Fission Product Nuclides Research Articles

Evaluation of Improved Contribution Function Method in the Compression of CINDER90 Depletion Library

The improved contribution function (ICF) method based on generalized perturbation theory (GPT) is applied to the compression of CINDER90 depletion library. A series of representative problems for PWR are defined with different fuel materials and operation conditions. According to the characteristics that the reactivity indicator is only meaningful in the neutron radiation calculation while the decay heat indicator is affected by both radiation and decay stage, the ICF method is adopted in the selection of candidate nuclides, then actinides and fission product nuclides of compressed system are obtained. In addition, the absorber nuclides of the compressed system are identified by analyzing the depletion process of the poison with a large absorption cross section separately. Finally, a compressed depletion library consisting of 489 nuclides is generated. The numerical results show that the compressed library maintains high precision in the calculation of k‐inf, nuclide concentration, decay heat power, and pin power. Meanwhile, the use of compressed depletion library leads to a significant reduction of computing resources.

Method for predicting lifetime of ion exchange resin in PWR primary loop based on nuclide distribution measurement

Ion exchange resins are commonly utilized for water treatment in pressurized water reactors' primary loop. These resins have the ability to adsorb radioactive cations present in the primary loop coolant, leading to a purification effect. However, the resin becomes radioactive post-purification and cannot be regenerated, thus requiring regular replacement. As the process of resin replacement exposes personnel to radiation, it is crucial to enhance the resin's utilization. This paper aims to investigate a non-destructive testing method for evaluating the lifetime of resin in pressurized water reactor purification units. It also aims to offer guidance for safe resin replacement. The resin is contained in a column (commonly referred to as ion exchange columns) and consumed from top to bottom, with radioactivity decreasing in intensity from top to bottom. The study utilizes the longitudinal radioactivity distribution characteristics of ionic resins to build a gamma-ray detection system with high resistance to interlayer interference and good longitudinal resolution. In this study, a two-stage resin column was utilized to capture common fission product nuclides and corrosion-activated nuclides, such as 60Co, 137Cs, 54Mn, and 124Sb, from the effluent of a pressurized water reactor. The distribution of these nuclides was measured using this device. The results indicate that the activity levels of 60Co, 137Cs, and 54Mn reached their peak in the column before rapidly declining. The height of the resin layer corresponding to the highest activity level of each nuclide varied, indicating the resin's adsorption selectivity. After processing an additional 10 tons of wastewater, a second measurement was taken which showed that the failure range of the resin expanded downwards as the treated water volume increased. The entire distribution curve of the nuclides within the exchange column also shifted upwards and to the right. By comparing the first and second observations, we discovered that the bottom part of the resin will fail when the activity of the most penetrating nuclide in the wastewater surpasses a specific threshold. This research suggests a novel approach to predicting the lifetime of resin, which generalizes the trend of radionuclide intensity inside the ion exchange column.

Validation and application of the Dragon5 lattice code for neutronics and burnup analysis of VVER-1000 pin cell and assembly model

This work deals with the validation and assessment of neutronics and burnup analysis capacity of the deterministic lattice code Dragon5 for VVER-1000 hexagonal pin and assembly model configuration. Various VVER-1000 hexagonal pin-cell benchmark cases including the fresh uranium fuel and mixed oxide fuel of ORNL are simulated and compared with the HELIOS and OpenMC code. The data library differences of certain nuclides will lead to large deviations of the results. Accurate results can be presented by the Dragon5 lattice code when compared with OpenMC code. The OECD LEU and MOX assembly benchmarks are computed to further validate the prediction ability for VVER-1000 assembly, reasonable agreement between Dragon5 solutions and published values of other codes is observed for burnup parameters including infinite multiplication factors, concentrations of important actinides and fission product nuclides. Moreover, the post-irradiation of UO2 sample experiment of the Russian Kalinin-1 VVER-1000 reactor is simulated and compared with laboratory measured values to provide the experiment validation of the Dragon5 lattice code. The predicted concentrations of actinides up to Cm244 are compared with HELIOS and Serpent Monte Carlo code. The results show that the Dragon5 lattice code presents better prediction accuracy for Calculated-to-Experimental ratios (C/E's) of most actinide concentrations than the HELIOS and Serpent code. Therefore, the École Polytechnique de Montréal Dragon5 lattice physics code can be applied to the neutronics and burnup analysis of VVER-1000 hexagonal pin-cell and assembly model.

