Nuclear Data Libraries Research Articles

Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs

Abstract The purpose of the present work is to determine the neutronic characteristics of annular cylindrical prismatic Very High Temperature Gas-cooled Reactor (VHTR) with different advanced fuel designs. The study includes detailed neutronic investigations of the eight types of TRISO (containing SiC layer) and TRIZO (containing ZrC layer) fuel particles for annular VHTR core. All calculations were performed using continuous energy Monte Carlo code MVP2.0 and MVPBURN with JENDL4.0 nuclear data library. The obtained results indicated that SiC + PyC and PyC layers in TRISO fuel appeared to be not excellent choice for the annular VHTR core due to its shorter operation year and, lower fuel burnup. The annular cylindrical prismatic core fueled with TRIZO fuel particle shows good neutronic characteristics compared to those designed with other TRISO fuel particles.

Accurate measurements of delayed neutron data for reactor applications: methodology and application to 235U(nth,f)

Large inconsistencies still exist in nuclear data libraries regarding the kinetic parameters of delayed neutron (DN) precursors. As an example, there is a 17% gap between ENDF-B/VIII.0 and JEFF-3.3 on the average lifetime T1/2 of DN precursors from thermal fission of 235U. This parameter is of major importance for reactivity predictions of nuclear reactors in nominal or accidental configurations. In this context, CEA is actively participating to the ALDEN project (Average number and Lifetime of DElayed Neutrons) which aims at providing the nuclear data community with new data sets of DN from thermal and fast neutron induced fission of various actinides. A dedicated experimental setup was designed and optimized for that purpose and is presented in this paper. It consists of a “long counter” detector containing 16 proportional counters filled with 3He, embedded in a high density polyethylene matrix. The detector surrounds a fissile target prepared in the form of a miniature fission chamber, containing a few hundreds of micro-grams of fissile material. This set-up is connected to fast and efficient neutron shutters that can produce step-irradiations of variable durations. The equations driving the DN counting following step-irradiations of the fissile target are established and discussed in the perspective of DN yield or group parameter measurement. A comprehensive analysis of the different steps of data reduction is detailed: dead time characterization, Region of Interest (ROI) determination, absolute and relative efficiency calibration, fission rate estimation, irradiation time and background determination, DN decay curve production and physical parameter fitting. Following a prototype experiment performed in 2018 at the PF1b cold neutron beam line of Institute Laue Langevin (ILL, Grenoble, France), we discuss here the analysis of two campaigns occurring in 2019 and 2021 in which significant improvements were achieved in terms of background minimization, counting statistics and fission rate determination. The achievements of this work are the measurement of the delayed neutron emission per fission for the thermal neutron induced fission of 235U, estimated at (1.625 ± 0.010) % and the group parameters leading to an estimated lifetime of T1/2\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$${T}_{1/2}$$\\end{document} = (8.87 ± 0.10) s. Those results are consistent with the values recommended by the IAEA/CRP work and they come with reduced uncertainties compared with previously published results.

Comparison of Neutronic Analysis for 250 MWth Molten Salt Reactor using Monte Carlo OpenMC Code with Different Nuclear Data Libraries of JENDL 5.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF 3.3

