Nuclear Data Improvement Research Articles

Development of Fast Decay-Energy Spectroscopy With Magnetic Microcalorimeters

Decay energy spectroscopy (DES) is an increasingly important radiometric technique arising from the unique thermal detection physics and the precision of low-temperature microcalorimetry. DES can enable high-precision analysis of very small amounts of radioactive material with simple and rapid sample preparation, making it a key new tool for nuclear safeguards, nuclear forensics, the improvement of nuclear data, measurements of absolute activity, and other applications. For problems in nuclear safeguards and nuclear forensics it is highly desirable to increase the count rates achievable in DES, while maintaining simple and rapid sample preparation. In this report we describe progress in a new project to achieve high count rates in DES by taking advantage of the physics of magnetic microcalorimetry. We describe the results of initial modeling and experiment to validate the idea, a sample fabrication technique suitable for nuclear materials laboratories, the preparation and initial tests of paramagnetic Au-Er alloy in thin foil form, and the development of a sensing coil geometry that will perform well with the increased-thickness sensor used in the new technique.

Uncertainty quantification and target accuracy assessment of nuclear data to effective neutron multiplication factor of heat pipe cooled reactor

With the increasing improvement of the core physics calculation method, the uncertainty introduced by the calculation software and model approximation in the process of nuclear reactor core physics design is becoming smaller and smaller, while the uncertainty of the calculation results caused by the uncertainty of the input parameters itself is becoming more and more important. As an important input parameter of core physics calculation, the uncertainty of nuclear data will be transmitted, which will affect the accuracy of core physics calculation results. Quantifying the uncertainty of the core physics calculation results caused by nuclear data is an important element in the nuclear reactor safety design process. On the other hand, according to the target accuracy requirements of the nuclear reactor design, the key nuclear data and the priority of the improvement when the physical parameters of the nuclear reactor core reach the target accuracy requirements are evaluated, which can provide reference standards for the improvement of nuclear data and guide the development of specific experimental measurements work.

ARIEL & SANDA nuclear data activities

Nuclear data are fundamental quantities for developing nuclear energy concepts and research. They are essential for the simulation of nuclear systems, safety and performance calculations, and reactor instrumentation. Nuclear data improvement requires a combination of many different know-hows that are distributed over many institutions along Europe. In the EURATOM call for Nuclear Fission and Radiation Protection NFRP-2018, two nuclear data projects were started in September 2019: the Coordination and Support Action ARIEL (Accelerator and Research reactor Infrastructures for Education and Learning) and the Research and Innovation Action SANDA (Solving Challenges in Nuclear Data for the Safety of European Nuclear facilities). The ARIEL project integrates education and training of young scientists and technicians with access to neutron beam research infrastructures and supports scientific visits to conduct short-term research projects relevant to thesis works. The SANDA project is focuses on research innovation actions, including detector and nuclear target development, important nuclear data measurements, nuclear data evaluation, and validation. A description of these ongoing projects, including the first results, is the subject of this article.

JADE, a new software tool for nuclear fusion data libraries verification & validation

• Automatic verification and validation of nuclear data libraries. • Benchmarks for cross sections verification. • Fusion related nuclear data libraries verification and validation. • Spotting inconsistencies and errors in cross sections. Nuclear data improvement is a key objective in the nuclear field and, in particular, for fusion application. Due to their complex production process, new nuclear library releases must undergo systematic and extensive Verification and Validation (V&V) procedures. A novel software tool, JADE, is presented in this work. Its aim is to lead the path to a full standardization and automatization of V&V processes for neutron-gamma Monte Carlo transport for fusion applications, as well as saving substantial user time and avoid error-prone editing operations. JADE is a Python 3 based software that will be able to automatically run and post-process an exhaustive suite of computational and experimental benchmarks, together with the incorporation of pre-existing quality checks strategies. A first computational benchmark, the Leakage Sphere benchmark, has been implemented, a series of nuclear libraries were tested with it and the results have been discussed. The capability of JADE for spotting anomalies in the cross sections has been conclusively proven by being able to highlight inconsistencies already documented in the related literature. Moreover, some previously unknown anomalies have been identified and discussed, giving evidence of JADE’s applicability for the validation of future nuclear libraries releases. Finally, future developments and objectives are discussed.