Nondestructive and destructive assay for forensics characterization of weapons-grade plutonium produced in LEU irradiated in a thermal neutron spectrum

A post-irradiation examination (PIE) of an enriched uranium sample was performed to advance a nuclear forensics methodology, based on intra-elemental nuclide ratios. The PIE was carried out on a few milligrams of low enriched uranium dioxide (LEUO2) sample, irradiated to a burnup of 1 GWd/MTU. The PIE of the LEUO2 sample provided the concentration of selected plutonium and fission product nuclides (239Pu, 240Pu, 241Pu, 91Y, 95Zr, 95Nb, 103Ru, 133Cs, 134Cs, 135Cs, 137Cs, 136Ba, 138Ba, 140Ba, 140La, 141Ce, 144Ce, 149Sm, 150Sm, 152Sm, 153Eu, and 154Eu). The PIE supported validation of the MCNP (Monte Carlo N-Particle neutronics simulation code) predicted concentration of these nuclides in the irradiated LEUO2 sample. This study showed that most of the MCNP predicted nuclide concentrations were within 15% and several within 10% compared to the PIE results. Gamma and mass spectrometry of nuclides in irradiated LEUO2 enabled the determination of uranium burnup and time since irradiation.

Fission product depletion chain simplification method based on quantitative contribution function

Recently, an advanced neutronics lattice code called Kylin-2 has been developed by Nuclear Power Institute of China. In order to develop multi-group library for Kylin-2, a fission product depletion chain simplification method was proposed in this paper to generate burnup data for the multi-group library. This method constructed two quantitative contribution functions based on the thought of transmutation trajectory analysis (TTA), according to these contribution functions, a code was developed to identify important fission product nuclides and then a simplification depletion chain was reconstructed using these important nuclides. Verification and validation showed that this method successfully identifies important fission product nuclides and improve the depletion accuracy.

Supplementing fuel behaviour analyses via coupled Monte Carlo neutronics and fission product solution

The accuracy of power distribution solution and burnup algorithms in standalone fuel behaviour codes is limited. For improved accuracy, fuel behaviour codes can be coupled with external reactor physics solvers. A coupled calculation system between the Monte Carlo neutronics and burnup calculation solver Serpent and the fuel behaviour solver TRANSURANUS was developed in an earlier work. The data exchanged between the solvers included fuel pin axial and radial power and temperature distributions, the axial fast flux distribution in the cladding and the changes in the pin radii. In this work the coupling was further improved by developing the capability to transfer and utilize Serpent calculated nuclide compositions in TRANSURANUS within the coupled calculations. Additionally, support for corrector type burnup algorithms in the coupled solution was now implemented for increased accuracy of the nuclide solution. The new capabilities were demonstrated with a coupled single rod burnup calculation utilizing power and coolant history data from the Loviisa nuclear power plant. The focus at this stage was on fission product nuclides as their accumulation adversely affects the fuel performance, making their accurate solution important to ensure safe yet economical use of fuel. The demonstration showed small but visible differences when compared to simulations that were run without using fission products calculated by Serpent. The developed capabilities will now facilitate import of further fission product nuclides to be taken into account in TRANSURANUS.

