PWR Fuel Research Articles

Calculation of Isotope-Specific Fission Rates for Neutrino Source Calculation in PWR Fuel Diversion Safeguarding

The emission of antineutrinos from nuclear reactors offers a potential avenue for the international enforcement of reactor safeguards. A variety of frameworks have been proposed for detecting these particles, with the objective of verifying an agreed-upon composition of fuel in the operating reactor. More specifically, these frameworks should identify the diversion of a “significant quantity” of fissile material from an agreed upon core loading. For any quantitative analysis of these frameworks, isotope-specific fission rates of a nuclear reactor are required to calculate the reactor neutrino source. Unfortunately, the calculation of isotope-specific fission rates for a realistic core is nontrivial and can require significant simulation efforts. Therefore, this work uses industry-standard simulation tools (CASMO-4/SIMULATE-3) to provide isotope-specific fission rates for a set of 15 plutonium diversion scenarios for a mixed-oxide-loaded pressure water reactor. These diversion scenarios span a wide range of diverted fuel amounts, from 2.17 to 655.19 kg of fissile plutonium. The isotope-specific fission rates reported in this paper can be combined with a neutrino emission model for the direct calculation of the reactor neutrino source. This work can be considered a dedicated effort toward the calculation of realistic isotope-specific reaction rates for use in the development and analysis of safeguarding frameworks. As such, these isotope-specific fission rates are provided over three cycles with realistic fuel loading and shuffling patterns. In this way, this work can act as a standard neutrino source reference for the development and comparison of safeguarding frameworks.

Effect of Protactinium-231 (Pa-231) Addition on Effective Multiplication Factor Value of ThN-U233N Fueled PWR Reactor

This study aims to determine the optimum percentage of addition of protactinium-231 in a ThN-U233N fueled PWR reactor.In ThN-U233N, uranium-233 is used as the fissile material. Pa-231 as burnable poisson is added to ThN-U233N fuel to reduce the keff value. In this research conducted benchmarking keff values ​​of OpenMC and SRAC codes, homogeneous and heterogeneous core configurations, and the addition of Pa-231 in ThN-U233N fuel. Neutronic analysis was performed using the OpenMC code. In the homogeneous core configuration, the enrichment percentage of U-233 is varied by 1%-15%. Meanwhile, the heterogeneous core configuration uses 3 variations of ring geometry and fuel. The results of the benchmarking of the OpenMC and SRAC code keff values ​​show a maximum error of 1,906%. The homogeneous core configuration produces the optimum value for U-233 enrichment of 3%. Meanwhile, the optimum value of the heterogeneous core configuration is achieved in ring geometry F1 = F2 = F3 = 3 rings with enrichment of U-233 at F1 = 0,5%, F2 = 3%, F3 = 5,5% and addition of Pa-231 of 1.5%. The optimum calculation results has an excess reactivity value 8.37%.

Numerical Investigation of Cladding Ballooning and Burst in VVER and PWR Fuel Rods in Experiments with Various Loading Conditions

The paper presents the results of numerical modeling for ballooning and burst of fuel rods claddings made of domestic and foreign alloys. The integral code SOCRAT-V1/V2 is used as a means of modeling. The uncertainty analysis of the calculation results to input uncertainties of the temperature and pressure measurements was performed. The modeling results demonstrate a good qualitative and quantitative compliance with measured times of cladding failure under partial core uncovery conditions. The results of SOCRAT-V1/V2 validation evidence on the importance of performing new experiments for domestic fuel rod cladding ballooning and burst.

Comparative Study of UO2 and (Th,U-233)O2 Performance in Small Long-Life PWR Fuel Cell

A study was performed comparing the performance of UO2 and (Th,U-233)O2 fuel in small long-life PWR. The neutronic calculation carried out by PIJ module of SRAC2006 was done to a fuel cell in 10 years of operation. The calculation was conducted by varying the enrichment of U-235 in UO2 and U-233 in (Th,U-233)O2 for 1% - 20% and also by varying the fuel volume fraction for 40%, 45%, 50%, 55%, and 60%. The performance was observed by comparing the enrichment needed by each fuel type to gain criticality in 10 years, the infinite multiplication factor (k-inf) value, and the conversion ratio (CR) value. The calculation results showed that 60% fuel volume fraction gave critical conditions with the lowest infinite multiplication factor and highest conversion ratio for both fuel types. While in terms of fissile nuclide enrichment needed, (Th,U-233)O2 had better performance than UO2, because only 5% U-233 was needed in (Th,U-233)O2 while UO2 needed 9% U-235 to gain criticality in 10 years of operation.

