Rod Worth Research Articles

Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence

This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.

Validation of DDC-3D code system for neutronics and thermal-hydraulics coupling analysis using BEAVRS benchmark

The direct whole-core calculations can provide accurate results and insights to the physics phenomena of the reactor. It can also capture the local effects of temperature and density fields on fuel depletion. However, the computational cost of the direct whole-core calculations is expensive. To compromise between computational cost and accuracy, the DDC-3D code system has been developed to perform neutronics and thermal-hydraulics coupling analysis. The DDC-3D code system couples the open-source codes DRAGON & DONJON based on two-step method and subchannel code COBRA-EN. The Picard iteration method is applied to ensure the stability of numerical calculation. Then the BEAVRS benchmark is used to validate the computational capabilities of DDC-3D code system. The critical boron concentrations, control rod worths and fission rate distributions are calculated in HZP condition. The results show a good agreement with measured data. The results demonstrate that the two-step method is applicable and valid for multiphysics simulations. For the result of HFP condition for cycle 1, the results also agree well with measured data, including the trend of the critical boron concentrations and power distributions throughout the cycle 1. Although the detailed thermal–hydraulic experimental values are not available, the thermal-hydraulics analysis of the hot fuel assemblies indicates that the calculation results are reasonable. In general, the results demonstrate the feasibility and accuracy of DDC-3D code system for neutronics and thermal-hydraulics coupling calculations and life cycle simulation of PWRs.

Neutronic analysis of new generation TVS-2M fuel assemblies in VVER-1000 reactor type

The Russian fuel assembly TVS-2M is a newly proposed fuel configuration that contains gadolinium-oxide at different concentrations mixed with enriched UO2 at different U-235 enrichment values. The purpose of this study is to investigate the neutronic behavior of the Bushehr VVER-1000 reactor core by replacing the current standard Fuel Assemblies (FAs) called TVS with the new generation of TVS-2M FAs. In such manner, the reactor core was simulated under both modes of TVS-2M and TVS FAs arrangement using the WIMS, CITATION, MCNPX, and SuperMC codes in equal terms. Neutronic parameters of the reactor core such as effective multiplication factor, excess reactivity, neutron flux distribution, power peaking factor, and control rod worth are calculated and discussed within a comparative framework of both configurations of FAs; TVS and TVS-2M.

Analysis of initial core and time dependent fuel burnup for high temperature testing reactors (HTTRs)

Abstract HTTRs are high temperature testing reactors which are considered generation IV nuclear power reactors. The reactor core startup and fuel burnup time dependent are simulated and analyzed. For the present computer model core startup, fuel loading with approach to criticality, control rod critical positions, scram reactivity and control rod worth, different core excess reactivity are evaluated. The results are compared with both typical measurements and other previously published methods. Moreover, the time dependent core model and fuel burn up are simulated to calculate the core multiplication factor, cycle length, discharged burnup, core isotopic evolution with time and burnable poisons behavior.

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.

Performance evaluation and validation of OpenMC code for criticality analysis of an MTR-type research reactor

This paper assesses the performance of OpenMC, an open-source Monte Carlo code for particle transport, through criticality analysis of an MTR-type research reactor. The evaluation involves validation against experimental data from the NUR research reactor's first fresh fuel core and a code-to-code comparison with MCNP5, using a consistent methodology ensuring identical input conditions. Key parameters, including excess reactivity, control rod worths, shutdown margin, critical configurations, kinetics parameters, neutron flux, power distributions, and peaking factors, were examined. Results indicate good agreement between OpenMC simulations, experimental data, and MCNP5 results, with marginal differences well within acceptable ranges. Discrepancies with experimental data are below 5 %, considering measurement accuracy, while code-to-code comparisons reveal relative differences below 1 % in most cases. This study supports OpenMC's reliability for MTR reactor criticality analysis, affirming the validity of the developed NUR reactor model for future applications and highlighting the code's credibility and applicability in the nuclear reactor simulations.

