UO2 Core Research Articles

Fuel loading pattern with 50% MOX fuel in HPR1000 core

Fuel loading pattern optimization with MOX (mixed oxide of plutonium and uranium) fuel is an important aspect for GIII LWR. For the HPR1000 core, an optimized MOX fuel is designed. Based on which, two loading patterns with 50% MOX fuel are designed. One is 1/4 core annually reloading pattern, and the other is 18-month reloading pattern. The core neutronic parameters are investigated by comparing with the 100% UO2 core. The MOX fuel has impacts on the nuclear characteristic parameters, but the 50% MOX core still satisfies the nuclear and safety criteria samilar to that considered in 100% UO2 core. The results of this study prove that the GIII LWR HPR1000 has excellent capability while loading with 50% MOX fuel.

Feasibility to convert an advanced PWR from UO2 to a mixed U/ThO2 core – Part I: Parametric studies

This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235U, as LEU (20w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO2-68%ThO2); the second with (24% UO2-76% ThO2), and the third with (20% UO2-80% ThO2), using 235U LEU (20w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45w/o; 3.40w/o; 2.35w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO2 core, such a lower power density, and keeping the same 18months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

Thorium-based mixed oxide fuel in a pressurized water reactor: A beginning of life feasibility analysis with MCNP

Thorium is an asset the nuclear industry does not use, and plutonium is a liability that much of the world would like to be rid of. By incorporating a thorium–plutonium mixed oxide fuel (Th-MOX) into the fuel cycle, pressurized water reactors could provide a means for the United States to address both of these issues – but only if key reactor safety parameters are not affected.The feasibility of utilizing Th-MOX fuel in a pressurized water reactor is examined under steady-state, beginning of life conditions. With a three-dimensional MCNP model of a Westinghouse-type 17×17 PWR, many possibilities for replacing one-third of the UO2 assemblies with Th-MOX assemblies were considered. The excess reactivity, critical boron concentration, and centerline axial and radial flux profiles for several configurations and compositions of a one-third Th-MOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5wt% PuO2 was determined to be the best candidate for further analysis. Therefore, this configuration was compared to a 100% UO2 core using the following parameters: delayed neutron fraction (βeff), temperature coefficient, shutdown margin (SDM), and axial and radial nuclear hot channel factors (FZN and FRN).The one-third Th-MOX configuration showed an undesirable reduction in βeff from 0.00716±4.60E−07 for the 100% UO2 configuration to 0.00607±4.30E−07. The reduction in βeff would perhaps be ameliorated by the one-third Th-MOX configuration’s temperature coefficient of reactivity, which at −2.05±0.02pcm°F−1 is more favorable than the corresponding value of −1.42±0.02pcm°F−1 for the 100% UO2 configuration. The SDM of the one-third Th-MOX configuration is estimated to be 4079±7pcm, which is 28% lower than value of the 100% UO2 configuration. The FZN for the two cores were virtually identical. However, FRN for the one-third Th-MOX configuration (1.67±0.28) was 20% higher than the corresponding value for the 100% UO2 configuration (1.39±0.23).These preliminary results are encouraging. However, additional investigations are required to study the impact of thermal fluid feedback and the effect of burnup and poison buildup.

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (keff's) to neutron cross sections used for the reactivity analysis of UO2 and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in keff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO2 core is mainly attributed to the decreases in the capture cross sections of 238U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in keff are decreases in the capture cross sections of 238U,238Pu, 239Pu, and those to decrease in keff are decreases in ν of 239Pu and increases in the capture cross sections of241Am. They compensate each other, and the difference in keff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.

Isolation of the Large {Actinide}38 Poly-oxo Cluster with Uranium

By controlling the water content, a new poly-oxo-metalate species containing 38 uranium centers has been solvothermally synthesized in the presence of benzoic acid in tetrahydrofuran (THF). The {U38} motif contains a distorted UO2 core of fluorite type, stabilized by benzoate and THF molecules. This compound is analogous to the {Pu38} motif and was characterized by X-ray photoelectron spectroscopy and magnetic analyses.

