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In a typical pressurized water reactor, neutron detectors located outside the reactor core monitor reactor power. In addition, they are also used to measure the reactivity of the control rods. A novel approach to calculate the ex-core... more
In a typical pressurized water reactor, neutron detectors located outside the reactor core monitor reactor power. In addition, they are also used to measure the reactivity of the control rods. A novel approach to calculate the ex-core neutron detector response in a typical pressurized water reactor using the Monte Carlo technique is presented. A detailed ex-core model of the Krško nuclear power plant was developed using the Monte Carlo neutron transport code MCNP. Due to the location of the ex-core neutron detectors, the hybrid code ADVANTG is used to generate variance reduction parameters to accelerte the convergence of the results outside the reactor core. To use ADVANTG, the fixed neutron source had to be reconstructed from the criticality core calculation. This paper presents the sensitivity analysis of the response of the ex-core detectors to the neutron data libraries used, the description of the fixed neutron source and the ADVANTG parameters. It was found that a pin-wise des...
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There has been a continued effort since 2019 within the IAEA INDEN collaboration to improve the evaluation of neutron induced reactions on iron isotopes. The reason for the 30% underestimation of the neutron leakage spectrum from a thick... more
There has been a continued effort since 2019 within the IAEA INDEN collaboration to improve the evaluation of neutron induced reactions on iron isotopes. The reason for the 30% underestimation of the neutron leakage spectrum from a thick iron sphere was found primarily to be due to the overestimation of the inelastic cross sections in the 56Fe evaluated data file produced within the CIELO project of the OECD/NEA Data Bank. The over-estimation of the neutron flux between the resonances near 300 keV was traced to neglecting the fluctuating nature of the total cross section of 57Fe in the fast neutron energy range, since the evaluated resolved resonance range of 57Fe extended only up to 190 keV. The added 1/v background in the "iron window" below 28 keV is in excellent agreement with the independently evaluated one in the JENDL-5.0 library that included the direct capture component in the evaluation. Performance of the updated 56,57Fe evaluations was tested on a set of critic...
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Previous work done on reactor kinetics and control in load-following operation modes available in open literature is reviewed. The analysis is focused on, however not limited to pressurized water reactors. Different approximations of the... more
Previous work done on reactor kinetics and control in load-following operation modes available in open literature is reviewed. The analysis is focused on, however not limited to pressurized water reactors. Different approximations of the time-dependent neutron transport problem as well as different control algorithms are described in detail and compared. Due to lack of published information the majority of the comparisons was done on qualitative level. In order to facilitate future testing and intercomparisons of models and algorithms, two so-called reference scenarios with time-dependent power demand are defined: a scenario to test the limitations of the load-following capabilities of the nuclear facilities and a second, quasi-realistic scenario.
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Graphite and selected fluoride salts have their place in historic, as well as in most advanced nuclear reactors. Integral benchmarking of materials used in nuclear technology is very important for validation of calculation codes used at... more
Graphite and selected fluoride salts have their place in historic, as well as in most advanced nuclear reactors. Integral benchmarking of materials used in nuclear technology is very important for validation of calculation codes used at advanced reactor modelling. Discovered discrepancies can also point at inaccurate description of nuclear data. Benchmark experiments have been carried out in zero power light water reactor, LR-0, which is suited for these tasks thanks to the well-defined criticality. Criticality and neutron spectrum in graphite and fluoride salt has been benchmarked in experimental phase. The results of critical benchmark for graphite are showing good coincidence with calculation in ENDF/B-VII.0 nuclear library, however discrepancies exceeding 3a interval of uncertainties have been demonstrated with fluoride salt. Large discrepancies are apparent from C/E-1 comparison of neutron spectrum in fluoride salt. These discrepancies are likely caused by improper description ...
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The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty... more
The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty quantification is being developed by the International Atomic Energy Agency and expected to be released in January 2020; a preliminary version IRDFF-1.05 was released in 2014 and is being tested in this work. Validation of the cross-section evaluations proposed for this library is a high priority task. The validation can be realized using integral cross sections measured in standard and/or reference neutron benchmark fields. Integral quantities feature significantly lower uncertainties than differential nuclear data. If the neutron spectrum where the cross section is measured is well characterized, then the Spectrum Averaged Cross Section can be used for validating of existing evaluations.
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Reactor dosimetry is a method used to determine neutron flux or fluence in experimental or power reactors by measuring irradiated samples activities. Flux is then computed by solving Batemans equations of evolution with the appropriate... more
Reactor dosimetry is a method used to determine neutron flux or fluence in experimental or power reactors by measuring irradiated samples activities. Flux is then computed by solving Batemans equations of evolution with the appropriate nuclear data. Samples are often pure metal materials with different reactions of interest (capture or inelastic). Some samples may also be diluted alloys to limit neutron self-shielding. By combining information from different foils (with reactions having different energy responses function), a multigroup neutron spectrum can be determined by using unfolding methods. CEA and IJS have decided to validate methodologies of unfolding flux through collaboration between the two institutes [1]. One major step is the realization of an experimental dosimetry program in TRIGA Mark-II at IJS. This paper will focus on the analysis of the experimental measurements (reaction rates determination) and compare at last the experimental results to the monte-carlo comput...
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Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Joef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the... more
Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Joef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range ...
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FENDL nuclear data libraries are planned to be adjourned soon with a new release (v3.2). A beta version, together with previous FENDL releases and the ENDF/B-VIII.0, have been tested on the sphere leakage and ITER 1D computational... more
FENDL nuclear data libraries are planned to be adjourned soon with a new release (v3.2). A beta version, together with previous FENDL releases and the ENDF/B-VIII.0, have been tested on the sphere leakage and ITER 1D computational benchmarks using JADE, a new and under development verification and validation (V&V) tool. The consistency checks performed on the new FENDL v3.2 beta release did not spot any formal inconsistency and the comparison of the results show that, in general, the new FENDLv3.2 beta behaviour is quite similar to the FENDL v3.1d one. Nevertheless, a few significant differences with respect to the previous FENDL version and to the ENDF/B-VIII.0 results have been highlighted by the tool and discussed in the paper. The work proves how JADE has the potential to become an important player in the V&V procedures of nuclear data libraries.