Abstract Nuclear data libraries playian important role in the accuracy of neutronic aspect calculations, which determine various factors in a nuclear reactor design. Likewise, deterministic or Monte Carlo methods significantly affect the calculation results. Thus, the combination of nuclear data library and calculation methods needs to be studied carefully. This research will conduct neutronic analysis for the Molten Salt Reactor (MSR) design, one of the Generation IV reactor types with advanced features for future energy resources, using the Monte Carlo-based program OpenMC and different nuclear data libraries. MSR will be designed for a 250 MWth power capacity, operable for five years. The fuel used will be LiF-BeF2 as a coolant and ThF4-UF4 as the fuel. The compositions for each fuel will be optimized to achieve critical conditions during five years of operation. This research aims to evaluate the use of differentinuclear data libraries on reactor criticality conditions, using the nuclear data libraries ENDF/VIII-B, ENDF/VII.0, JEFF 3.3, and JENDL 5.0. To perform the calculation and analysis, MSR chain depletion modification has to be done for different nuclear data libraries. Then, evaluation will be conducted on effective multiplicationifactor (k-eff), neutron flux spectrum, and cross-section. The results of this research show that the use of JENDL 5.0 provides more optimal critical conditions. This is due to the completeness of JENDL 5.0 nuclear data covering 795 nuclide data and higher fission cross-section values. While ENDF/VIII-B has 557 nuclide data evaluations compared to ENDF/VII.0’s 423 nuclide data, more stable k-eff values are shown for the use of ENDF nuclear data libraries. This research is expected to significantly contribute to meeting Indonesia’s energy needs by providing efficient solutions in MSR design in neutronic aspects using the most accurate nuclear data library.

Design study of gas-cooled fast reactor with natural uranium as fuel employing modified CANDLE shuffling strategy in the axial direction

Abstract This study investigated gas-cooled fast reactor performance utilizing natural uranium as the fuel employs modified constant axial shape of neutron flux, nuclide densities, and power shape during life of energy production (CANDLE) shuffling strategy in the axial direction. The performance has been carried out on a reactor with various power levels in the range of 2,700–3,100 MWt and refueling every 10 years of burnup. The main importance of the modified CANDLE (MCANDLE) is that it can utilize natural uranium as fuel without needs reprocessing or enrichment plant. During this work, the core has been partitioned into 10 regions of similar volume along the axial direction. The fuel was filled into the first region, then transported to the second region after 10 years of burnup, and then to the third region 10 years later. This technique was practiced to all 10 regions, and the fuel in the tenth region was discharged in the reactor core. 10 % of natural thorium, natural uranium and enriched nitride (15 N) were used as fuel input. The calculations were performed employing a standard reactor analysis code (SRAC). The collision probability method (PIJ) module of SRAC was used to calculate cell burnup, and the reactor core design calculations were done with the CITATION module of SRAC, which used JENDL-4.0 as a nuclear data library. The result shows that a high power level causes an increasing rate of burnup level and effective multiplication factor. The highest average discharge burnup level is about 31.3 % HM for case E and the lowest average discharge burnup is about 27.2 % HM for case A.

Benchmark Analysis of a Prismatic Type-high Temperature Gas-cooled Reactor with the ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 Nuclear Data Libraries

In this work we performed a benchmark analysis of the High Temperature Engineering Test Reactor (HTTR) fully-loaded start-up critical core with a 30-bundle loading configuration using the Monte Carlo code Serpent 2 with the recently released nuclear data libraries ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0. The purpose of the work is to reveal the impacts of using different ENDF nuclear data libraries on neutronics calculations of a prismatic high temperature gas-cooled reactor (HTGR). The benchmark results obtained with Serpent 2 were compared against the available experimental values and those attained by different computer codes (MCNP5 and SCALE) using the nuclear data library ENDF/B-VII.0. The comparative results showed good agreement between Serpent 2, the experimental values, MCNP5 and SCALE. The results also exhibited a notable difference between using ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 in predicting the neutronics parameters of the HTTR that suggests further investigation in future work.

Californium-252 production at the High Flux Isotope Reactor - II: Comparison between the highly enriched uranium and a proposed low-enriched uranium core

This is the second paper on a 252Cf production study performed in support of efforts to convert the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The first paper primarily focuses on validating computational tools and nuclear data. This companion paper evaluates another critical aspect: the 252Cf production capability with a proposed LEU core. HFIR must maintain its world-class performance and missions following conversion and because 252Cf is a vital, multipurpose neutron-emitting radioisotope, the ability to efficiently produce 252Cf must be preserved. In this study, the HFIRCON transport and depletion tool, several nuclear data libraries, and Campaign 78 data were used to compute 252Cf production, sensitivity, and safety metrics. Results indicate the 252Cf production and production rates are slightly higher with a 95MWth LEU core compared with those obtained with the 85MWth HEU core. Additionally, the target peak fission rate densities, discharge cumulative fission densities, and heat deposition rates with the LEU core are within a few percent of those calculated with the HEU core. The findings suggest HFIR’s 252Cf production capability can be effectively maintained with an LEU core without adversely affecting the safety metrics.

