Evaluated Nuclear Data File Research Articles

Excitation functions of (n, p) reactions for the production of iodine isotopes used in medical applications

The cross sections of 124Xe (n, p)124I, 125Xe (n, p)125I, 126Xe (n, p)126I, 128Xe (n, p)128I, 129Xe (n, p)129I, 130Xe (n, p)130I, 131Xe (n, p)131I, 132Xe (n, p)132I, 133Xe (n, p)133I, 134Xe (n, p)134I and 135Xe (n, p)135I for production of iodine isotopes were calculated at neutron energy ranges between 1 and 20 MeV with the help of three computer nuclear codes (EMPIRE 3.2.3, ALICE-ASH and TALYS 1.95). Furthermore, the cross sections were calculated by the empirical formula derived in some previous studies at 14.5 MeV. The previous calculated data from the evaluated data from the Evaluated Nuclear Data File (ENDF), such as ENDF/B-VIII.0; JEFF-3.3; ROSFOND-2010; CENDL-3.2; JENDL-4.0; JENDL-5.0; and TENDL-2021, as well as the experimental nuclear reaction cross-section data (EXFOR) about the above reactions, have been collected. In general, the calculated results by computer codes are in good agreement when compared with earlier empirical formulas, the measured values in the EXFOR library, and the evaluated data (ENDF/B-VIII.0; JEFF-3.3; JENDL-4.0; JENDL-5.0; ROSFOND-2010; CENDL-3.2; and TENDL-2021) with some exceptional.

Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor

The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.

Validation of the SCALE/Polaris−PARCS Code Procedure with the ENDF/B-VII.1 AMPX 56-Group Library: Pressurized Water Reactor

This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors.

NEBOAS: A Neutron yiElds Based On AcceleratorS application

The web-based application NEBOAS was developed to calculate the neutron yield of the accelerator-based neutron source MONNET (MONo-energetic NEutron Tower) at JRC-Geel. Neutrons are produced with p and d-induced reactions on particular targets. NEBOAS provides double differential neutron yields, as a function of the angle of neutron emission and energy. It provides also integrated neutron yields, and neutron energy spectra. Any projectile energy may be utilized, as long as it is covered by the nuclear data available on thin targets (that just degrade the projectiles' energy) or thick targets (that stop the projectiles). The calculation employs reaction cross-section evaluations from Evaluated Nuclear Data File (ENDF) or compilations of experimental measurements from the Experimental Nuclear Reaction Data (EXFOR), and stopping powers as recommended in the National Institute of Standards and Technology (NIST) or the Stopping and Range of Ions in Matter (SRIM-2013) databases.

Early Nuclear Fusion Cross-Section Advances 1934–1952 and Comparison to Today’s ENDF Data

