Neutron Multiplication Factor Research Articles

Implementation of thorium-based fuel for Indonesia Micro Reactor (IMR)

Indonesia has committed to reducing greenhouse gas emissions in the framework of net zero emissions, and nuclear energy will be a part of the energy mix. Indonesia has been developing a 10 MWt micro reactor of high temperature gas-cooled reactor (HTGR) with pebble fuel containing 17 % enriched UO2 kernel with OTTO cycle scheme. For the government’s interest to use thorium, this paper assesses the possibility of replacing uranium with thorium. The assessment is performed by calculating the effective neutron multiplication factor (keff). Neutrons captured by 232Th lead to the formation of two intermediate isotopes, namely 233Th and 233Pa, the latter being a significant neutron absorption cross-section. Neutronics analysis is performed using parameters of thorium mass fraction, core temperature, and active core height. The thorium mass fractions varies from 0 % to 90 % Th. The core temperatures are in the range of the normal operating temperature at 900 K, assumed superheat temperature at 1200 K, and shutdown temperature at 300 K. The active core height varies at 197, 180, 150, 130, and 125 cm. This paper aims to determine the keff of the IMR at maximum thorium/uranium mass fraction composition, maximum active core height, core temperature, and fuel burn-up. The MCNP code is used for analysis. The results show that keff decreases with increasing thorium mass fraction and core temperature, and decreasing active core height. Calculations with thorium mass fraction of 15 %, core active height of 197 cm in 20 steps show that the reactor may be continuously operated for more than 920 days without refueling. The analysis results show that IMR design has negative temperature reactivity, inherent safety characteristics, and the ability to use thorium-based fuels.

SAFETY JUSTIFICATION FOR TRANSPORTATION OF NUCLEAR MATERIALS

The Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan conducts a large number of studies to justify the safety of designed reactors and already operating reactors. Experiments are carried out both at in-pile and out-of-pile research installations. To implement these studies, there is a need to transport nuclear materials to the place where they are performed. Thus, it is necessary to justify the safety of nuclear material transportation. This paper presents neutronic calculations of the effective neutron multiplication factor to justify the safety of nuclear material transportation in accordance with the legislation of the Republic of Kazakhstan. The calculations used the most commonly transported type of nuclear materials – uranium dioxide (UO2) fuel pellets of various enrichments. Transportation was conducted using a transport packaging set IP-1, in accordance with the rules for the transportation of nuclear materials. Both normal and emergency conditions for the transportation of nuclear materials were modeled, taking into account the possibility of water leaking into the packaging and water getting into canisters with uranium dioxide fuel pellets. To determine the effective neutron multiplication factor (keff) of the transport and packaging set, the Monte Carlo calculation method was used in the MCNP5 program, with the ENDF/B-VII nuclear data library.

Neutronic and Proliferation Resistance Analysis of Small Modular Pressurized Water Reactor with Various Fuel Types using SRAC 2006

One strategy for emission reduction in Indonesia is the dedieselization program, replacing diesel power plants with renewables. Pressurized Water Reactor (PWR) type Small Modular Reactor (SMR) is also a consideration. This study aims to compare fuel performance in SMRs. Four fuel types with varying fissile and fertile nuclides but equal fissile mass percentages were assessed: U235- U238, U235-Th232, U233-U238, and U233-Th232, using PIJ SRAC 2006 and JENDL 4.0. Parameters include neutron multiplication factor, conversion ratio, and isotope production. Results indicated that U233-U238 has a highest initial neutron multiplication factor, 1.55. Substituting uranium-238 with thorium-232 boosts neutron flux and conversion, reducing nuclear proliferation risk. While initial neutron multiplication may not be highest, thorium-232 can produce uranium-233 with high initial neutron multiplication factor. This research aids fuel selection for Indonesia SMR dedieselization program.

