Nuclear Data File Research Articles

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.

МАТЕМАТИЧЕСКАЯ МОДЕЛЬ ПЕРВИЧНЫХ РАДИАЦИОННЫХ ПОВРЕЖДЕНИЙ БОРОСИЛИКАТНОГО СТЕКЛА, ПРЕДНАЗНАЧЕННОГО ДЛЯ ИММОБИЛИЗАЦИИ РАДИОАКТИВНЫХ ОТХОДОВ

The paper describes the physical basis of the mathematical model of radiation-ion damage in solid matter. The following stages of defect generation are considered: primary knocked out atoms, atomic collision cascades (dynamic stage of damage formation) based on 6 types of cascade functions. The software implementation of the described model is included in the RTS&T code designed for statistical modeling of multi-particle (200 types of particles, resonances and ions) radiation transport in heterogeneous 3D geometries in the energy range up to 20 TeV. In the region of low energies, the RTS&T transfer model is based on direct use of all the information contained in the evaluated nuclear data files submitted in the ENDF-6 format. The report compares the results of numerical modeling with experimental data.

ABSOLUTE MEASUREMENT OF 14.57MeV NEUTRON FLUENCE RATE USING PROTON RECOIL NEUTRON TELESCOPE.

A single stage vacuum-type proton recoil neutron telescope (PRT) was used for accurate measurement of 14.57MeV neutron fluence rate from an indigenously developed D-T neutron generator at Purnima, BARC. The telescope consists of a polyethylene radiator having 4cm diameter and CsI (Tl) scintillation crystal having thickness 1.5mm and 4cm diameter separated by 20.5cm kept in a vacuum chamber. The neutron detection efficiency of the telescope for 14.57 MeV neutrons was calculated analytically using n-p scattering cross section data from Evaluated Nuclear Data File VII and also evaluated using fluka simulation. The relativistic transformation of n-p differential scattering cross section from centre-of-mass to laboratory system was used for calculating the efficiency of PRT. The 14.57 MeV neutron fluence rate was also measured using copper foils. The comparison of fluence rate measured using PRT and copper foil activation techniques is presented in this paper. The total uncertainty in measurement using PRT and copper foil activation technique is found to be 3.93 and 6.97%, respectively.

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.

Excitation function calculations in proton-induced nuclear reactions on Zn and Cu upto 80 MeV

Nuclear reaction model codes like the EMPIRE-3.2, the ALICE-1991 and the TALYS-1.9 have been used to calculate the excitation functions of 63Cu(p,n)63Zn, 65Cu(p,n)65Zn, 66Zn(p,n)66Ga, 67Zn(p,n)67Ga, 67Zn(p,2n)66Ga, 68Zn(p,n)68Ga and 68Zn(p,2n)67Ga reactions for incident proton energy range of 1–80 MeV. These codes compute reaction cross section using the direct, the compound and the pre-equilibrium (PEQ) processes as major nuclear reaction mechanisms. The computed results have been analyzed and compared with the available experimentally measured data and with the evaluated nuclear data files. Good agreements have been found between theoretically calculated and measured data.

Monte Carlo calculations of different types of burnable poison rods used in PWR

Abstract The effect of burnable poisons on neutronic parameters of PWR is an important issue in nuclear fuel design. Burnable poisons are used to control the reactivity of the core and to improve fuel utilization. It classified to Burnable Poison Rods (BPRs) and Integral Burnable Absorbers (IBAs). This work focuses mainly on Burnable Poison Rods. The main purpose with this paper is to study and analyze the effect of different materials of Burnable Poison Rods (BPRs). The investigation was performed using MCNP6 code together with the Evaluated Nuclear Data Files, (ENDF/B)-VII.1. The investigation included study of the effect of: presence of BPRs and the changes of BPRs number. The calculations were carried out for two types of BPRs; Pyrex and WABA rods. The reactivity and radial power distribution and its variations vs. burnup were carried out up to50 Gwd/MTU. The results evidence that at BOC, the Pyrex and WABA have large compensation of reactivity; Pyrex has more negative reactivity than WABA. Lower power is obtained at the pin near the poisons rods due to large absorption cross section of boron material. The maximum pin power was located at the edge of the assembly due to the absorbing effect of BPRs.

