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DOE OSTI2017

AGR-5/6/7 Irradiation Monitoring Web Site Design

The Nuclear Data Management and Analysis System (NDMAS) manages irradiation experiment data collected under the Advanced Gas Reactor (AGR) Fuel Development and Qualification program within the Advanced Reactor Technologies Technology Development Office. The AGR-5/6/7 irradiation experiment combines the fifth, sixth, and seventh planned irradiation campaigns of the U.S. Department of Energy AGR Fuel Development and Qualification Program. These combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions (PNL-5245). Experimental irradiation measurements for this experiment (i.e., thermocouple temperature, gas flow rate, fission product release rate, gross gamma count, and Advanced Test Reactor (ATR) operational parameters) are captured automatically to the NDMAS database. Data derived from numerical simulation or other mathematical analyses (i.e., fast fluence, burnup, temperature, and fission product release to birth ratio) are captured to the database when provided by the researchers. The NDMAS database is a relational database accessed via structured query language and managed with Microsoft SQL Server. While select users can directly access the database, that is not the most convenient way to interact with the data. The NDMAS team constructs numerous tables that combine related datasets, and delivers those data to researchers via graphs and spreadsheets on a Microsoft SharePoint-based website. For the AGR-5/6/7 experimental data, a new subsite has been developed, along with a set of web pages. These pages will be used by members of fuel irradiation team for monitoring irradiation conditions according to specification (SPC-1749), by the physics modelers to access necessary model inputs, and by the nuclear engineering community to view and access data such as fuel temperatures, fission product release from capsules, and ATR operational parameters. The deliverable represented in this document is a SharePoint website containing tables of the AGR-5/6/7 measured and calculated data, graphs displaying the data, and links to source documents and references.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Validation of MCNP Critical Benchmark Models of PU-MET-FAST-016

A new centralized repository of high-quality MCNP models of critical benchmark experiments is currently under development at Los Alamos National Laboratory (LANL). The repository is the combined effort of the Nuclear Criticality Safety, Nuclear Data, and Monte Carlo code development/application organizations at LANL. The initial set of benchmark models in the repository are derived from the Whisper Suite provided with MCNP6.2 and are evaluated against the benchmark experiments described in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The goal is to build a single LANL benchmark collection that is up to date with the latest ICSBEP revision, has a formal review and revision process, is contained in an open-source repository, and utilizes new Python tools for improved input and output file review. This paper describes the validation of the models associated with PU-MET-FAST-016, “Flooded 3X3X3 Arrays of 3-kg Plutonium (Pu) Metal Cylinders – Phase I”. In this critical experiment, twenty-seven 3-kg Pu metal cylinders with a height of 46.33 ± 0.15 mm and diameter of 65.25 ± 0.05 mm were arranged in a cubic array. The Pu cylinders were sealed in aluminum cans, and three cans were placed in each of nine perforated aluminum sleeves, as shown in Fig. 1. The nine sleeves were supported by an aluminum frame that allowed for control of the lateral spacing of the cylinders, both in the X and Y directions. The benchmark experiment measured thirteen configurations. Of these configurations, six were modeled using MCNP.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

TREAT neutronics analysis and design support, Part II: Multi-SERTTA-CAL

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. An integral aspect of prior TREAT transient testing was the incorporation of calibration experiments to experimentally evaluate and validate test conditions in preparation of the actual fuel testing. The calibration experiment package established the test parameter conditions to support fine-tuning of the computational models to deliver the required energy deposition to the fuel samples. The calibration vehicle was designed to be as near neutronically equivalent to the experiment vehicle as possible to minimize errors between the calibration and final tests. The Multi-SERTTA-CAL vehicle was designed to serve as the calibration vehicle supporting Multi-SERTTA experimentation. Models of the Multi-SERTTA-CAL vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries; these results were then compared against those performed for Multi-SERTTA to determine the similarity and possible design modification necessary prior to construction of these experiment vehicles. The estimated reactivity insertion worth into the TREAT core is very similar between the two vehicle designs, with the primary physical difference being a hollow Inconel tube running down the length of the calibration vehicle. Calculations of PCF indicate that on average there is a reduction of approximately 6 % for PWR fuel rodlets irradiated under both wet and dry conditions. Changes to the primary or secondary vessel structure in the calibration vehicle can be performed to offset this discrepancy and maintain neutronic equivalency. Current possible modifications to the calibration vehicle include reduction of the primary vessel wall thickness, swapping Zircaloy-4 for stainless steel 316 in the secondary containment, or slight modification to the temperature and pressure of the water environment within the primary vessel. Removal of some of the instrumentation within the calibration vehicle can also serve to slightly increase the PCF. Future efforts include further modification and optimization of the Multi-SERTTA and Multi-SERTTA-CAL designs in preparation of actual TREAT transient testing. Experimental results from both test vehicles will be compared against calculational results and methods to provide validation and support additional neutronics analyses. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Isomeric yield ratios of fission products: A missing piece in reactor antineutrino summation calculations