Investigation of the production routes of 81,82m,83,84 Rb isotopes through proton-induced nuclear reactions on 82,83,84,86 Kr

Investigations using nuclear reaction models are significant and have considerable consequences in terms of understanding nuclear reaction processes and their nature. The nuclear data on radionuclides created by the cyclotron find application in nuclear diagnostics and therapeutic treatments. The excitation function of proton-induced nuclear reactions on 82,83,84,86 Kr for the generation of medically beneficial short-lived radioisotopes 81,82m,83,84 Rb from threshold energy to 50 MeV has been investigated using Koning–Delaroche optical model potential with nuclear level density (NLD) from both the phenomenological and microscopic nuclear models incorporated in TALYS1.96 code. The phenomenological constant temperature + Fermi gas model, back-shifted Fermi gas model, generalised superfluid model, and microscopic NLD from the Skyrme–Hartree–Fock–Bogoliubov, Gogny–Hartree–Fock–Bogoliubov, and temperature-dependent Gogny–Hartree–Fock–Bogoliubov NLD tables have been used to calculate the excitation function. The calculated excitation functions are compared with the experimental data from EXFOR and the TALYS evaluated nuclear data library (TENDL2023). An analysis is carried out by estimating the mean deviation D-value between the theoretical and experimental cross-sections, and the NLD model corresponding to the best result is used for the estimation of activity and production yield for each reaction channel.

Systematic study of theoretical estimation of nuclear reaction cross section of proton-induced reactions on Neodymium target upto 65 MeV using TALYS-1.96 code

Abstract Systematic study of theoretical estimation of nuclear cross section of charged particle induced reactions on rare earth nuclei has been carried out. The production cross section of the $^{150,149,148,146,144,143,141}$Pm and $^{149,147}$Nd nuclei were calculated theoretically via proton induced reaction on neodymium using TALYS (version 1.96) code in the default mode, with different combination of nuclear models as well as using adjusted nuclear model parameters from reaction threshold to the proton energy upto 65 MeV. The theoretically computed results were compared with the experimental results taken from EXFOR database and literature, reported by various research groups. Moreover, the effect of the various Level Density Models (LDMs), Preequilibrium Models (PEs), Optical Model Potentials (OMPs), gamma Strength Functions ($\gamma$SFs) in the cross section calculation were taken into consideration. The present theoretical analysis will help to understand the theory of nuclear reaction models and improve the evaluated nuclear data libraries.

Characterization of neutron spectrum of polyethylene-moderated AmBe neutron source using a passive single-cylindrical neutron spectrometer

The changes in the neutron spectrum produced by the AmBe neutron source moderated in the polyethylene sphere need to be characterized. A passive single-moderator neutron spectrometer with indium foil activation neutron detector has been developed to characterize the neutron spectrum of an AmBe source inside a polyethylene sphere. The detector response function was calculated using the MCNPX 2.7 program with IRDFF-II nuclear data library. Measurements were conducted by placing the Single-Cylindrical Neutron Spectrometer (SCNS) on the outer surface of the sphere for 5 h. The 116mIn activity due to neutron activation in each indium foil was measured using a gamma spectrometer with an HPGe detector. By identifying the count value at the peak energy of 1294 KeV and considering an HPGe detector efficiency of 3.2% at the foil position 1 mm above the detector surface, the activity of 116mIn was obtained. The activity value of 116mIn from each indium foil was compared with the MCNPX simulation results. The neutron spectrum was unfolded using the UMG 3.3 program with activity data input of 116mIn for each foil and detector response. A neutron spectrum was obtained with a total neutron flux of 634 ± 60 n/cm2·s, consisting of 25% thermal neutrons, 16% epithermal neutrons, and 61% fast neutrons. When compared with the simulation results, the total neutron flux in the spectrum produced by SCNS-In showed only a small difference of 1%. Based on these neutron spectrum measurements, it was determined that placing the AmBe neutron source inside a 15″ diameter PE-sphere will reduce the fast neutron flux by 78%.