Comparative analysis on small modular reactor (SMR) with uranium and thorium fuel cycle

The utilization of Small Modular Reactors (SMRs) and micro-reactors holds significant potential advantages. The primary benefits include their smaller size, accessibility to remote regions, and reduced land requirements. One noteworthy SMR is the NuScale Power Module (NPM), which has a thermal capacity of 160 MWt and operates on a 2-year cycle. However, there is a growing interest in exploring alternative fuels to enhance SMR performance and safety, and thorium has emerged as an appealing option due to its superior neutron characteristics and potential for higher fuel efficiency. This research aims to do preliminary research into adapting fuel for a thorium fuel cycle in the integral PWR (iPWR). The thorium cycle excels with an initial neutron multiplication factor of 1.134, almost the same as uranium (1.129). Additionally, the thorium cycle boasts superior neutron flux and Conversion Ratio (CR) in the thermal neutron energy range, enhancing sustainability with an average CR of 0.765, compared to 0.574 and 0.692 for uranium and uranium–thorium cycle. The addition of thorium improves efficiency in converting fertile material into fissile material with an enough criticality, where fuels incorporating thorium have a higher CR (Conversion Ratio). Furthermore, thorium fuel cycle has significantly lower production of plutonium isotopes compared to other fuel cycles. However, it faces a challenge with a high power peaking factor (PPF) of 1.923 at the Beginning of Cycle (BOC), requiring further research for mitigation.

Subcritical neutron noise measurements for plutonium systems with varying geometry and mass

Recently, experiments were performed at the National Criticality Experiments Research Center (NCERC) with Zero Power Physics Reactor (ZPPR) plates in two very different geometries. The intent of this set of experiments was to provide the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project team measured data of multiple radiation detector responses to be used in an adjustment of plutonium nuclear data. The work shown in this paper will focus on multiplicity results measured for subcritical arrangements of ZPPR plates in the two geometries. The Hage–Cifarelli formalism of the Feynman variance-to-mean method, a moments-based time-correlated neutron noise approach, was used. A validation of the methodology used to estimate uncertainties in the moments was performed and is presented. The effective neutron multiplication factor (keff) value were inferred from this analysis. The keffvalues were compared to simulated values using the radiation transport code Monte Carlo N-Particle® (MCNP). The results shown in this work provide evidence of larger bias in keffvalues as the geometric buckling of the system increases. This effect will be important for nuclear data validation and future adjustments of plutonium nuclear data.

Core nuclear design and heat transfer analysis of a megawatt-level SiC coated particle based vehicular micro reactor

As a versatile and diversified source of new energy, transportable micro reactors can offer high flexibility and the ability to provide power services even in remote areas. When utilizing coated fuel particles and supercritical carbon dioxide (S-CO2) coolant, the transportable micro reactor core achieves exceptional inherent safety and high power generation efficiency, rendering it incredibly competitive within the nuclear market. The Vehicular Micro Reactor (VMR) previously investigated is an integrated micro-reactor cooled by S-CO2, which employs TRistructural ISOtropic (TRISO) fuel particles dispersed in a conventional graphite matrix to generate 5 MW of power output. However, its drawback lies in the low fuel loading capacity. In recent year, the application of high-strength SiC materials in the field of nuclear energy has been developing continuously. In this paper, a new type of double-layered SiC coated fuel particle is designed, based on which a SiC-based Vehicular Micro Reactor (SVMR) is proposed and explored. It represents an improved design of VMR, with the main difference being its utilization of SiC as a matrix and filling by the two-layered SiC-coated fuel particle to achieve higher fuel loading than the VMR. In this paper, both the neutronic and heat transfer characteristics of the SVMR are discussed. The neutronic simulation is conducted using SCALE code based on Monte Carlo method, the module KENO IV was used to analyze the effective neutron multiplication factor (keff), and SCALE/TRITON (Transport Rigor Implemented with Time-Dependent Operation for Neutronic Depletion) coupled by 2-D radiation transport code NEWT and ORIGEN for burn-up calculation. By optimizing the assembly size and SiC-coated fuel particle size, the SVMR core achieves high fuel loading capacity and superior fuel utilization efficiency. The results of the heat transfer analysis also demonstrate that the SVMR core, which is entirely SiC-based, exhibits exceptional heat transfer capability. The excellent neutron economy and inherent safety make SVMR a more competitive micro reactor concept.

