Nuclear Data Research Articles

The impact of fast spectrum graphite moderated systems on thermal spectrum graphite moderated reactors through the resonance physics of n+12C

Understanding the nuclear data used in predictive modeling and simulation of advanced reactors is vital for their development, licensing, and deployment. In this paper, we use the Kairos gFHR and the X-Energy Xe-100 as demonstration examples to test the scientific hypothesis that the resonance physics of n+12C allows for integral experiments with fast-neutron-spectra to be useful in helping to reduce nuclear data uncertainty on k-effective of new reactor systems with thermal-neutron spectra. This work hypothesizes that legacy criticality experiments hold value for these new reactors. HMF071 (highly enriched, metal-fuel, fast-spectrum critical experiment) (Parkey et al., 2014; Nuclear Science Committee of The Nuclear Energy Agency, 2016) is a series of legacy criticality benchmarks with graphite reflectors which were selected to test this scientific hypothesis by leveraging the large correlation visually identified in the covariance matrix of n+12C. This hypothesis will be tested through calculating the similarity and uncertainty-reduction provided from evaluating the effect of a fast-neutron-spectrum system, HFM071, on a thermal-neutron spectrum system. The sensitivity with respect to elastic scattering of carbon had greater than 70% similarity for one of the cases of HMF071 compared against each of the advanced reactor designs considered. This result quantifies the relevance that the fast spectrum experiments can have for thermal spectrum applications due to the resonance physics of 12C. With this connection quantified, the uncertainty reduction from the fast-spectrum experiments on these thermal-spectrum reactors was calculated using the SCALE module TSURFER (Rearden and Jessee, 2018; Wieselquist and Lefebvre, 2023). The HMF071 series of critical experiments were able to lower the nuclear data uncertainty associated with carbon on the Xe-100 and gFHR only by about 6 pcm using SCALE 6.2.4 and the ENDF/B-VII.1 nuclear data library. In addition, the HMF071 series of critical experiments were able to lower the nuclear data uncertainty associated with carbon on the Xe-100 and gFHR by about 11 pcm and 16 pcm respectively using SCALE 6.3.1 and the ENDF/B-VIII.0 nuclear data library. When all-materials were used in the TSURFER evaluation, the uncertainty reduction on the gFHR was about 39 pcm using SCALE 6.3.1 and the ENDF/B-VIII.0 nuclear data library. This is very small; however, due to the fact that carbon is such a well-studied element, any further reduction of uncertainty for thermal-spectrum systems is difficult (but useful) to obtain. These null results disproved our hypothesis and do not allow us to demonstrate this specific use case for evaluating legacy fast-neutron-spectrum experiments, which are otherwise often neglected, for modeling modern thermal-neutron spectrum graphite moderated/reflected reactors.

Measurement of the neutron capture cross section of 185Re in the keV energy region

ABSTRACT Despite various applications of Rhenium in nuclear science and engineering, there have been limited measurements of the neutron capture cross section of 185Re in the keV energy region, with existing data showing discrepancies. To address this, the study employed the time-of-flight method to measure the neutron capture cross section of 185Re. The prompt γ-rays from the neutron capture reactions were detected using a NaI(Tl) detector. The pulse-height weighting technique was employed to determine the neutron capture yield. The measured cross-section values were determined relative to the neutron capture cross section of 197Au in the energy range from 15 to 90 keV. Additionally, the obtained results were corrected for neutron transport effects and impurities. The results of the measurement provided improved accuracy compared to previous studies, contributing to the nuclear data for the capture cross section of 185Re in the keV energy range.