Neutronic Analysis of Modular Gas-cooled Fast Reactor for 5-25% of Plutonium Fuel using Parallelization MCNP6 Code

Modular Gas-cooled Fast Reactor (GFR) is one of six advanced reactor concepts set by the generation IV international forum. Modular GFR has the potential for use actinide recycling and closed fuel cycle as well as applying fast reactor, using helium gas as the main coolant, high working temperature and low void reactivity effect. The neutronic analysis of nuclear reactor means behavior study of subatomic particles that interact with matter. In this paper, the feasibility of plutonium fuel in modular Gas-cooled Fast Reactor (GFR) was investigated. The Monte Carlo method has advantages in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data but requires a highly computation time. Since the progress of high performance computing, the reactor physicist community began proposing to use Monte Carlo method for nuclear reactor simulation through the parallelization of calculations. The GFR feasibility design study will carried out with plutonium fuel as fuel cycle inputs with 5-25% of fissile contain. The most important neutronic parameters characterizing of GFR core are determined for beginning of life (BOL) and during burnup calculation conditions. The results of calculation series in parallel computing give the good agreement will be faster of calculation time when more threads. Materials of (U-Pu)C and (U-Pu)N fuel are the good candidates to be chosen in GFR research that give keff more than 1.2 in fissile contain 20%Pu. The variation fissile contain gives the linearity with keff. In depletion simulation, the core reactor still in critical during 20 years operation, burn up values linear with operation time and mass evolution of plutonium and uranium from start-up core to equilibrium core.

Assessment of neutron kerma coefficients and its calculation methodologies using recent basic ENDF-6 libraries

The generation of heat due to interaction of neutrons with matter is an important aspect in the designing of nuclear reactors. The estimation of neutron heating depends profoundly on the evaluated nuclear reaction parameters. Hence, within the purview of radiation damage of materials, it is necessary to investigate the effects of improvements in nuclear databases on the neutron heating rates under various neutron spectra. There are two well established methods to quantify the neutron kerma coefficients in matter: energy balance method and direct method. The well established code like NJOY21 gives the energy balance neutron kerma. In this study, the importance and performance of the direct method has been investigated, as the recent general purpose basic evaluated nuclear data files provide more complete database. Though both the methods predict closely the reported measured values of neutron kerma coefficients for structural elements, significant differences are observed between these two methods at higher energies where contributions from charged particle emitting reactions become important. The direct method is found to be more meaningful as it always ensures positive kerma coefficients and helps to separately identify the contributions from light charged particles and recoil nucleus. The neutron spectrum dependence of neutron kerma from both the methods are also investigated. Further, the nuclear data uncertainties in neutron heating are quantified by the Total Monte Carlo (TMC) methodology using the random ENDF-6 files in TENDL-2015 and the results show that the nuclear data uncertainties are in the range of 9–29%.

The High-Energy Intra-Nuclear Cascade Liège-based Residual (HEIR) nuclear data library

It is standard practice for nuclear data files to include tabulated data for distinct reaction channels for incident energies up to 20-30 MeV. Above these energies, the assumptions implicit in the definition of individual channels break down and event generators are typically used within codes that simulate nuclear observables in applications. These offer robust simulation of the physics but increase the computational burden. So-called ‘high-energy’ nuclear data files have been produced, but the well-known libraries are more than a decade old and rely upon models developed many years before their release. This presentation describes a modern library with a high level of production automation that offers regular updates as the models it is based upon are improved. The most recent versions of the intra-nuclear cascade and de-excitation models available within Geant4 were used to generate tabulated data of residual nuclide production. For the first released library, the INCL++5.3 and ABLA version within Geant4 v10.3 were used to calculate over 1012 incident protons over 2095 target isotopes with incident energies up to 1 GeV. These were collated into tabulated data in the international-standard ENDF-6 format. The resulting files were provided as group-wise files and were distributed as HEIR-0.1 with the FISPACT-II version 4.0 release. A second library, HEIR-0.2, has been generated using the new INCL++6.0 and C++ translation of the ABLA07 model available within Geant4 v10.4. Simulations were performed using incident protons, neutrons, deuterons and π±. An improved agreement is observed in the comparison to experimental data not only between the two versions, but against the other well-known high-energy nuclear data files and models available within Geant4. This benchmark includes mass and isotopic distributions, as well as incident-energy dependent cumulative and independent cross sections from the EXFOR database.