Use of shielding integral benchmark archive and database for nucleardata validation

Shielding benchmarks were extensively used for the validation and improvement of nuclear data since many years. Recent evaluations however mostly rely on the validation using critical benchmarks, which can introduce biases and compensation effects. A new Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPEC SG47) entitled "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation" was started in June 2019 with the objective to promote further development and use of Shielding Integral Benchmark Archive and Database (SINBAD) and thus contribute to the diversification of the nuclear data validation practice by including more extensively other types of integral measurements, such as shielding benchmarks, in the validation and evaluation procedure. Use of shielding benchmarks is expected to provide a wider-scope test of the performance of the evaluated nuclear data and would ultimately contribute to a production of more general-purpose cross-section evaluations.

HPRL – International cooperation to identify and monitor priority nuclear data needs for nuclear applications

The OECD-NEA High Priority Request List (HPRL) is a point of reference to guide and stimulate the improvement of nuclear data for nuclear energy and other applications, and a tool to bridge the gap between data users and producers. The HPRL is application-driven and the requests are submitted by nuclear data users or representatives of the user’s communities. A panel of international experts reviews and monitors the requests in the framework of an Expert Group mandated by the NEA Nuclear Science Committee Working Party on International Nuclear Data Evaluation Cooperation (WPEC). After approval, individual requests are classified to three categories: high priority requests, general requests, and special purpose requests (e.g., dosimetry, standards). The HPRL is hosted by the NEA in the form of a relational database publicly available on the web. This paper provides an overview of HPRL entries, status and outlook. Examples of requests successfully completed are given and new requests are described with emphasis on updated nuclear data needs in the fields of nuclear energy, neutron standards and dosimetry.

Nuclear data research supported by EURATOM: CHANDA, ERINDA and EUFRAT

Nuclear data and associated tools are critical elements of the nuclear energy industry and research, playing an essential role in the simulation of nuclear systems, safety and performance calculations and interpretation of the reactor instrumentation. Nuclear data improvement requires a combination of much different know-hows that are distributed over many small- and medium-sized institutions along Europe. The Euratom programs have facilitated the setup of pan European collaborations getting together the required experience inside the projects CHANDA, ERINDA and the JRC action EUFRAT. The paper describes the holistic and inclusive approach of these projects that have also worked together to coordinate the European nuclear data research capabilities to improve the facilities, detectors, models and evaluation, validation and simulation tools. It also shows examples of success histories and summary of results of these projects and of their impact on the EU nuclear safety and industry, together with an outlook to the future.

Validation of actinide nuclear data based on reactivity worth experiments in a MOX-LWR spectrum

OSMOSE was a CEA/DOE collaborative experiment dedicated to the improvement of actinide nuclear data. Based on reactivity worth measurements of pure isotopic samples of various actinides from 232Th to 245Cm, it represents a unique set experimental data to validate evaluated nuclear data.This paper presents the analysis of the experiments conducted in a MOXLWR-type core configuration, with TRIPOLI-4® Monte-Carlo calculations. The comparison between calculated and measured reactivity worth shows a good agreement with JEFF-3.1.1 for all the actinides, except 241Am. The combination of the C/E results with the sensitivity coefficients and prior covariances in a re-estimation process suggests that some improvement is required on the 241Am (n,g), towards an increase of +6% of the JEFF-3.1.1 thermal capture cross section. Based on the same reestimation process, improvements in the nuclear data covariances are pointed out, with posterior uncertainties on the reactivity worth significantly reduced for 238Pu, 239Pu, 241Am, 237Np and 244+245Cm.

An Experimental Programme optimized with Uncertainty Propagation: PETALE in the CROCUS Reactor

The PETALE experimental programme in the CROCUS reactor at EPFL intends to contribute to the validation and improvement of neutron nuclear data in the MeV energy range for stainless steel, particularly in the prospect of heavy reflector elements of pressurized water reactors. It mainly consists of several transmission experiments: first, through metallic sheets of nuclear-grade stainless steel interleaved with dosimeter foils, and, successively, through its elemental components of interest – iron, nickel, and chromium. The present article describes the study for optimizing the response of the dosimetry experiments to the nuclear data of interest.

