Burnup Analysis Research Articles

Burnup analysis of the pebble-bed fluoride-salt-cooled high-temperature reactor based on the Chord Length Sampling method

The neutronics calculation of the PB-FHR with the fraction ranging from 5% to 30% are carried out by the CLS method and ERM method. It can be observed that the infinite multiplication factor obtained with the CLS method are consistently smaller than those obtained with the ERM method when the fraction is less than 15%. Moreover, the differences of the two methods basically increases with the increasing burnup level when the fraction is less than 15%. For the fraction up to 30%, the CLS method shows significant bias of 520 pcm with the explicit method. After the CLS method with a modified packing fraction correction is adopted, the largest difference drops to 220 pcm when the fraction is 30%.

Analysis of Depletion and Inventory of the Fuel for the MNSR Research Reactor Using the Deterministic Code DRAGON5

The Syrian miniature neutron source reactor (MNSR), a 30-kW, 90.0% highly enriched uranium fueled (U-Al) MNSR-type reactor has gone critical. Under operating conditions of 2 h per day for 5 days a week at a peak thermal neutron flux of 1.0 × 1012 n/cm2·s, the estimated core life is 10 years. After the fuel is depleted, the full spent-fuel assembly will be replaced with new low-enriched uranium. This study presents the results of a multigroup fuel burnup and depletion analysis of the MNSR fuel lattice using the DRAGON5 transport lattice code. Furthermore, infinite multiplication factor k∞ and several two-group macroscopic parameters, including scattering cross section, fission cross section, total cross section, and diffusion coefficient, and the transport mean free path have been studied. In addition to this, fuel isotopic composition dependency on burnup was calculated as a part of this study. The results contained in this study can be used as a microscopic database for performing criticality safety analysis and shielding computations for the design of a spent-fuel storage cask for the MNSR reactor core.

A Transport-Free Method for Predicting the Post-Depletion Spatial Neutron Flux Distribution

An adjoint-based method to predict the variation in spatial flux distribution during a depletion interval is presented in this paper. Burnup analyses require dividing a fuel cycle into multiple time intervals. At the start of each interval, the neutron transport equation is solved, and a subsequent depletion calculation is performed to obtain isotopic concentrations at the end of the interval. The most common approaches are to assume that either the flux or the power are constant through this depletion interval. In reality, changes in material compositions cause the flux and power distribution to change instantaneously, and thus, these assumptions are not valid in general except in the limit of infinitesimally small time steps. To overcome these assumptions, a method for predicting the spatial flux variation (SFV) due to changes in material compositions is derived, implemented, and verified. The formulation relies on the first-order perturbation formulation in conjunction with the forward and adjoint moments of the fission source, obtained from the fission matrix. Moreover, multiple adjoint modes are used to better predict the flux variation following materials transmutations. Such a prediction is capable of mimicking a transport calculation across a depletion interval based on the beginning-of-step transport solution and could be used to extend the simulated time between transport simulations in depletion and fuel cycle analysis. The SFV method is applied to a single three-dimensional fuel pin, depleted using a variety of depletion step sizes and verified against a reference simulation. The results show that the method produces accurate prediction of the end-of-step spatial flux distribution.

Burnup analysis for HTR-10 reactor core loaded with uranium and thorium oxide

Abstract We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO2 and (Th+Pu)O2 mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O2, achieved about 24% higher fuel cycle length than the UO2 case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found.