We describe the advancing knowledge of fusion cross sections from 1934 through the development of the first thermonuclear tests fielded by Los Alamos (the singular entity denoted Los Alamos Laboratory/Los Alamos Scientific Laboratory/Los Alamos National Laboratory at different times is designated “Los Alamos” in this paper) in the Pacific in 1951–1952; this technical history has not been previously documented. We compare these nuclear reaction cross sections to the current state of their knowledge as codified in the Evaluated Nuclear Data File (ENDF) databases, focusing on the Big Five reactions: 3H ( d , n ) 4 He, 3He ( d , p ) 4 He, 2H ( d , n ) 3 He, 2H ( d , p ) 3 H, and 3H ( t , 2 n ) 4 He. At Oppenheimer’s July 1942 University of California, Berkeley, “galaxy of luminaries” conference, Konopinski suggested that the cross section for 3H ( d , n ) 4 He “DT” could be large, and although Teller described this as an “inspired guess,” we provide evidence instead suggesting that Konopinski knew of a 1938 measurement by Ruhlig that secondary DT reactions were “exceedingly probable.” Bethe’s direction that the DT cross section should be measured at Purdue University (Purdue) in 1943 led to the remarkable and unexpected finding that the DT cross section exceeds deuteron-deuteron (DD) by a factor of 100. This was a game-changing result, making Teller’s dream, i.e., the terrestrial production of fusion energy, feasible. Eyewitness accounts are transcribed from the earliest discoveries of the large magnitude of the resonant DT cross section. A description is given of the Manhattan Project’s early 1942–1944 DD measurements at the University of Chicago, the 1943 DT measurements at Purdue, and the subsequent 1945–1946 DD and DT measurements at Los Alamos. The Los Alamos experiments, led by Bretscher, were the first to extend to very low incident ion center-of-mass energies in the 6- to 50-keV range needed in applications and the first to identify, characterize, and document the 3/2+ “Bretscher state” responsible for the resonance-enhanced DT cross section. The early measurements were based on thick-target experiments that required a knowledge of hydrogen-isotope stopping powers, much of which was informed by 1930s German studies. We end with the high-accuracy APSST (named for Arnold, Phillips, Sawyer, Stovall, and Tuck) measurements at Los Alamos, 1951–1952. The very first 1942–1946 measurements were accurate to about 50% or somewhat better, but by the early 1950s, the cross sections were determined much more accurately, to within a few percent of our best values today, which come from R-matrix Energy Dependent Analysis (EDA) code analyses of the data, most notably the very accurate 1980s–1990 Los Alamos DT and DD fusion data from Jarmie and Brown. We show that Fermi, in his 1945 Los Alamos lectures, anticipated the S-factor (for the DT cross section), which is a concept widely used later in nuclear astrophysics. To this long abstract, we add a final tidbit: Marshall Holloway, a coauthor on the first-ever 1943 DT cross-section measurement at Purdue, went on to lead the engineering and fabrication of the first H-bomb test, Ivy Mike.

Empirical formula for (n, f) reaction cross sections of Uranium isotopes at 1–20 MeV neutrons

This article describes how to calculate neutron-induced fission reaction cross sections using a proposed empirical formula and the EMPIRE 3.2.3 and TALYS 1.95 computer codes for Uranium isotopes up to the third fission plateau. In this study, the excitation functions of 232U (n, f), 233U (n, f), 234U (n, f), 235U (n, f), 236U (n, f), 237U (n, f) and 238U (n, f) nuclear reactions were calculated at 1–20 MeV neutron energies. The results were contrasted to measured values from the Experimental Nuclear Reaction Data (EXFOR) as well as the evaluated data from Evaluated Nuclear Data File (ENDF) such as (EAF-2010, JEFF-3.3, ENDF/B-VIII.0 and TENDL-2019). Overall, the calculated, experimental, and evaluated fission cross-sections are in concordance.

Computational and experimental determination of slow neutron flux by using Li-6 foil and nuclear track detector

In the present study, slow neutron flux was determined with an experimental method based on 6Li(n, t)4He reaction. 6Li coated foils were mounted on track etch detectors without any gap and they were irradiated with an Am–Be neutron source for 6 h. The track detectors were affixed on a polyethylene plate which has a thickness of 2.0 cm for neutron moderation. Following irradiation, the track detectors were etched, the tritium tracks were counted with an in-house developed track counting system, and the total number of reactions was computed. The difference in particle ranges and track diameters on the Cr-39 detector allowed for the precise distinguishing of tritium, alpha, and neutron particles. The neutron spectrum at the 6Li coated foil position was simulated with MCNP5 using Evaluated Nuclear Data File (ENDF) cross sections of 6Li(n, t)4He reaction. The total reaction numbers obtained with experimental and computational methods are compatible with each other.