The calculation of the spectrum of antineutrinos ($\bar{v}_e$) from a reactor is a complicated problem requiring several nuclear data and physics inputs. Many of these have been investigated thoroughly to improve calculations and properly account for uncertainties. One input which has heretofore escaped consideration is the fission-yield distribution between ground and isomeric states. Here, in this work, we explore the effect of incorporating newly evaluated isomeric yield ratios (IYR) for 43 fission products into summation calculations and identify the disproportionate importance of certain isotopes, particularly at higher energies. Our analysis indicates that updated IYRs contribute to a significant increase in the $\bar{v}_e$ spectrum around and above 7 MeV, with increases of more than 50% at higher energies. Through a detailed sensitivity study, we highlight a number of isotopes for which the IYR has a substantial effect on the $\bar{v}_e$ spectrum. This work stresses the critical role of isomeric yields in calculations of reactor $\bar{v}_e$ spectra and points to the necessity for their accurate experimental determination, especially for key fission products, in order to refine our understanding and address the observed anomalies between measured and calculated $\bar{v}_e$ spectra.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Gamma-ray Measurements from Neutron Pulsed Die-Away Experiments

Pulsed-neutron die-away experiments are a promising experimental method that complement criticality benchmarks for nuclear data validation. The experiments have several advantages that include compatibility with non-fissile material, low benchmark uncertainties, and high sensitivity to absorption and scattering cross sections. When validating thermal neutron scattering laws, small targets are desirable for their large sensitivity to the scattering cross section and angular distribution. Unfortunately, room return or scattering from a shielding box limits the use of the very small targets needed to maximize sensitivity to thermal scattering laws. To address this problem, we propose changing the observable of the experiment from leaked thermal neutrons to γ-rays produced by (n, γ) reactions in the target. If feasible, this change allows removal of the shielding box, achieving higher sensitivity to thermal neutron scattering laws. This study focuses on replicating benchmark pulsed-neutron die-away experiments with γ-rays. We compare the integral parameters with γ-rays to those obtained with neutrons. We also discuss experimental design choices such as detector placement, shielding, and the presence of Cd. The results show the integral parameter with γ-rays is sensitive to detector location with respect to the target when Cd shielding is present. In addition, while Pb shielding around the γ-ray detectors do not seem to affect the integral parameter, the presence of Cd shielding does and adds background γ-rays to the die-away curves making it difficult to calculate the integral parameter.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

On-the-fly Doppler Broadening with Probability Table Interpolation for Unresolved Resonance Region in RMC

With the rapid development of computational power and parallel algorithms, Monte Carlo method has been widely investigated and used in neutron transport simulations due to its advantages in geometry modelling and usage of continuous energy point-wise cross sections. Nowadays the road has been paved to couple Monte Carlo codes with thermal-hydraulics codes to make multi-physics calculations. One of the most important aspects of thermal-hydraulics feedback is the detailed temperature distribution, which demands temperature dependent cross sections in neutron transport simulations. The conventional pre-generated cross section libraries at each temperature point are no longer feasible for detailed temperature distributions because of tremendous memory footprint. Hence, different on-the-fly Doppler broadening methods have been proposed for resolved resonance energy region and thermal energy region, such as the interpolation, regress model fitting, multipole representation (MPR), target motion sampling (TMS) and different integration algorithms based on Doppler broadening equation. Compared to resolved resonance energy region and thermal energy region, the representation of cross sections in unresolved resonance region is slightly different. As a result of the limitation of experiment measurements, the overlapped resonances cannot be distinguished from each other and the statistic average resonance parameters are given in the evaluated nuclear data file. In early applications, infinite-dilute pointwise cross sections were obtained from average parameters and were directly used in neutron simulations. However, it was proven that these infinite-dilute pointwise cross sections cannot capture the energy self-shielding effect, thus leading to an overestimate of reaction rates and usually an underestimate of k{sub eff} owing to the increased absorption rates and increased scattering probability to energy below unresolved resonance region where the capture-to-fission ratio is typically higher. This phenomenon is of great importance especially for problems where there is a considerable flux in unresolved resonance region. Then the probability table method was put forward by Levitt to take energy self-shielding effect into account, which provides probability distribution function of cross sections in unresolved resonance region instead of specific cross section values. The probability distribution function is presented in the form of tables and the mean values are equal to the infinite dilution cross sections. To obtain the temperature dependent cross sections in unresolved energy region, there are two different methods. One is to calculate cross sections directly from average resonance parameters and their statistical distributions , which is effective and precise but time-consuming. The other is to interpolate between pre-generated probability table files, which is much more efficient. In this paper, on-the-fly Doppler broadening of cross sections in unresolved resonance regions based on probability tables interpolation was implemented in RMC codes as a supplement to on-the-fly Doppler broadening for resolved resonance energy region and thermal region, which is based on the integration of Doppler broadening equation. The results of criticality calculation and depletion calculation show that on-the-fly Doppler broadening for the whole energy region is accurate and efficient. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Technical Evaluation Report for the NDMAS AGR-5/6/7 TRISO Irradiation Monitoring System

