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Journal Articles

$$^{rm 208,207,206,nat}$$Pb(p,x)$$^{207}$$Bi and $$^{209}$$Bi(p,x)$$^{207}$$Bi excitation functions in the energy range of 0.04 - 2.6 GeV

Titarenko, Yu. E.*; Batyaev, V. F.*; Pavlov, K. V.*; Titarenko, A. Yu.*; Malinovskiy, S. V.*; Rogov, V. I.*; Zhivun, V. M.*; Kulevoy, T. V.*; Chauzova, M. V.*; Lushin, S. V.*; et al.

Nuclear Instruments and Methods in Physics Research A, 984, p.164635_1 - 164635_8, 2020/12

 Times Cited Count:0 Percentile:0.02(Instruments & Instrumentation)

The paper presents the $$^{207}$$Bi production cross-sections measured by the direct gamma-spectrometry technique in the samples of lead enriched with isotopes 208, 207 and 206, as well as in the samples of natural lead and bismuth, irradiated by protons of 11 energies in the range from 0.04 to 2.6 GeV. The obtained experimental results are compared with the previous measurements, with the TENDL-2019 data-library evaluations and the simulated data by means of the high-energy transport codes MCNP6.1 (CEM03.03), PHITS (INCL4.6/GEM), and Geant4 (INCL++/ABLA). The observed discrepancies between model predictions and experimental data are discussed.

Journal Articles

HPRL; International cooperation to identify and monitor priority nuclear data needs for nuclear applications

Dupont, E.*; Bossant, M.*; Capote, R.*; Carlson, A. D.*; Danon, Y.*; Fleming, M.*; Ge, Z.*; Harada, Hideo; Iwamoto, Osamu; Iwamoto, Nobuyuki; et al.

EPJ Web of Conferences, 239, p.15005_1 - 15005_4, 2020/09

 Times Cited Count:0 Percentile:0.19

Journal Articles

The Joint evaluated fission and fusion nuclear data library, JEFF-3.3

Plompen, A. J. M.*; Cabellos, O.*; De Saint Jean, C.*; Fleming, M.*; Algora, A.*; Angelone, M.*; Archier, P.*; Bauge, E.*; Bersillon, O.*; Blokhin, A.*; et al.

European Physical Journal A, 56(7), p.181_1 - 181_108, 2020/07

 Times Cited Count:73 Percentile:99.41(Physics, Nuclear)

The Joint Evaluated Fission and Fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}$$U, $$^{238}$$U and $$^{239}$$Pu, on $$^{241}$$Am and $$^{23}$$Na, $$^{59}$$Ni, Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yileds, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 is excellent for a wide range of nuclear technology applications, in particular nuclear energy.

Journal Articles

Sensitivity and uncertainty analysis of $$beta_{rm eff}$$ for MYRRHA using a Monte Carlo technique

Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*

European Physical Journal; Nuclear Sciences & Technologies (Internet), 4, p.42_1 - 42_7, 2018/11

This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction $$beta_{rm eff}$$ for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo transport code MCNP. The $$beta_{rm eff}$$ sensitivities are calculated by the modified $$k$$-ratio method proposed by Chiba. Comparing the $$beta_{rm eff}$$ sensitivities obtained with different scaling factors $$a$$ introduced by Chiba shows that a value of $$a=20$$ is the most suitable for the uncertainty quantification of $$beta_{rm eff}$$. Using the calculated $$beta_{rm eff}$$ sensitivities and the JENDL-4.0u covariance data, the $$beta_{rm eff}$$ uncertainties for the critical and subcritical cores are determined to be 2.2 $$pm$$ 0.2% and 2.0 $$pm$$ 0.2%, respectively, which are dominated by delayed neutron yield of $$^{239}$$Pu and $$^{238}$$U.