ANALYSIS OF THORIUM PIN CELL BURN UP OF THE PWR USING WIMS CODE

A thorium-fueled benchmark comparison was made in this study between state-of-the-art codes, WIMSD-5B code to MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations as part of efforts to examine the possible benefits of using thorium in PWR fuel. WIMSD-5B calculations employ the same model as a reference, MOCUP, and CASMO, however, there are some variances in methodology and cross-section libraries. On a PWR pin cell model, eigenvalue and isotope concentrations were examined up to high burnup. The eigenvalue comparison as a function of burnup is good, with a maximum difference of less than 5% and an average absolute difference of less than 1%. The isotope concentration comparisons outperform a set of ThO2-UO2 fuel benchmarks and are comparable to a set of uranium fuel benchmarks previously published in the literature. As a function, the eigenvalue comparison The actinide and fission product data sources for a typical thorium fuel are reported in the WIMSD-5B burnup calculations. The reasons for discrepancies in coding are examined and explored.Keywords: Thorium, PWR Fuel, Burn up, Pin Cell, WIMSD-5B

Thermo-Hydraulic Assessment of a Pwr Fuel Assembly and its Single Rod: Investigation of the Onset Nucleate Boiling Possibility by Cfd and Analytical Methods

Thermo-Hydraulic Assessment of a Pwr Fuel Assembly and its Single Rod: Investigation of the Onset Nucleate Boiling Possibility by Cfd and Analytical Methods

Drag Coefficient Estimation in Fsi for Pwr Fuel Assembly Bowing

Drag Coefficient Estimation in Fsi for Pwr Fuel Assembly Bowing

Defective PWR Fuel Rods Detection and Characterization Using an Artificial Neural Network

Defective PWR Fuel Rods Detection and Characterization Using an Artificial Neural Network

Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C

Based on the conventional “two-step” scheme for the PWR fuel management, an advanced PWR fuel management software Bamboo-C has been developed by the most advanced methodologies in the reactor-physics field. Bamboo-C consists of three main functional codes: LOCUST code for the heterogeneous modeling and simulation and homogenization calculation of 2D assemblies; SPARK code for 3D core steady-state and transient analysis; and LtoS code for assembly homogenization parameter function, which links LOCUST and SPARK. Bamboo-C has all the necessary analysis functions for the fuel management and nuclear design of PWRs, mainly including the start-up physics tests, calculations of the neutron-kinetics parameters, differential and integral worth of rod cluster control assemblies (RCCAs), and power-operation following simulation. Finally, the engineering validations of Bamboo-C have been completed according to the operation data from the reactors CNP300, CNP650 and CNP1000 designed by China independently. The validation results show that the errors between the values of such key parameters of cores as critical boron concentration, temperature coefficient, RCCA worth, and power distributions, calculated by Bamboo-C, and their measured values satisfy the corresponding engineering criterion limits.

Investigation of the CRUD Layer in Failed PWR Fuel Using an EPMA

This study characterizes a failed discharged fuel rod with 53 000 MWd/tonne U from a nuclear power plant in Korea. Chalk River Unidentified Deposits (CRUD) and the oxide layer were observed using an electron probe micro-analyzer (EPMA, SX-50 R, CAMECA, France) with wavelength dispersive (X-ray) spectroscopy. A normally irradiated cladding specimen was analyzed for comparison with the failed fuel rod. The analysis revealed an oxide layer with a thickness of about 10 μm and double-stratified agglomerates of CRUD species shapes. In contrast, sound fuel rods irradiated under conditions similar to failed fuel showed clusters in which Fe, Ni, and Cr were distributed. The main elements constituting the CRUD material, notably Ni and Fe, were located in the same position. Moreover, the thickness of the oxidized layer of the failed fuel rod was found to be significantly different from the thickness of the sound fuel rod. Consequently, EPMA techniques offer the possibility of identifying and analyzing the CRUD phases and segregations in spent pressurized water reactor fuel. Although phases and segregations are small in terms of the amount expected to be present in background radiation, they nevertheless present a significant analytical challenge.