FNG copper benchmark evaluation for the SINBAD database

The FNG-Copper benchmark mock-up is among the first SINBAD (Shielding Integral Benchmark Archive and Database) evaluations being prepared after a long pause. The neutronics benchmark experiment on a pure 60 × 60 × 69.946 cm3 Copper assembly was performed in the frame of the European Fusion Programme (Fusion for Energy -F4E) at the 14-MeV Frascati neutron generator (FNG) of ENEA Frascati with the objective to validate the copper nuclear cross-section data relevant for ITER design calculations, including the related uncertainties. A detailed evaluation of the experimental configuration, measurement system and results was performed for the SINBAD database with a particular focus on a realistic, complete and consistent estimation of uncertainties involved in the measurements and the calculations. Several input models for Monte Carlo (MCNP-5, -6) and deterministic (DORT) transport and sensitivity/uncertainty (SUSD3D) codes are provided using both very detailed and simplified geometry models and different neutron source modelling. To facilitate the use of the experimental data for nuclear data validation and improvement, the SINBAD compilation includes the energy dependent sensitivity profiles of the detector reaction rates with respect to the underlying cross-section data. The evaluation is to serve as an example to promote further SINBAD evaluations with the message that the benchmark experiment is not completed when the measurements are done, but only when all the detailed analysis is performed including the uncertainty assessment, and the complete information is preserved and stored for future use by the international community.

Innovative Hybrid Pile Oscillator Technique in the Minerve Reactor: Open Loop Versus Closed Loop

Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiment has long been implemented in the Minerve experimental reactor, operated by CEA Cadarache, Saint-Paul-les-Durance, France. A hybrid technique mixing reactivity worth and flux variations surrounding the oscillator is presented here. It was made possible after developing high-sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the separation of mixed neutron effects, such as fission and capture or scattering and capture. The experimental results are presented and discussed in this paper, which focus on comparing two measurement setups, one using a power control system (closed-loop technique) and another one in which reactor power is free to drift (open-loop technique). First, it is demonstrated that the open loop is equivalent to the closed loop, and both techniques bring similar results. It is shown that measuring the flux change around oscillated samples provides valuable information on partial neutron cross sections. Second, uncertainty management and methods’ reproducibility are discussed. As expected, both techniques can achieve similar performances. In this paper, uncertainty as low as 0.007 pcm for samples’ reactivity worth and $1.2 \times 10^{-4}$ for local flux change were achieved.

Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop) and another one where the power is free to drift (open loop). First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

Beta-effective sensitivity and uncertainty analysis of MYRRHA reactor for possible use in nuclear data validation and improvement

New reactor concept studies, such as those of the MYRRHA Accelerator Driven System (ADS), with their extended nuclear data needs relaunched the interest in sensitivity and uncertainty (S/U) studies. S/U analysis play an essential role in assuring the accuracy and reliability of nuclear reactor computations and provide an insight in the physical phenomena involved in the neutron transport. This paper presents the study of the MYRRHA critical core, which is designed to operate at 100 MW thermal power. The previously presented analysis of the MYRRHA effective multiplication factor (keff) was extended to study the sensitivity and uncertainty in kinetic parameters, such as the effective delayed neutron fraction (βeff). The sensitivity and uncertainty computations were performed by means of the SUSD3D code based on forward and adjoint-flux first-order perturbation theory. The sensitivity coefficients of βeff with respect to the basic nuclear data were calculated by the k-ratio derivation method proposed in 2011. The main components of the uncertainties in βeff were estimated using the JENDL-4.0u covariance data. The energy dependent sensitivity profiles of keff and βeff with respect to nuclear data are compared in view of exploiting the differences among them for a physically more complete, comprehensive and consistent nuclear data validation and improvement procedure. The accuracy of the keff and βeff measurements and calculations has to be carefully evaluated in order to conclude on the nuclear data status and possible improvements.