Fuel burnup analysis and fuel management for TRIGA Mark III research reactor

The fundamental advantage in using Monte Carlo methods for burnup calculations is to formulate an effective optimal fuel management strategy for the TRR-1/M1 research reactor. The core management study has been performed by utilizing the essentially parameters including multiplication factor, power peaking, neutron flux and burnup calculation based on the Monte Carlo calculation. The fuel element burnup was calculated after reshuffling the reactor core. The fuel cycle length and core parameters such as core excess reactivity, neutron flux, axial and radial power factors and other parameters are determined. The core excess reactivity was calculated as a function of burnup. The maximum excess reactivity shall not exceed 6.3% Δk/k. The maximum fuel temperature shall not exceed 930 °C during steady-state operation. Typically, a core loading operated with the maximum burnup between 100 to 200 MWD depending on the utilization requirements. The thermal neutron flux in the irradiation positions is within the order of 1011 - 1013 n/cm2-sec. The study gives valuable results into the behaviour of the TRR-1/M1 research reactor and will ensure optimized utilization and operation of the reactor during its life time. It will establish the strategic planning for fuel management in the reshuffling and reloading schemes patterns and its safe implementation in the future.

Monte Carlo fuel burnup analyses of the Accelerator-Driven Subcritical systems

This paper describes the Monte Carlo fuel burnup analysis performed for the Argonne National Laboratory (ANL) Accelerator-Driven Subcritical (ADS) conceptual design. The ANL ADS conceptual design was developed to dispose of the 131 tons of Minor Actinides (MA) fuel of the current United States Spent Nuclear Fuel (SNF) inventory. It uses a neutron source from 1 GeV 25 MW proton beam with the liquid lead target to drive the subcritical fission assembly. It utilizes the liquid metal slurry fuel concept to load the MAs fuel. The subcritical fission assembly has a neutron multiplication factor of 0.98. The total fission power generated in the system is about 2.67 GW. The system is designed to operate for 35 full power years. Monte Carlo fuel burnup analyses were performed to calculate the amount of MAs consumed by the ANL ADS conceptual design. The MCNP6 computer program was used to simulate the high energy physics of the protons striking the liquid lead target for generating the neutron source. The MCB5 and SERPENT Monte Carlo burnup computer programs were used to simulate the fuel burnup during the ADS operation. In the ANL ADS conceptual configuration, the reactivity loss during the fuel burnup is compensated by adding fresh fuel into the subcritical assembly. In this paper, both batch and continuous fueling modes are examined. The fresh fuel particles are loaded into the slurry fuel carrier at the beginning of each burnup step of the batch-fueling mode and loaded continuously into the slurry fuel carrier during the operation in the continuous fueling mode. Monte Carlo fuel burnup analyses show that one ANL ADS conceptual design can consume about 25.7 tons of MAs fuel during the 35 full power years with the continuous fueling mode. About five units is required to dispose of the 131 tons of MAs.

Development and validation of depletion code system IMPC-Burnup for ADS

Depletion calculation is important for studying the transmutation efficiency of minor actinides and long-life fission products in accelerator-driven subcritical reactor system (ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEA-ADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover, the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.

Numerical calculation of fuel burn-up rate in a cylindrical nuclear reactor

The prediction of nuclear reactor fuel burn-up rates throughout a reactors lifetime is an important problem in reactor core design. In this study, we present a novel algorithm called BuCal code. This code was used to perform burn-up analysis for a pressurized water reactor fuel with $${\text{UO}}_{2 }$$ whose concentration is 19.5% enriched. Simulation results indicate that the total estimate of 235U consumption in 225 days with high neutron fluence is approximately 99.9% of the initial value. The study further showed that the microscopic absorption and fission cross sections decreases with increasing temperature and the concentrations of 235U and 238U decreases as the numbers of days increases while 236U build-up as the number of days increases.

Depletion chain optimization of lattice code STREAM for LWR fuel assembly burnup analysis

