17 research outputs found

    Comparison of resonance integrals of cross sections from JEFF-3.2 library for some problematic reactions

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    The quality of the capture cross sections in JEFF-3.2 for a selection of nuclides has been assessed in comparison to other evaluated nuclear data libraries (ENDF/B-VII.1, JENDL-4.0, TENDL-2014 and IRDFF v1.05). The incident neutron capture reactions of this nuclides have been compared to experimental data from the EXFOR database in terms of resonance integrals and, where available, energy dependent data. Recommendations for next version of the JEFF library have been given. For 55Mn, JEFF-3.2 is strongly recommended. For 58Fe and 176,178Hf, JEFF-3.2 is recommended. For 93Nb and 148Nd, JEFF-3.2 is not recommended. For those two nuclides, the capture cross section from JENDL-4.0 is recommended.JRC.D.4-Standards for Nuclear Safety, Security and Safeguard

    Evaluation of neutron induced reaction cross sections on gold

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    A new evaluation of neutron induced reactions on 197Au nucleus in the energy regions below 500 eV and from 4 keV and 100 keV is presented. Complete evaluated data files in ENDF-6 format have been produced by joining the evaluation with corresponding files from the ENDF/B-VII.1 library. The evaluation in the unresolved resonance region between 4 keV and 100 keV is based on a generalized single-level representation compatible with the energy-dependent option of the ENDF-6 format. The average partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser-Feshbach statistical reaction theory including width fluctuations. The transmission coefficients have been obtained from a combined analysis of the capture cross section resulting from the cross section standards evaluation project and theoretical non-fluctuating cross sections derived from a dispersive coupled channel optical model. The evaluated cross sections have been validated by a comparison with transmission and capture data obtained at the time-of-flight facility GELINA. The evaluated files have been processed with the latest updates of NJOY.99 to test their format and application consistency as well as to produce a continuous-energy data library in ACE format for use in Monte Carlo codes. The ACE files have been utilized to study the effect of the evaluated resonance parameters on results of lead slowing-down experiments. The evaluated files will be implemented in the next release of the JEFF-3 library which is maintained by the Nuclear Energy Agency of the OECD.JRC.D.4-Standards for Nuclear Safety, Security and Safeguard

    Observables of interest for the characterisation of Spent Nuclear Fuel

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    The characterisation of Spent Nuclear Fuel (SNF) in view of intermediate storage and final disposal is discussed. The main observables of interest that need to be determined are the decay heat, neutron and -ray emission spectra. In addition, the inventory of specific nuclides that are important for criticality safety analysis and to verify the fuel history has to be determined. Some of the observables such as the decay heat and neutron and -ray emission rate can be determined by Non-Destructive Analysis (NDA) methods. Unfortunately, this is not always possible especially during routine operation. Hence, a characterisation of SNF will rely on theoretical calculations combined with results of NDA methods. In this work the observables of interest, also referred to as source terms, are discussed based on theoretical calculations starting from fresh UO2 and MOX fuel. The irradiation conditions are representative for PWR. The Serpent code is used to define the nuclides which have an important contribution to the observables. The emphasis is on cooling times between 1 a and 1000 a.JRC.G.2-Standards for Nuclear Safety, Security and Safeguard

    Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

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    A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK-CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data

    Recommendations for MYRRHA relevant cross section data to the JEFF project

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    Within the framework of Work Package 10 of the EC FP7 CHANDA project, nuclear data of importance for the operation of MYRRHA, a lead-bismuth cooled accelerator driven reactor under development at SCK‚ÄĘCEN (BE), were studied. Based on data in the main nuclear data libraries, i.e. JEFF, JENDL, ENDF/B and BROND, and in the TENDL and CIELO libraries and on experimental data reported in the literature, recommendations to the JEFF project were made for several nuclides of interest to the MYRRHA reactor.JRC.G.2-Standards for Nuclear Safety, Security and Safeguard

    A non-destructive method to determine the neutron production rate of a sample of spent nuclear fuel under standard controlled area conditions