Sensitivity and uncertainty analyses of fission product nuclide inventories for passive gamma spectroscopy

ABSTRACT The passive gamma spectroscopy (PGS) is a useful technique to extract information on spent nuclear fuels without any destructive actions. This method requires a correlation between number densities (NDs) of target nuclides, and it is generally estimated by numerical simulation. Therefore, the prediction accuracy of these nuclide generations is one of the key issues in PGS. Nuclear data used in nuclear fuel depletion calculations is one of the dominant uncertainty sources, so we quantify nuclear data-induced uncertainties of NDs of six fission product nuclides, which are important in PGS: Ce-144, Cs-134, −137, Ru-106, Sb-125, and Eu-154. Generation mechanisms of these nuclides are quantitatively investigated through sensitivities of these NDs to nuclear data. With the sensitivities and covariance data of nuclear data, uncertainties of NDs of these nuclides are quantified. The uncertainties of Ce-144, Cs-137, and Ru-106 are less than 2%, and that of Sb-125 is around 6%. In these uncertainties, fission yield uncertainties are dominant. On the Cs-134 and Eu-154 generations, total uncertainties are around 5% and uncertainties of (n,) cross-sections are dominant.

CENDL-3.2: The new version of Chinese general purpose evaluated nuclear data library

A new version of Chinese Evaluated Nuclear Data Library, namely CENDL-3.2, has been completed under the joint efforts of CENDL working group. This library is constructed with the general purpose to provide high-quality nuclear data for the modern nuclear science and engineering. 272 nuclides from light to heavy are covered in CENDL-3.2 in total and the data for 134 nuclides are new or updated evaluations in energy region of 10-5 eV-20 MeV. The data of most of the key nuclides in nuclear application like U, Pu, Th, Fe et al. have been revised and improved, and various evaluation techniques have been developed to produce the nuclear data with good quality. Moreover, model dependent covariances data for main reaction cross sections are added for 70 fission product nuclides. To assess the accuracy of CENDL-3.2 in application, the data have been tested with the criticality and shielding benchmarks collected in ENDITS-1.0.

ИСПОЛЬЗОВАНИЕ ПРЯМЫХ ИНТЕГРАЛЬНЫХ ТЕСТОВ СО СТАНДАРТИЗИРОВАННЫМ СПЕКТРОМ ДЛЯ ТЕСТИРОВАНИЯ СЕЧЕНИЯ ЗАХВАТА ОСКОЛКОВ ДЕЛЕНИЯ

The paper presents the results of validation for the capture cross section of the main fission product (FP) nuclides from different neutron data libraries on the results of measurements in integral experiments. A list of the main fission products for various types of fast reactors with various burnups of fuel is defined. When fissioning fuel nuclides in a nuclear reactor, numerous fission products are formed. The most important characteristic of fission products is the neutron capture cross section, as a result of which the valuable neutron for maintaining the fission chain reaction disappears. During long campaigns, the accumulated fission products can lead to a decrease in the criticality of the reactor to ~ 7 %. The evaluated neutron data for the main set of FP nuclides are contained in national libraries of evaluated neutron data. Evaluation of neutron data, as a rule, is carried out on the basis of experimental data from the international EXFOR database using various computational models of nuclear interactions. Often, the evaluated data is very different from each other. Evaluated data can be tested using integral experiments. To test the capture cross section for FP nuclides, direct integral experiments at the CFRMF reactor and experiments in which the neutron spectrum is close to the Maxwell spectrum with a spectrum temperature of kT = 30 keV were used. The first of the above experiments is a dosimetric benchmark experiment, given in the ENDF-202 benchmark experiment handbook. The second type of experiment underlies the recommendations for the MACS capture cross section for nuclides from the KADONiS astrophysical database.