Neutronic characteristic analysis of the GAMA microreactor

The GAMA Microreactor is a low-power nuclear power plant dedicated for remote area with a power output of 300 kWe. The primary characteristics of the GAMA Microreactor are its system compactness, design simplicity, and safe operation without moving parts. The fuel is uranium hydride powder, which works simultaneously as a moderator, contained within a stainless-steel vessel and surrounded by a graphite reflector. Liquid sodium-filled heat pipes are embedded in the fuel powder to extract fission-generated heat via passive coolant flow due to capillary action. As a part of the design process, neutronic parameters of the GAMA Microreactor have been determined using SCALE6.2 and OpenMC codes. Among the analyzed parameters are the excess reactivity as a function of burn-up, shutdown margin, control rod worth, temperature coefficient of reactivity (TCR), fuel density coefficient of reactivity (DCR), and reactivity coefficient due to hydrogen dissociation. From the calculation results, the GAMA Microreactor has a maximum excess reactivity of 0.049 at the operational temperature (700 °C) and 0.0788 at ambient temperature. The TCR value is negative (−2.267 pcm/°C) at the beginning of the cycle (BOL) and −4.9618 pcm/°C at the end of the cycle (EOC). Hydrogen dissociation imposes negative reactivity on the reactor both at the BOL (−0.2631 Δk/k-%dissociation) and EOC (−0.3524 Δk/k-%dissociation). Therefore, the GAMA Microreactor maintains its inherent safety characteristics over a considerably long operation time.

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient ΔρBU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

Effects of Neutron Flux Distribution and Control Rod Shadowing on Control Rod Calibrations in the Oregon State TRIGA® Reactor

Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four OSTR control rods is measured using the rod-pull method. Reactor power and period measurements in this method rely on the fission chamber power detector on the north side of the reflector. It is proposed that the location of the fission chamber and the neutron flux distribution in the core may result in an inaccurate reactor period measurement due to the asymmetry of the neutron flux distribution in the OSTR core. The asymmetry of the flux is believed to be more pronounced during super-criticality, resulting in errors in the time-of-power-rise measurements. As a result, control rod calibration experiments may under-predict or over-predict the reactivity worth of certain control rods. A time-independent Monte–Carlo method for the quantification of these effects is presented. Thermal flux maps at the core axial mid-plane are obtained from the model to inform discrepancies between predicted and observed results.

Benchmarking of the SPERT-III E-core experiment with the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC

This paper presents a code-to-code verification of the SPERT-III reactor in its E-core configuration using the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC. A SPERT-III E-core model was originally developed for the Monte Carlo code TRIPOLI-4® and will be the baseline of our analysis The simulation results obtained for the main reactor physics parameters (neutron effective multiplication factor, rod worth, reactivity coefficients and kinetics parameters) with TRIPOLI-4®and OpenMC using different nuclear data libraries are compared and analyzed. Additional verifications are carried out with the next-generation Monte Carlo code TRIPOLI-5®, which is currently under development by CEA and IRSN (France). This paper summarizes a joint study that was conducted within an international research collaboration between CEA Paris-Saclay (France) and the Nuclear Research Center Negev (NRCN, Israel), and is a stepping stone towards future multi-physics simulations exploring the SPERT-III reactor power excursions, including thermal-hydraulics feedback.

Unearthing the Chicago Piles: A Monte Carlo perspective

The divergence of the Chicago Pile 1 (CP-1) on December 2nd, 1942, marks the birth of the nuclear age. Built under the West Stands of the Stagg Field in Chicago in slightly over two weeks, CP-1 was the first man-made nuclear reactor, conceived as a proof-of-principle physics experiment to demonstrate the feasibility of a self-sustained neutron chain reaction. The fuel consisted in a mixture of uranium oxide and metal uranium, with lumps of several sizes and shapes stacked within bricks of graphite used as moderator to form a cubic lattice. After only three months of operation, CP-1 was dismantled due to safety reasons and rebuilt at the Argonne Site in twenty-one days, under the name of Chicago Pile 2 (CP-2): the new pile had a longer life, contributing vastly to research on material science and nuclear reactor theory, and was finally decommissioned in 1954. Prompted by the eightieth anniversary of the Chicago Pile(s), in this talk we celebrate these astonishing scientific artifacts by building fully detailed three-dimensional digital twins of CP-1 and CP-2 using TRIPOLI-4®, the Monte Carlo code developed at CEA. Technological data for the models have been retrieved from published and unpublished documents of the Manhattan Project. We compare the simulation results to the available experimental measurements, encompassing the reactivity of the piles, the control rod worth, the kinetics parameters and several reactivity coefficients. For each, we examine the associated uncertainties due to missing or incomplete technological data as well as the bias due to nuclear data libraries.