Fast-thermal coupled cores in zero power reactors: A demonstration of feasibility and pertinence for the ZEPHYR project

The development of GEN-IV fast reactor technology at an industrial scale will need significant improvement of nuclear data and related uncertainties to cope with awaited target uncertainties (Palmiotti and Salvatores, 2011). However, GEN-IV dedicated Zero Power Facilities are now sparse. In France, both the MASURCA Zero Power Reactor (ZPR) and the future ZEPHYR multipurpose ZPR to be built in Cadarache belong to this type of facilities. For this latest one, fast/thermal coupled configurations performed in MINERVE during the 1970’s have been recently revisited using modern calculation tools. The principle of a fast zone fed by a thermal booster, known from decades as the concept of coupled cores regained interest in recent years. This paper describes the concept of fast/thermal coupled core, a deeper description of the physical characteristics of the adaptation zone, and its use in the improvement of fast-spectrum nuclear data thanks to integral experiments. An optimized configuration is then presented. The preliminary design phase is performed using deterministic tools, a posteriori validated against a full Monte Carlo calculation. Particular attention is paid on the impact of the buffer zone on the spectral characteristics in the oscillation channel, as well as the sample mass characteristics required to ensure a proper signal analysis. Moreover, preliminary studies dedicated to the experimental realization of such a configuration are described, including the reactivity stability in case of an accidental flooding of the fast central zone.

Fission yield covariance matrices for the main neutron-induced fissioning systems contained in the JEFF-3.1.1 library

Nuclear data improvement, validation and completion are one of the major concerns for ensuring safety-by-design standards and economic optimization for modern nuclear facilities. Uncertainty and sensitivity analyses must be carried out to estimate the nuclear data influence on integral reactor parameters, highlighting margins of improvement needed to get full benefit from modern advanced modeling. The quality of any uncertainty quantification analysis strongly depends on the covariance matrices made available in the evaluated nuclear data libraries. In modern nuclear data libraries (e.g. JEFF, ENDF/B or JENDL) no correlations for fission yields (FY) are provided, so in the present work we propose a covariance generation methodology which yielded reasonable results for application-relevant thermal and fast neutron-induced fission product yields. The main goal was not to perform a new evaluation but to reproduce coherently the JEFF-3.1.1 fission yield library and add covariance information. We did so using the Generalized Least Square Method (GLSM) and the Marginalization techniques implemented in the CONRAD code, developed at CEA-Cadarache. Results on the thermal neutron-induced fission of 235U,239Pu,241Pu and on the fast-neutron induced fission of 238U,239Pu and 240Pu are presented and tested on elementary fission decay heat uncertainty quantifications.

A hybrid pile oscillator experiment in the Minerve reactor

Pile oscillator techniques are powerful methods to measure small reactivity worth of material of interest for nuclear data improvement. In the Minerve reactor, operated by CEA Cadarache, experiments are being held using a reactivity oscillator method based on the measurement of the global reactivity perturbation associated to samples motion in the core center. The technique has recently been extended by introducing new miniature fission chambers close to the oscillation location, in order to measure local neutron flux perturbations. New detectors and associated acquisition systems were developed for that purpose by CEA. First results are presented and discussed in this paper. Data processing techniques are presented in details which demonstrate that global and local perturbations can be unraveled. Experiments also demonstrate that measuring the local flux depression around oscillated samples indeed provide valuable information regarding partial cross sections. The preliminary results are found to be very sensitive to capture cross section at the expense of scattering cross section, making the technique very useful to measure small capture effects of highly scattering samples and improve nuclear data knowledge, as foreseen in the future CEA’s ZEPHYR critical facility to be built in Cadarache around 2026.

Research and development for accuracy improvement of neutron nuclear data on minor actinides

To improve accuracy of neutron nuclear data on minor actinides, a Japanese nuclear data project entitled “Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC)” has been implemented. Several independent measurement techniques were developed for improving measurement precision at J-PARC/MLF/ANNRI and KURRI/LINAC facilities. Effectiveness of combining the independent techniques has been demonstrated for identifying bias effects and improving accuracy, especially in characterization of samples used for nuclear data measurements. Capture cross sections and/or total cross sections have been measured for Am-241, Am-243, Np-237, Tc-99, Gd-155, and Gd-157. Systematic nuclear data evaluation has also been performed by taking into account the identified bias effect. Highlights of the AIMAC project are outlined.