Transition cycle analysis of light water-cooled SMR core loaded with MOX (TRU) and FCM (TRU) fueled PWR assemblies

In this study, a light water-cooled small modular reactor (SMR) core loaded with special fuel assemblies composed of mixed oxide (MOX) fuel rods using UO2-TRUO2 and fully ceramic micro-encapsulated (FCM) fuel rods using TRUO2 was suggested and analyzed for transuranic (TRU) transmutation as an alternative option before the commercialization of sodium-cooled fast reactor (SFR). Especially, the FCM rods were loaded with only TRUO2 for deep burning of TRU. This study performed neutronic analysis of the transition cycle cores with the new fuel assemblies. In addition to the neutronic analysis, detailed analyses on two important issues for the light water-cooled reactor (LWR) core loaded with TRU were performed in fuel assembly level calculation. First, an effective burnable absorber design was searched to mitigate the degradation in effectiveness of the burnable absorber caused by the hardened neutron spectrum. Secondly, a reactivity decomposition analysis was performed for understanding quantitative nuclide-wise and reaction-wise contributions to the void reactivity which is a key safety parameter of TRU-bearing fuel assemblies. The analysis on the TRU mass flow through the transition cycles showed that a high net TRU consumption rate of 14.7% can be achieved at the equilibrium cycle.

Investigation of Pellet-Clad Mechanical Interaction in Failed Spent PWR Fuel

Investigation of Pellet-Clad Mechanical Interaction in Failed Spent PWR Fuel

ANALYSIS OF CS-137 TO CS-134 ACTIVITY RATIO FOR FAILED FUEL EXPOSURE ESTIMATION

The Cs-134 to Cs-137 activity ratio of the Cs-134 and Cs-137 fission products released from failed fuel rods into primary coolant is very useful to identify the exposure along with the fuel batch of the failed fuel. The calculated and measured Cs-137 to Cs-134 radioactivity ratios of failed BWR and PWR fuels are compared and analyzed for better understanding of their relationship. Moreover, the impacts of power uprate and fuel reload outage on calculated Cs-137 to Cs-134 activity ratios are studied and the physics behind the impacts are provided.

Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles.One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties.From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively.An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

New calculation method for PWR control rod assemblies with APOLLO3®

In this paper, we present recent advances on PWR core calculations schemes with the most advanced features of the new deterministic neutronic transport code APOLLO3®. We focus mostly on reactivity effects of control rod sub-assemblies representation. Two kinds of representation are being studied: a representation in which the control rod sub-assembly is surrounded by standard sub-assemblies, calculated using 2D TDT-MOC solver associated with the fine structure self-shielding method in 281 groups at the Lattice calculation level; and a semi-heterogeneous modeling (3×3 zones) of the control rod sub-assembly at the Core calculation level. Taking advantage of the possibility of subdividing a sub-assembly with the MINARET core solver which uses an unstructured conforming triangular spatial mesh (Discontinuous Galerkin Finite Elements), the core calculation represents much better the control rod shadowing effects within the control rod sub-assembly. Tests to demonstrate the ability of such a calculation scheme have been carried out on an UOX fueled PWR reactor. The accuracy of the new APOLLO3® scheme has been tested against TRIPOLI-4® and has been found extremely accurate without requiring any equivalency method. This calculation scheme lays down the foundations for a new upgrading approach for deterministic calculations to study PWR cores.

Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

Optimum Erbium Isotopes Composition and Distribution for Power Flattening in Advanced PWR Fuel Assembly

Optimum Erbium Isotopes Composition and Distribution for Power Flattening in Advanced PWR Fuel Assembly

PWR Fuel Element Neutronic Analysis with Burnable Poison Rods Using Zircaloy and Hi-Nicalon Type S Claddings

ABSTRACTThe alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of rising the burning and maintaining the safety of nuclear plants, is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and a low absorption cross-section for thermal neutrons. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with burnable poison variations, the impact on multiplication factor and reactivity coefficients to both claddings: zircaloy and silicon carbide. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This code system is widely accepted and used worldwide for safety analysis, and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR.

Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative) inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for UO2 , whereas the latter is reduced to 9% when more accurate irradiation data are used (core-follow flux data instead of life-average flux data).

Study on the relation between Doppler reactivity coefficient and resonance integrals of Thorium and Uranium in PWR fuels

Doppler reactivity coefficient of Thorium fuel is negatively larger than that of Uranium fuel, although resonance integral of Th-232 is smaller than that of U-238.The purpose of this study is to reveal the mechanism of the opposite tendency of Doppler reactivity coefficient and resonance integral between Thorium fuel and Uranium fuel.Larger Doppler reactivity coefficient is caused by larger relative change of capture cross-section of Th-232, and the larger change results from narrower resonance width. In addition, narrower resonance width causes smaller resonance integral.It was found that resonance width is the key parameter to understand the opposite tendency of Doppler reactivity coefficient and resonance integral between Thorium fuels and Uranium fuels.