In search of the best nuclear data file for proton induced reactions: Varying both models and their parameters

A lot of research work has been carried out in fine tuning model parameters to reproduce experimental data for neutron induced reactions. This however is not the case for proton induced reactions where large deviations still exist between model calculations and experiments for some cross sections. In this work, we present a method for searching both the model and model parameter space in order to identify the ’best’ nuclear reaction models with their parameter sets that reproduces carefully selected experimental data. Three sets of experimental data from EXFOR are used in this work: (1) cross sections of the target nucleus (2) cross sections of the residual nuclei and (3) angular distributions. Selected models and their parameters were varied simultaneously to produce a large set of random nuclear data files. The goodness of fit between our adjustments and experimental data was achieved by computing a global reduced chi square which took into consideration the above listed experimental data. The method has been applied for the adjustment of proton induced reactions on 59Co between 1 to 100 MeV. The adjusted files obtained are compared with available experimental data and evaluations from other nuclear data libraries.

Benchmark experiment with iron slab by Time-of-flight technique at CIAE

In order to validate the evaluated nuclear data, leakage spectra in the range of 0.8 to 15 MeV from samples were measured by time-of-flight (TOF) technique using a D-T neutron source. An experimental system for benchmark validation of nuclear data with slab samples has been set up at China Institute of Atomic Energy (CIAE). In this study, test samples are iron slabs, of which the thickness are 5cm, 10cm and 15cm, and the measured angles were chosen to be about 60° and 120°. By comparing measured leakage spectrum with calculated ones by MCNP-4C code, using the data from the CENDL-3.1, ENDF/B-VIII.0, JENDL-4.0 and JEFF-3.3 nuclear data files, and the comparison was made by the spectrum shape and by the C/E values in different energy regions.

International network of nuclear structure and decay data evaluators

Compilation, evaluation and dissemination of nuclear data are arduous tasks that rely on contributions from experts in both the basic and applied sciences communities whose efforts are coordinated by the International Atomic Energy Agency (IAEA). The Evaluated Nuclear Structure Data File (ENSDF) includes the most extensive and comprehensive set of nuclear structure and decay data evaluations performed by the international network of Nuclear Structure and Decay Data evaluators (NSDD) under the auspices of the IAEA. In this report we describe the recent NSDD activities supported by the IAEA and provide some future perspectives.

Method and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files: Major findings of the NEA WPEC Subgroup 39

Nuclear data adjustments using integral experiments play since several decades a crucial role in providing reactor designers and fuel cycle analysts with nuclear data with reduced and validated uncertainties in order to allow design optimization while meeting safety and economics requirements. Adjustment methods, choice of integral experiments and covariance data assessment should be carefully defined in order to produce credible and physically acceptable adjustments, applicable to a wide range of systems. The OECD-NEA has established a series of Expert Groups since 2005 to investigate the key scientific issues and to provide recommendations for applications. The present paper discusses the major results of the most recent of these Expert Groups together with an indication of the path forward .

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.

New features and improvements in the NEA nuclear data tool suite

The OECD Nuclear Energy Agency (NEA) has developed and maintains several products that are used in the verification and validation of nuclear data, including the Java-based Nuclear Data Information System (JANIS) and the Nuclear Data Sensitivity Tool (NDaST). These integrate other collections of the NEA, including the International Handbooks of benchmark experiments on Criticality Safety and Reactor Physics (ICSBEP and IRPhEP) and their supporting relational databases (DICE and IDAT). Recent development of the JANIS, DICE and NDaST systems have resulted in the ability to perform uncertainty propagation utilising Legendre polynomial sensitivities, calculation of case-to-case covariances and correlations, use of spectrum weighting in perturbations, calculation of statistical results with suites of randomly sampled nuclear data files and new command-line interfaces to automate analyses and generate XML outputs. All of the most recent, major nuclear data libraries have been fully processed and incorporated, along with new visualisation features for covariances and sensitivities, an expanded set of reaction channel definitions, and new EXFOR data types defined by the NRDC. Optimisation of numerical methods has also improved performance, with over order-of-magnitude speed-up in the case of sensitivity-uncertainty calculations.