Neutron-induced reaction cross section measurements for palladium isotopes at 14.77 ± 0.17 MeV with covariance analysis

The nuclear reaction cross sections for 110Pd(n,2n)109mPd, 102Pd(n,2n)101Pd, 106Pd(n,p)106mRh, 110Pd(n,2n)109gPd and 105Pd(n,p)105Rh have been measured at the neutron energy of 14.77 ± 0.17 MeV by using the method of activation and off-line γ-ray spectrometry. The uncertainties in the various nuclear parameters in the measured reactions and their correlations were estimated by using covariance analysis. The present data were compared with literature data from the EXFOR database and nuclear data libraries, including ENDF/B-VIII.0, JEFF-3.3, JENDL-5.0 and TENDL-2021. Theoretical model calculations for the cross-sections are performed with the TALYS-1.96 nuclear model code with optimized parameters. There is good agreement between the measured, literature and evaluated data. The results are important for shielding applications in the radiation fields involving neutrons.

The MUSIC Critical Benchmark and Nuclear Data

The Measurement of Uranium Subcritical and Critical (MUSIC) experiment was a series of measurements of critical and subcritical configurations of bare highly enriched uranium. The goal was to compare measurement methods, analysis techniques, and simulation methods across regimes of criticality and to provide high-quality validation of 235U nuclear data. A benchmark evaluation of the two critical configurations of the MUSIC experiments will soon be published in the release of the International Criticality Safety Benchmark Evaluation Project Handbook. The recent execution of the experiment aids in proper quantification of model simplifications and all uncertainties associated with the experiment. Historical benchmark evaluations are heavily relied on for uranium nuclear data validation despite the fact that the same level of documentation and comparable uncertainty analysis may not be present. The MUSIC evaluation is less likely to include “unknown unknowns” that could impede accurately modeling the system. Presented are both highly detailed and very simplified models, which represent the experimental configurations accurately, aiding the users of the benchmark for nuclear data or transport code validation. The sensitivities of k eff to nuclear data and nuclear data–related uncertainties are very similar between this experiment and previous bare uranium sphere experiments. In addition, the nuclear data uncertainties to any nuclides other than 235 U are small. For all these reasons, the recently evaluated MUSIC benchmark critical configurations could prove very useful for 235 U nuclear data validation. Currently, major libraries have good agreement with the experimental results, within 200 pcm for all nuclear data libraries, and within one standard deviation of the experimental result for most. Suggested nuclear data adjustments based on MUSIC and Lady Godiva are also presented, with posterior improvements to both the agreement in k eff and the uncertainty associated with the nuclear data.

Delayed neutron library analysis using dynamic control rod calibration method based on results from the VR-1 reactor experiment and Serpent transient calculation

This study is focused on a comparison of experimental neutron flux data from the VR-1 reactor with results from Serpent calculation. Serpent data were calculated by the dynamic part of the code and the steady-state criticality calculation. The Serpent calculation was performed with the JEFF 3.3 and ENDF/B-VIII.0 nuclear data library. This paper expands previous work and focuses on the integral control rod worth. Comparison of the ENDF/B-VIII.0 and JEFF 3.3 nuclear data libraries with experimental data is especially valuable. The comparison shows better performance of JEFF 3.3 nuclear data instead of ENDF/B-VIII.0 for the transient calculation of the cases presented in this article. In addition, the comparison of static and dynamic reactivity provided an interesting difference, especially for small rod insertion. Moreover, the study confirms that the Serpent dynamic external source simulation mode provides an accurate description of reactor behavior, and that the shape of the integral control rod worth is close to one experimentally measured.