Steady-state analysis of the 1970 Windscale nuclear criticality incident

This paper presents several steady-state neutron transport models which are used to determine the behaviour of the nuclear criticality incident that occurred at the Windscale Works in 1970. The transfer vessel involved in this incident contained two immiscible fissile liquids (an aqueous phase and an organic phase) which had been disturbed such that an emulsion layer formed between them. Infinite domains of the aqueous, organic, and emulsion phases were modelled to enable the understanding of the effects of nuclear data library, neutron transport code and material on the calculated kinetics data, effective multiplication factor and macroscopic neutron cross sections. Six possible configurations of the 1970 Windscale criticality incident were simulated using three-dimensional models in MCNP, two-dimensional models in academic research code EVENT, and one-dimensional models in a collision probability (CP) code. Quantities of interest, such as reactivity, scalar neutron flux, neutron absorption and production rates, and kinetics data are presented. System reactivity, mean neutron generation time and neutron absorption and production rates are shown to increase with increasing emulsion thickness, whereas the total scalar neutron flux is shown to decrease. It is determined that the volume of organic phase present during the transient was likely to be around 39.0 L, as the emulsion thickness required to reach criticality (6.5 cm) was within the range cited in the literature. EVENT and the CP code showed deficiencies when calculating the scalar neutron flux through the plenum gas at the top of the transfer vessel, with the CP code overestimating the total scalar neutron flux in the system by as much as 19 %. The intrinsic neutron source, subcritical multiplication and subcritical power of the system were computed using Monte Carlo simulations. It was shown, by use of the Hansen criteria, that the subcritical system was likely to be in a strong neutron source regime (such that stochastic variations in neutron population are negligible when regarding nuclear criticality transients). In addition, it was computed that a vessel containing 39.0 L of organic solvent had a subcritical power of 10.81 ±0.03μW.

Caryophyllene-type sesquiterpenes from the endophytic fungus Pestalotiopsis lespedezae through an OSMAC approach.

Two new caryophyllene-type sesquiterpenes pestalotiopsins U and V (1 and 2) and three known compounds pestalotiopsin B (7), pestaloporinate B (8), and pestalotiopsin C (9) were isolated by the cultivation of the endophytic fungus Pestalotiopsis lespedezae on solid rice medium, while four additional new caryophyllene pestalotiopsins W-Z (3-6) were obtained when 3.5% NaI was added to the fungal culture medium. The structures of the new compounds were determined by HRESIMS and 1D/2D nuclear magnetic resonance data. Compounds 1-9 were tested for cytotoxicity against the mouse lymphoma cell line L5178Y, but only 6 displayed significant activity with an IC50 value of 2.4 μM.

JENDL-5.0 nuclear data sensitivity, uncertainty, and similarity analyses on the criticality of RSG GAS multipurpose research reactor

The criticality, sensitivity, uncertainty and similarity analysis of the clean, first core of the G. A. Siwabessy Multipurpose Reactor (RSG GAS), using the newly released Japanese Nuclear Data Library version 5 (JENDL-5) was conducted to contribute the validation effort of the JENDL-5, especially for the application on beryllium reflected, light-water moderated, low-enriched uranium (19.75 wt%) fueled material testing reactors. The JENDL-5 sensitivity coefficients and covariance matrices were prepared by MCNP6 and AMPX-6, respectively. The keff uncertainty was then evaluated with TSUNAMI-IP module of SCALE-6. The same module was used for similarity analysis by comparing the sensitivity data of 3,690 criticality experiments. The criticality (keff) analysis results ([C/E-1]) showed maximum overestimation of 801 pcm, 477 pcm and 547 pcm, for JENDL-5, JENDL-4.0 and ENDF/B-VIII.0, respectively. The keff uncertainty originated from the nuclear data was estimated to be 620 pcm, 644 pcm and 637 pcm for JENDL-5 only, JENDL-5 & ENDF/B-VIII.0, and ENDF/B-VIII.0 covariance data, respectively, which are comparable with the keff ([C/E-1]) values. About 125 experiment cases were found to have correlation factors of > 0.8, however, no experiment showed a strong correlation factor (>0.9), therefore, in the future criticality experiments similar to the RSG GAS should be conducted.

Extension of recommended cross section database for production of therapeutic isotopes

Radionuclide-based diagnostics and therapy require proper selection of production nuclear reaction based on knowledge of the production excitation functions and the achievable yields completed with data on the formation of possible impurities. In the present work the existing IAEA recommended cross section data database for production of therapeutic isotopes is extended to production of the 47Sc,47Ca(47Sc), 58mCo, 71As(71Ge), 71Ge, 77Br, 77Kr(77Br), 80mBr, 103Pd, 103Pd(103mRh), 103Ru(103mRh), 105Rh, 117mSn, 119Sb, 119mTe(119Sb), 134Ce, 135La, 149gTb, 161Tb, 165Er, 165Tm(165Er), 167Tm, 197mHg, 197gHg, 198gAu, and 230Pa(230U) radioisotopes. Nearly 60 nuclear reactions are presented and discussed. The new recommended cross-section data and their uncertainties for the production of these 21 radionuclides will be available on the Web page of the IAEA Nuclear Data Section at https://nds.iaea.org/radionuclides and also at the IAEA medical portal https://nds.iaea.org/medportal.

