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

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

A Distributed Algorithm for Bandwidth Allocation in Stable Ad Hoc Networks

We propose a distributed algorithm for allocating bandwidth in stable ad hoc networks. After having discussed the problem of bandwidth allocation in such networks, we define a sequence of feasible solutions to this problem. This sequence has the property to be an increasing sequence in terms of overall used bandwidth. After a theoretical analysis of the sequence, we design a distributed algorithm based on this sequence. We test our algorithm by simulations on different topologies like chains, rings, meshes and geometric random graphs.We compare our solutions with the optimal solution in terms of global bandwidth allocation that presents the smallest standard deviation and with the the fairest solution regarding to max-min fairness. The simulations show that the global used bandwidth is less than $25\%$ from optimality in the worst case and the standard deviation is the smallest of the three tested solutions

Pojednostavljeni račun sparivanja u poligrafovima

Matching polynomial and perfect matchings for fasciagraphs, rotagraphs and twisted rotagraphs are treated in the paper. Classical transfer matrix approach makes it possible to get recursions for matching polynomial and perfect matchings, but the order of the matrix grows exponentially in the number of the linking edges between monographs. Novel transfer matrices are introduced whose order is much lower than that in classical transfer matrices. The virtue of the method introduced is especially pronounced when two or more linking edges end in the same terminal vertex of a monograph. An example of a polyacene polygraph with extended pairings is given where a novel matrix has only 16 entries as compared to 65536 entries in the classical transfer matrix. However, all pairings are treated here on equal footing, but the method introduced can be applied to selected types of pairings of interest in chemistry.U radu se razmatraju polinomi sparivanja i savršena sparivanja u fascia- i rotagrafovima te izvijenim rotagrafovima. Iako klasični postupak transfer matrice omogućava izvođenje rekurzija za polinom sparivanja i savršena sparivanja, red ove matrice eksponencijalno raste s brojem veza me|u monografovima. Ovdje su uvedene nove transfer matrice čiji je red mnogo ni`i od onoga za klasične transfer matrice, i to posebice kada jedna ili više veza me|u monografovima završava u jednom te istom čvoru. Postupak je ilustriran na primjeru poliacenskih poligrafova gdje ovdje uvedena matrica ima samo 16 elemenata u usporedbi s 65536 elemenata klasične transfer matrice. Iako se ovdje uvedeni postupak primjenjuje istovremeno na sva moguća sparivanja u poligrafovima, on je otvoren za primjenu na odabrana sparivanja od posebnoga kemijskoga interesa

Evaluation of neutron induced reaction cross sections on gold

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

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

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

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•CEN (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