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Resonance behavior is a feature of nuclear reaction cross sections. Resonance density increases with increasing incident particle energy and they begin to overlap, until they can no longer be resolved experimentally, but they still... more
Resonance behavior is a feature of nuclear reaction cross sections. Resonance density increases with increasing incident particle energy and they begin to overlap, until they can no longer be resolved experimentally, but they still contribute to self-shielding and must be accounted for. This is usually done by representing them with statistical average parameters according to methods and approximations described in standard text-books. Self-shielding factors are commonly used in deterministic transport codes, while statistical Monte Carlo codes use probability tables or multiband parameters. An exercise was conducted at the International Atomic Energy Agency (IAEA) to validate codes and methods for generating data that account for self-shielding in deterministic and Monte Carlo codes. A simple numerical model problem was defined, considering a sphere of 1 m radius with a 20 MeV isotropic neutron source at the center. The chosen material for testing was 139La from the ENDF/B-VIII.0 l...
ABSTRAC On December 2011 the ENDF/B-VII.1 library has been released. The ENDF/B-VII.1 library includes a re-evaluation of the neutron interaction data for Manganese-55. Many application files presently include nuclear data for Mn-55 that... more
ABSTRAC On December 2011 the ENDF/B-VII.1 library has been released. The ENDF/B-VII.1 library includes a re-evaluation of the neutron interaction data for Manganese-55. Many application files presently include nuclear data for Mn-55 that rely on previous evaluations. This is the case of the ADS-2.0 nuclear data library, available from the International Atomic Energy Agency (IAEA) for the calculations of accelerator-driven systems (ADS) and based on the ENDF/B-VII.0 evaluated nuclear data library. The new Mn-55 nuclear data were processed at IJS with the NJOY system in the ADS 421-group and the VITAMIN-J 175-group energy structures. The multi-group data were processed through TRANSX and ANISN for the simulation of a simple integral experiment with a Manganese shell that was performed at the OKTAVIAN facility. The deterministic calculation has been cross-validated with MCNP calculations adopting consistent nuclear data. Afterwards, the ANISN model has served for comparison with the 19...
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ABSTRACT This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation,... more
ABSTRACT This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.
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Abstract The new paradigm of nuclear reaction evaluation postulates that differential and integral experiments along with the reaction models should be used in concert to produce the evaluated data files. The results of a previous... more
Abstract The new paradigm of nuclear reaction evaluation postulates that differential and integral experiments along with the reaction models should be used in concert to produce the evaluated data files. The results of a previous assimilation project are summarized as a proof of principle that adjusting reaction model parameters to the results of integral experiments is feasible. The new paradigm requires vast modernization of the nuclear data infrastructure. Such modernization is actually carried out by several national and international efforts. We describe those dedicated to: (i) handling of differential and integral data, (ii) providing and operating the sensitivity profiles, (iii) ensuring reproducibility of the evaluations, and (iv) enabling automated verification of the entire library. We also point to the importance of the nuclear reaction modeling and discuss advantages and disadvantages of the new paradigm.
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Neutron activation analysis is the reference method used for offline determination of the neutron flux density in defined positions. It can be used in the nuclear energy industry-as well as in medical- or space applications. For accurate... more
Neutron activation analysis is the reference method used for offline determination of the neutron flux density in defined positions. It can be used in the nuclear energy industry-as well as in medical- or space applications. For accurate neutron flux evaluation, well-known and reliable cross sections are needed. In the thermal and fast energy region, many reliable monitoring reactions exists, however, in case of the epithermal and intermediate energy region, there are practically no dosimetry nuclear reactions sensitive specifically in this energy range. Due to this fact, both new data are being measured and methodologies are under development to describe and test this energy region. It was found that various neutron filters can be used to cut parts of neutron spectra and thus methodology based on spectrum filtering could potentially be employed to survey cross sections of interest. It this paper, the use of 3 different filters - B4C, Cd, and In is studied, on the case of the 55Mn(n...
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An international effort has produced evaluations of the neutron data standards. Evaluations were obtained for the cross section standards: the H(n,n), 6Li(n,t), 10B(n,067), loB(vx), natc(n,n,) Au(n ...
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Abstract We present details of the prompt fission gamma property evaluations included in the released ENDF/B-VIII.0 files. The average prompt gamma multiplicity, the average prompt gamma spectrum, the average total prompt gamma energy... more
Abstract We present details of the prompt fission gamma property evaluations included in the released ENDF/B-VIII.0 files. The average prompt gamma multiplicity, the average prompt gamma spectrum, the average total prompt gamma energy released and the prompt gamma multiplicity distributions, included in the ENDF/B-VIII.0 files, are presented for 235U(n,f), 238U(n,f), and 239Pu(n,f) reactions, as a function of incident neutron energy up to 30 MeV. The evaluation is based on available experimental data and model calculations. Notable data include recent measurements by experimental groups at Los Alamos and Lawrence Livermore National Laboratories, at Geel and Budapest, at CEA labs, as well as seminal historical gamma production measurements by Drake at Los Alamos.
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ABSTRACT CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction... more
ABSTRACT CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction data community to identify and document discrepancies among existing evaluated data libraries, measured data, and model calculation interpretations, and aims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted files. The focus will initially be on a small number of the highest-priority isotopes, namely 1H, 16O, 56Fe, 235,238U, and 239Pu. This paper identifies discrepancies between various evaluations of the highest priority isotopes, and was commissioned by the OECD’s Nuclear Energy Agency WPEC (Working Party on International Nuclear Data Evaluation Co-operation) during a meeting held in May 2012. The evaluated data for these materials in the existing nuclear data libraries — ENDF/B- VII.1, JEFF-3.1, JENDL-4.0, CENDL-3.1, ROSFOND, IRDFF 1.0 — are reviewed, discrepancies are identified, and some integral properties are given. The paper summarizes a program of nuclear science and computational work needed to create the new CIELO nuclear data evaluations.
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The detailed information on neutron source characteristics is important to meet the demand of research and applications on neutron activation and transport phenomena. Fast monoenergetic neutrons can be produced by two body reactions... more
The detailed information on neutron source characteristics is important to meet the demand of research and applications on neutron activation and transport phenomena. Fast monoenergetic neutrons can be produced by two body reactions induced by accelerated particles. In a number of laboratories much progress has been made in producing intense neutron sources based on the Deuteron – Tritium reaction to study the interaction of 14 MeV neutrons with structural materials of fusion reactors, to increase the sensitivity of the activation analysis, to measure the low cross sections and to use them in neutron therapy. At present the MCNP family of codes cannot model the D – T reaction for Deuteron energies in the KeV range explicitly. The physical processes involved in the neutron generation were modelled by defining the SOURCE and SRCDX subroutines of the MCNP code. The properties of a neutron source from a D – T reaction were investigated, starting from the ENDF/B-VII nuclear data library ...