Generating multi-group cross-sections using continuous-energy Monte Carlo method for fast reactor analysis

The deterministic two-step method, comprising multigroup cross-section generation and core calculation, is widely applied in fast reactor design and analysis. Monte Carlo (MC) methods with continuous energy and fine geometry provide high-precision cross-sections essential for advanced fast reactor neutronics analysis. This paper presents an analysis of integrating MC-generated homogenized cross-sections with various core solvers, demonstrating their effectiveness and potential improvements in fast reactor simulations. For diffusion core calculations, the superhomogénéisation (SPH) technique reduces control rod worth overestimation from 13.5% to 0.35% in the MET-1000 benchmark, improving power distribution predictions. In transport core calculations, the flux-moment homogenization technique (MHT) addresses reactivity overestimation by incorporating cross-section anisotropy, reducing error by 698 pcm. For Method of Characteristics (MOC) core calculations, transport-corrected multigroup cross-sections yield high precision in pin-by-pin power distribution for a 100 MWe lead-based fast reactor benchmark. While MC methods require significant computational resources, such as 62 CPU-hours for the MET-1000 core and 85.5 CPU-hours for the 100 MWe lead-based fast reactor core, they offer flexibility in geometry modeling. This work highlights MC multigroup cross-section generation methods applicable to diffusion, MOC, and transport core calculations for fast reactor analysis, suggesting further exploration into their performance in various reactor parameters and computational efficiency.

Characterization of neutronic parameters and radiation shielding design for an aqueous homogeneous reactor

In this research, an aqueous homogeneous reactor is simulated and its neutronic parameters are investigated. Due to the negative reactivity caused by temperature change and void formation, the cold excess reactivity is considered equal to 4 βeff. Also, calculations of radiation shielding and neutronic characterization of the reactor are performed. For different heights of uranyl sulfate fuel, the effective multiplication factor (keff) of the reactor is calculated under cold zero power and hot full power conditions to assess the required excess reactivity. The amount of fuel burnup and changes in the keff are investigated for 500 days, finally 200 operating days are chosen as the working duration of the reactor. The control rods worth (CRW) for different diameters of boron carbid and reactor safety parameters including shot down margin (SDM) and safety reactivity factor (SRF), temperature and void coefficients of reactivity are analyzed. In order to design a suitable radiation shield for this reactor, two different shields are considered. The first shield is a combination of polyethylene thicknesses and barite concrete, and the second shield is made only of barite concrete. The total thickness of the first and second shield was 230 and 180 cm, respectively. Due to the problems of using polyethylene (e.g. problems related to machining, shield dimensions and low melting temperature), the second shield is chosen as the suitable radiation shield for this reactor. Using this radiation shield, the total dose rates of neutron and gamma rays satisfy the ICRP dose limits.

A practical subcritical rod worth measurement technique based on the improved neutron source multiplication method

The control rod worth is a key safety parameter required to be measured in commercial pressurized water reactors (PWRs). Conventionally, the control rod worth is measured after reaching the critical state, which occupies the considerable time in the zero-power physics test. In this study, an efficient control-rod worth measurement technique has been proposed based on the improved neutron-source multiplication method, which can be implemented with the source-range detector count rates in the subcritical states. Moreover, the noise reduction technique has been adopted to smooth the large fluctuation existing in the original signals. In order to verify the engineering performance of the proposed measurement technique, the measured source-range detector count rates during the rod withdrawal process before reaching critical state in a CNP1000 reactor have been employed. It demonstrated that almost all estimated results of control rod worth satisfy the engineering acceptance criteria, except one control rod with the relative difference over 10 %, which indicates the capability of the proposed method in estimating control rod worth.

Research on Neutronics Safety Parameters of the AP1000 Nuclear Reactor under Different Conditions

Changes in temperature during reactor operation may cause changes in physical parameters, leading to core overheating and accidents. It is essential to analyze and assess the safety parameters of the core under different operating conditions. This paper investigates the effects of fuel temperature, moderator density, boron concentration, and control rod state on AP1000 safety parameters. The study uses RMC and NJOY to calculate the changes in reactivity factor, effective delayed neutron fraction, and neutron generation time of the AP1000 reactor under different operating conditions. The changes in reactivity coefficients, neutron fluxes, and relative power densities of AP1000 reactors are analyzed for normal and accidental operating conditions. The results indicated that the reactivity coefficient remained negative under accident conditions, which ensured the safe operation of the reactor. The delayed neutron fraction, neutron flux, and power density distributions are affected by fuel temperature, moderator density, and control rod position. The control rod worth was sufficient for the emergency shutdown of the reactor under accidental conditions. It is demonstrated that the operation of the AP1000 reactor under study conditions is safe and controllable.

