptable agreement with boron letdown curve and measurements were obtained. Finally, conclusions and recommendations for future research were presented.

Download Full-text

DIAGNOSIS OF THE UNRESOLVED DOMAIN TREATMENT IN MONTE CARLO TRANSPORT CALCULATIONS THROUGH THE IDENTIFICATION AND MODELLING OF CRITICALITY SAFETY EXPERIMENTS

EPJ Web of Conferences ◽

10.1051/epjconf/202124710003 ◽

2021 ◽

Vol 247 ◽

pp. 10003

Author(s):

N. García-Herranz ◽

J. Rodríguez ◽

A. Jiménez-Carrascosa ◽

O. Cabellos

Keyword(s):

Monte Carlo ◽

Neutron Transport ◽

Resonance Region ◽

Nuclear Data ◽

High Fidelity ◽

Continuous Energy ◽

Critical Facilities ◽

Nuclear Systems ◽

Monte Carlo Transport ◽

Criticality Safety

Monte Carlo neutron transport codes can be used for high-fidelity predictions of the performance of nuclear systems. However, validation against experiments is required in order to establish the credibility in the results and identify the inaccuracies due to the used calculation scheme and associated databases. The International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) contains criticality safety benchmarks derived from experiments that have been performed at various nuclear critical facilities around the world and are very valuable for validation purposes. The main objective of this work is the identification and modelling of experimental benchmarks included at ICSBEP in support of the validation of Monte Carlo neutron transport calculations when applied to fast systems, and in particular, KENO-VI and associated AMPX-formatted continuous-energy libraries from SCALE package. In such systems, the predicted k-eff values can be very sensitive to the treatment of nuclear data in the Unresolved Resonance Region (URR). Consequently, benchmarks with intermediate and fast spectra are identified and modelled with KENO-VI. Then, calculated results with and without probability tables in the URR are compared with each other in order to identify the most sensitive configurations to the URR. As a result of the proposed study, recommendations are given about the benchmarks that should be modelled and analysed to qualify the processed continuous-energy libraries before their use in Monte Carlo transport codes for practical fast reactor applications.

Download Full-text

ORNL BENCHMARK EXPERIMENT FOR NEUTRON TRANSPORT IN THICK SODIUM.

10.2172/4604477 ◽

1972 ◽

Author(s):

R.E. Maerker ◽

F.J. Muckenthaler ◽

C.E. Clifford

Keyword(s):

Neutron Transport ◽

Benchmark Experiment

Download Full-text