International Conference N N Nu u uc c cl l le e ea a ar r r E E En n ne e er r rg g gy y y f f fo o or r r N N Ne e ew w w E E Eu u ur r ro o op p pe e e 2 2 20 0 00 0 09 9 9 ABSTRACT The mock-up of the EU Test Blanket Module (TBM) based... more
International Conference N N Nu u uc c cl l le e ea a ar r r E E En n ne e er r rg g gy y y f f fo o or r r N N Ne e ew w w E E Eu u ur r ro o op p pe e e 2 2 20 0 00 0 09 9 9 ABSTRACT The mock-up of the EU Test Blanket Module (TBM) based on the Helium Cooled Lithium Lead (HCLL) concept was irradiated at Frascatti, Italy, in order to assess the uncertainty in the tritium production rate (TPR) due to the uncertainty in the nuclear data and the computational methods. The benchmark should contribute also to the validation of the new nuclear cross-section and covariance data evaluations. This paper presents the final design of the benchmark and the analysis using the deterministic transport, sensitivity and uncertainty code system. The analysis includes the calculation of the tritium production rate (TPR) in LiPb layers and the neutron reaction rates, which were measured in the experimental set-up. The SUSD3D cross-section sensitivity and uncertainty code together with the 2D/3D determi...
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Influence of various up-to-date cross section libraries on the multiplication factor of TRIGA benchmark as well as the influence of fuel composition on the multiplication factor of the system composed of various types of TRIGA fuel... more
Influence of various up-to-date cross section libraries on the multiplication factor of TRIGA benchmark as well as the influence of fuel composition on the multiplication factor of the system composed of various types of TRIGA fuel elements was investigated. It was observed that keff calculated by using the ENDF/B VII cross section library is systematically higher than using the ENDF/B-VI cross section library. The main contributions (~220 pcm) are from 235U and Zr. 1
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The LOADF package has been designed for on-line reactor core monitoring of the Krsko NPP and for off-line prediction-mode calculations. Special features of the code are axial offset trajectory plots in which the axial offset difference... more
The LOADF package has been designed for on-line reactor core monitoring of the Krsko NPP and for off-line prediction-mode calculations. Special features of the code are axial offset trajectory plots in which the axial offset difference between power and xenon is plotted versus the difference between the axial offsets of iodine and xenon. The trajectories are elliptic tilted spirals, which can be parameterised. They measure the deviation from steady state condition with respect to xenon transients and can be used to predict the axial offset trends under transient conditions. An application of the axial offset trajectory diagram is described on the example of a power reduction transient.
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The objective of SINBAD (Shielding Integral Benchmark Archive Database) is to archive and make available the information on high–quality benchmark experiments for validating radiation transport codes and nuclear data. The quality... more
The objective of SINBAD (Shielding Integral Benchmark Archive Database) is to archive and make available the information on high–quality benchmark experiments for validating radiation transport codes and nuclear data. The quality assessment of the information available therein is based on the following issues: • completeness of the experimental specifications; • accuracy of the computational models; • utilization of recent nuclear data evaluations; • availability of sensitivity and uncertainty analyses. Experiments have been performed on facilities worldwide with materials of major interest for confined fusion reactors, like the forthcoming ITER. A process is under way to review the database in the fusion blanket neutronics section of SINBAD. The ranking process and the general principles for improving the analysis models are illustrated by the experiments on the OKTAVIAN, the FNG and the FNS facilities. The refinement of MCNP5 Monte Carlo computational models is aimed at reducing t...
The computer code Dragon is a free deterministic code developed by various organizations. It is a property of École Polytechnique de Montréal. Dragon contains a collection of various models which can describe the neutron transport in a... more
The computer code Dragon is a free deterministic code developed by various organizations. It is a property of École Polytechnique de Montréal. Dragon contains a collection of various models which can describe the neutron transport in a given geometry of a unit cell, reactor fuel assembly or in a reactor core. To obtain the final solution it is necessary to link together different modules at each step and any compromise at any level can lead to poor final results. For a nuclear engineer it is crucial to maintain the accuracy when reducing computational time. In the past the advanced self shielding models which were incorporated in the Dragon code Version4 were analysed. The conclusion obtained in that analysis was that the computational time of the burnup calculations was too long to be used for routine calculations. With the additional research and analysis presented in this paper satisfactory results were obtained that maintain the accuracy and reduce the computational time. In thi...
In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French... more
In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)
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SINBAD, the Shielding Integral Benchmark Archive Database, includes the experimental data (radiation shielding and dosimetry) and the computational models relative to integral benchmark experiments relevant for shielding applications. In... more
SINBAD, the Shielding Integral Benchmark Archive Database, includes the experimental data (radiation shielding and dosimetry) and the computational models relative to integral benchmark experiments relevant for shielding applications. In the SINBAD community a discussion has been initiated to establish the criteria for assessing the quality of the experiments. The reactor physics experiments Nesdip-2, Nesdip-3, Janus-1 and Janus-8 are here considered. An independent review of the experimental information was carried out with the aim to spot out which experimental data are incomplete, inconsistent or inaccurate. A note is finally released on each experiment, which provides an easy and quick interpretation for the SINBAD users.
This paper presents the validation and verification of the calculational model of TRIGA research reactor using Serpent 2 Monte Carlo code. Detailed geometrical model of TRIGA research reactor was developed to benchmark future burnup... more
This paper presents the validation and verification of the calculational model of TRIGA research reactor using Serpent 2 Monte Carlo code. Detailed geometrical model of TRIGA research reactor was developed to benchmark future burnup calculations. Results of flux and reaction rates calculations using Serpent 2 code were compared to the experimental results and to calculations performed with the Monte Carlo code MCNP. The calculated normalized reaction rates in the core are in very good agreement with the calculated MCNP and experimental results, therefore the computational model describes very well the neutron flux and reaction rate distribution in the reactor core while in the reflector the distribution of the flux in comparison with the measurements differ by maximum 3 percent.
The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases... more
The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases with burn-up due to consumption of burnable poison. 33 fuel elements with burn-up values between 3 % and 14 % were investigated. The experiments showed that variations in the initial fuel composition significantly influence the reactivity and, consequently, increase the inaccuracy of the burn-up measurements. Particularly important are variations in the initial concentration of erbium, which is used as burnable poison in FLIP fuel
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The large number of randomly distributed coated fuel particles (TRISO particles) inside the fuel pebbles in a high temperature reactor (HTR) makes their accurate modelling in MCNP very complicated and time-consuming. A simpler method is... more
The large number of randomly distributed coated fuel particles (TRISO particles) inside the fuel pebbles in a high temperature reactor (HTR) makes their accurate modelling in MCNP very complicated and time-consuming. A simpler method is to create an infinite cubic lattice of fuel particles and filling the pebble cells with said lattice. This technique introduces a systematic error in the fuel mass, due to the truncation of the fuel particles on the surface which defines the fuel region boundary in a pebble. This error, depending on the lattice pitch, can reach a few percent, and may thus introduce inaccuracies in the calculation of keff. The exact dependence of the fuel mass on the fuel kernel lattice pitch has been established for three lattice types. With these relationships the lattice pitch may be adjusted and the error in the fuel mass can be corrected.