Green and low-carbon upcycling of ludwigite: Prepared shields against nuclear radiation hazards and shielding mechanism

The complex composition of ludwigite makes its comprehensive usage difficult. Given the rapid development of ores in nuclear shielding materials, ludwigite was proposed to prepare epoxy resin matrix nuclear shielding composites at room temperature, which could achieve green and low-carbon upcycling. The shielding performance of neutron and gamma-ray sources was evaluated using the 241Am–Be neutron source, the Co-60 gamma-ray source and the Evaluated Nuclear Data File (ENDF). Scanning electron microscopy was used to examine the microstructures. Increasing the ludwigite weight and thickness of the shielding composite increases the shielding percentage of the prepared composites. The prepared shielding composites have macroscopic cross-sections of 0.92–1.89 cm−1, which are comparable to SWX-201, which contains 5% boron. Boron absorption and hydrogen elastic scattering are the primary shielding mechanisms for neutron attenuation in 10−11–10−6 MeV. The shielding composites’ gamma-ray shielding performance is comparable to that of water and ordinary concrete. The work is significant not only for using ludwigite but also for the shielding mechanism analysis of complex multi-element composites.

Complete β -decay patterns of Cs142,Ba142, and La142 determined using total absorption spectroscopy

Background: The β decays of fission products produced in nuclear fuel are important for nuclear energy applications and fundamental science of reactor antineutrinos. In particular, nuclear reactor safety is related to the decay modes of radioactive neutron-rich nuclei, primarily via the emission of γ rays, neutrons, and electrons. Additionally nuclear reactors are the most powerful man-made source of antineutrinos emitted during the β decay of fission products. These antineutrinos are used to inspect fundamental properties of leptons as well as informing reactor operation. However, the majority of data on complex decays of fission products collected in the evaluated nuclear data repositories like Evaluated Nuclear Structure Data File (ENSDF) and Evaluated Nuclear Data Files (ENDF) are based on low-efficiency and often incomplete measurements resulting in questionable reference reactor antineutrino flux predictions, see the analysis by [Nichols, J. Nucl. Sci. Technol. 52, 17 (2015)]. Various assessments like the one done under the auspices of the [Yoshida et al., Assessment of Fission Product Decay Data for Decay Heat Calculations: A report by the Working Party on International Evaluation Co-operation of the Nuclear Energy Agency Nuclear Science Committee (Nuclear Energy Agency, Organization for Economic Co-operation and Development, Paris, France, 2007), Vol. 25], as well as by [Sonzogni, Johnson, and McCutchan, Phys. Rev. C 91, 011301(R) (2015)] and [Dwyer and Langford, Phys. Rev. Lett. 114, 012502 (2015)], list the A=142 isobars with high cumulative fission yield among the important nuclei where data for reactor decay heat and/or antineutrino production should be verified and/or improved. Purpose: Our goal is to improve the quality of β-decay measurements and evaluate the impact of modified decay schemes on reactor decay heat and antineutrino energy spectra, for fission products along the A=142 isobaric chain. This work is an in depth follow-up on [Rasco et al., Phys. Rev. Lett. 117, 092501 (2016)]. which presented briefly the impact of the corrected decay scheme of 142Cs. Here, we extend the data to full isobaric decay chain including the daughter nuclei, 142Ba and 142La, and present more details on the 142Cs results. Method: The decays of neutron-rich isobars of mass A=142 produced by means of proton-induced fission of U238 were measured using the Modular Total Absorption Spectrometer (MTAS) array on-line at the mass separator and Tandem accelerator at Oak Ridge National Laboratory. Results: The β-decay schemes for 142Cs and 142La were modified with respect to the nuclear data repositories. A small β-delayed neutron branching ratio for 142Cs emitter was remeasured as 0.10−3+5 %. Improved precision on the measured half-lives is reported. Small corrections to the low-energy decay of 142Ba are made. The β-decay patterns for 142La and 142Cs are presented. The decay heat release and cross section for the detection of reactor antineutrinos are deduced and compared to earlier results. Conclusions: The β-feeding pattern for 142Cs having decay energy value Qβ of over 7 MeV was substantially modified with respect to the current ENSDF entry. Smaller changes were encountered for 142La, but since this A=142 isobar also has a large cumulative yield in fission, the changes influence both decay heat and the antineutrino spectra. The previously known β intensities for 142Ba decay (Qβ value of 2.2 MeV) were verified and slightly modified. Overall, increased decay heat values and lower flux of antineutrinos interacting with matter are presented.12 MoreReceived 13 October 2022Accepted 14 February 2023DOI:https://doi.org/10.1103/PhysRevC.107.034303©2023 American Physical SocietyPhysics Subject Headings (PhySH)Research AreasBeta decayFissionLifetimes & widthsNuclear reactorsNuclear structure & decaysProperties90 ≤ A ≤ 149TechniquesSpectrometers & spectroscopic techniquesNuclear Physics

