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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Developing Source Term Database for Advanced Reactors

A source term database is crucial to informing nuclear emergency response measures, enabling emergency responders to assess the potential severity of nuclear and radiological consequences. In recent times, various advanced reactor designs have come into operation, are under construction, or are being designed and developed. This report documents an effort carried out to develop a source term database for advanced reactors. The report covers key design features of these reactors and discusses radioactivity buildup and source term inventories of dose-significant radionuclides in the reactor core. For neutronic and depletion analyses, we used the SCALE code system, a computational suite for reactor physics, depletion, criticality, and sensitivity/uncertainty quantification. We used SCALE/TRITON to perform depletion calculations to predict cycle length and discharge burnup and to generate the ORIGEN reactor library. Subsequently, we used SCALE/ORIGAMI to calculate radioactivity buildup and, thereby, the source term inventories at the targeted discharge burnup, using the ENDF/B-VII.1 nuclear data library. This report covers several advanced reactors, including the KLT-40S, RITM-200N, VOYGR, and eVinci. However, other reactors, such as the RITM-200S and ARC-100, have yet to be investigated and will be explored in future efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Scattering matrix pole expansions for complex wave numbers in R -matrix theory

In this followup article to Ducru et al., we establish new results on scattering matrix pole expansions for complex wave numbers in R-matrix theory. In the past, two branches of theoretical formalisms emerged to describe the scattering matrix in nuclear physics: R-matrix theory and pole expansions. The two have been quite isolated from one another. Recently, our study of Brune's alternative parametrization of R-matrix theory has shown the need to extend the scattering matrix (and the underlying R-matrix operators) to complex wave numbers. Two competing ways of doing so have emerged from a historical ambiguity in the definitions of the shift S and penetration P functions: the legacy Lane and Thomas's “force closure” approach versus analytic continuation (which is the standard in mathematical physics). The R-matrix community has not yet come to a consensus as to which to adopt for evaluations in standard nuclear data libraries, such as ENDF. Here, in this article, we argue in favor of analytic continuation of R-matrix operators. We bridge R-matrix theory with the Humblet-Rosenfeld pole expansions, and discover new properties of the Siegert-Humblet radioactive poles and widths, including their invariance properties to changes in channel radii a c . We then show that analytic continuation of R-matrix operators preserves important physical and mathematical properties of the scattering matrix—canceling spurious poles and guaranteeing generalized unitarity—while still being able to close channels below thresholds.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Criticality safety enhancements for SCALE 6.2 and beyond

SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. Since 1980, regulators, industry, and research institutions around the world have relied on SCALE for nuclear safety analysis and design. SCALE 6.2 provides several new capabilities and significant improvements in many existing features for criticality safety analysis. Enhancements are realized for nuclear data; multigroup resonance self-shielding; continuous-energy Monte Carlo analysis for sensitivity/uncertainty analysis, radiation shielding, and depletion; and graphical user interfaces. An overview of these capabilities is provided in this paper, and additional details are provided in several companion papers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Whole core analysis of OECD/NEA TVA-WB1 benchmark exercise 1 using continuous-energy Monte Carlo code MVP with JENDL-4.0u and ENDF/B-VII.0

The TVA WB1 benchmark problem was precisely modeled in accordance with the geometry and material specifications, and benchmark exercise 1 was solved with the MVP code and the nuclear data libraries based on JENDL-4.0u (JENDL-4.0 updated file) and ENDF/B-VII.0. The criticality, the control rod bank worth, the isothermal temperature coefficient (ITC), the integral rod worth of bank D, and differential soluble boron worth (DBW) were obtained. The criticality and the control rod bank worth were compared with the measured data and published calculation results with other codes, Serpent and KENO. The criticality and the control rod bank worth calculated by MVP and the libraries agree with the measured and the published calculation results within the criteria specified in the ANSI/ANS standard. From these results, the calculation model for the MVP and the libraries was verified. The other parameters such as ITC, the integral rod worth, and DBW were also compared between JENDL-4.0u and ENDF/B-VII.0. The comparison of the results shows JENDL-4.0u and ENDF/B-VII.0 give comparable results within the standard deviation, the difference between libraries is small. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Analysis of SCALE Criticality and Sensitivity Calculations for Reflected HEU Cylinders [Abstract]