The Deuteron Deuteron (D D) fusion reactions are foreseen for the next stage fusion reactors. The Frascati Neutron Generator (FNG), located near Rome, can produce neutrons of about 3 MeV by accelerating Deuterons onto solid targets... more
The Deuteron Deuteron (D D) fusion reactions are foreseen for the next stage fusion reactors. The Frascati Neutron Generator (FNG), located near Rome, can produce neutrons of about 3 MeV by accelerating Deuterons onto solid targets containing Deuterium. A computational model is developed for the simulation of the FNG neutron source. It requires the specification of the beam and target characteristics (e. g. Deuteron energy, Deuterium atomic fraction). The model is implemented in the subroutines of the MCNPX and MCNP5 codes, which need so far to be recompiled. The Deuterons are transported inside the solid target by a Monte Carlo method. The neutrons are generated with the angle energy distribution as defined by the laws and nuclear data for the Deuteron Deuteron reaction in the ENDF/B-VII library. The sensitivity studies on the input parameters of the D D model are presented. The modelling of experiments with the D D neutron source is feasible concerning the main features and associ...
On December 2011 the ENDF/B-VII.1 library has been released. The ENDF/B-VII.1 library includes a re-evaluation of the neutron interaction data for Manganese-55. Many application files presently include nuclear data for Mn-55 that rely on... more
On December 2011 the ENDF/B-VII.1 library has been released. The ENDF/B-VII.1 library includes a re-evaluation of the neutron interaction data for Manganese-55. Many application files presently include nuclear data for Mn-55 that rely on previous evaluations. This is the case of the ADS-2.0 nuclear data library, available from the International Atomic Energy Agency (IAEA) for the calculations of accelerator-driven systems (ADS) and based on the ENDF/B-VII.0 evaluated nuclear data library.
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Tentative design and implementation of an irradiation facility are described, which will add a 14 MeV component to the neutron spectrum in a thermal research reactor irradiation facility. The idea is to exploit the Li(n,T)He reaction in a... more
Tentative design and implementation of an irradiation facility are described, which will add a 14 MeV component to the neutron spectrum in a thermal research reactor irradiation facility. The idea is to exploit the Li(n,T)He reaction in a blanket of lithium in a deuterated environment, placed in a thermal reactor irradiation channel. Existing literature describing similar devices is reviewed, theoretical basis is presented and plans for the design and implementation are described. Potential uses of such a device are activation studies of new materials developed for fusion reactors like ITER and DEMO, evaluated nuclear reaction data validation at energies above 10 MeV, fast neutron activation analysis, etc.
Nuclear constants like the cross sections exhibit a strong energy-dependence, which may require several hundred-thousand points to describe accurately. Also, the database must be complete: parameters (e.g. the cross sections) must be... more
Nuclear constants like the cross sections exhibit a strong energy-dependence, which may require several hundred-thousand points to describe accurately. Also, the database must be complete: parameters (e.g. the cross sections) must be defined at all energies, even when no experimentally measured data are available. It is the task of the evaluator to assess the most probable value of a parameter at any energy, resolving issues of discrepant measurement, assigning values (by an educated guess or based on model calculations) where no data are available and providing data in computer-readable format.
In this modern age of powerful computers and availability of large computer clusters it is common to use a Monte Carlo method as a neutronic solver of reasonably large reactor systems, like research reactors. It is the only approach... more
In this modern age of powerful computers and availability of large computer clusters it is common to use a Monte Carlo method as a neutronic solver of reasonably large reactor systems, like research reactors. It is the only approach capable of giving detail insight of neutron transport phenomena in complex geometries. Due to its high demand of computer power and memory, the method has been mainly used in criticality calculations. One of the possible ways of using Monte Carlo method is for the generation of neutron homogenized multigroup cross sections, which are later used in deterministic codes to provide neutron solution on a coarse mesh. These types of applications are implemented in the Monte Carlo code SERPENT. One is based on a simple homogenization method with volume and flux weighting (FVH) of cross sections and the other is based on the B1 method. While the first method suffers in the presence of the strong absorbers the second method is applicable only for the cases with f...
To predict accurately the formation of plutonium within a reactor lattice fuel cell it is necessary to divide the fuel into rings, because the build-up of plutonium is affected by the spatial resonance self shielding in U, which results... more
To predict accurately the formation of plutonium within a reactor lattice fuel cell it is necessary to divide the fuel into rings, because the build-up of plutonium is affected by the spatial resonance self shielding in U, which results in the enhanced formation of plutonium near the surface of the fuel cell. The so called ‘rim effect’ has already been studied in the past. Our goal was to check the prediction of the same effect with the WIMS-D5 deterministic code employed in the CORD-2 system, developed at the Reactor Physics Department of the Jožef Stefan Institute, and used for core design calculations of pressurized water reactors. The calculations were performed on an array of 3×3 lattice cells. The results show an overestimated prediction of the quantity of plutonium (0.1 – 3.5%) with burnup, compared to the reference case obtained with the Serpent code, while the rim effect in the outer layer of the fuel cell is not reproduced at all. Since the WIMS-D5 code failed in predictin...
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Key reactions have been selected to compare JEFF-3.3 (CIELO 2) and IAEA CIELO (CIELO 1) evaluated nuclear data files for neutron induced reactions on 235U and 238U targets. IAEA CIELO evaluation uses reaction models to construct the... more
Key reactions have been selected to compare JEFF-3.3 (CIELO 2) and IAEA CIELO (CIELO 1) evaluated nuclear data files for neutron induced reactions on 235U and 238U targets. IAEA CIELO evaluation uses reaction models to construct the evaluation prior, but strongly relied on differential data including all reaction cross sections fitted within the IAEA Neutron Standards project. The JEFF-3.3 evaluation relied on a mix of differential and integral data with strong contribution from nuclear reaction modelling. Differences in evaluations are discussed; a better reproduction of differential data for the IAEA CIELO evaluation is shown for key reaction channels.