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Impact of Americium-241 (n,γ) Branching Ratio on SFR Core Reactivity and Spent Fuel Characteristics

An accurate prediction of core physics and fuel cycle parameters largely depends on the order of details and accuracy in nuclear data taken into account for actual calculations. {sup 241}Am is a major gateway nuclide for most of the minor actinides and thus an important nuclide for core physics and fuel-cycle calculations. The {sup 241}Am(n,γ) branching ratio (BR) is in fact energy dependent, therefore, it is necessary to taken into account the spectrum effect on the calculation of the average BR for the full-core depletion calculations. Moreover, the accuracy of the BR used in the depletion calculations could significantly influence the core physics performance and post irradiated fuel compositions. The BR of {sup 241}Am(n,γ) in the ENDF/B-VII.0 library is relatively small and flat in the thermal energy range, gradually increases within the intermediate energy range, and even becomes larger in the fast energy range. This indicates that the properly collapsed BR for fast reactors could be significantly different from that of thermal reactors. The evaluated BRs also differ from one evaluation to another. As seen in Table I, average BRs for several evaluated libraries calculated by means of a fast spectrum are similar but have some differences. While some of widely available and used depletion codes, such as ORIGEN in SCALE, have a capability to calculate a variable BR based on the system neutron spectrum, most of the currently available depletion codes still use a pre-determined single value BR for each library. However, ideally it should be determined on an on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different {sup 241}Am(n,γ) BRs on core and spent fuel parameters. This paper investigates the impact of the {sup 241}Am(n,γ) BR on the results of sodium-cooled fast reactor (SFR) full-core based fuel-cycle calculations. The analysis is performed by gradually increasing the value of BR from 0.15 to 0.25 and studying its impact on the core reactivity and characteristics of SFR spent fuels over extended storage times (∼10,000 years). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

A Solid of Revolution Time Study using COG11.1 and MCNP6.1

The performance of two continuous energy Monte Carlo transport codes, COG11.1 and MCNP6.1, has been compared for the {sup 239}Pu Jezebel Benchmark using cross sections based on the ENDF/B-VII.1 evaluated nuclear data library. The purpose of this study is to determine the length of time it takes each code to complete a series of criticality calculations, using two different methods to model a solid of revolution. A solid of revolution is essential in many fields of engineering and mathematics, and has been a standard feature in computer-aided design and manufacturing since the 1970's. In the field of criticality safety, a solid of revolution can be used to reproduce and simplify complex engineering drawings and systems that would otherwise require excessive use of geometric primitives and computational resources to achieve a comparable level of fidelity. There are two basic methods to model a solid of revolution: Method 1 (COG) specifies a set of x-y pairs or polar coordinates and rotate the curve about the x-axis. Method 2 (COG and MCNP) creates a set of end-to-end conical frustums and/or cylinders. Method 1 is only available in COG and can be implemented using one surface and sector specification. Method 2 can be implemented in both COG and MCNP. This time study was initially conceived as a result of criticality safety analysis performed for a complex system with a high degree of rotational symmetry. Each COG input deck typically took 1-2 days to run in serial mode. COG input decks were then converted to MCNP format and run on the same computer cluster. Instead of expected 1-2 day runtimes, it took MCNP6.1 more than one week to complete each run. Since MCNP typically runs faster than many codes, these lengthy runtimes were considered unusual. This study attempts to investigate the runtime discrepancy between the two codes. The methods and benchmark in this time study were selected to provide a consistent, generalized approach to modeling an arbitrary solid of revolution. While there are many different ways to model the {sup 239}Pu Jezebel Benchmark, the focus of this study is on investigating the practical difference between the two basic methods to model a surface of revolution. Throughout this study, the only code that maintains reasonably low runtimes is COG11.1 (Method 1). Based on the observed performance of each code and method, it is likely that the divergent runtimes are the result of tracking particles across planes that lie between conical frustums, as opposed to defining the object boundary as a single surface. Slight fluctuations in graphed results were also observed, due to random sampling and CPU loading. This was confirmed by re-running input decks and comparing average runtimes to a preliminary curve fit. From the perspective of a criticality safety analyst, the results of this study are expected to contribute towards understanding one of the practical trade-offs between codes and methods as they apply to real-world analysis. Ultimately, it's about choosing the right tool for the job.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