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.

Bayesian updating for data adjustments and multi-level uncertainty propagation within Total Monte Carlo

In this work, a method is proposed for combining differential and integral benchmark experimental data within a Bayesian framework for nuclear data adjustments and multi-level uncertainty propagation, using the Total Monte Carlo method. First, input parameters to basic nuclear physics models implemented within the TALYS code, were randomly varied to produce a large set of random nuclear data files. Next, a probabilistic data assimilation was carried out by computing the likelihood function for each random nuclear data file based first on only differential experimental data and then on integral benchmark data. The individual likelihood functions from the two updates were then combined into a global likelihood function. The proposed method was applied for the adjustment of n+208Pb in the fast energy region below 20 MeV. The adjusted file was compared with available experimental data as well as evaluations from the major nuclear data libraries and found to compare favourably.

Determination of out-core fuel burnup in TRIGA PUSPATI

The investigation was conducted on the out-core neutron flux and burn-up at irradiated fuel stored in TRIGA PUSPATI research reactor tank. This is required to examine whether the thermal and/or fast neutron flux can influence burn-up of the irradiated fuel stored in the same vicinity of the reactor core, the fuel rack being located 1 m above the core. MCNPX code was used to simulate fast and thermal neutron flux for the reactor operating at 750 kW. In this work, the computational model was created using MCNPX version 2.7 with the evaluated nuclear data file for thermal neutron scattering law data (ENDF7) cross-section data library and using a 10 cm x 10 cm x 10 cm mesh model. The results showed that the axial distribution for thermal neutrons occurred at energy lower than 1 x 10-6 MeV. Thermal neutron traveled at the maximum distance of 78 cm due to thermalization by moderator. Based on the maximum distance traveled by the thermal neutron, the thermal neutron did not reach the storage rack located 1 m from the core, hence there was no burn-up occurring at the irradiated fuel since burn-up can only occur in the thermal neutron region. For fast neutron, the axial distribution energy was higher than 1 x 10-6 MeV and traveled more than 158 cm. The reaction time for the fast neutron was too short to result in burn-up due to its fast travel.

Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo reactor physics codes for the soluble-boron-free SMR cores

Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in reactor physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15% 235U enriched homogeneously mixed all-UO2 fuel and 18% 235U enriched micro–heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13×13 arrangement. The eigenvalue/reactivity (k∞) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B-VII, JEF2.2 and JEF3.1. A good agreement in k∞ values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are ~200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ~450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes.

МОДЕЛИРОВАНИЕ ШУМОВЫХ ЭКСПЕРИМЕНТОВ ДЛЯ ВЕРИФИКАЦИИ ПАРАМЕТРОВ КИНЕТИКИ, ПОЛУЧЕННЫХ ПО ПРОГРАММЕ САПФИР_РФ

New multi-group neutron data libraries and a new scheme of fuel burnup have been developed on the basis of the modern ROSFOND evaluated nuclear data files. The upgraded version has a working title - the SAPFIR_RF code. Noise experiments for defining the ratio of total delayed neutron yield to neutron lifetime, βeff/l, were performed to verify the computational scheme of kinetics parameters implemented in the SAPFIR_RF. These experiments have been performed using the homogeneous assemblies with various isotopic composition of uranium-plutonium fuel. In order to analyze and interpret the experimental data correctly, numerical investigations simulating the experimental conditions and features of measurement results processing were performed; these investigations highlighted the need for systematic correction to be made for measurement results. Taking into account the correction, the calculation results are qualitatively and quantitatively in agreement with measurement data within the measurement accuracy of ±8 %. This paper draws on the modern data of Russian neutron data library, ROSFOND, as well as on G.R. Keepin’s kinetic parameters and evaluated nuclear data of the foreign libraries, ENDF/B-VI and ENDF/B-VII. Kinetic parameters obtained by G.R. Keepin and ENDF/B-VI library give two extreme evaluations of βeff/l ratio; the difference between them averages ~4,5 %. It should be noted that the calculation results received using Keepin's parameters are closer to experimental data in the most cases.