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Abstract The use of Monte Carlo transport method with the Serpent code for generating unit cell cross sections of a light-water reactor is investigated. The geometry is a 3 × 3 array of cells, where homogenization is performed over the... more
Abstract The use of Monte Carlo transport method with the Serpent code for generating unit cell cross sections of a light-water reactor is investigated. The geometry is a 3 × 3 array of cells, where homogenization is performed over the central cell, while the neighboring cells represent a kind of color-set scheme to model the radial leakage from the central cell. Instead of the default homogenization method of Serpent, the Effective Diffusion Homogenization method is applied externally, which conserves reaction rates, as well as the boundary partial currents of the central cell. The exercise serves to explore the potential of the Monte Carlo method for core design calculations and to validate and improve the existing computational scheme in which unit-cell calculations are based on the 1-D deterministic transport model in the WIMSD code. The cross sections by both methods are compared and applied to predict the hot-zero-power critical boron concentration and radial power distribution of the Krsko NPP in comparison with measured values. The results confirm applicability of Monte Carlo transport calculations with EDH homogenization at the unit-cell level and warrant further extension to burnup and whole-assembly Monte Carlo modeling, at least for validation purposes due to present computational time constraints.
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Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response... more
Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. These data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm3). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which were recently irradiated in the Jozef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia, and provides recommendations on how they can be overcome. The paper concludes with a discussion on the renormalization of the results from the second step calculations, to obtain accurate physical values.
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Reactivity coefficient method offers a way to correct macroscopic diffusion cross-sections when operating conditions change. The method is based on reactivity coefficients. Coefficients are not dependent on a reactor operating history.... more
Reactivity coefficient method offers a way to correct macroscopic diffusion cross-sections when operating conditions change. The method is based on reactivity coefficients. Coefficients are not dependent on a reactor operating history. They can be calculated in advance and stored in a library. The library needs to be updated only when drastic changes in the fuel design (different dimensions, new type of burnable poisons, etc.) are introduced. Two ways of the implementation are investigated in the paper. In the first one the nodal diffusion solver is extended with the power feedback option. This allows calculation of a reactor core at arbitrary conditions with only one basic set of macroscopic cross-sections. The second one is a possibility to generate a cycle specific macroscopic cross-section p rovider. The cross-section library is generated with the standard calculational procedures at three specific operating conditions (hot full power, hot zero power and cold zero power) and at ...
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ABSTRACT We generated a preliminary set of resonance parameters for 182,183,184,186W in the neutron energy range of thermal up to several keV. The evaluation methodology uses the Reich-Moore approximation to fit with the R-matrix code... more
ABSTRACT We generated a preliminary set of resonance parameters for 182,183,184,186W in the neutron energy range of thermal up to several keV. The evaluation methodology uses the Reich-Moore approximation to fit with the R-matrix code SAMMY, the high-resolution measurements performed in 2010 and 2012 at the Geel linear accelerator facility. For 183W, the transmission data and capture cross sections calculated with the set of resonance parameters are compared with the experimental values, and some of the average properties of the resonance parameters are discussed. In the analyzed energy range, this work almost doubles the existing resolved resonance evaluations in the ENDF/B-VII.1 library. A preliminary analysis of the performance of the calculated cross sections based on Lead slowing-down benchmarks is presented and briefly discussed.
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The EURACOS Na and Fe experiments have been re-evaluated in the scope of the OECD/NEA SINBAD project to determine if these benchmarks can be useful for the validation of the modern nuclear data evaluations. The measured activation rates... more
The EURACOS Na and Fe experiments have been re-evaluated in the scope of the OECD/NEA SINBAD project to determine if these benchmarks can be useful for the validation of the modern nuclear data evaluations. The measured activation rates of different foils as well as unfolded proton recoil spectra at different depths in sodium and iron have been compared to the Monte Carlo calculated parameters. The results show good agreement with the experiment only for the high-threshold 32 S(n,p) reaction detectors. The comparison of the lower threshold 103 Rh(n,n’) and 115 In(n,n’) reactions reveals large discrepancy between the experimental and calculated spatial distributions for these reaction rates. On the other hand, the unfolded neutron spectra are in relatively good agreement with the calculated ones. The overal results show that the EURACOS Na and Fe experiments cannot be considered of benchmark quality mainly due to unsatisfactory information on the neutron source. Nevertheless, some ac...
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ABSTRACT The evaluation procedures for 55Mn are elaborated, focusing on the issues related to the fluctuations in the cross sections above the resolved resonance range. Smooth cross sections are defined in the unresolved resonance range,... more
ABSTRACT The evaluation procedures for 55Mn are elaborated, focusing on the issues related to the fluctuations in the cross sections above the resolved resonance range. Smooth cross sections are defined in the unresolved resonance range, based on the resolution-broadened total cross section measurements, where relevant. Above this energy fluctuations in the measured total cross section are introduced by scaling all reaction cross sections, but preserving the resolution-broadened total cross section. Special procedures are designed to match the observed structure in the average cosine of scattering by adjusting the ratio of the shape-elastic and compound-elastic contributions to the elastic scattering cross sections. The evaluated data file is being assembled and subjected to rigorous testing, verification and validation.
The precise value of the thermal capture cross section of ²³â¸U is uncertain, and evaluated cross sections from various sources differ by more than their assigned uncertainties. A number of the original publications have been reviewed... more
The precise value of the thermal capture cross section of ²³â¸U is uncertain, and evaluated cross sections from various sources differ by more than their assigned uncertainties. A number of the original publications have been reviewed to assess the discrepant data, corrections were made for more recent standard cross sections and other constants, and one new measurement was analyzed. Because of the strong correlations in activation measurements, the gamma-ray emission probabilities from the β⻠decay of ²³â¹Np were also analyzed. As a result of the analysis, a value of 2.683 {+-} 0.012 b was derived for the thermal capture cross section of ²³â¸U. A new evaluation of the gamma-ray emission probabilities from ²³â¹Np decay was also undertaken.
ABSTRACT Prompt fission neutron spectra (PFNS) have proven to have a significant effect on criticality of selected benchmarks, in some cases as important as cross-sections. Therefore, a precise determination of uncertainties in PFNS is... more
ABSTRACT Prompt fission neutron spectra (PFNS) have proven to have a significant effect on criticality of selected benchmarks, in some cases as important as cross-sections. Therefore, a precise determination of uncertainties in PFNS is desired. Existing PFNS evaluations in nuclear data libraries relied so far almost exclusively on the Los Alamos model. However, deviations of evaluated data from available experiments have been noticed at both low and high neutron emission energies. New experimental measurements of PFNS have been recently published, thus demanding new evaluations. The present work describes the effort of integrating Kalman and EMPIRE codes in such a way to allow for parameter fitting of PFNS models. The first results are shown for the major actinides for two different PFNS models (Kornilov and Los Alamos). This represents the first step towards reevaluation of both cross-section and fission spectra data considering both microscopic and integral experimental data for major actinides.