The SCALE code package offers several nuclear data libraries to support Monte Carlo (MC) transport, as well as MC-based derivation of $\kappa$ eff sensitivity and uncertainty (S/U) data. The CSAS sequence using the KENO MC code can utilize continuous-energy (CE) cross sections, or pre-generated multigroup (MG) cross section libraries. The use of MG libraries introduces bias into calculations in exchange for faster transport solutions. The TSUNAMI-3D sequence also utilizes KENO MC calculations. TSUNAMI-3D has two CE calculational methods: the Iterated Fission Probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. Previous work has shown poor agreement between CLUTCH and confirmatory direct perturbation calculations in specific applications, e.g., fissionable and polyethylene reflectors

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

A study on the criticality safety for the direct disposal of used nuclear fuel in Japan: Application of burnup credit to the criticality safety evaluation for the disposal canister - 14049

Japan has recently started the technical development of the direct disposal of used nuclear fuel (UNF) to prepare various disposal options. Unlike the disposal of radioactive waste from UNF reprocessing, criticality safety is important here because of the presence of a certain amount of fissile nuclides in UNF. This paper provides an outline of the research to be addressed in this field and the relevant studies in Japan. It especially includes the first result of a criticality safety evaluation of a disposal canister model adopting burnup credit. In this evaluation, it was assumed that a steel disposal canister loaded with fuel assemblies was surrounded by a clay buffer material, and the fuel assemblies and the disposal canister were intact. Conditions representative of discharged pressurized-water reactor (PWR) fuel were applied in this study - initial U-235 enrichment of 4.5 wt% and burnup of 45 GWd/t. The isotopic composition of the burnt fuel was evaluated using SWAT3.1 code. In the depletion calculation, conservative parameters were applied within the typical PWR operation range. The effective neutron multiplication factor (k{sub eff}) was then calculated using a continuous-energy Monte Carlo code MVP-2.0, adopting the JENDL-4.0 nuclear data library. Uncertainty in k{sub eff} due to depletion calculation errors was evaluated using the correction factors of isotopic composition, which were obtained via multiple comparisons between the calculated and measured isotopic concentrations of irradiated fuels. In addition, the effects of axial burnup profile and horizontal burnup gradient on k{sub eff} were evaluated using the open burnup distribution database. The k{sub eff} including the uncertainties and conservatism was found to be below 0.95 for both actinide-only and actinide and fission product considerations when the fuel assemblies and the disposal canister were assumed to be intact. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

NCSP supports 240 Pu prompt fission neutron spectrum (PFNS) evaluation

A new 240 Pu PFNS evaluation was recently undertaken at LANL as a strategic priority. It truly is an NCSP end-to-end product. It factors in a new differential experiment funded by NCSP, builds on theoretical work coming out of a previous NCSP nuclear data evaluation milestone and was validated with an NCERC experiment that was recently evaluated as an integral benchmark with NCSP funds.

240Pu↗
DOE OSTI2021

First Penning trap mass measurement of 36 Ca

Background: Isobaric quintets provide the best test of the isobaric multiplet mass equation (IMME) and can uniquely identify higher order corrections suggestive of isospin symmetry breaking effects in the nuclear Hamilto nian. The Generalized IMME (GIMME) is a novel microscopic interaction theory that predicts an extension to the quadratic form of the IMME. Only the A = 20, 32 T = 2 quintets have the exotic T z = –2 member ground state mass determined to high-precision by Penning trap mass spectrometry. Purpose: To establish A = 36 as the third high-precision T = 2 isobaric quintet with the T z = –2 member ground state mass measured by Penning trap mass spectrometry and provide the first test of the predictive power of the GIMME. Method: Here, a radioactive beam of neutron-deficient 36 Ca was produced by projectile fragmentation at the National Superconducting Cyclotron Laboratory. The beam was thermalized and the mass of 36 Ca + and 36 Ca 2+ measured by the Time of Flight - Ion Cyclotron Resonance method in the LEBIT 9.4 T Penning trap. Results: We measure the mass excess of 36 Ca to be ME= –6483.6(56) keV, an improvement in precision by a factor of 6 over the literature value. The new datum is considered together with evaluated nuclear data on the A = 36, T = 2 quintet. We find agreement with the quadratic form of the IMME given by isospin symmetry, but only coarse qualitative agreement with predictions of the GIMME. Conclusion: A total of three isobaric quintets have their most exotic members measured by Penning trap mass spectrometry. The GIMME predictions in the T = 2 quintet appear to break down for A = 32 and greater.

