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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)