ABSTRACT The aim of the current bilateral project between CEA Cadarache and JSI is to improve the accuracy of the online thermal power monitoring at the JSI TRIGA reactor. Simultaneously, a new wide range multichannel acquisition system... more
ABSTRACT The aim of the current bilateral project between CEA Cadarache and JSI is to improve the accuracy of the online thermal power monitoring at the JSI TRIGA reactor. Simultaneously, a new wide range multichannel acquisition system for fission chambers, recently developed by CEA, is tested. In the paper, calculational and experimental power calibration methods are described. The focus is on use of multiple detectors in combination with pre-calculated and pre-measured control-rod-position-dependent correction factors to improve the reactor power reading. The system will be implemented and tested at the JSI TRIGA reactor in 2014.
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The International Atomic Energy Agency has initiated a Coordinated Research Project (CRP) for the development of nuclear techniques for landmine detection. Out of the fourteen institutes participating in the CRP, twelve are working on... more
The International Atomic Energy Agency has initiated a Coordinated Research Project (CRP) for the development of nuclear techniques for landmine detection. Out of the fourteen institutes participating in the CRP, twelve are working on neutron-based techniques. Small isotope neutron sources and D-T neutron generators have been used by the researchers. The techniques used include neutron scattering by explosives as well
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A consistent, effective diffusion homogenization method for cross sections for pressurized water reactors (PWRs) is presented. It can be applied to obtain cell-averaged as well as assembly-averaged cross sections. Since no additional... more
A consistent, effective diffusion homogenization method for cross sections for pressurized water reactors (PWRs) is presented. It can be applied to obtain cell-averaged as well as assembly-averaged cross sections. Since no additional parameters are necessary, standard diffusion codes can be used. It is shown that a few-group diffusion calculation over a fuel assembly compares favorably with a transport calculation. This
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ABSTRACT This work was carried out in view of the possible use of diamond detectors as high resolution neutron spectrometers for the ITER project. An MCNP5(X) based computational tool has been developed to simulate the fast neutron... more
ABSTRACT This work was carried out in view of the possible use of diamond detectors as high resolution neutron spectrometers for the ITER project. An MCNP5(X) based computational tool has been developed to simulate the fast neutron response of diamond detectors. The source neutrons are generated by a source routine, developed earlier, that includes deuteron beam energy loss, angular straggling, and two-body relativistic kinematics. The diamond detector routine calculates a pulse height spectrum that is built up by elastic and inelastic scattering, (n,a), (n,p), and (n,d) reaction channels. A combination of nuclear data from ENDF/B-VII.0, TENDL-2010, and ENSDF is used. The simulated spectra are compared with measured spectra. It is shown that the simulation tool allows an interpretation of most of the characteristic features in the spectrum. This is an important step towards the use of diamond detectors for spectral analysis and fluence measurements. © 2001 Elsevier Science. All rights reserved.
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Fast quasi-monoenergetic neutrons can be produced by accelerating charged deuterons on tritium solid targets. Benchmark experiments were performed in many laboratories with intense D-T neutron sources. The aim is to validate the... more
Fast quasi-monoenergetic neutrons can be produced by accelerating charged deuterons on tritium solid targets. Benchmark experiments were performed in many laboratories with intense D-T neutron sources. The aim is to validate the computational models and nuclear data for fusion applications. The detailed information on the neutron source term is highly important for the benchmark analyses. At present, the MCNP family of codes cannot explicitly model the D-T reaction for Deuterons in the KeV energy range. The physics for the D-T neutron production was modelled at ENEA (Italy) in the SOURCE and SRCDX subroutines to compile with the MCNP source code. Some improvements to the original subroutines were introduced. The differential cross-sections for the D-T reaction from the ENDF/B-VII library were built into the code. The relativistic approach was implemented for neutron kinematics. The new D-T neutron source model was applied to the MCNP5 simulation of the tungsten integral experiment p...
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ABSTRACT Issues in evaluation methodology of the prompt fission neutron spectra (PFNS) and neutron multiplicity for the thermal-neutron-induced fission of the 235U are discussed. The inconsistency between the experimental differential and... more
ABSTRACT Issues in evaluation methodology of the prompt fission neutron spectra (PFNS) and neutron multiplicity for the thermal-neutron-induced fission of the 235U are discussed. The inconsistency between the experimental differential and integral data is addressed. By using differential data as ”shape data” good consistency was achieved between available sets of differential data. Integral dosimetry data have been used to define the PFNS slope at high outgoing neutron energies, where the quality of the differential data is poor. The inclusion into the fit of measured integral (spectrum-averaged) cross sections had a very small impact in the region where differential PFNS data are abundant and accurate, but removed the discrepancy with integral data at higher neutron emission energies. All experimental data are consistently fitted giving a PFNS average energy of 2.008 MeV. The impact on criticality prediction of the newly evaluated PFNS was tested. The highly enriched 235U solution assemblies with high leakage HEU-SOL-THERM-001 and HEU-SOL-THERM-009 benchmarks are the most sensitive to the PFNS. Criticality calculations for those solutions show a significant increase in reactivity if the average neutron energy of the fission neutrons is reduced from the ENDF/B-VI.5 value of 2.03 MeV. The proposed reduction of the PFNS average energy by 1.1% can be compensated by reducing the average number of neutrons per fission at the thermal energy to the Gwin et al. measured value. The simple least-squares PFNS fit was confirmed by a more sophisticated combined fit of differential PFNS data for 233,235U, 239Pu and 252Cf nuclides with the generalised least-squares method using the GMA and GANDR codes.
ABSTRACT Deuteron–deuteron (D–D) fusion reactions are foreseen for the next stage fusion reactors. A computational model is developed for simulating the neutron production by accelerating deuterons with energy less then 10MeV onto solid... more
ABSTRACT Deuteron–deuteron (D–D) fusion reactions are foreseen for the next stage fusion reactors. A computational model is developed for simulating the neutron production by accelerating deuterons with energy less then 10MeV onto solid targets containing deuterium. It requires the specification of the beam and target characteristics (e.g. deuterons energy, deuterium atomic fraction). The model is implemented in the subroutines of the MCNPX and MCNP5 codes, which need so far to be recompiled. The deuterons are transported inside the solid target by a Monte Carlo method. The neutrons are generated with the angle – energy distribution as defined by the laws and nuclear data for the deuteron–deuteron reaction in the ENDF/B-VII.0 library. The sensitivity studies on the input parameters of the D–D model are presented. The D–D source model is finally validated by an experiment, which has been performed by the FNG team at the IRMM with a high energy resolution detector. The results of the simulations indicate that the source model may be useful for the evaluation of the D–D neutron source term and associated uncertainties in experimental facilities.