20 ≤ A ≤ 38↗
DOE OSTI2022

Investigating Fission Reaction Rate Ratio Sensitivities [Slides]

One EUCLID project goal is to create a NEW capability in the MCNP6 code. This is being tested on reaction rate sensitivities and compared with SENSMG results. Reaction rate ratio sensitivities were added to EUCLID sensitivity library and investigated to determine value toward designing optimized experiments aimed at reducing compensating errors in nuclear data. The use of adjustment using multiple responses (and underlying/required capabilities) may benefit many applications. These include a tool for adjustment using multiple integral responses (ND adjustment and validation community); sensitivities of additional responses (ND adjustment and validation community); sensitivity capabilities (users in many application areas including safeguards/nonproliferation, criticality safety, etc.); and experiment optimization capability (applications which can benefit from integral experiments). Team is working to build and test future tools and to understand compensating errors

235U↗
DOE OSTI2025

Summary of FY25 activities on the project Realistic Reaction Evaluations for Fission Products Off Stability (RREFPOS)

Applications such nonproliferation, post-detonation forensics, spent-fuel assay, reactor burnup and design, as well as astrophysics, rely on the accurate description of the neutron interaction with unstable fission products. Current neutron-induced cross-section descriptions for these nuclei are either non-existent or based on simplified assumptions, leading to unquantified impacts on predicted cross-sections. The goal of the Realistic Reaction Evaluations for Fission Products Off Stability (RREFPOS) project is to develop a reproducible methodology leading to a more predictive and realistic new set of evaluated nuclear data, improving the reliability of applications involving fission products off stability. The current report documents the activities related to the RREFPOS project in its first year, corresponding to the Fiscal Year of 2025 (FY25).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Measured Reaction Rate Ratios of 235 U, 238 U, 237 Np, 239 Pu Samples in the PFUNS 235 U Prompt Fission Neutron Spectrum Criticality Experiment

The Prompt Fission Uranium Neutron Spectrum experiment, an experiment to reduce uncertainties in the high energy tail of the 235 U prompt fission spectrum, achieved success by performing two separate irradiations measuring approximately 40 different IRDFF-II reactions total using over 20 different foil materials at the National Criticality Experiments Research Center in February 2024. The criticality experiment utilized a set of highly enriched uranium hemispherical shells of increasing diameters with a large void in the center where the samples were located. The focus of this work is the first of two PFUNS irradiations focused on irradiating two of each fission foils, one bare and one cadmium covered, along with metal activation foils containing reaction products with short half-lives, such as indium, iron, and gold along with nickel fluence monitors. This work focuses on presenting the initial reaction rate ratio results of the fission foils from the aforementioned first irradiation to assist in nuclear data validation of those species in a nearly pure 235 U prompt fission neutron spectrum and compares to the Lady Godiva and Flattop-25 historic experiments at the Los Alamos Critical Experiments Facility. Future work will combine fission foil and metallic activation foil reaction rate ratio results from both the first and second higher power irradiation and reaction rates for a final spectral adjustment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Sensitivity studies for the weak r process: neutron capture rates

Rapid neutron capture nucleosynthesis involves thousands of nuclear species far from stability, whose nuclear properties need to be understood in order to accurately predict nucleosynthetic outcomes. Recently sensitivity studies have provided a deeper understanding of how the r process proceeds and have identified pieces of nuclear data of interest for further experimental or theoretical study. A key result of these studies has been to point out the importance of individual neutron capture rates in setting the final r-process abundance pattern for a ‘main’ (A ∼ 130 peak and above) r process. Here we examine neutron capture in the context of a ‘weak’ r process that forms primarily the A ∼ 80 r-process abundance peak. We identify the astrophysical conditions required to produce this peak region through weak r-processing and point out the neutron capture rates that most strongly influence the final abundance pattern.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Doppler Broadening and Other Temperature Effects