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    A method to determine the neutron production rate of a sample of spent nuclear fuel by means of non-destructive analysis conducted under controlled-area conditions is described, validated and demonstrated. A standard neutron well-counter designed for routine nuclear safeguards applications is applied. The method relies on a transfer procedure that is adapted to the hot-cell facilities at the laboratory for high and medium level activity of the SCK CEN. The sample transfer and measurement procedures are described together with results of Monte Carlo simulations. Experiments with radionuclide sources were carried out at the Joint Research Centre to test the procedures and to determine the performance characteristics of the detection device. Finally, measurements of a segment of a spent nuclear fuel rod were carried out at the SCK CEN to validate and demonstrate the method.JRC.G.2-Standards for Nuclear Safety, Security and Safeguard

    Status of evaluated data files for 238U in the resonance region

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    Experimental data and evaluated data libraries related to neutron induced reaction cross sections for 238U in the resonance region are reviewed. Based on this review a set of test files is produced to study systematic effects such as the impact of the upper boundary of the resolved resonance region (RRR) and the representation of the infinite diluted capture and in-elastic cross section in the unresolved resonance region (URR). A set of Benchmark experiments was selected and used to verify the test files. Based on these studies recommendations to perform a new evaluation have been defined. This report has been prepared in support to the CIELO (Collaborative International Evaluated Library Organisation) project. The objective of this project is the creation of a world-wide recognised nuclear data file with a focus on six nuclides, i.e. 1H, 16O, 56Fe, 235U, 238U and 239Pu. Within the CIELO project, the Joint Research Centre (JRC) at Geel (B) is in charge of the production of an evaluated cross section data file for neutron induced reaction of 238U in the resonance region.JRC.D.4-Standards for Nuclear Safety, Security and Safeguard

    Nuclear data uncertainty propagation to integral responses using SANDY

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    SANDY is a nuclear data sampling code compatible with nuclear data files in ENDF-6 format. Exploiting the basic theory of stochastic sampling, SANDY generates random nuclear data samples that reproduce the covariance information stored in the ENDF-6 files. Such random data are rewritten in perturbed ENDF-6 or PENDF files and can be used as nuclear code inputs to produce perturbed responses. After the statistical analysis of the sampled responses, the sample mean and variance are quantified. Not only can SANDY be used for the study of the response sample variance, but it can also estimate global sensitivity indices for correlated and uncorrelated input parameters using a variance-based decomposition method. SANDY was tested for the generation of random ENDF-6 files with perturbed cross sections and resonance parameters. These files were used for the uncertainty quantification of integral responses, such as the one-group integrated cross section and the resonance integral of several isotopes. Then, the response variance was apportioned to the several inputs and the parameters producing the largest impact were identified.JRC.G.2-Standards for Nuclear Safety, Security and Safeguar

    Systematic effects on cross section data derived from reaction rates in reactor spectra and a re-analysis of 241Am reactor activation measurements

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    Methodologies to derive cross section data from spectrum integrated reaction rates were studied. The Westcott convention and some of its approximations were considered. Mostly measurements without and with transmission filter are combined to determine the reaction cross section at thermal energy together with the resonance integral. The accuracy of the results strongly depends on the assumptions that are made about the neutron energy distribution, which is mostly parameterised as a sum of a thermal and an epi-thermal component. Resonance integrals derived from such data can be strongly biased and should only be used in case no other data are available. The cross section at thermal energy can be biased for reaction cross sections which are dominated by low energy resonances. The amplitude of the effect is related to the lower energy limit that is used for the epi-thermal component of the neutron energy distribution. It is less affected by the assumptions on the shape of the energy distribution. When the energy dependence of the cross section is known and information about the neutron energy distribution is available, a method to correct for a bias on the cross section at thermal energy is proposed. Reactor activation measurements to determine the thermal 241Am(n,ő≥) cross section reported in the literature were reviewed. In case enough information was available, the results were corrected to account for possible biases and included in a least squares fit. These data combined with results of time-of-flight measurements give a capture cross section 720 (14) b for 241Am(n,ő≥) at thermal energy.JRC.G.2-Standards for Nuclear Safety, Security and Safeguard
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