Modernization efforts for the R-Matrix code SAMMY

The R-Matrix code SAMMY [1] is a widely used nuclear data evaluation code focused on the resolved range, which includes corrections for experimental effects. The code is still mostly written in FORTRAN 77 and uses a memory management system suitable for the time of its initial writing in 1984. A modernization effort is underway to update the code to modern software development practices. A continuous-integration testing framework was added to automate the large existing set of test cases. Improvements in memory management were implemented to make the code easier to maintain and enable enhancements. The resonance parameters and covariance information are now stored in C++ objects shared by SAMMY and AMPX [2], which is the processing code that generates nuclear data libraries for SCALE [3]. Further plans include switching to the Evaluated Nuclear Data File (ENDF) reading and writing routines in AMPX because these routines are more robust, easier to maintain, and support more features. Support for the new Generalized Nuclear Database Structure (GNDS) format [4] is also of interest. GNDS will share not only the resonance parameters but also the parameters associated with experimental correction in GNDS. The data are currently available in a binary SAMMY format, and the ability to export them to GNDS would make them more widely available and shareable. The next step will be to use the same resonance processing code at 0K in AMPX and SAMMY as an available formalism. Then, any improvements in the formalism can immediately be tested in SCALE because the reconstruction in AMPX will use the same cross section model. The new data library can then be used for testing using the VALID Benchmark suite [5] or other suitable benchmark suites.

IMPACT OF THE ENDF/B-VIII.0 LIBRARY ON MCNP6.2-COMPUTED AMBIENT AND PERSONAL DOSE EQUIVALENT CONVERSION COEFFICIENTS.

The 2018 release of the Evaluated Nuclear Data File (ENDF)/B-VIII.0 library initiated several examinations of the impact of the revised nuclear data on Monte Carlo models. This study used version 6.2 of the Monte Carlo N-Particle® code (MCNP6.2) with the ENDF/B-VI and ENDF/B-VIII.0 libraries to compute conversion coefficients for fluence-to-ambient and fluence-to-personal dose equivalent from neutron energy groups described by an International Atomic Energy Agency 53-bin structure and 47 monoenergetic neutron sources listed in International Commission on Radiological Protection (ICRP) 74. The MCNP6.2 models with ENDF/B-VI data were validated against results published in 2005 by Veinot and Hertel. Conversion coefficients computed with MCNP6.2 and ENDF/B-VIII.0 slightly underestimated the ICRP 74 values but were within ICRP-specified tolerances and do not justify revising the ICRP coefficients. Conversion coefficients for personal dose equivalent were computed with MCNP6.2 and ENDF/B-VIII.0 at four angles of incidence greater than those available in ICRP 74. These new coefficients are relevant to radiation protection studies of interest to US defence-focused organisations.