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Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in (252)Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and... more
Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in (252)Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and by L. Green. The experiments were simulated with MCNP5 using cross section data from the ENDF/B-VII.0 library. The determination of neutron backscattering was calculated with the use of neutron flagging. Correction factors to experimentally measured values were determined to obtain average cross sections in a pure (252)Cf spontaneous fission spectrum. Influence of concrete wall thickness, air moisture and room size on the average cross section was analyzed. Correction factors amounted to about 30%. Corrected values corresponding to average cross sections in a pure (252)Cf spectrum were calculated for (197)Au, (232)Th, (181)Ta, (98)Mo, (65)Cu and (84)Sr. Average cross sections were also calculated with the RR_UNC software using IRDFF-v.1.05 and ENDF...
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The development and production of various types of nuclear data represents an important activity of the IAEA Nuclear Data Section, as demanded for a wide range of energy-based applications. Recent worldwide initiatives to study innovative... more
The development and production of various types of nuclear data represents an important activity of the IAEA Nuclear Data Section, as demanded for a wide range of energy-based applications. Recent worldwide initiatives to study innovative designs of nuclear power systems require extensive and accurate nuclear data for materials to which little attention has been devoted in the past. Four highly
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The objective of SINBAD (Shielding Integral Benchmark Archive Database) is to archive and make available the information on high–quality benchmark experiments for validating radiation transport codes and nuclear data. The quality... more
The objective of SINBAD (Shielding Integral Benchmark Archive Database) is to archive and make available the information on high–quality benchmark experiments for validating radiation transport codes and nuclear data. The quality assessment of the information available therein is based on the following issues: • completeness of the experimental specifications; • accuracy of the computational models; • utilization of recent nuclear data evaluations; • availability of sensitivity and uncertainty analyses. Experiments have been performed on facilities worldwide with materials of major interest for confined fusion reactors, like the forthcoming ITER. A process is under way to review the database in the fusion blanket neutronics section of SINBAD. The ranking process and the general principles for improving the analysis models are illustrated by the experiments on the OKTAVIAN, the FNG and the FNS facilities. The refinement of MCNP5 Monte Carlo computational models is aimed at reducing t...
Multinational efforts have ensured the development and maintenance of a number of important databases to cover the worldwide demand for access to nuclear reaction cross sections and nuclear structure and decay data. Compilations for EXFOR... more
Multinational efforts have ensured the development and maintenance of a number of important databases to cover the worldwide demand for access to nuclear reaction cross sections and nuclear structure and decay data. Compilations for EXFOR (EXchange FORmat (for nuclear reaction cross sections)) and evaluations for ENSDF (Evaluated Nuclear Structure Data File) were placed on international footings in 1970 and 1974, respectively. Both have been maintained ever since through the combined efforts of a number of individuals, research groups and national/international data centres brought together collectively under the auspices of the International Atomic Energy Agency (IAEA). ENDF (Evaluated Nuclear Data File) databases are normally sponsored on a national basis, and their format (ENDF-6) and contents constitute the backbone of the necessary input data for a wide range of nuclear applications. The most recent studies and activities to improve the contents and user-friendliness of the thr...
The TRIGA Mark II reactor at the Jožef Stefan Institute in Ljubljana, Slovenia, features several in-core and ex-core irradiation facilities. The in-core facilities, i.e. several irradiation channels located in the reactor core and in the... more
The TRIGA Mark II reactor at the Jožef Stefan Institute in Ljubljana, Slovenia, features several in-core and ex-core irradiation facilities. The in-core facilities, i.e. several irradiation channels located in the reactor core and in the surrounding reflector, have already been thoroughly characterised, the ex-core facilities, however, have not been characterized in such detail yet. The facilities are as follows: • two channels, the radial beam port (RBP) and the radial piercing thruport (RPP) which extend from the outer and the inner boundary of the graphite reflector, which surrounds the reactor core, to the outer concrete reactor structure, respectively. • a third, tangential channel (TangCh) which pierces the graphite reflector and the reactor
ABSTRACT This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of... more
ABSTRACT This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of this method in the TRIGA Mark II of Slovenia led to the assessment of three energy groups neutron fluxes in central irradiation locations within reactor core.
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... 2. However, there is some contradiction be-Electronic address: Andrej.Trkov@ijs.si -15 -10 -5 0 5 10 15 0 10 20 30 40 ... Issue 12, December 2008, Pages 2905-2909 Special Issue on Workshop on Neutron Cross Section Covariances June... more
... 2. However, there is some contradiction be-Electronic address: Andrej.Trkov@ijs.si -15 -10 -5 0 5 10 15 0 10 20 30 40 ... Issue 12, December 2008, Pages 2905-2909 Special Issue on Workshop on Neutron Cross Section Covariances June 24-28, 2008, Port Jefferson, New York ...
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EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including... more
EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including heavy-ions) or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions (∽ keV) and goes up
Research Interests: Mathematical Physics, Physics, Theoretical Physics, Particle Physics, Graphic User Interface Design, and 13 moreData Collection, Nuclear Reactions, WKB approximation, Cross Section, Moment of Inertia, Nuclear Data, Physical Model, Energy Levels, Monte Carlo Method, Gamma radiation, Heavy Ions, Angular Momentum, and Modular System
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Experimental results of pulse parameters and control rod worth measurements at TRIGA Mark 2 reactor in Ljubljana are presented. The measurements were performed with a completely fresh, uniform, and compact core. Only standard fuel... more
Experimental results of pulse parameters and control rod worth measurements at TRIGA Mark 2 reactor in Ljubljana are presented. The measurements were performed with a completely fresh, uniform, and compact core. Only standard fuel elements with 12 wt% uranium were used. Special efforts were made to get reliable and accurate results at well-defined experimental conditions, and it is proposed to
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ABSTRACT Inherently positive parameters with large relative uncertainties (typically ≳30%≳30%) are often considered to be governed by the lognormal distribution. This assumption has the practical benefit of avoiding the possibility of... more
ABSTRACT Inherently positive parameters with large relative uncertainties (typically ≳30%≳30%) are often considered to be governed by the lognormal distribution. This assumption has the practical benefit of avoiding the possibility of sampling negative values in stochastic applications. Furthermore, it is typically assumed that the correlation coefficients for comparable multivariate normal and lognormal distributions are equivalent. However, this ideal situation is approached only in the linear approximation which happens to be applicable just for small uncertainties. This paper derives and discusses the proper transformation of correlation coefficients between both distributions for the most general case which is applicable for arbitrary uncertainties. It is seen that for lognormal distributions with large relative uncertainties strong anti-correlations (negative correlations) are mathematically forbidden. This is due to the asymmetry that is an inherent feature of these distributions. Some implications of these results for practical nuclear applications are discussed and they are illustrated with examples in this paper. Finally, modifications to the ENDF-6 format used for representing uncertainties in evaluated nuclear data libraries are suggested, as needed to deal with this issue.
