Effective Neutron Multiplication Factor Research Articles

Physical start-up of the Nuclear Subcritical Facility “Neutron Source” of NSC KIPT

National Science Center “Kharkov Institute of Physics and Technology” jointly with the Argonne National Laboratory developed, designed and constructed the Subcritical Nuclear Facility “Neutron source based on the subcritical assembly driven by electron linear accelerator”. The article presents the main results obtained during the physical start-up of the plant, the program of which involved the staged loading of fuel assemblies into the core with subsequent measurements of its neutron-physical characteristics. In total, 37 fuel assemblies of WWR-M2 type were loaded into the facility core. The values o f the effective neutron multiplication factor keff and reactivity ρ of the subcritical assembly were measured after each loading step using the multiplication and area ratio methods. In addition, the neutron flux to electron current ratio method was tested for on-line reactivity monitoring. The investigations were performed when the accelerator was operated with 100 MeV electron beam energy, 2.7 μs electron pulse duration, 20 Hz pulse repetition rate and ≤ 40 mA beam pulse current. The obtained results were analyzed in terms of ensuring the facility nuclear safety at the next stage of its commissioning which is the pilot operation.

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.

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.

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

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.

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.

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.

Shedding light on the 239Pu fission source term with new high-precision experiments and advanced fission modeling

In the last decade, there has been a renaissance of fission research resulting in new high-precision experiments and advanced fission modeling. For instance, the Chi-Nu and CEA teams supplied, for the first time, the 239Pu prompt fission neutron spectrum (PFNS) for broad ranges of incident and outgoing neutron energies. The CEA team also measured 239Pu average prompt neutron multiplicities, ν̄p, with lower statistical uncertainties and a technique significantly different than the one used in the past. The NIFFTE collaboration provided 239Pu(n,f)/235U(n,f) cross section shape ratios with uncertainties below 1% utilizing a novel detector type. Advanced fission event generators were developed, among them CGMF, FIFRELIN, FREYA, and GEF, which calculate post-scission fission observables in a correlated manner. These new experimental data and more consistent fission models change the evaluated PFNS, ν̄p, and (n,f) cross sections, some only modestly, compared to ENDF/B-VIII.0. In turn, the individual new nuclear data distinctly change simulated effective neutron multiplication factors of fast critical assemblies, but their combined impact is small, while affecting the prediction of LLNL pulsed sphere neutron leakage spectra and reaction rates only within experimental uncertainties. Also, the parameters obtained from fitting to ν̄p reproduce various post-scission fission observables within the uncertainties of experimental data. This indicates that new differential experiments and consistent fission modeling reduce compensating errors present in ENDF/B-VIII.0.

Data Assimilation using Non-invasive Monte Carlo Sensitivity Analysis of Reactor Kinetics Parameters

Accurately predicting the criticality of an experiment before interacting with the experimental components is very important for criticality safety. Radiation transport software can be utilized to calculate the effective neutron multiplication factor of a nuclear system. Because of the integral nature of the effective neutron multiplication factor, the value calculated contains various sources of nuclear-data induced uncertainty. The sensitivity analysis and data assimilation technique presented in this paper exhibit one possible method of identifying and reducing the effective neutron multiplication factor nuclear-data induced uncertainty. The results presented in this work show that it is possible to use relative sensitivity coefficients of the prompt neutron decay constant and the effective delayed neutron fraction to 239Pu nuclear data to reduce nuclear-data induced uncertainties in the effective neutron ultiplication factor. This work has been utilized by members of the Los Alamos National Laboratory project EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) for optimally designing a new experiment, which will be used to reduce compensating errors in 239Pu nuclear data.

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

How can a diverse set of integral and semi-integral measurements inform identification of discrepant nuclear data?

Nuclear data are used for a variety of applications, including criticality safety, reactor performance, and material safeguards. Despite the breadth of use-cases, the effective neutron multiplication factor, keff, of ICSBEP critical assemblies are primarily used for nuclear data validation; these are sensitive to specific energy regions and nuclides and are unable to uniquely constrain nuclear data. As a consequence, general-purpose nuclear data libraries, such as ENDF/B-VIII.0 [1], may have deficiencies that, while not apparent in criticality applications, negatively impact other applications, such as non-destructive analysis of special nuclear material [2, 3] and neutron diagnosed subcritical experiments [4]. Recent work by the Experiments Underpinned by Computational Learning for Improvements in Nuclear Data (EUCLID) project developed a machine learning tool, RAFIEKI, which uses random forests and the SHAP metric to determine which nuclear data contribute most to predicted bias between measured and simulated responses (e.g. keff). This paper contrasts RAFIEKI analysis applied to keff only against RAFIEKI analysis with keff paired with either LLNL pulsed sphere measurements or subcritical benchmarks. Two examples show that a) including pulsed sphere measurements substantially increases 9Be nuclear data importance to bias between 2 and 15 MeV, and b) including subcritical benchmarks has the potential for disentangling compensating errors between 240Pu (n,el) and (n,il) cross-sections between 0.1 and 10 MeV. These results show that RAFIEKI analysis applied to response sets that include, but go beyond, keff can aid nuclear data evaluators in identifying issues in nuclear data.