Journal Articles

High-energy nuclear data uncertainties propagated to MYRRHA safety parameters

Stankovskiy, A.*; Iwamoto, Hiroki; $c{C}$elik, Y.*; Van den Eynde, G.*

Annals of Nuclear Energy, 120, p.207 - 218, 2018/10

 Times Cited Count:6 Percentile:72.04(Nuclear Science & Technology)

Propagation of high-energy (above 20-MeV) nuclear data uncertainties on the safety related neutronic responses in accelerator driven systems has been assessed. The total core power and production of radionuclides contributing to radiation source terms were focused on. The article features a method based on the Monte Carlo sampling of random nuclear data files from the covariance matrices generated from the sets of reaction cross sections obtained with model calculations of high-energy particle interactions with matter or picked up from already existing nuclear data libraries. It has been demonstrated that nuclear data uncertainties do not need to be propagated through particle transport calculations to obtain uncertainties on the responses. This advantage allowed to investigate the convergence of the sample average to the best estimate. The number of random nuclear data file sets needed to obtain reliable uncertainty on the total core power is around 300 that results in the uncertainty of 14%. The uncertainties on the concentrations of nuclides most important for the safety assessment that are accumulated in lead-bismuth eutectic during irradiation, range from 5 to 60%. Concentrations of some nuclides exemplified by Tritium converge much slower than neutron multiplicities so that several thousands of samples are needed to ensure reliable uncertainty estimates.

Journal Articles

Monte Carlo uncertainty quantification of the effective delayed neutron fraction

Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*

Journal of Nuclear Science and Technology, 55(5), p.539 - 547, 2018/05

 Times Cited Count:6 Percentile:72.04(Nuclear Science & Technology)

The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction $$beta_{rm eff}$$ is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of $$beta_{rm eff}$$ uncertainty. For validation of the calculated $$beta_{rm eff}$$, a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified $$k$$-ratio method is superior to Bretscher's prompt $$k$$-ratio method in terms of reducing the statistical uncertainty in calculating not only $$beta_{rm eff}$$ but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified $$k$$-ratio method is the most practical for uncertainty quantification of $$beta_{rm eff}$$. Finally, total $$beta_{rm eff}$$ uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of $$^{235}$$U.

Journal Articles

Sensitivity and uncertainty analysis of the VENUS-F critical core

Iwamoto, Hiroki; Stankovskiy, A.*; Uyttenhove, W.*

Energy Procedia, 71, p.33 - 41, 2015/05

 Times Cited Count:1 Percentile:75.48

Journal Articles

Nuclear data sensitivity/ uncertainty analysis for XT-ADS

Sugawara, Takanori; Sarotto, M.*; Stankovskiy, A.*; Van den Eynde, G.*

Annals of Nuclear Energy, 38(5), p.1098 - 1108, 2011/03

 Times Cited Count:11 Percentile:69.09(Nuclear Science & Technology)

The XT-ADS, an accelerator-driven system for an experimental demonstration, has been investigated in the framework of IP EUROTRANS FP6 project. In this study, the sensitivity and uncertainty analyses were performed to comprehend the reliability of the XT-ADS neutronic design. For the sensitivity analysis, it was found that the sensitivity coefficients were significantly different by changing the geometry models and calculation codes. For the uncertainty analysis, it was confirmed that the uncertainties deduced from the covariance data varied significantly by changing them. The uncertainties deduced from the covariance data for the XT-ADS criticality were 0.94%, 1.9% and 1.1% by the SCALE 44-group, TENDL-2009 and JENDL-3.3 data, respectively. When the target accuracy of 0.3%dk for the criticality was considered, the uncertainties did not satisfy it. To achieve this accuracy, the uncertainties should be improved by experiments under an adequate condition.

Journal Articles

Impact of nuclear data uncertainties on neutronics parameters of MYRRHA/XT-ADS

Sugawara, Takanori; Stankovskiy, A.*; Sarotto, M.*; Van den Eynde, G.*

NEA/NSC/DOC(2011)4, p.27 - 35, 2011/00

A flexible fast spectrum research reactor MYRRHA able to operate in subcritical (driven by a proton accelerator) and critical modes is being developed in SCK$$cdot$$CEN. In the framework of IP EUROTRANS programme the XT-ADS model has been investigated for MYRRHA. In this study, the sensitivity and uncertainty analysis was performed to comprehend the reliability of the XT-ADS neutronics design. The calculated sensitivity coefficients for neutronics parameters varied significantly between calculation models. The uncertainties deduced from the covariance data strongly depend on the original covariance data. The calculated nuclear data uncertainties could not meet the target accuracy. To improve uncertainties, the integral experiments in adequate conditions are expedient.

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