A depletion chain simplification method is applied to UNIST lattice code STREAM (Steady state and Transient REactor Analysis code with Method of Characteristics) in this paper to alleviate the computational burden of depletion calculation associated with a large depletion matrix. A simplified burnup matrix (burnup chain) of 464 nuclides and 10,638 transitions is thus created from a large detailed burnup matrix containing 3,837 nuclides and 43,416 transitions from ENDF/B-VII.0 nuclear decay data. The simplified burnup matrix is optimized for the purpose of calculating effective neutron multiplication factors (keff) and power distributions with reduced computation time and memory usage. The nuclide selection method relies on the Generalized Perturbation Theory (GPT): by exploiting the adjoint function of nuclide number densities as derived with GPT, a set of nuclides which must be included in the simplified chain (nuclides with important contribution to reactivity) is determined. Numerical verification of the simplified burnup chain is conducted using the deterministic neutron transport analysis code STREAM, developed to perform whole light water reactor (LWR) core calculations with the direct transport analysis method and the two-step method. The simplified burnup chain is tested on the 16 LWR fuel assembly depletion problems from the Virtual Environment for Reactor Application (VERA) benchmarks, including burnable poison (BP) pin cells commonly used in nuclear reactor design such as gadolinia, Pyrex, AIC, B4C and IFBA, and one OPR-1000 fuel assembly with gadolinium bearing fuel depletion problem. For burnup up to 80 MWd/kg and 235U enrichment ranging from 2.1 w/o to 4.6 w/o, the calculations with the simplified burnup chain predict the keff variations within 10 pcm and identical power distributions, with a speed-up factor from 20 to 40 in depletion calculation and a reduction factor by 3–4 of the total simulation time compared to the ones with the detailed burnup chain.

Fuel Breeding Analysis On Low Moderated Fuel Ratio Based On Actinides Closed Water-Cooled Thorium Reactor

Utilization of spent nuclear fuel and some fuel breeding capabilities of nuclear fuels to extend the sustainability aspect of nuclear fuel become more important issues to be optimized. Thorium fuel utilization based on water-cooled reactor is one of the possible options to be used and optimized as well as uranium fuel utilization. Some schemes of accumulated spent nuclear fuels can be used as recycled fuel in water-cooled reactor based on thorium fuel. In the present analysis, fuel sustainability aspect of nuclear fuel will be evaluated, which is based on a water-cooled reactor. As a fuel basis, thorium is used with can be mixed with additional recycled spent nuclear fuels. Some minor actinides (MA) as recycled fuels are used as doping material to be loaded to the water cooled reactors with thorium fuel as fuel basis and heavy water as moderator and coolant. The evaluation has been made by adopting a computational simulation of an equilibrium burnup analysis method, which was coupled with cell calculation of computer code of SRAC with JENDL.32 as nuclear data library. Several survey parameters have been evaluated to evaluate some effect of MA doping rate, different moderation ratio and power density levels to the reactor performance including fuel-breeding capability and void reactivity coefficient. Effect of some actinide composition to fuel breeding capability as well as safety aspect, which is based on void reactivity coefficient have been investigated. Fuel breeding capability can be obtained by the present reactor systems; as well as negative void reactivity has been show for more moderator ratio and less power density. Low portion of moderation to fuel ratios (MFR) are used to have a better fuel breeding capability as well as some from contribution from recycled fuel of minor actinides (MA) and less power density. A negative void reactivity can be obtained in this system and it becomes less negative for doping MA and more power density as well as a positive void reactivity coefficient value for much less moderation ratio.

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems (ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of (1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history, (2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and (3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces the validation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover, the proton irradiation module was verified by a comparison with the analytic method of calculating the 210Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.

Burnup analysis for the pebble-bed fluoride-salt-cooled high-temperature reactor based on the energy-dependent elastic scattering cross-sectional model

To carry out accurate burnup calculations for a pebble-bed fluoride-salt-cooled high-temperature reactor, the energy-dependent cross-sectional model based on the Doppler broadening rejection correction method has been proposed to develop the energy-dependent elastic scattering cross-sectional model. In this study, the Monte Carlo continuous energy code PSG2/Serpent was used to examine the difference between the constant cross-sectional model and the energy-dependent cross-sectional model during burnup. For the cases analyzed in this study, numerical simulations show that the multiplication coefficient was improved by hundreds pcm and 239Pu concentration was improved by approximately 1–2% during burnup when the energy-dependent elastic scattering cross-sectional model is considered.