In this paper I attempt to document what I have learned and still remember about temperature effects, particularly Doppler broadening, on nuclear data, as it applies to ENDF formatted data. My focus is on the SIGMA1 method of Doppler broadening that I developed over 50 years ago. However, I do realize that there are currently many different methods used in computer codes, and I have tried to keep my discussion general as it applies to ALL of these methods. In particular, ALL of the MYTHS I describe below apply to all Doppler broadening methods: they all solve the parabolic diffusion equation in spherical geometry in (speed, reaction rate) versus temperature. These methods differ only in how the cross section is represented. So, at 85 years old here is what I still remember.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

TREAT Neutronics Analysis and Design Support, Part I: Multi-SERTTA

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. Calculations were performed to support analysis of a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.74 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sensitivity and uncertainty analysis of PWR spent fuel observables to operational and model parameters

Sensitivity and uncertainty analyses of spent nuclear fuel (SNF) observables - decay heat, neutron and γ-ray emission rate - to operational and model parameters have been performed. A 2D model representing a typical PWR 17x17 UO{sub 2} fuel assembly has been taken as reference. The Serpent code and ENDF/B-VII.1 evaluated nuclear data library have been used for the analyses. Relative uncertainty of decay heat, neutron and γ-ray emission rates due to Monte Carlo counting statistics, as well as the number of fuel pin radial regions required for an accurate characterization of SNF neutron emission, have been estimated. A parametric study has been carried out to investigate the impact of the homogenization to an average value of the fuel and moderator temperature, boron concentration power and burnup on the prediction of the selected observables. In addition, linear sensitivity coefficients have been calculated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

IAEA CRP on HTGR Uncertainties in Modeling: Assessment of Phase I Lattice to Core Model Uncertainties

Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented by the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1994

Gamma-ray and millimeter-wave emissions from the 1991 June X-class solar flares

We have studied the spectacular 1991 June X-class flares using gamma-ray data from the Charged Particle Detectors (CPDs) of the Burst and Transient Source Experiment (BATSE) on the Compton Gamma Ray Observatory (CGRO) and 80 GHz millimeter data from Nobeyama, Japan. The CPDs were the only CGRO instrument that did not saturate during the extremely intense 1991 June 4 flare. We have shown that for this flare the CPDs respond to MeV photons, most of which are due to bremsstrahlung produced by relativistic electrons at the Sun. We have further shown that the gamma-ray and millimeter observations agree numerically if the 80 GHz radiation is gyrosynchrotron radiation produced by trapped electrons and the gamma rays are thick-target bremsstrahlung due to electrons precipitating out of the trap. The requirement that the trapping time obtained from the numerical comparison be consistent with the observed time profiles implies a magnetic field between about 200 and 300 G and an electron spectral index between about 3 to 5. By comparing the CPD observations with both the 80 GHz data and nuclear line data from the Energetic Gamma Ray Experiment Telescope (EGRET) and the Oriented Scintillation Spectroscopy Experiment (OSSE) on CGRO for the flares of June 4, 6, 9, and 11, we found that the ratio of the CPD counts to both the millimeter flux densities and the nuclear line fluences decreases with decreasing flare heliocentric angle. All of these flares were produced in the same active region. We interpreted this result in terms of a loop model in which the gyrosynchrotron emission is produced in the coronal portion of the loop where the electrons are kept isotropic by pitch angle scattering due to plasma turbulence, while the bremsstrahlung is produced by precipitating electrons that interact anisotropically. We found that the trapping time in the coronal portion is time dependent, reaching a minimum of about 10 s at the peak of the CPD count rate. We suggested the damping of the turbulence as a possible reason for the variation of the trapping time. turbulence as a possible reason for the variation of the trapping time.

Ramaty, Reuven↗
DOE OSTI2015

Reducing uncertainties for short lived cumulative fission product yields

Uncertainties associated with short lived (halflives less than 1 day) fission product yields listed in databases such as the National Nuclear Data Center’s ENDF/B-VII are large enough for certain isotopes to provide an opportunity for new precision measurements to offer significant uncertainty reductions. A series of experiments has begun where small samples of 235 U are irradiated with a pulsed, fission neutron spectrum at the Nevada National Security Site and placed between two broad-energy germanium detectors. The amount of various isotopes present immediately following the irradiation can be determined given the total counts and the calibrated properties of the detector system. The uncertainty on the fission yields for multiple isotopes has been reduced by nearly an order of magnitude.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