Understanding the impact of nuclear-data covariances on various integral responses using adjustment

The EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project created a library of sensitivities for nine different integral responses with respect to nuclear data. These integral responses were obtained from measurements at LLNL (Lawrence Livermore National Laboratory) pulsed spheres, critical and sub-critical assemblies. At the same time, covariances for ENDF/B-VIII.0 were processed at LANL (Los Alamos National Laboratory). The combination of these data allow us to study the impact of nuclear-data covariances on various integral responses, either by forward-propagating covariances via sensitivities, or by using nuclear data, integral responses, and sensitivities for adjustment. Here, we will present: the impact of 1H, 9Be, 12C, 27Al, 56Fe, 235,238U, and 239,240Pu ENDF/B-VIII.0 covariances on simulated bounds of the following integral responses: LLNL pulsed-spheres neutron-leakage spectra, the effective neutron multiplication factor, reaction rates, and reactivity coefficients of ICSBEP critical assemblies. Also, adjustment results with the same nuclear-data covariances and responses will be discussed. LA-UR-22-31309

Producing Evaluation-quality 239Pu Average Prompt Fission Neutron Multiplicities using a Correlated Fission Model

An evaluation of the average prompt fission neutron multiplicity, ν¯ p, of 239Pu(n,f) is shown. This evaluation includes (a) the correlated fission model CGMF, and (b) a detailed analysis of past and recently published experimental data. Using CGMF-calculated ν¯ pas prior enables to link, through the use of evaluated model input parameters, ν¯ p to other fission observables such as the prompt fission neutron spectrum (PFNS), pre-neutron emission fission yields as a function of mass, and the average total kinetic energy of the fragments. These evaluated parameters produce realistic predictions of many fission observables, while the evaluated ν¯ p agrees well (χ2 ≈ 1) with data. Moreover, with the new evaluated ν¯ p, the effective neutron multiplication factor of fast Pu ICSBEP critical assemblies are predicted with a mean bias of 58 pcm compared to 18 pcm with ENDF/B-VIII.0, when paired with a new 239Pu PFNS and fission cross section. Due to these encouraging validation results, the evaluated ν¯ p is currently part of a release candidate for the 239Pu ENDF/B-VIII.1 file. Hence, a correlated fission model was used for the first time for evaluating ν¯ p that is of evaluation quality. This is an important step towards consistent evaluations of prompt fission observables.

Alternatives to the burnup strategy and fuel material of the EM2 modular reactor

The Energy Multiplier Module (EM2) is a modular high-temperature helium-cooled fast reactor developed by General Atomics (GA). It contemplates a convert-and-burn core, in which fertile fuel is converted to fissile fuel for on-site burning. These features allow the reactor to achieve a core lifetime of 32 years without refueling with different types of fuel materials, like thorium, natural uranium, depleted uranium (DU), and spent fuel. The objective of this study is to evaluate the neutronic performance of the EM2 reactor under four types of fuel material combinations (Low enriched uranium [LEU]/DU, LEU/Th, U-Pu/DU and U/Th) with two burnup strategies. The first strategy was taken from the GA model (configuration 1) and the second was proposed in this paper (configuration 2). The Serpent Monte Carlo code with ENDF/B-VII cross section library was used to perform the calculations. The following parameters were analyzed for each scenario: the effective neutron multiplication factor (keff), the conversion ratio, reactivity coefficients, and the effective delayed neutron fraction (βeff). The fissile inventory and fission rate contribution of the main isotopes were also discussed. The results obtained show that the scenarios with configuration 2 had a higher peak excess reactivity than those with configuration 1. For the scenarios with DU as a fertile fuel, 238U produced ∼20 % of the fission throughout the fuel cycle. In contrast, in the scenarios with thorium as fertile fuel, the 232Th fission rate was negligible. As for the reactivity coefficients, all scenarios presented negative Doppler coefficient and positive void reactivity; both values increased along burnup. On the other hand, the βeff decreased with burnup on and did not vary significantly between both configurations.

A feasibility study of SMART reactor power performance optimizations-part 1: Steady-state and burn-up analysis

SMART is an integral small pressurized water nuclear reactor design with a rated power output of 100 MWe from 330 MWth, but it needs a higher power output for the United Kingdom energy market. This study applies Monte Carlo code OpenMC to build a full-core model and innovatively adjust the simulation coefficients to approach the reactor operating conditions. The analysis results point out the reasonable optimization’s technical direction. The model’s sensitivity to ENDF and JEFF nuclear data libraries and spatial division is tested and verified. Then it performs a series of simulations to obtain the core’s neutronic parameters, such as neutron energy and spatial distributions, effective neutron multiplication factor keff and its variation versus depletion. The analysis found that the initially designed core’s keff is 1.22906, and the temperature reactivity defect is 11612 pcm. In 1129 full-power operating days, the keff will decrease to 0.99126, and the reactor depletes 8.524 × 1026235U atoms. However, the outermost fuel assemblies’ 235U depletion rate is lower than 45% in this extended refuelling cycle, and their ending enrichment is higher than 2.4%. That means the fuel economy of the original design’s two-batch refuelling scheme core layout is insufficient. Improving the thermal neutron fluence in these assemblies may optimize the SMART power performance effectively.