High precision analysis of isotopic composition for samples used for nuclear cross-section measurements

For the accuracy improvement of nuclear data of minor actinides and long-lived fission products in the project of “Research and development for Accuracy Improvement of neutron nuclear data on Minor Actinides”, the isotopic compositions of two Am samples (241 Am sample and 243 Am sample) were analyzed by thermal ionization mass spectrometry. Only the peak of 241 Am was observed in the analysis of 241 Am sample, and level of isotopic impurities were improved from 0.1% to 0.0004%. In the analysis of 243 Am sample, the peak of unreported isotope of 242m Am was observed in addition to the peaks of 243 Am and 241 Am. The mass spectrometry also showed that 243 Am sample has other unreported impurities such as 239 Pu and 240 Pu.

Combined use of k-effective and beta-effective measurements for nuclear data validation and improvement

Nuclear cross-sections are being re-evaluated time and again since decades, and in the recent years they are “tuned” to match better in particular the k eff measurements. However, the k eff parameter as a very global parameter provides simply too many possible combinations and variations, all more or less in agreement with the differential cross section data measurements and the associated uncertainties. Adding other types of integral measurements, such as shielding and kinetics benchmarks, on the cross-section evaluation menu is expected to provide a complementary and a more complete view on the problem. This paper discusses the advantages of exploiting the effective delayed neutron fraction (β eff ) measurements for nuclear data (ND) evaluation and testing. The cross-section sensitivity and uncertainty analysis revealed some crucial differences in the k eff and β eff sensitivity profiles which are favorable for ND validation.

Analysis of criticality benchmark experiments with beryllium reflectors

Beryllium is an important nuclear material, and the reliability of the data for neutron-induced nuclear reactions of beryllium is of significant importance for nuclear engineering. The evaluated nuclear data for beryllium have been improving from ENDF/B-VI to ENDF/B-VⅡ.0 and then to ENDF/B-VⅡ.1. The comparisons between the calculated and experimental results of the criticality benchmark experiments are the essential means to test the reliability of nuclear data and indicate the direction of the improvement of nuclear data. There are several series of criticality benchmark experiments with beryllium reflector available for testing beryllium nuclear data. However, the calculated results are not consistent across these benchmarks. Two series of these benchmarks that are similar to each other, namely HMF058 and HMF066, are selected for discussion. HMF058 and HMF066 are both highly enriched metal fast benchmarks, with five cases of experiments in HMF058 benchmark and nine in HMF066. With ENDF/B-VⅡ.1 cross sections, a clearly increasing C/E keff bias is observed with increasing beryllium reflector thickness for the five cases in HMF058 benchmark, while using ENDF/B-VⅡ.0 cross sections, all the C/E values for keff remain within the experimental uncertainty. However, HMF066 are calculated very well with ENDF/B-VⅡ.1 cross sections, but a bias of about 500 pcm is observed with ENDF/B-VⅡ.0 data. These results are particularly puzzling since there is little difference between the configurations of HMF058 and HMF066, so the quality of beryllium nuclear data cannot be evaluated and the direction for improvement cannot be figured out either. The similarity method, based on the use of sensitivity coefficients calculated by sensitivity and uncertainty code SURE, is used to analyze the similarity between two series of benchmark experiments. First, the neutronics similarity index between each pair of the total of fourteen cases of experiments from the two benchmarks is calculated. Then, the most similar experiments from HMF066 to each case of the five experiments from HMF058 are selected by similarity index, and the experiments are grouped into five similarity suites, each with one from HMF058 and the others from HMF066. The experiments in the same similarity suite are highly similar to each other on neutronics. In a similarity suite, the deviations of calculated results and experimental values are disagreed for experiments from different series, but the deviations agree with each other for experiments from the same series. This shows that the agreement between the calculated results and experimental values cannot be improved by revising the nuclear data. It is necessary to carry out the detailed reevaluation of the benchmark experiments, or to develop reliable new integral experiments to exclude unreliable experiments, in order to avoid the misleading of the nuclear data testing.