New directions in nuclear data research for accelerator-based production of medical radionuclides

Extensive nuclear data studies have been carried out over the last 30 years in the context of accelerator-based production of radionuclides, especially at energies below 30 MeV, and the achieved database is fairly good. Yet there are some deficiencies or new needs of data. Those needs are generally associated with new emerging clinical applications of radionuclides, e.g. theranostic approach, bimodal imaging, radioimmuno-therapy, etc. This article gives an overview of on-going nuclear data research utilizing charged-particle accelerators in four directions, namely low-energy region, intermediate energy range, use of the α-particle beam, and utilization of fast neutrons generated at accelerators. Wherever possible, a comparison of experimental data with theoretical estimates is presented and evaluated (standardised) data, if available, are also briefly discussed.

Initial Verification and Validation of a New CASMO5 JENDL-5 Nuclear Data Library for Typical LWR Applications

A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expanded for general-purpose applications. At the same time, many essential nuclides for conventional light water reactor (LWR) analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5–based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1–based library as possible with JENDL-5. This study performs initial verification and validation of the new library under typical LWR conditions. Verifications were performed based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark phase III-C, and the calculated kinf and fuel compositions of the boiling water reactor fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, a tank-type critical assembly critical experiment and Takahama-3 postirradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1–based library in predictions of reactivity and fuel compositions for typical LWR systems.

Calculations and analysis of n+52Cr reactions below 20 MeV

ABSTRACT Chromium is one of the important structural materials in the process of nuclear reaction, and the precise nuclear data of n + 52Cr reaction is very important. On the basis of the spherical optical model, the unified Hauser–Feshbach theory and exciton model, all reaction cross sections, angular distributions, energy spectra, and double-differential cross sections were calculated and analysed for n + 52Cr reactions below 20 MeV. The theoretical calculations were compared with the available experimental data and other evaluation data from ENDF/B-VIII.0 and JENDL-5.

Status of the 226Ra nuclear data library and its impact on the production amount of 225Ac via the 226Ra (n,2n) reaction

ABSTRACT 225Ac is a promising alpha emitter for targeted alpha therapy. However, the current capability of 225Ac supply is limited to approximately 63 GBq/y, primarily relying on the natural source of 229Th stocked at a few institutes. Therefore, alternative 225Ac production methods are highly desired. The research and development of 225Ac and its parent nuclide production methods using accelerators and reactors to target 226Ra and 232Th are actively pursued worldwide. Hence, the authors focused on the 226Ra(n,2n)225Ra method using fast neutron spectra as an application of the experimental fast reactor Joyo. This study investigated the status of nuclear data libraries for 226Ra, an essential target for 225Ac production, and evaluated the impact of different nuclear data libraries on the amount of 225Ac produced. Consequently, cross-sections with covariance data were stored in TENDL-2021, JEFF-3.3, and EAF-2010 but not in ENDF or JENDL, the major nuclear data libraries. Furthermore, no consistency occurred among the respective nuclear data, and the 225Ra production amount varied.

Determination of the absolute intensity of the 1205 keV [formula omitted]-ray emission from 91Y

A radiochemically pure solution of 91Y was produced by the thermal neutron fission of 235U followed by successive chemical separations to remove fission product impurities. The gamma emission rate of the 91Y 1205 keV gamma was measured using multiple high purity germanium gamma spectrometers previously calibrated for counting efficiency using a certificated mixed nuclide gamma standard. The activity concentration of the 91Y was subsequently standardised by liquid scintillation counting. From the combination the activity concentration and gamma emission intensity, the absolute intensity of the 1205 keV gamma emission was derived as 0.2297(39)%. This data agrees within the quoted uncertainties with the absolute intensity of 0.26(4)% published in nuclear data sheets A=91 (Baglin, 2013), but reduces the uncertainty by an order of magnitude.

Need for precise nuclear structure data for reactor studies

This paper highlights the strong need for precise nuclear structure and decay data measurements in order to perform high-quality modelling on nuclear reactors and other applications. The context of nuclear data evaluation, as well as the importance of low uncertainty evaluations, will be first presented. The importance of such data for interpreting nuclear data experimental measurements is stressed throughout. To demonstrate this, we will explain how mass and charge-dependent fission yields, decay data (in particular for the purpose of residual heat calculations), and inelastic neutrons scattering cross section rely on nuclear structure and decay information and how new and higher quality in such data can lead to improved accuracy in the precision of evaluated nuclear data.