Preliminary neutronics analysis for generation III+ VVER design - VVER-1200

The preliminary neutronics analysis of a new generation, first-of-its-kind VVER-1200 reactor has been carried out to develop an independent physics analysis capability that will support its future safety evaluations and licensing processes. An open-source Monte Carlo (OpenMC) code, utilizing LANL released ENDF/B-VIII.0 nuclear data, has been employed for this purpose. Initially, the validation of the lattice and core models of OpenMC simulations is established by comparing the effective neutron multiplication factor (Keff) estimated for a Normal Operating Condition (NOC) with the literature information. The comparison demonstrates that the prediction capability of the OpenMC code is on par with other international state-of-the-art codes. Once sufficient confidence is obtained in the modeling, the analysis is extended to investigate the inherent safety parameters, namely the reactivity coefficients of fuel and moderator temperatures, as well as the reactivity coefficient of boric acid concentration in the moderator at both the lattice (fuel assembly) and core levels. Behavioral changes of these parameters of interest, from the lattice to the core level, are examined and reported.

Validating Geant4 with Monte Carlo simulations in the context of nuclear disarmament verification

In this paper neutron transport simulation results of experimental configurations used during the 2019 IPNDV measurement campaign at the Belgian Nuclear Research Centre SCK CEN in Mol are presented. As these are a good basis for benchmarking simulations, results of the Monte-Carlo simulation codes Geant4, SCALE/KENO-VI, MCNP and openMC are compared with the aim of validating Geant4. With fissile material present as plutonium–uranium mixed oxide, (α,n) emissions are included in the simulations. A Geant4 extension to calculate neutron multiplication factors keff was developed and applied to the IPNDV configurations and a set of OECD/NEA benchmark experiments. Neutron fluxes of the IPNDV configurations through a reference volume are simulated. The Geant4 (α,n) simulation toolkit SaG4n is compared to the SCALE code ORIGEN and small deviations are observed. Geant4 calculations of keff for criticality benchmark experiments deviate in the mean by ＋0.4%, for MCNP and openMC deviations are ≤6%, for KENO-VI ≤8%. Geant4’s total neutron fluxes through a reference volume next to the IPNDV configurations agree within a margin of ±5% with MCNP and openMC and −14 % with KENO-VI. Considerable differences between all codes were observed for thermal neutron scattering but Geant4 results are still within the other codes’ variability. Differences to Geant4 for epithermal energies can be addressed to a not yet included “Doppler Broadening Rejection Correction”. By comparing Geant4 simulations of a complex set-up with various reference codes and demonstrating its good performance as well as some discrepancies, this study contributes to the validation of Geant4 neutron physics in fissile material systems and for nuclear disarmament verification simulations.

Modelling of spent fuel dissolution process for assessment of criticality safety margin

Criticality safety margin assessment is a challenging task for a large dissolution tank under uncertain process conditions. The Monte-Carlo based neutron transport code (MCkeff) was used to model geometry, material and process conditions of dissolution tank containing fissile material. This manuscript presents a sampling-based probabilistic technique for estimating single-sided tolerance limit on criticality safety parameter due to random variation of process parameters. Based on their effect on fissile mass, fissile concentration, neutron moderation and neutron reflection, potential uncertain parameters impacting the criticality safety margin were identified. By withdrawing random samples from uncertain domain of process parameter, a simulation test matrix was formed to study the variability of the neutron multiplication factor. The Wilk's formula, which corresponds to the 95th percentile and 95% confidence level, was used to estimate required sample size to minimise statistical error (i.e., Type-I error associated to the rejection of the genuine sample and Type-II error connected to the acceptance of the false sample). In order to cut down the computation costs associated with requirement of large computer runs, a meta-model for the criticality safety problem was also constructed using a linear regression model. Additionally, associated model uncertainty was quantified. The variation in neutron multiplication factor caused by changes in process variables like pellet diameter, linear pellet mass, temperature, etc. was also investigated as an input parameter uncertainty. It is found that model uncertainty plays significant role while assessing the criticality safety margin using meta-model.