Monte Carlo burnup and its uncertainty propagation analyses for VERA depletion benchmarks by McCARD

Abstract For an efficient Monte Carlo (MC) burnup analysis, an accurate high-order depletion scheme to consider the nonlinear flux variation in a coarse burnup-step interval is crucial accompanied with an accurate depletion equation solver. In a Seoul National University MC code, McCARD, the high-order depletion schemes of the quadratic depletion method (QDM) and the linear extrapolation/quadratic interpolation (LEQI) method and a depletion equation solver by the Chebyshev rational approximation method (CRAM) have been newly implemented in addition to the existing constant extrapolation/backward extrapolation (CEBE) method using the matrix exponential method (MEM) solver with substeps. In this paper, the quadratic extrapolation/quadratic interpolation (QEQI) method is proposed as a new high-order depletion scheme. In order to examine the effectiveness of the newly-implemented depletion modules in McCARD, four problems in the VERA depletion benchmarks are solved by CEBE/MEM, CEBE/CRAM, LEQI/MEM, QEQI/MEM, and QDM for gadolinium isotopes. From the comparisons, it is shown that the QEQI/MEM predicts kinf's most accurately among the test cases. In addition, statistical uncertainty propagation analyses for a VERA pin cell problem are conducted by the sensitivity and uncertainty and the stochastic sampling methods.

One candidate method based on short-lived indicator for burnup analysis

One novel gamma spectrometry method based on short-lived fission products was developed that can infer the level of burnup without complex computations or knowing irradiation history. The relationship between the burnup information and equilibrium concentration of short-lived indicators was derived. In addition, new candidate indicators for gamma spectrum analysis, such as 88Kr, 91Sr, 92Sr, 135I and 142La were screened out. The results of three cases simulated with ORIGEN and MCNP were in good agreement with the theoretical values.

Burnup characteristics analyses of graphite impurities in HTGR fuel element

In the fuel element of high-temperature gas cooled reactors (HTGRs), there exist various impurities in the graphite matrix, which are mainly originated from the raw materials, such as naturally occurring graphite, artificial graphite and resin. These impurities can reduce the excess reactivity of the reactor core, both in the initial criticality and during the overall operation history. In the practical physics design of HTGR, the equivalent boron content (EBC) method is usually used to describe the graphite impurities, where the impurity isotopes are simply converted to some amounts of boron by preserving the neutron absorption. Although it is found to be effective in criticality calculations, EBC may introduce errors in burnup analyses. This paper investigates the burnup characteristics of the graphite impurities. According to their behaviors during burnup, the impurities are classified into three categories, i.e. the high-burnable isotopes, low-burnable isotopes and unburnable isotopes. Moreover, a revised equivalent boron content (REBC) is proposed to describe the absorption rates of impurities, which can improve the computational accuracy in the burnup analyses of HTGR fuel.

Conceptual design study of beam window for accelerator-driven system with subcriticality adjustment rod

The design of the beam window of the spallation target is one of the critical issues in the accelerator-driven system (ADS) design. This study aims to perform the coupled analysis for the feasible beam window concept. To mitigate the design condition, namely to reduce the necessary proton beam current, subcriticality adjustment rod (SAR) was installed to the ADS core. The burnup analysis was performed for the ADS core with SAR and the results indicated that the maximum proton beam current during the burnup cycle was reduced from 20 to 13.5 mA.Based on the burnup analysis result, the coupled analysis; particle transport, thermal hydraulics and structural analyses, was performed. As the final result, the most robust beam window design; the hemisphere shape, the outer radius = 235 mm, the thickness at the top of the beam window = 3.5 mm and the factor of safety for the buckling = 9.0, was presented. The buckling pressure was 2.2 times larger than the previous one and more feasible beam window concept was presented through this study.