A passive single-cylindrical neutron spectrometer for photoneutron spectrometry in 10 MV Elekta Precise Linac head

Measuring the neutron spectrum on the Linac head using a passive type neutron detector is relatively easier than an active type detector because it does not require the involvement of operators who work in a high radiation area. Neutron spectrum measurements with passive foil activation neutron detectors in a single-moderator are more practical and easier than multi-foil techniques. This technique is quite simple, easy to make, low cost, and the measurement time is relatively short. This spectrometer is designed by placing seven passive Au detectors at varying depths in a single-cylindrical moderator. The response function for each detector is calculated by MCNPX 2.7 program and used ENDF7 as a nuclear data library. Detector foil activity was measured using a gamma spectrometer, and the neutron spectrum was obtained using the unfolding technique using the UMG 3.3 program. The results of measuring the neutron spectrum measurements have been carried out at a position 10 cm below the head of the Elekta Precise 10 MV Linac was 2.73 × 105 n/cm2·s, where the fast neutron from these results are 4% higher when compared to the simulation results using PHITS.

Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture

In this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dual-purpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF–ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.

Evaluation of cross sections for fast ion reactions with beryllium in helium and hydrogen fusion plasmas

To computationally support hydrogen and helium plasma discharges in the early stages of tokamak operation and to support the commissioning of the neutron detectors during these operational phases, creation of a realistic neutron and gamma ray particle source for Monte Carlo simulations will be needed. One of the most important parts of creating the particle source is calculating the reaction rates of the particle-emitting reactions to determine the emission profile in the plasma and the energy spectra of the emitted particles. In this paper the analysis and evaluation of cross sections for important neutron-emitting reactions, namely, 9Be(p,nγ)9B, 9Be(3He,nγ)11C, and charged-particle emission reactions 9Be(p,d)2α and 9Be(p,α)6Li that cause neutron emission in the next step of interactions are presented. The reaction cross sections were evaluated based on experimental measurements and empirical models describing the interaction of two charged particles. Evaluation of the associated uncertainties was also performed. The main goal of the work is to propose the newly evaluated cross sections for inclusion in the FENDL nuclear data library, thus making the cross section available to other researchers studying the above listed reactions.

Reaction Rate Ratios for Recent Fast Metal Experiments with Large Plutonium Masses

Reaction rate ratios are integral responses that are used within the criticality experiments field because they contain spectral information. While these types of measurements have been utilized for nuclear data validation with historic experiments, few experiments of this type have been utilized for recent experiments, as few exist. This work focuses on measured reaction rate ratios for two nearly bare plutonium critical assemblies with different geometries: one that is cube like (with a Pu mass of 40 kg) and one that is slab like (with a Pu mass of 109 kg). Irradiations were performed with both configurations in which foils were placed near the center of the assembly. Plutonium, highly enriched uranium, depleted uranium, and Au foils were included in the irradiation and counted via high-purity germanium detectors. From these measurements, reaction rate ratios were calculated. Measured and simulated values and uncertainties are presented for the reaction rate ratios. Ratios utilizing the following reactions are given in this work: 197Au(n, γ ), 197Au(n,2n), 235U(n,fission), 238U(n,fission), 238U(n,2n), 238U(n, γ ), and 239Pu(n,fission). Uncertainties for the measured reaction rate ratios ranged from 4% to 7%, and the contribution of various parameters to this uncertainty was investigated. The results are compared to historical experiments and should be used for nuclear data validation for future nuclear data library releases. These measurements are part of the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project, which utilizes measurement responses in addition to keff (such as these reaction rate ratios) to help reduce uncertainties in 239Pu nuclear data.

Experiments on Criticality and Reactivity Worths in the FCA-XXII-1 Assembly Simulating Highly Enriched MOX Fueled Tight Lattice LWR Cores

A series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio). Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keff values overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%. Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values.

Nuclear Logging in Geological Probing for a Low-Carbon Energy Future – A New Frontier?