Sensitivity and uncertainty analysis of effective multiplication factor of KRITZ-1 benchmark using ENDF/B-VIII.0 and JEFF3-3 nuclear data libraries

Sensitivity and uncertainty analysis have been performed to estimate the uncertainty in the effective neutron multiplication factor, keff, of the KRITZ-1 benchmark and to identify which nuclear reactions and isotopes of interest could be significantly influencing it. First, the adjoint-weighted perturbation method using the KSEN card of the MCNP6 code is applied to compute the sensitivity vectors for 235U, 238U, 1H, 16O, and 10B isotopes with the JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries. Then, the sensitivity vectors and covariance matrices were combined to calculate the uncertainty in keff due to uncertainties in nuclear data. The 44 group covariance matrices have been processed with the NJOY21 system.It was found that the uncertainty in keff is mostly due to uncertainties on ν‾p , fission cross section of 235U, the capture cross section of the two isotopes 238U and 1H in the aforementioned data libraries

Neutronics study of chloride fuel salts for experimental design of small Molten Salt Reactor

From the Indian perspective, molten salt reactor technology is a promising option for thorium utilization in the third stage of the nation's nuclear programme. Recently, the physics design of a fluoride salt fuelled small molten salt reactor has been initiated as an experimental core to demonstrate the circulating fuel reactor technology. However, the literature suggests that chloride fuel salts have better breeding characteristics, particularly in fast-spectrum MSRs. The thermal-hydraulic and neutronics properties of the chloride fuel salts could considerably differ from that of fluoride. Therefore, the applicability of chloride fuel salts in a small MSR core has been investigated for the neutronics feasibility of the fuel type. In the study, criticality and safety parameters such as neutron multiplication factors and fuel temperature coefficient of reactivity are assessed for three options of chloride-based molten fuel salts. The predicted results for the experimental MSR thermal core fuelled with chloride-based fuel salts are reported in the paper. A brief review on the comparison of chloride fuel salts with the fluoride salts in MSR designs is also presented.

Deterministic sampling method using simplex ensemble and scaling method for efficient and robust uncertainty quantification

ABSTRACT Uncertainty quantification (UQ) of the neutron multiplication factor is important to investigate the appropriate safety margin for a target system. Although the random sampling method is a practical and useful UQ method, a large computational cost is required to reduce the statistical error of the estimated uncertainty. Furthermore, if an input variable follows a normal distribution with a large standard deviation, the perturbed input variable by the random sampling method may become a physically inappropriate or negative value. To address these issues for the efficient and robust UQ, a modified deterministic sampling method using the simplex ensemble and the scaling method is proposed. The features of the proposed method are summarized as follows: The sample size is r + 2 , where r corresponds to the effective rank of the covariance matrix between the input variables; depending on a situation of target UQ, the amounts of perturbations for the input parameters can be arbitrarily given by the scaling factor method; the scaling factor can be updated to avoid physically inappropriate in the perturbed input variables. The effectiveness of the proposed method is demonstrated through the UQ of the neutron multiplication factor due to fuel manufacturing uncertainties for a typical PWR pin-cell burnup calculation.