The ENDF/B-VII. 1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII. 0. These advances... more
The ENDF/B-VII. 1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII. 0. These advances focus on neutron cross sections ...
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ABSTRACT The EURACOS experiment series (sodium, iron, and no shielding) has been re-evaluated in the scope of the OECD/NEA SINBAD project to determine how useful these benchmarks can be for the validation of the modern nuclear data... more
ABSTRACT The EURACOS experiment series (sodium, iron, and no shielding) has been re-evaluated in the scope of the OECD/NEA SINBAD project to determine how useful these benchmarks can be for the validation of the modern nuclear data evaluations. The measured activation rates of different foils as well as unfolded proton recoil spectra at different depths in sodium and iron have been compared to the Monte Carlo calculated parameters. The results suggest good agreement with the experiment only for the high-threshold 32S(n,p) reaction detectors. The comparison of the lower threshold 103Rh(n,n′) and 115In(n,n′) reactions, and the (n,γ) reaction on 197Au under cadmium cover, reveals large discrepancy between the experimental and calculated spatial distributions for these reaction rates which is expected to be due to the uncertainties in the source and background description rather than due to the cross section uncertainties. On the other hand, the unfolded neutron spectra shapes are in relatively good agreement with the calculated ones. The overall results show that the EURACOS Na and Fe experiments cannot be considered of benchmark quality mainly due to unsatisfactory information on the neutron source. Nevertheless, some activation measurements from the EURACOS sodium and iron experiments such as the high threshold reactions and the spectra measurements are potentially useful for modern nuclear data and computer code benchmarking.
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The prompt fission neutron spectra covariance matrices of 235U, 238U and 239Pu were evaluated using the Monte Carlo method for two empirical spectra formulations, Watt and Kornilov, and compared to the existing evaluations such as... more
The prompt fission neutron spectra covariance matrices of 235U, 238U and 239Pu were evaluated using the Monte Carlo method for two empirical spectra formulations, Watt and Kornilov, and compared to the existing evaluations such as JENDL-3.3. The fission spectra, together with the covariance matrices and the methods for the calculation of the corresponding sensitivity profiles were validated on several applications,
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ABSTRACT Covariance data in the existing evaluated nuclear data libraries often include large relative uncertainties and mathematical inconsistencies, which arise especially in combination with random sampling. The 232Th evaluation from... more
ABSTRACT Covariance data in the existing evaluated nuclear data libraries often include large relative uncertainties and mathematical inconsistencies, which arise especially in combination with random sampling. The 232Th evaluation from the ENDF/B-VII.1 library has been taken as an example. Possible solutions for mathematically impossible correlation matrices with negative eigenvalues and too low correlation coefficients between inherently positive parameters with large relative uncertainties are proposed. Convergence of the random sampling for lognormal distribution with extremely high relative standard deviations is slow by nature. Using weighted sampling, single parameters or a limited number of correlated parameters with large uncertainties can be sampled. Efficient sampling of a large number of correlated parameters with extremely large relative uncertainties remains unsolved.
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ABSTRACT Two relatively new approaches to neutron cross section data evaluation are described. They are known collectively as Unified Monte Carlo (versions UMC-G and UMC-B). Comparisons are made between these two methods, as well as with... more
ABSTRACT Two relatively new approaches to neutron cross section data evaluation are described. They are known collectively as Unified Monte Carlo (versions UMC-G and UMC-B). Comparisons are made between these two methods, as well as with the well-known generalized least-squares (GLSQ) technique, through the use of simple, hypothetical (toy) examples. These new Monte Carlo methods are based on stochastic sampling of probability functions that are constructed with the use of theoretical and experimental data by applying the principle of maximum entropy. No further assumptions are involved in either UMC-G or UMC-B. However, the GLSQ procedure requires the linearization of non-linear terms, such as those that occur when cross section ratio data are included in an evaluation. It is shown that these two stochastic techniques yield results that agree well with each other, and with the GLSQ method, when linear data are involved, or when the perturbations due to data discrepancies and nonlinearity effects are small. Otherwise, there can be noticeable differences. The present investigation also demonstrates, as observed in earlier work, that the least-squares approach breaks down when these conditions are not satisfied. This paper also presents an actual evaluation of the 55Mn(n,γ)56Mn neutron dosimetry reaction cross section in the energy range from 100 keV to 20 MeV, which was performed using both GLSQ and UMC-G approaches.
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A continuous 60-year record (19381998) of stable isotope compositions of carbon and oxygen, as well as trace metal (Mg, Sr, Ba) concentrations in a laminated calcite crust precipitated in a short artificial tunnel on a non-equilibrium... more
A continuous 60-year record (19381998) of stable isotope compositions of carbon and oxygen, as well as trace metal (Mg, Sr, Ba) concentrations in a laminated calcite crust precipitated in a short artificial tunnel on a non-equilibrium groundwater-fed karstic river is presented. ...
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Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the... more
Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities.
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ABSTRACT The precise value of the thermal capture cross section of238U is uncertain, and evaluated cross sections from various sourcesdiffer by more than their assigned uncertainties. A number of theoriginal publications have been... more
ABSTRACT The precise value of the thermal capture cross section of238U is uncertain, and evaluated cross sections from various sourcesdiffer by more than their assigned uncertainties. A number of theoriginal publications have been reviewed to assess the discrepant data,corrections were made for more recent standard cross sections andotherconstants, and one new measurement was analyzed. Due to the strongcorrelations in activation measurements, the gamma-ray emissionprobabilities from the beta decay of 239Np were also analyzed. As aresult of the analysis, a value of 2.683 +- 0.012 barns was derived forthe thermal capture cross section of 238U. A new evaluation of thegamma-ray emission probabilities from 239Np decay was alsoundertaken.
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The Energy Citations Database (ECD) provides access to historical and current research (1948 to the present) from the Department of Energy (DOE) and predecessor agencies.