| 21. |
- Pomp, Stephan, et al.
(författare)
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Experiments and Theoretical Data for Studying the Impact of Fission Yield Uncertainties on the Nuclear Fuel Cycle with TALYS/GEF and the Total Monte Carlo Method
- 2015
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Ingår i: Nuclear Data Sheets. - : Elsevier BV. - 0090-3752 .- 1095-9904. ; 123:SI, s. 220-224
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Tidskriftsartikel (refereegranskat)abstract
- We describe the research program of the nuclear reactions research group at Uppsala University concerning experimental and theoretical efforts to quantify and reduce nuclear data uncertainties relevant for the nuclear fuel cycle. We briefly describe the Total Monte Carlo (TMC) methodology and how it can be used to study fuel cycle and accident scenarios, and summarize our relevant experimental activities. Input from the latter is to be used to guide the nuclear models and constrain parameter space for TMC. The TMC method relies on the availability of good nuclear models. For this we use the TALYS code which is currently being extended to include the GEF model for the fission channel. We present results from TALYS-1.6 using different versions of GEF with both default and randomized input parameters and compare calculations with experimental data for U-234(n,f) in the fast energy range. These preliminary studies reveal some systematic differences between experimental data and calculations but give overall good and promising results.
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| 22. |
- Rochman, D., et al.
(författare)
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Nuclear data uncertainty for criticality-safety : Monte Carlo vs. linear perturbation
- 2016
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 92, s. 150-160
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Tidskriftsartikel (refereegranskat)abstract
- This work is presenting a comparison of results for different methods of uncertainty propagation due to nuclear data for 330 criticality-safety benchmarks. Covariance information is propagated to key using either Monte Carlo methods (NUSS: based on existing nuclear data covariances, and TMC: based on reaction model parameters) or sensitivity calculations from MCNP6 coupled with nuclear data covariances. We are showing that all three methods are globally equivalent for criticality calculations considering the two first moments of a distribution (average and standard deviation), but the Monte Carlo methods lead to actual probability distributions, where the third moment (skewness) should not be ignored for safety assessments.
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| 23. |
- Rochman, D., et al.
(författare)
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Radiative neutron capture : Hauser Feshbach vs. statistical resonances
- 2017
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Ingår i: Physics Letters B. - : ELSEVIER SCIENCE BV. - 0370-2693 .- 1873-2445. ; 764, s. 109-113
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Tidskriftsartikel (refereegranskat)abstract
- The radiative neutron capture rates for isotopes of astrophysical interest are commonly calculated on the basis of the statistical Hauser Feshbach (HF) reaction model, leading to smooth and monotonically varying temperature-dependent Maxwellian-averaged cross sections (MACS). The HF approximation is known to be valid if the number of resonances in the compound system is relatively high. However, such a condition is hardly fulfilled for keV neutrons captured on light or exotic neutron-rich nuclei. For this reason, a different procedure is proposed here, based on the generation of statistical resonances. This novel technique, called the "High Fidelity Resonance" (HFR) method is shown to provide similar results as the HF approach for nuclei with a high level density but to deviate and be more realistic than HF predictions for light and neutron-rich nuclei or at relatively low sub-keV energies. The MACS derived with the HFR method are systematically compared with the traditional HF calculations for some 3300 neutron-rich nuclei and shown to give rise to significantly larger predictions with respect to the HF approach at energies of astrophysical relevance. For this reason, the HF approach should not be applied to light or neutron-rich nuclei. The Doppler broadening of the generated resonances is also studied and found to have a negligible impact on the calculated MACS.
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| 24. |
- Rochman, D., et al.