Validation of MCkeff code using ICSBEP results

Validation of the indigenously developed Monte Carlo program, MCkeff, is carried out by comparing the results of International Criticality Safety Benchmark Evaluation Project (ICSBEP) problems with that of standard legacy code MCNP (general-purpose Monte-Carlo N-particle transport code). Code-to-code criticality validation is essential before the usage of MCkeff to model various kinds of fissile systems. The Evaluated Nuclear Data Files (ENDF/B VII.I in ACE Format) used in MCNP is expended in the present study, thereby avoiding the differences in the results due to cross-section data. Thus, the deviations in the computed results will reflect only the differences arising out of differences in algorithms employed in these codes. Plutonium metal benchmark problems of different geometrical configurations were selected for validation purposes. All problems were simulated using MCkeff with the same inputs as in MCNP validation, and the computed results of keff (neutron multiplication factor) of various systems were compared. The results from MCkeff agree with that of MCNP, and the maximum deviation between them is less than 0.1%. Hence, the code MCkeff can be used to analyze other metal systems problems. Further validation work on the different fissile systems with other physical forms is in progress.

Utilization of ACE nuclear data file toolkit ACEtk to calculate relative sensitivity coefficients of point-kinetics parameters

Sensitivity and uncertainty methods are quintessential for nuclear criticality safety and experiment design. This type of analysis relies on calculations of sensitivity coefficients; sensitivity coefficients of the effective neutron multiplication factor with respect to some nuclear data are predominantly calculated and used. As a part of the Laboratory Directed Research & Development project EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) at Los Alamos National Laboratory, sensitivity coefficients of many radiation detector measurement responses with respect to nuclear data were investigated. Specifically, this paper outlines a method to calculate point-kinetics parameters relative sensitivity coefficients with respect to nuclear data. Point-kinetics parameters such as the prompt neutron decay constant, effective delayed neutron fraction, and neutron generation time are especially important to experimenters and reactor operators designing systems with dynamic neutron populations. This method couples capabilities of the ACE (A Compact ENDF) nuclear data file toolkit, ACEtk, with the ability to load cross sections into the radiation transport code Monte Carlo N-Particle (MCNP). Key aspects of optimizing this method for a particular application and sensitivity profiles of the Jezebel criticality experiment are examined and discussed.

Characterization of the GBC-32 Fuel Assembly Source Terms

This paper presents burnup/depletion calculations of the typical Westinghouse 17x17 fuel assembly to be used as a radioactive waste package in a Generic Burnup Credit cask benchmark problem with 32 elements (GBC-32). This first phase is addressing spent fuel source terms calculation while evaluation of the shielding performance of the GBC-32 cask is planned for the second phase. The TRITON-NEWT methodology of the SCALE6.1.3 program package was used in a tandem with ORIGEN-S code for deterministic 2D calculation of the GBC-32 fuel assembly neutron multiplication factor, providing spatial-temporal fluxes and isotopic concentration change. The burnup simulation was done up to 60 GWd/tU with sensitivity analysis of relevant physical parameters influenced by the working cross-section library. This approach also allowed generation of the specific user-defined collapsed cross-section libraries as a function of fuel enrichment and burnup level. Calculation of isotopic concentrations, decay heat, neutron-gamma spectra and major actinides activity for different fuel assembly cooling periods was performed using ORIGEN-ARP module.

A Novel, Low-Cost, Position-Sensitive Neutron Detector to Support Thick-Target Inverse Kinematics Experiments for Nuclear Data Measurements

High quality nuclear data lie at the heart of accurately modelling stellar systems and terrestrial nuclear reactors. However, some key reaction cross sections have large uncertainties, which limit such models in predicting isotopic abundances and other aspects of stellar evolution, along with key operational parameters for nuclear reactors. Reactions involving neutrons are particularly difficult to measure experimentally in laboratories, not least due to the unique challenges involved when detecting neutrons. We present a new approach to measuring nuclear reactions involving neutrons by exploiting the Thick-Target Inverse Kinematics (TTIK) approach. For such measurements, a new detector called ATTIKUS (A Thick-Target Inverse Kinematics detector by Universities in Sheffield) is under construction. Here we present designs and Geant4 Monte-Carlo simulations of the detector. The simulations indicate that a neutron position reconstruction resolution of 10 cm is obtainable and demonstrate how this device could be applied to the 13C(α,n) reaction, which is considered to be the main neutron source for the s-process in low-mass Asymptotic Giant Branch stars. In the TTIK method, the emission position of the neutron (the nuclear interaction position in a gaseous target) is directly linked to the centre-of-mass energy of the reaction. Therefore, a position resolution will translate into an energy resolution, depending on the beam-target combination. The inverse reaction, 16O(n,α), causes a large uncertainty in calculating the effective neutron multiplication factor, Keff in nuclear reactors, so improvements are required here.