This paper examines the potential of nuclear logging techniques, ubiquitous in the petroleum industry, to extract geological information needed to support the transition to the low-carbon energy future being envisioned to replace fossil fuels and explores the technological advances needed. Assessment to date, using Monte Carlo modeling and available measurements shows promise in (1) monitoring injected CO2 for carbon capture and sequestration (CCS) to mitigate climate change, (2) assessing sites to bury high-level waste to support nuclear power generation and later monitor buried radioactive waste, and (3) in geothermal, a renewable option. For each case, standard techniques, while promising, also show technological gaps. Additionally, two postulated low-carbon areas supporting renewable energy generation are briefly examined—downhole quantification of strategic minerals and prediction and detection of naturally occurring hydrogen in the geology. Three related technology areas are also examined: transition from the current dual-track philosophy of nuclear logging tool design to a compact, more universally applicable advanced accelerator-based multiple-parameter tool concept, incorporation of artificial intelligence-guided physical health management systems to minimize generator failure, and advances in generation/detection hardware and software. Software advances would include codes with dynamic visualization and improved nuclear data libraries to design, calibrate, and assess tools, especially to provide a priori space-time profiles of attendant multiple radiation types, which would arise in the novel systems being postulated. Two exotic concepts, recently suggested also for downhole use, associate-particle imaging to monitor casing integrity of wells storing methane and hydrogen and muon-based deep density probing, are briefly explored.

Cross-section induced uncertainties in neutron source mission rate calculations

Analysis presented in the paper is focused on the characterization of uncertainties induced by cross-section data, which contributing to the overall uncertainty of the correction factors. Advances in computational methods and computational power shifted the calculation of correction factors among the standard steps of the manganese bath method what decrease an interest about this part of the method. Anyway, with the development of computational tools, the cross-section data were also improved, and new evaluations of nuclear data libraries include more information about the cross-section data covariances. Therefore, the propagation of uncertainties induced by neutron cross-sections can be carried out within standard transport calculation of the manganese sulphate bath model. In this paper, the super-sequence SAMPLER module that implements stochastic techniques is used to assess the uncertainty in computed results. Reaction rates on all nuclides of the solution are computed in 400 cases with uncertain parameters and the results are evaluated by an auxiliary tool. Consideration of nuclear data uncertainty in calculations is a general trend that requires the attention of nuclear researchers and should draw attention in metrology. Additional 1.5 % (one-sigma) uncertainty is introduced to the overall uncertainty through correction factor.

Accuracy Evaluation of Monte Carlo Simulation Results Using ENDF/B-VIII.0 and JENDL-5 Libraries for 10 MWth Micro Heat Pipe-Cooled Reactor

The micro heat pipe-cooled reactor is an innovative type of reactor that utilizes heat pipes to cool its core. It consists of a reactor core, an energy conversion system, shielding, and a heat removal system. This reactor shows great potential as a viable option for supplying electricity in remote areas. By incorporating a monolithic core with heat pipes and an efficient heat conversion system, this reactor design eliminates the need for a main pipeline, circulating pump, and auxiliary equipment, resulting in a cost-effective, compact, and transportable system. The monolithic reactor design has undergone significant advancements in neutronics and thermal hydraulics. This article focuses on evaluating the impact of the latest released nuclear data libraries, ENDF/B-VIII.0 and JENDL-5, on calculated neutronics and kinetics parameters. The total keff uncertainty was propagated and found to be significant for both recently evaluated nuclear data libraries (678.52 pcm for ENDF/B-VIII.0 and 525.91 pcm for JENDL-5, respectively). The total uncertainty originated from nuclear data was evaluated for total ν, reaction cross sections, and angular distributions in the case of JENDL-5, and for ENDF/B-VIII.0, uncertainty from angular distributions was not included because of the unavailability of its multigroup structure covariance matrices. The results reveal that the largest contributor for ENDF/B-VIII.0 is 235U total (409.18 pcm), while that for JENDL-5 is 56Fe capture cross section (361.93 pcm). For the kinetic parameter’s uncertainty, the impact on the total βeff, leff, and λeff simulation results was found to be not significant (about 1%).