A Method to Estimate Fission Product Concentration Uncertainty in a Multi-Time-Step MCNP6 Code Nuclear Fuel Burnup Calculation

The Monte Carlo N-Particle (MCNP6) radiation transport code is widely used to perform material transmutation and depletion calculations using the embedded module CINDER90. CINDER90 is capable of obtaining fission product and transuranic nuclide concentrations with a high level of accuracy in irradiated nuclear fuel. This information is very useful for many nuclear applications including reactor design and analysis, nuclear safeguards, nuclear security, and nuclear forensics, to name a few. However, at present the MCNP6 code does not estimate the overall statistical uncertainty in the nuclide concentrations reported at the end of a depletion calculation. We report our approach using a random sampling method to estimate stochastic uncertainty in fission product nuclide concentration using various parameters reported in MCNP6 output and how these uncertainties are affected by the calculation parameters.

Spatially-dependent nuclear reactor kinetic calculations with the explicit fission product model

In order to simulate complicated physical phenomena induced by fission product nuclides directly, we have developed the explicit fission product nuclides (EFP) model and have applied to one-point nuclear reactor kinetics problems in the past. In the present study, we extend the EFP model to spatially-dependent kinetics problems. Explicit representations of neutron diffusion kinetic equations with the EFP model are provided, and a new numerical method, a partial matrix exponential (PME) method, is proposed to solve differential equations for FP nuclides number densities analytically. Verification tests of the PME method against the θ method with fine time discretization are carried out, and the PME method is well verified. Numerical tests reveal that accuracy of the PME method is better than the θ method in coarse time mesh discretization, but accuracy of these two methods is comparable in typical time mesh discretization in actual kinetic calculations.Also to demonstrate effectiveness and usefulness of the EFP model, we carry out several two-dimensional kinetics calculations with the EFP model considering leakage of gaseous FP nuclides from nuclear fuel. It is shown that gaseous FP leakage affects reactor power transient and this effect is dependent on magnitude of the leakage. Time transient of number densities of some FP nuclides is presented, and this shows different trend specific to each of FP nuclides.

Validation of lattice physics code STREAM for predicting pressurized water reactor spent nuclear fuel isotopic inventory

This work investigates the depletion capability implemented in lattice physics code STREAM for the prediction of pressurized water reactor (PWR) uranium dioxide (UO2) spent nuclear fuel (SNF) isotopic inventory. The validation of this capability is performed by comparison of STREAM calculation results to measured SNF assay data obtained from PWRs Takahama-3, Calvert Cliffs and GKN II. The depletion analysis is conducted with the ENDF/B-VII.0 library and uses a pin cell model of the fuel rods from which the fuel samples were taken. The Chebyshev Rational Approximation Method (CRAM) is used to solve the depletion equation with about 1300–1600 isotopes in the depletion chain. 16 actinides and 23 fission products are analyzed in 14 spent UO2 fuel samples. The actinides are isotopes of uranium, neptunium, plutonium, americium and curium. The fission products nuclides include isotopes of cesium, neodymium, europium, samarium as well as 106Ru, 144Ce, 155Gd, 99Tc, 90Sr, 109Ag, and 103Rh. The sensitivity of some of the nuclides to the details of the power history and the adjustment of the fuel sample burnup is discussed. The impact of using ENDF/B-VII.0 library instead of ENDF/B-VI.8 is also discussed. Most of the nuclides analyzed are well predicted within ±7% of the experiment for actinides and fission products. STREAM depletion results are also compared to the codes SWAT, HELIOS and SCALE results based on publicly available information in literature, to check the performance of STREAM relative to other codes for the prediction of SNF isotopic inventory. The comparison to other code systems shows that the implementation in STREAM is of comparable accuracy. Overall, this paper demonstrates that the depletion capability in STREAM can be reliably applied to predict the isotopic inventory of PWR UO2 SNF for burnup ranging from 14 to 54 GWd/t and initial enrichment ranging from 3.0 to 4.1 wt% 235U.

Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion

ABSTRACT Nuclear data-induced uncertainties of infinite neutron multiplication factors (k ∞) during fuel depletion are quantified in a single cell and a 3×3 multi-cell including burnable absorbers. Uncertainties of reaction cross sections, fission yields, decay half-lives and decay branching ratios provided in the JENDL libraries are taken into account. Hundred percent uncertainties are assumed to nuclear data to which uncertainty information are not provided in JENDL. Uncertainties propagation calculations are carried out with the adjoint-based procedure, and required sensitivity profiles of k ∞ with respect to these nuclear data are efficiently calculated by the depletion perturbation theory. Covariance matrices for fission yields and decay data in a simplified burnup chain are successfully generated by the stochastic-based procedure. k ∞ uncertainties of about 0.6% during fuel depletion are obtained, and it is shown that actinoids reaction cross sections are dominant contributors. Nuclide-wise decomposition of the uncertainties and observation of component-wise sensitivity profiles provide physical interpretations. By virtue of the adjoint-based procedure, several parametric surveys are also conducted. Contributions of uncertainties in fission products (FPs) nuclides are quantified, and important nuclides and energy ranges are identified for further evaluation of nuclear data of FP nuclides. Effect of cooling period on k ∞ uncertainties is also discussed.

An improved DeCART library generation procedure with explicit resonance interference using continuous energy Monte Carlo calculation

DeCART multi-group cross section library generation procedure based on the KAERI library generation system was improved using the new alternative procedures based on the Monte Carlo slowing down calculations. In the new alternative procedures, the resonance integral tables were generated by solving the slowing down equation with continuous energy Monte Carlo calculations and the new procedures for the library generation were incorporated for the additional improvements. Most of fission product and actinide nuclides are processed as resonant nuclides to reduce the errors which are resulted from the approximate treatment of their resonance cross sections, and the one-group effective FPYs (Fission Product Yield) which are calculated by fission reaction rates are incorporated for improving the accuracy of the burnup calculations. Using the new procedure and ENDF/B-VII.1 evaluated nuclear data library, new DeCART cross section libraries are generated and they are tested for various benchmark calculations for pin-cell, FA, and core depletion problem against the McCARD reference solutions. The results confirmed that the new generated libraries can be used in the PWR SMR (Small Modular Reactor) core design.

High burnup calculation characteristics of traveling wave reactor

The traveling wave reactor (TWR) is an innovative nuclear system, whose discharge burnup is 400 GWd/tHM, about three to four times of existing fast reactor and six to eight times of pressurized water reactor. High discharge burnup forces the present codes to face the great challenges in precision.This paper studies the high burnup calculation characteristics of traveling wave reactor from energy spectrum, importance of fission production and calculation error accumulation of fuel consumption by KYLIN-1 code. The analysis results of typical traveling wave reactor hexagonal assembly show that the low enriched uranium assemblies have different energy spectra at beginning and end of life, which cause big calculation error ofkinf by existing code system. To ensure the correctness of the traveling wave reactor burnup calculations, the burn chain should contain 70 kinds of important fission product nuclides. The calculation error accumulation of fuel consumption is small when the burnup step is increased. The calculation error ofkinf is about 0.001% each burnup step.

Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

ABSTRACTMeasured isotopic compositions of UO2 and MOX fuel samples taken from irradiated light water reactor fuel assemblies were analyzed by CASMO5 coupled with a JENDL-4.0 base library to assess the uncertainties in the calculated isotopic compositions on heavy and fission product nuclides. The burnup calculations for the analysis were performed based on a single-assembly model taking into account the detail fuel assembly specifications and irradiation histories. For the MOX fuel samples, a multiple-assembly model was also adopted taking into account the effect of the surrounding UO2 fuel assemblies. The average and standard deviation of the biases (C/E − 1's (here C and E are calculated and measured results, respectively)) were calculated for each nuclide separately on the PWR and BWR UO2 fuel samples. The averaged biases for 235U, 236U, 239Pu, 240Pu, 241Pu and 242Pu were 2.7%, −0.9%, 0.3%, 0.7%, −2.4% and −1.7% for PWR UO2 samples, and 6.7%, −1.5%, 2.5%, −0.6%, 0.4% and −0.1% for BWR UO2 samples, respectively. The biases with the single-assembly model on the MOX fuel samples showed large positive values of 239Pu, and application of the multiple-assembly model reduced the biases as reported in our previous studies.