(författare)
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Re-evaluation of the thermal neutron capture cross section of Nd-147
- 2016
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 94, s. 612-617
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Tidskriftsartikel (refereegranskat)abstract
- In this paper we are proposing a re-evaluation of the thermal-neutron induced capture cross section of Nd-147. A unique measurement exists from which this cross section was calculated in 1974. This original calculation is based on an assumed value for a specific gamma-ray fraction (called F-2), taken from the neighboring nucleus Nd-145. With the availability of reaction codes such as TALYS, such fraction can nowadays be calculated using specific reaction models and parameters. The new value of F-2 indicates a decrease of the thermal cross section by 45%, leading to 243 barns, instead of the 440 barns previously reported. This new cross section impacts the calculation of the number density for the well-known burn-up indicator Nd-148, but as shown, the change is close to the usual experimental uncertainties for the 148Nd number densities, thus having a limited impact on burn-up calculation.
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| 25. |
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| 26. |
- Sjöstrand, Henrik, et al.
(författare)
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Propagation of nuclear data uncertainties for ELECTRA burn-up calculations
- 2014
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Ingår i: Nuclear Data Sheets. - : Elsevier BV. - 0090-3752 .- 1095-9904. ; 118, s. 527-530
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Tidskriftsartikel (refereegranskat)abstract
- The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low -power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor life using the Total Monte Carlo approach(TMC). Within the TENDL project the nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty in the long-term radiotoxicity, decay heat, gas pressureand volatile fission products were found to be insignificant. However, the uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.
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| 27. |
- Sjöstrand, Henrik, 1978-, et al.
(författare)
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Propagation Of Nuclear Data Uncertainties For Fusion Power Measurements
- 2016
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Konferensbidrag (refereegranskat)abstract
- Fusion plasmas produce neutrons and by measuring the neutron emission the fusion power can be inferred. Accurate neutron yield measurements are paramount for the safe and efficient operation of fusion experiments and, eventually, fusion power plants.Neutron measurements are an essential part of the diagnostic system at large fusion machines such as JET and ITER. At JET, a system of activation foils provides the absolute calibration for the neutron yield determination. The activation system uses the property of certain nuclei to emit radiation after being excited by neutron reactions. A sample of suitable nuclei is placed in the neutron flux close to the plasma and after irradiation the induced radiation is measured. Knowing the neutron activation cross section one can calculate the time-integrated neutron flux at the sample position. To relate the local flux to the total neutron yield, the spatial flux response has to be identified. This describes how the local neutron emission affects the flux at the detector. The required spatial flux response is commonly determined using neutron transport codes, e.g., MCNP.Nuclear data is used as input both in the calculation of the spatial flux response and when the flux at the irradiation site is inferred. Consequently, high quality nuclear data is essential for the proper determination of the neutron yield and fusion power. However, uncertainties due to nuclear data are generally not fully taken into account in today’s uncertainty analysis for neutron yield calibrations using activation foils. In this paper, the neutron yield uncertainty due to nuclear data is investigated using the so-called Total Monte Carlo Method. The work is performed using a detailed MCNP model of JET fusion machine. In this work the uncertainties due to the cross sections and angular distributions in JET structural materials, as well as the activation cross sections, are analyzed. It is shown that a significant contribution to the neutron yield uncertainty can come from uncertainties in the nuclear data.
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| 28. |
- Sjöstrand, Henrik, et al.
(författare)
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Total Monte Carlo evaluation for dose calculations
- 2014
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Ingår i: Radiation Protection Dosimetry. - : Oxford University Press (OUP). - 0144-8420 .- 1742-3406. ; 161:1-4, s. 312-315
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Tidskriftsartikel (refereegranskat)abstract
- Total Monte Carlo (TMC) is a method to propagate nuclear data (ND) uncertainties in transport codes, by using a large set of ND files, which covers the ND uncertainty. The transport code is run multiple times, each time with a unique ND file, and the result is a distribution of the investigated parameter, e.g. dose, where the width of the distribution is interpreted as the uncertainty due to ND. Until recently, this was computer intensive, but with a new development, fast TMC, more applications are accessible. The aim of this work is to test the fast TMC methodology on a dosimetry application and to propagate the 56Fe uncertainties on the predictions of the dose outside a proposed 14-MeV neutron facility. The uncertainty was found to be 4.2 %. This can be considered small; however, this cannot be generalised to all dosimetry applications and so ND uncertainties should routinely be included in most dosimetry modelling.
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