Isotopic composition of a sample enriched in 93 Zr

A project to determine the neutron-capture cross section of long lived fission products and minor actinides has been started by using a beam-line at Japan Proton Accelerator Research Complex (J-PARC). We prepared one of the target nuclides is Zr-93, which in Oak Ridge National Laboratory. Qualitative and quantitative analyses on the sample were performed at Kyoto University. The isotopic composition of (m) Zr (m 90, 91, 92, 93, 94, and 96) was precisely determined by multi-collector thermal ionization mass spectrometry with < 0.1 % of 2 sigma uncertainty. We determined that the atomic abundance of Zr-93 in the sample to be 18.86 ± A 0.05 %.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

Analysis of the time-of-flight neutron scattering cross-section data for light water measured at the SEQUOIA spectrometer, Spallation Neutron Source (SNS)

Thermal neutron scattering cross-section data for light water available in the major nuclear data libraries observes significant differences especially at reactor operating temperatures. During the past few years there has been a renewed interest in reviewing the existing thermal scattering models and generating more accurate and reliable thermal scattering cross sections using existing experimental data and in some cases based on Molecular Dynamics (MD) simulations. There is a need for performing new time-of-flight experiments at high temperatures and pressures, to have a better understanding of the physics involved in the scattering process that could help improve the existing TSL data.Lack of experimental thermal scattering data for light water at high temperatures led to a new measurement campaign within the INSIDER project at the Institut de radioprotection et de sûreté nucléaire (IRSN). Double differential scattering cross section for light water have been measured at the SEQUOIA spectrometer based at the Spallation Neutron Source (SNS), Oak Ridge National Laboratory, United States. Several measurements have been carried out at different temperatures and pressures corresponding to liquid light water. Measurements at five different incident neutron energies Ei (8, 60, 160, 280 and 800 meV) have been carried out to help exploring different regions of the frequency spectrum. This paper presents the analysis of the dynamic structure factor and the derived frequency spectrum of light water. The analysis of the experimental data would provide one with better confidence, the behavior of thermal scattering cross sections for light water at high temperatures, knowledge of which is very important for the design of novel reactors as well as existing pressurized water reactors.

Jaiswal, V↗

DOE OSTI2016

Surviving Scientific Academia . . . and Beyond

It's been 16 years since I first took a physics class at Weber State University. Since them, I've survived graduate school in Nuclear Engineering, and a postdoc appointment doing nuclear nonproliferation. Now I'm a Technical Staff Member at Los Alamos National Laboratory working with nuclear data, the physics behind the numerical simulations of nuclear reactors and nuclear weapons. Along the way, I've learned a few things. First, scientific computing is everywhere in science. If you are not writing codes, you will be analyzing their output, and generally there will be more output than a human can correctly and accurately interpret in a timely manner. Second, a career in science or engineering can be very rewarding with opportunities to collaborate with and generate friendships with very bright people from all over the world.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

Final Technical Report for "Nuclear Technologies for Near Term Fusion Devices"

Over approximately 18 years, this project evolved to focus on a number of related topics, all tied to the nuclear analysis of fusion energy systems. For the earliest years, the University of Wisconsin (UW)’s effort was in support of the Advanced Power Extraction (APEX) study to investigate high power density first wall and blanket systems. A variety of design concepts were studied before this study gave way to a design effort for a US Test Blanket Module (TBM) to be installed in ITER. Simultaneous to this TBM project, nuclear analysis supported the conceptual design of a number of fusion nuclear science facilities that might fill a role in the path to fusion energy. Beginning in approximately 2005, this project added a component focused on the development of novel radiation transport software capability in support of the above nuclear analysis needs. Specifically, a clear need was identified to support neutron and photon transport on the complex geometries associated with Computer-Aided Design (CAD). Following the initial development of the Direct Accelerated Geoemtry Monte Carlo (DAGMC) capability, additional features were added, including unstructured mesh tallies and multi-physics analysis such as the Rigorous 2-Step (R2S) methodology for Shutdown Dose Rate (SDR) prediction. Throughout the project, there were also smaller tasks in support of the fusion materials community and for the testing of changes to the nuclear data that is fundamental to this kind of nuclear analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Fukushima Daiichi Radionuclide Inventories