Population structure of long-snout seahorse Hippocampus reidi in Southwestern Atlantic and implications for management

Abstract Hippocampus reidi represents the most abundant species of the genus Hippocampus along the Brazilian coast. Despite being charismatic, the species is globally threatened due to habitat degradation and commercial exploration, especially in Brazil, which is the leader in exportation and consumption of the species. Through mitochondrial (cytochrome b and control region) and nuclear (1st intron S7) data, the current study investigates the variation and genetic structure of H. reidi along the Brazilian coast, from Pará to Santa Catarina states. The mitochondrial data indicate the presence of two lineages: (1) North/Northeast and (2) South/Southeast, which was partially recovered by nuclear data. This scenario could be related to temperature differences and circulation patterns of the Brazil and North-Brazil currents, which define these groups into biogeographic sub-provinces. The lineages occur in sympatry in Bahia state, which can be explained by the occurrence of secondary contact during the last glacial maximum. Despite presenting two lineages, for management and conservation, three units are indicated: (1) North/Northeast, (2) Bahia, and (3) South/Southeast. The North/Northeast unit proved to be more vulnerable, presenting the lowest genetic diversity indices, representing a priority for future conservation actions.

The road to a modernized NJOY

The modernized version of the NJOY Nuclear Data Processing System is being built from a series of components that enable the traditional work required of the production version of NJOY while also providing a much more interactive user experience through a modern interface. This new version of NJOY is being developed in C++. Python bindings on top of the C++ interface are provided as well. Standard Python packages such as Matplotlib, NumPy or SciPy can thus be seamlessly integrated with NJOY components. We distinguish two component types: format components such as ENDFtk and ACEtk that read and write formatted nuclear data and processing components such as scion which provides many of the common operations needed during processing of nuclear data. The backbone of the modernized NJOY is dryad which provides a format agnostic nuclear data interface that insulates the modernized processing components and modules from the various data formats that NJOY needs to serve.

The fast neutron fluence and the activation monitor activity calculations using fine multigroup and continuous nuclear data

The fast neutron fluence and the activation monitor activities are routinely calculated with TORT deterministic code and BUGLE-B7 nuclear data library with 47 broad energy groups. The objective of the paper is to analyse options to improve reactor dosimetry transport calculations. There are two paths to improve reactor dosimetry calculations. Increasing geometry, angular and energy mesh size is applicable for TORT code while using newer nuclear data libraries is relevant for both deterministic and Monte Carlo codes. Two new calculation options (improved TORT and Monte Carlo MCNP6) were compared with the standard TORT calculation for VVER-440 Dukovany Unit 3 Cycle 31. The fast neutron fluence with 0.5 MeV threshold as well as activity of Fe. Ni, Ti, Cu. Mn and Nb monitors were evaluated. Standard TORT calculations were improved from S16P3 to S30P3 with three times finer axial mesh size. 120° core symmetry r-ϑ mesh size with 0.5° step and fine multigroup libraries VUAMIN-B7 with 199 neutron energy groups and ENDF/B-VH.1 with 200 neutron energy groups. Both ENDFB and IRDFF activation cross sections were used. The drawback of expanded mesh size is raised calculation runtime since TORT deterministic code is not parallelized and one calculation can require multiple weeks of CPU time. An alternative option of using MCNP6 Monte Carlo code with continuous ENDF/B-VH.1 nuclear data with detailed 3-D geometry and pin-wise effective neutron source prepared by MOBY-DICK diffusion code reactor analysis wras explored. It was found that using finer mesh size affects reactor dosimetry’ tallies less than the choice of nuclear data library. BUGLE-B7 and VTTAMIN-B7 produce results typically within 1% difference. ENDF/B-VH.1 calculations with 200 neutron energy’ groups with TORT code are even in better agreement with MCNP6 calculations with continuous nuclear data libraries. The largest differences of around 2% were observed between VTTAMIN-B7 library based onENDFB-VH.O nuclear data and ENDF/B-VH.1 library. Nuclear data library’ has larger impact on the results with up to 7 % difference between all 0.5 MeV fast neutron fluence calculations. The largest intact of nuclear data was observed for Mn(n.2n) monitor.

Target Accuracy Requirements Exercise within WPEC/SG46 and Feedback on Nuclear Data Needs

The Target Accuracy Requirements (TAR) exercise was launched in the framework of the WPEC/S46 (2018-2022) on “Efficient and Effective Use of Integral Experiments for Nuclear Data Validation” [1]. This Exercise aims to quantify priority nuclear data needs or uncertainties reduction to meet design integral parameters target accuracies. To perform this work, the status and methodology of WPEC/SG26 (2005-2008) [2] was updated and reviewed. The overall objective of WPEC/SG46 Exercise on TAR to assess the effect of nuclear data uncertainty reduction on output relevant for reactor design and operation was successfully achieved. This work will be contributing in the short term to provide new entries in the NEA High Priority Request List (HPRL) [3].