Analysis of Fission Fragments Contributors on Total Decay Heat of Thermal Neutron-Induced Fission of U-235

Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made together with the analysis of the most important isotopes affecting the decay heat. They are useful in designing the necessary nuclear safety for spent fuels, and their importance cannot be overlooked in the designs of transporting fuel storage containers as well as in the management of the radioactive waste generated. In this paper, by using MATLAB, the decay heat after the thermal fission of a U-235 nucleus was numerically calculated by solving linear differential equations for all the buildups/decays of the fission products. Also, the most contribution of radioactive isotopes to the decay heat was analyzed by using Microsoft Excel. The most influential isotopes were deduced in two ways; either by calculating the most influential isotopes at specific times, or by determining the largest influences in a cumulative manner. All required nuclear data such as decay constants their branching ratios, independent fission yield, and average α-, β-, and γ-energies released per disintegration of any nuclide, have been extracted from the latest version of the Evaluated Nuclear Data Files (ENDF) database ENDF/B-VIII.0. The two different methods used showed a difference in the contributing isotopes, which is logical for the difference in the method of calculation. The first method is suitable for instantaneous data while the second method is more suitable when there is a need to know the cumulative calculations. In sum, we can say that both methods complement each other, and neither of them can be dispensed with in the accurate calculations related to transportation and storage of spent fuel.

Comparison of MCNP thermal scattering laws to inelastic neutron scattering data

Neutron transport computations employ thermal scattering laws (TSLs) to account for the influence of material structure and collective excitations at low energy (≲ 10 eV). Recent work has been undertaken to improve TSL accuracy and expand the number of available materials in the Evaluated Nuclear Data File (ENDF). However, comparison of TSL to measured data is often sparse. We present a method to evaluate TSL accuracy in the Monte Carlo N-Particle (MCNP) code by simulating inelastic neutron scattering (INS) measurement and comparing to data. The method is briefly described with example application to the recently released ENDF-VIII TSL library for polyethylene as well as a cryogenic PE TSL we are separately developing.

Uncertainty propagation of prompt fission neutron spectrum for physics analysis of fast and thermal reactors

This paper presents a strategy for processing the covariance matrix of the prompt fission neutron spectrum (PFNS) and propagating the PFNS uncertainties to the neutronic calculation of fast and thermal reactors. Due to their great significance to the accurate prediction of nuclear criticality, the sensitivity and uncertainty of core parameters for PFNS are analyzed in this study. The propagation of PFNS uncertainties from the evaluated nuclear data files (ENDF) to the core parameters is divided into two phases. In the first phase, the covariance data in the ENDF is transformed into the multi-group form, in which the “zero-sum” constraint is considered and a corresponding functional module is developed in the updated NECP-Atlas nuclear data processing code. In the second phase, the neutron transport code is coupled as a “black box” to propagate the uncertainties in the multi-group form to the core parameters. Sensitivity and uncertainty (S&U) analysis function is implemented in the NECP-UNICORN uncertainty quantification code. As for its application, S&U analysis for PFNS has been performed to the fast and thermal reactors. The numerical results indicate that the influence of PFNS uncertainties to the confidence interval of reactivity prediction and power distribution should be considered in the nuclear reactor analyses.

Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

An inter-comparison between ENDF/B-VIII.0-NECP-Atlas and ENDF/B-VIII.0-NJOY results for criticality safety benchmarks and benchmarks on the reactivity temperature coefficient

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutual-comparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

Development of a Fast-Spectrum Self-Powered Neutron Detector for Molten Salt Experiments in the Versatile Test Reactor

The self-powered neutron detector (SPND) is a widely used flux monitor in thermal nuclear reactors. Although this is a mature technology, the current state of the art is tuned for a thermal neutron spectrum, so many of the devices currently in use lack sensitivity to fast neutrons. Because current in SPNDs is produced through nuclear reactions with the neutron flux inside a reactor, sensitivity in SPNDs is determined by the neutron cross section of the neutron-sensitive portion of the detector, termed the emitter. This neutron cross section drops by orders of magnitude between thermal and fast neutron energies for many emitters in currently used SPNDs, with a corresponding drop in current from the detector. This paper discusses efforts to develop a fast-spectrum self-powered neutron detector (FS-SPND) that is sensitive to neutrons with energies ranging from 0.025 eV up to 1 MeV. An in-depth analysis of Evaluated Nuclear Data File (ENDF)/B-VII.1 neutron-capture cross sections was performed, and four new materials were identified that are suitable emitter candidates for use in measuring fast neutrons. All four materials are stable mid-shell nuclei in the region between doubly magic132Sn and208Pb. Each candidate was simulated with the Geant4 Monte Carlo simulation toolkit to optimize overall detector efficiency.