CENDL project, the chinese evaluated nuclear data library

The status of Chinese Evaluated Nuclear Data Library (CENDL) and the relevant CENDL project are introduced in this paper. Recently, a new version CENDL-3.2b0 was being prepared on the basis of the previous CENDL-3.1. The data in the light and actinide nuclide regions are updated from CENDL-3.1, and the new evaluations and calculations are performed mainly around structure and fission product nuclide regions. Covariance was also evaluated for structure and actinide nuclides. At the same time, the methodologies are systematically developed to fulfil the requirements of evaluations for CENDL-3.2b0. The updated nuclear reaction models for light and middle-heavy nuclides, non-model dependent nuclear data evaluation, covariance evaluation approaches, systematics, and integral validation system of nuclear data are incorporated in present CENDL project. The future developments are also planned.

Analysis of reactor-neutrino spectra fully based on gross theory of beta-decay emphasizing the special role of odd-odd FP nuclides

We applied the gross theory of β-decay to calculate the reactor antineutrino (νe ) spectra emitted from 235,238 U and 239,241 Pu samples under neutron irradiation by summing up all the contributions from a large number of decaying fission-products (FPs). Considering the special role of the odd(Z )-odd(N ) FPs in spectrum-shaping, we utilized the experimentally-known spin-parity of each odd-odd FP through the treatment proposed by Nakata, Tachibana and Yamada. Owing to this treatment, the consistency between calculated and experimental spectra was remarkably improved in a way expected from the nature of the gross theory of β-decay.

Consistent adjustment of radioactive decay and fission yields data with measurement data of decay heat and [formula omitted]-delayed neutron activities

Decay heat and delayed neutron yields, which are important physical quantities in the field of the nuclear engineering, are dependent on common nuclear data such as radioactive decay data and fission yields data of fission product nuclides. In the present study, correlations between uncertainties of these two quantities are investigated.Nuclear data relevant to uranium-235 and plutonium-239 fissions with thermal neutron are adjusted consistently with a procedure based on Bayes’ theorem using the measurement data of decay heat and delayed neutron activities. Numerical results suggest that the correlation between decay heat and delayed neutron activities uncertainties is not significant, and that independent treatments of decay heat or delayed neutron activities are possible. The effect of the consistent treatment of decay heat and delayed neutron activities is, however, observed in the adjustment results in some nuclear data such as uranium-235 thermal fission yields of yttrium-100m and zirconium-100.

Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

ABSTRACTComparing with the fission product nuclide (FP) decay heat summation calculation result in MeV/sec/fission based on the JENDL FP decay and yield data files 2011 for the burst fission, FP decay heat calculated by ORIGEN2.2 coupled with JENDL-4.0 base library ORLIBJ40 was verified at the cooling time from 1 sec to 108 sec for 235U (thermal), 238U (fast), 239Pu (thermal) and 241Pu (thermal). For these fission nuclides, FP decay heat calculated by CASMO5 at the same cooling time after a short irradiation (104 sec) was also compared with that of ORIGEN2.2. In the analysis of decay heat measurements at the cooling time from 2.3 years to 27 years consisting of four data sets on the fuel assemblies discharged from the US PWRs and BWRs, and the Swedish PWRs and BWRs, the average values of the ratios of the calculated to measured results (C/E's) were from 0.972 to 1.031 for ORIGEN2.2, and from 0.977 to 1.016 for CASMO5. The standard deviations of C/E's for the four data sets were from 0.02 to 0.03 for the both codes except for those of the US BWR fuel assemblies which were from 0.11 to 0.12. The obtained C/E's were similar to those in the precedent study.