Radionuclide inventories are generated to permit detailed analyses of the Fukushima Daiichi meltdowns. This is necessary information for severe accident calculations, dose calculations, and source term and consequence analyses. Inventories are calculated using SCALE6 and compared to values predicted by international researchers supporting the OECD/NEA's Benchmark Study on the Accident at Fukushima Daiichi Nuclear Power Station (BSAF). Both sets of inventory information are acceptable for best-estimate analyses of the Fukushima reactors. Consistent nuclear information for severe accident codes, including radionuclide class masses and core decay powers, are also derived from the SCALE6 analyses. Key nuclide activity ratios are calculated as functions of burnup and nuclear data in order to explore the utility for nuclear forensics and support future decommissioning efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

ORNL R-matrix Analyses for Non-Fissile Materials within NCSP [Slides]

This lecture covers Oak Ridge National Laboratory (ORNL) R-matrix analyses for non-fissile materials within Nuclear Criticality Safety Program (NCSP). This presentation includes a Strontium and Chlorine evaluation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Additions to the ICSBEP AND IRPHEP Handbooks Since NCSD 2013

Innumerable experiments have been performed worldwide to support criticality safety activities. Those experiments represent significant investments in cost, expertise, and infrastructure; as such, they represent valuable assets as the basis for recording, development, and validation of methods and nuclear data vital for modern and future criticality safety needs. Preservation and evaluation of these experimental data as benchmarks serves to provide for the best interest of the nuclear criticality safety community. High-quality integral benchmark experiments have always been a priority. There is ever growing interest in developing integral benchmark data to increasingly quantify and reduce calculational uncertainties to meet the demands of future criticality safety needs supporting next generation reactors and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent be-cause of dwindling availability of critical facilities worldwide and the high cost of per-forming new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Reactor Physics Benchmark Experiments (IRPhEP Handbook) are widely utilized. Significant benchmark data have been added to those two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Wilmingt on, North Carolina (September-October 2013). This paper highlights those additions. The 2016 edition of the ICSBEP Handbook now includes data for 570 evaluations containing benchmark specifications for 4,913 critical, subcritical, or near-critical configurations, representing contributions from over 20 countries. There are a total of 7 criticality-alarm-placement/shielding evaluations containing a total of 45 benchmark configurations, and 8 fundamental physics benchmark evaluations containing a total of 215 measurements relevant to criticality safety applications. The IRPhEP Handbook now includes data from 151 experimental series (representing 50 reactor facilities) and also represents contributions from over 20 countries. Of the 151 benchmarks, five are draft contributions. Both the ICSBEP and IRPhEP are collaborative efforts that involve numerous scientists, engineers, and administrative support personnel worldwide. Continued contribution to these handbooks is expected to further provide the high-quality benchmark data necessary to compliment criticality safety activities going forward.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Total Absorption Spectroscopy Study of ⁹²Rb Decay: A Major Contributor to Reactor Antineutrino Spectrum Shape

The accurate determination of the emitted reactor antineutrino flux is still a major challenge for actual and future neutrino experiments at reactors, especially after the evidence of a disagreement between the measured antineutrino energy spectrum by Double Chooz, Daya Bay, and Reno and calculated antineutrino spectra obtained from the conversion of the unique integral beta spectra measured at the ILL reactor. Using nuclear data to compute reactor antineutrino spectra may help understanding this bias, with the study of the underlying nuclear physics. Summation calculations allow identifying a list of nuclei that contribute importantly to the antineutrino energy spectra emitted after the fission of ²³⁹ , ²⁴¹Pu and ²³⁵ , ²³⁸U, and whose beta decay properties might deserve new measurements. Among these nuclei, ⁹²Rb exhausts by itself about 16% of of the antineutrino energy spectrum emitted by Pressurized Water Reactors in the 5 to 8 MeV range. In this Letter, we report new Total Absorption Spectroscopy (TAS) results for this important contributor. The obtained beta feeding from ⁹²Rb shows beta intensity unobserved before in the 4.5 to 5.5 MeV energy region and gives a ground state to ground state branch of 87.5 % ± 3%. These new data induce a dramatic change in recent summation calculations where a 51% GS to GS branch was considered for ⁹²Rb, increasing the summation antineutrino spectrum in the region nearby the observed bias.The new data still have an important impact on other summation calculations in which more recent data were considered

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