Calculations of astrophysical reaction rates using ENDF/B-VIII.0 library

Recent observations of neutron stars merger (GW170817) renewed interest in stellar nucleosynthesis. Stellar nucleosynthesis modeling requires fully traceable, unbiased, high fidelity nuclear data. The Evaluated Nuclear Data File (ENDF) libraries contain complete collections of reaction data sets over the nuclear industry standard 10−5 eV - 20 MeV energy span. For the first time, the (n,γ), (n,fission), (n,p), and (n,α) astrophysical reaction rates were computed within the 0.01-10 GK range of temperatures using the ENDF/B-VIII.0 library, and nuclear Reaction rate Library for astrophysics (REACLIB) fit parameters were deduced. The present results were used to estimate the slow neutron capture timescale. These findings demonstrate the potential astrophysical applications of the ENDF libraries and the complementary relations between the nuclear industry and astrophysics data developments.

The use of the ENDF library for nucleosynthesis studies

Stellar nucleosynthesis modeling studies would significantly benefit from the use of fully traceable, documented and unbiased nuclear data. The nuclear industry standard Evaluated Nuclear Data File (ENDF) libraries contain extensive collections of reaction data sets relevant to astrophysics. For the first time rapid-and slow-neutron capture, r- and s-process, respectively, abundances were computed from ENDF/B-VIII.0 and TENDL-2015 libraries and compared with available data. The current results highlight mutually beneficial relations between nuclear industry and nuclear astrophysics data developments.

Analytical determination of radioactive strontium and cesium by Thermal Ionization Mass Spectrometry

Thermal Ionization Mass Spectrometry (TIMS) has been evaluated for the detection of the radioactive isotopes of Sr and Cs. The commercial instrument (i.e., a Thermo Scientific Triton) was investigated for the analysis of isotopic ratios of 89Sr/90Sr and 135Cs/137Cs in the presence of atomic isobars (89Y and 90Zr for 89,90Sr analysis and 135,137Ba for 135,137Cs analysis). The decontamination achievable instrumentally was examined by isotopic ratio measurements of 89Y/88Sr and 90Zr/88Sr for Sr and 135Ba/133Cs and 137Ba/133Cs for Cs. The decontamination found was at or above 2.0E+8 for 90Zr from 88Sr, while the Y demonstrated a temperature dependence as it sublimed from the filament but remained better than ≈ 5E+7. The decontamination of Ba from Cs did not show any temperature dependence and remained above 5E+6 and 8E+6 for 135Ba and 137Ba from 133Cs, respectively. Two standard fusion procedures one with sodium hydroxide (NaOH) plus sodium peroxide (Na2O2) flux, and the second used lithium tetraborate (Li2B4O7) plus lithium metaborate (LiBO2) flux were evaluated for preparation of sample matrices prior to performing chemical separations. Ammonium molybdophosphate-polyacrylonitrile (AMP-PAN) and Sr-spec resin were used to isolate the Cs and Sr, respectively from a prepared background matrix (i.e., Montana Soil). A graded approach, increasing in stable background isotopes, was performed to monitor the chemical and instrumental response. The radioisotopes of Sr and Cs were produced by thermal neutron irradiation of a highly enriched uranium foil. Even though the irradiated sample was not a certified standard it does provide accurate expectation values via the published cumulative fission yield nuclear data in the Evaluated Nuclear Data Files (ENDF) [1]. The intra-element isotopic ratio results presented in this work for 89Sr/90Sr and 135Cs/137Cs agree with the published data at 1σ. Furthermore, the uncertainty of the isotopic ratio measurements with TIMS was a factor of 5–10 improved compared to these published values.