Burnup Performance of a PBR with an Accumulative Fuel Loading Scheme Utilizing Burnable Poison Particles in UO2 and ROX Fuels

The accumulative fuel loading scheme is a promising design for pebble bed reactors (PBRs). In a previous study, burnup analysis for such a reactor was performed using UO2 fuel and the results showed that the maximum burnup and operation period were 223 GWD/t and 10.2 years, respectively. In this study, rock-like oxide (ROX) fuel was analyzed in PBR with an accumulative fuel loading scheme. The results showed that a maximum burnup and operation period of 218 GWd/t and 8.4 years, respectively. However, high excess reactivity occurred in the initial condition for both UO2 and ROX fuels. In this study, burnable poison (BP) particles were used to suppress the initial excess reactivity. Several BP materials such as B4C, Gd2O3, Er2O3 and CdO were investigated. In case of UO2 fuel, the combination of B4C and Gd2O3 was found to be optimal for flattening the reactivity swing in the initial condition. However, due to the fissile density of ROX fuel being about five times lower than that of UO2 fuel, single-material BP particles (B4C) seemed to be optimal for flattening the reactivity swing. By utilizing the BP particles at the initial condition, the results of UO2 and ROX fuels showed that maximum keff was decreased from 1.423 to 1.068 and from 1.357 to 1.033, respectively. On the other hand, the maximum discharge burnup value and core lifetime were quite similar to the results for the reactor that did not use BP particles at the initial condition.

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

ABSTRACT Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of 237Np, 238Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of 237Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO2 or UO2–Gd2O3), which was similar to the case of PWR fuel. The differences in the number densities of 235U, 239Pu, 240Pu, 241Pu, 149Sm, and 151Sm have a large impact on k eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.

Thorium-based mixed oxide fuel in a pressurized water reactor: A burnup analysis with MCNP

A three-dimensional MCNP model of a Westinghouse-type 17×17 PWR was used to assess the feasibility of enhancing the current fuel supply with a thorium-plutonium mixed oxide fuel (ThMOX). A conventional UO2-fueled configuration was compared to a configuration which incorporated ThMOX fuel. The safety of the ThMOX configuration was compared to that of the UO2 configuration at several time steps of interest within all three cycles (beginning of cycle, peak excess reactivity of cycle, and end of cycle) using the following metrics: axial and radial nuclear hot channel factors FZNandFRN, moderator and fuel temperature coefficients (MTC and FTC), delayed neutron fraction (βeff), and shutdown margin (SDM). Additionally, the performance of the ThMOX configuration was assessed by tracking cycle lengths, the amount of plutonium destroyed, and fission product poison concentration.The ThMOX configuration’s FZN (1.455±0.222) was slightly less favorable than that of the UO2 configuration (1.515±0.246), though its FRN was statistically identical (3.623±0.179 compared to 3.430±0.066 for UO2). The delayed neutron fraction was another area of concern for the ThMOX configuration. It was lower than that of the UO2 configuration, but the lowest βeff occurred at the beginning of Cycle 3 (3.324E−03±8.919E−07). The βeff of the UO2 configuration was 46% higher (6.177E−03±1.510E−06) at this time step. Similarly, the ThMOX configuration’s SDMs were less favorable than those of the UO2 configuration, though even the lowest (3971±43pcm) remained well within the NRC prescribed limit of 1300pcm. However, the MTC of the ThMOX configuration was more favorable for all time steps except the beginning of Cycle 3, and even then it was negative as required by the NRC (−2.325±0.00098pcm/°F). The FTC results were also more favorable for the ThMOX configuration. Its highest FTC value (−6.718±0.001pcm/°F at the peak of Cycle 1) was more negative than all but the lowest FTC of the UO2 configuration (−7.335±0.002pcm/°F at the end of Cycle 3).Overall, incorporating ThMOX fuel into the fuel supply for a conventional PWR seems feasible from a neutronics standpoint. However, this analysis would benefit from a more detailed burnup simulation and optimization of the ThMOX fuel composition. Also, a full thermal hydraulic analysis in conjunction with a neutronics analysis is required.

Fuel burnup analysis for Thai research reactor by using MCNPX computer code

This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.