Data assimilation of decay heat experiments for feedback on nuclear data

Integral decay heat experiments can provide interesting feedback on particular nuclear data (decay data and fission yields mainly). After ensuring that the C/E discrepancies were mostly due to nuclear data discrepancies, a Bayesian inference approach can be applied. Nevertheless, the results strongly depend on the quality of the experiment and on our capability to estimate realistic experimental correlation matrices when considering several integral experiments in the assimilation process. A former study performed in 2019 was dedicated to the data assimilation of a large C/E dataset from the experimental validation database of fuel inventory calculations with the DARWIN2.3 package in order to provide feedback to the nuclear data evaluators. This paper is an attempt to exploit the General Electric decay heat experiments performed in the USA in the 1980s in order to confirm or not the trends on four particular cumulated fission yields: 235U(nth,f)133Cs, 235U(nth,f)137Cs, 239Pu(nth,f)106Ru and 239Pu(nth,f)144Ce.

New Physical Insight into High-Spin Physics: Double Helix Level Scheme A nuclear structure data approach

To present a new nuclear structure data concept of a level scheme, we redesign the high-spin level scheme of 171Yb nucleus. We start with the description of the property of repeatability of the Eγ data based on which Eγ can be parametrized by the 2c(2I + k − 1) formula with I the nuclear spin and k an integer band offset different from band to band. This describes a beam of parallel and equidistant lines whose 2c average slope and k numbers can be determined by a least-squares fit to the experimental bands, which characterize the average rotational behavior of the experimental bands. Based on repeatability the level scheme of 171Yb nucleus can be shaped as a discrete 3D double helix structure, with the experimental bands superimposed as decay paths. Double helix suggests that semiclassically nuclear rotation can be seen as a vortex motion.

NEA GitLab: A centralised platform for development and deployment of nuclear codes and data

The NEA Data Bank is an international reference centre for computer codes, nuclear and thermochemical data which has traditionally used simple file servers, and even DVDs, to deliver valuable content to end users across the globe. With the recent implementation of a self hosted GitLab system at the NEA it has enabled the Data Bank to streamline delivery, automate processes and testing, while empowering code owners and developers with a secure platform to collaborate and develop codes. In this paper we present the NEA Git- Lab system with some concrete examples of codes such as Kraken, PHITS and FISPACT-II taking advantage of many of the services and functionalities provided by GitLab. We illustrate methodologies on how to work effectively with third party software in a position of being a custodian of code, rather that a code owner, providing DevSecOps as a service.

Reanalysis of graphite nuclear data limits for AGR dosimetry with updated carbon scattering data and Magnox plant measurements

The Advanced Gas-cooled Reactors (AGR) operating in the UK are, like the Magnox reactors that preceded them, graphite moderated and cooled by pressurised CO2 gas. As such, the nuclear data for carbon are of primary importance in neutron dosimetry and nuclear heating calculations. There is evidence from benchmark data and plant measurements of underprediction of neutron fluxes at the lower end of the fast neutron range for thicknesses of graphite greater than ~50 cm. Furthermore, the range of validation is limited to ~70 cm of graphite, with components beyond this being allocated bounding damage rates. Following the previous ISRD, opportunities were taken to: Look again at the carbon scattering cross-sections by incorporating the ENDF/B-VIII.0 Beta 4 carbon data to assess the impact on the analysis of benchmark data. Reassess validation data from the earlier Magnox reactors to put them on a more appropriate equivalent graphite thickness scale to the AGRs. The investigation into graphite nuclear data libraries concluded that the incorporation of ENDF/B-VIII Beta 4 carbon data into the JEF2.2 and ENDF/B-VII.1 data libraries with BINGO collision processing had only a limited effect on the calculated responses. The UKNDL data, with DICE collision processing, currently used in dosimetry calculations for EDF Energy’s AGRs produces the closest agreement with measurement for all calculated responses for graphite thicknesses greater than 10 cm. For reactor plant validation, applying the sensitivity of the 93mNb(n,n’) response to graphite density resulted in a 15 cm increase to the validation database when compared to a simple “straight-line” distance. However, caution is advised in applying the benchmark equivalent distance derived from locations within a graphite column, to ex-core plant measurements.