Sensitivity of a control rod worth estimate to neutron detector position by time-dependent Monte Carlo simulations of the rod drop experiment

The control rod worth sensitivity to the neutron detector position in the rod drop experiment is studied by the time-dependent Monte Carlo (TDMC) neutron transport calculations for AGN-201K educational reactor and the Kyoto University Critical Assembly. The TDMC simulations of the rod drop experiments are conducted by the Seoul National University Monte Carlo (MC) code, McCARD, yielding time-dependent neutron densities at detector positions. The detector-position-dependent results of the total control rod worth calculated by the extrapolation, the integral counting, and the inverse methods are compared with the numerical reference using the MC eigenvalue calculations and the experimental results. From these comparisons, it is observed that the total control rod worth can be estimated with a considerable difference depending on the detector position through the rod drop experiment. The proposed TDMC simulation of the rod drop experiment can be applied for searching a better detector position or quantifying a bias for the control rod worth measurement.

Multi-physics coupled analyzes of research nuclear reactors based on steady-state and kinetics models

Nuclear research reactors are the workhorses of nuclear material irradiation and isotope production, but the complex assembly geometries and physical coupling effect adds difficulties to their accurate safety analysis. In this study, a multi-physics coupled method for assessing the steady-state and kinetic behaviors of the LVR-15 and MPRR research reactors is proposed. Two-dimensional whole-core Monte Carlo models are utilized to produce homogenized assembly cross-sections in order to describe complex core geometries. Utilizing a three-dimensional variational nodal transport method with unstructured triangular-z meshes, the neutronic field is calculated. Besides, a multi-channel model and a finite difference heat conduction model are employed to determine the temperature distributions of coolant and fuel. Based on these models, the MORPHY (Multi-purpose Reactor multi-PHYsics analysis) code platform is created and verified using the NEACRP neutronics/thermal-hydraulics benchmark. The steady-state and transient behaviors of LVR-15 and MPRR are analyzed using MORPHY. The steady-state results are found to agree well with Monte Carlo reference solutions. For LVR-15 and MPRR, the keff errors are 377 pcm/383 pcm, the maximum assembly power errors are 6.49%/10%, and the maximum control rod worth errors are 31 pcm/136 pcm, respectively. In the meantime, transient processes including unprotected loss-of-flow accidents and reactivity-initiated accidents are studied. Simulation results for both reactors exhibit reasonable negative feedback phenomena. Moreover, it is discovered that despite the fact that the two reactors have distinct core designs, their accident-related behaviors are similar due to the use of the same types of fuel material and water coolant.

Study of neutron flux redistribution and shadowing effect in rod worth measurements with the rod drop experiment

The main goal of this study is to quantify the neutron flux redistribution in detectors and the shadowing effect between control rods by proposing an instrumental calibration factor for measurement, based on computer simulations. The MCNP code was used to simulate the rod drop experiment in the IPEN/MB-01 research reactor. A study of the reactivity behavior was carried out during the most transient period of the rod drop experiment (t < 10 s), demonstrating great agreement between the simulation and the experiment carried out using the source jerk method. A methodology was developed to estimate the shadowing effect and the flux redistribution, and to propose a correction of the results obtained (calibration) in the simulations using the MCNP code and the rod drop experiment respectively. The maximum deviations between the reference value obtained from measuring the control banks via the stable period method and the simulated values were 0.5 % and 11.3 %, respectively.

Annular fuel for VVER-1000 reactor: Moderation impact on neutronic behavior

The impacts of moderation on the neutronic behavior of the VVER-1000 reactor assembly have been investigated by using annular fuel and compared with the conventional one. The study revealed that annular fuel shows a softer neutron spectrum, a better control rod worth and 16% less fissile content compared to the conventional fuel. After 60 MWd/kg burnup, annular fuel contains 6.45% and 255.41% more Pu-239 and Pu-241 respectively than the conventional fuel which can be a valuable source of fissile materials for the MOX fuel. Additionally, annular fuel shows a higher degree of symmetry in power peaking factor (PPF), higher scattering rate, 3.41% less tritium production at the beginning of cycle, and better shutdown margin without significant changes in axial & radial flux. From the neutronics point of view, annular fuel can be a propitious replacement of conventional fuel for the VVER-1000 reactor.