Application of Np–Am Mixture in Production of 238Pu in a VVER-1000 Reactor and the Reactivity Effect Caused by Loss-of-Coolant Accident in the Central Np–Am Fuel Assembly

This paper presents the results obtained from numerical evaluations for the possibility of large-scale 238Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of 238Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor.

Reactivity Measurement Methods and the First Results of the Physical Start-up for the Nuclear Subcritical Facility “Neutron Source”

The methods of the neutron-physical parameter measurements of the subcritical assembly neutron source driven with an electron linear accelerator during start-up are described in the article. The results of neutron multiplication factor and reactivity measurements by different methods are represented. The measurement results are compared with each other and with simulation results of MCNPX code. The measured results of the facility parameters are analyzed.

Uncertainty quantification and target accuracy assessment of nuclear data to effective neutron multiplication factor of heat pipe cooled reactor

With the increasing improvement of the core physics calculation method, the uncertainty introduced by the calculation software and model approximation in the process of nuclear reactor core physics design is becoming smaller and smaller, while the uncertainty of the calculation results caused by the uncertainty of the input parameters itself is becoming more and more important. As an important input parameter of core physics calculation, the uncertainty of nuclear data will be transmitted, which will affect the accuracy of core physics calculation results. Quantifying the uncertainty of the core physics calculation results caused by nuclear data is an important element in the nuclear reactor safety design process. On the other hand, according to the target accuracy requirements of the nuclear reactor design, the key nuclear data and the priority of the improvement when the physical parameters of the nuclear reactor core reach the target accuracy requirements are evaluated, which can provide reference standards for the improvement of nuclear data and guide the development of specific experimental measurements work.

Conceptual Neutronics Study of a Hybrid Fusion Neutron Source FNS-C with Aqueous Blanket

The modern challenges of nuclear energy are the replenishment of dwindling reserves of nuclear fuel and the development of a closed nuclear fuel cycle while complying with strict radiation safety requirements. A fusion neutron source has unique capabilities to solve these problems. The preliminary results of a neutronic analysis of the FNS-C fusion-fission hybrid neutron source with a thorium-uranium aqueous blanket by the Monte Carlo method computer simulation, using the MCNP-4 code with the ENDF/B-VII cross-section library, gives satisfactory results for the study of the possibility of creating a compact source of fusion neutrons based on a small spherical tokamak for commercial use. The obtained results show that the FNS-C hybrid blanket generates enough tritium to fully ensure the uninterrupted operation of the FNS-C throughout the year. The reproduction coefficient of 233U is 1.027 at a consumption of 1304 kg/year of the fissile material in the aqueous blanket containing 232Th enriched to 1.47% 233U. The FNS-C is operated with an effective neutron multiplication factor keff ~ 0.99 with reactivity ρ = –0.006249 in the presence of delayed neutrons, which corresponds to the safest state of the core of thermal neutron fission reactors. The thermal power of the FNS-C at keff ~ 0.99 is ~3 GW, which is comparable to the thermal power of fission reactors. This indicates the potential possibility of creating a safe thorium-uranium breeder power reactor based on a fusion neutron source. The results of the study were obtained for the simplified approximate geometrical FNS-C model. To confirm the preliminary results, it is necessary to develop a more accurate calculation model of the FNS-C machine.