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Calculation of LWR βeff kinetic parameters: Validation on the MISTRAL experimental program

This paper presents the analysis of the MISTRAL experimental program in EOLE on the determination of the effective delayed neutron fraction βeff for UO2 and MOX Light Water Reactor cores using the APOLLO2.8 code and JEFF-3.1.1 nuclear data library. The objective was to verify if the new 8 time-groups data in JEFF3 library (instead of the classical 6 groups) allow to reduce the calculation – experiment discrepancy observed using ENDF/B-VII.0 or the previous JEF-2.2 library. The present analysis has shown that the C/E bias is reduced from +2.8% with JEF2.2 to +0.8±1.6% for the UO2 cores and from +0.8% (with JEF2.2) to +0.2±1.6% for the MOX cores, using Discrete Ordinates 2D core calculations. The introduction of Method of Characteristics (MOC) in the flux calculation, through so-called Reference “SHEM-MOC” (281-groups – P3 for scattering) and optimized ”LWR2005” (26-groups – P3 and corrected P0 for scattering) calculation schemes improve slightly the results, with C/E discrepancies of +0.2±1.5% and +0.0±1.5% respectively for UO2, and +0.1±1.6% and −0.1±1.6% respectively for MOX.

Fuel residence time in BWRs with nitride fuels

This paper presents a neutronics study of a BWR core with uranium nitride fuels. Replacing the standard UO2 fuel with UN or UN-ZrO2 allows for a higher uranium content, which leads to an increase of the in-core fuel residence time. With the nitride fuels, the total void worth increases and the efficiency of the control rods and burnable poison deteriorates. Taking into account the higher amount of burnable poison needed at the beginning of life, the in-core fuel residence time increases by about 1.4year comparing to UO2 fuel with the same enrichment. This implies 1.4% higher availability of the plant and it is therefore of economic interest to the nuclear power plant operators. A similar increase of the fuel in-core lifetime in a UO2 core could be reached by an increase of the average enrichment of the oxide fuel by roughly 1%.

A study on fabrication technique of ZrB2 target

ZrB2 thin film is used as neutron burnable absorbers of nuclear reactor. This paper prepared ZrB2 target by cold isostatic press followed high temperature sintering in order to meet the requirement of preparing thin film by magnetron sputtering. The phase constituent and composition of ZrB2 target was analyzed by X-ray diffraction (XRD) and X-ray energy dispersion spectroscope(EDS). The microstructure of samples was observed by scanning electron microscope(SEM). The density was measured based on Archimedes drainage method; UO2 core was coated by ZrB2 films by magnetron sputtering using as-prepared ZrB2 target. The results indicate that 2000°C could sinter ZrB2 powder. The relative density of sintering product was 71%. The as-prepared ZrB2 might make films on the sample of UO2 surface by magnetron sputtering.

Uranium Oxo and Superoxo Cations Revealed Using Infrared Spectroscopy in the Gas Phase

The UO4+ and UO6+ cations are produced in a supersonic molecular beam by laser vaporization and studied with infrared laser photodissociation spectroscopy using rare gas atom predissociation. The argon complexes UO4+Ar2 and UO6+Ar2 are mass-selected in a reflectron time-of-flight spectrometer and excited with an IR-OPO laser system in the range of the O–U–O and O–O stretching vibrations. These same systems are studied with computational quantum chemistry. UO4+ is found to have a central UO2 core, with an additional η2 coordinated oxygen molecule. Charge transfer/oxidation gives the system the character of a UO22+, O2– ion pair. UO6+ has this same core structure, with an additional weakly bound oxygen molecule in an η1 coordination configuration. The O–U–O stretch is sensitive to the local environment and approximates the vibration of the isolated uranyl cation in these systems.

Study of the PWR REA pulse width for realistic UO2 and MOX core designs using 3D kinetics methods

For the development of new acceptance criteria for the analysis of Rod-Ejection-Accidents (REA) in Pressurized Water Reactor UO2 cores, full-width-at-half-maximum power pulse widths in the range 25–40 ms were employed in the analytical transient fuel behaviour studies. Due to the operation of MOX cores in Switzerland, a study using the CORETRAN 3D kinetics code was carried out at PSI to investigate and compare on the basis of real operated cycles MOX and UO2 cores, the pulse width magnitude as function of rod reactivity for postulated super-prompt critical REAs. This study is presented in a first part of this paper. Thereafter, to quantify the variation in pulse width when applying a different 3D kinetic solver, a benchmarking with the SIMULATE-3K code is carried out. This code-to-code comparison aims also to serve as an assessment of SIMULATE-3K for REA analyses of the Swiss cores since only CORETRAN had so far been benchmarked at PSI for this type of applications. Finally, to provide confidence in the results, particularly for pulse widths below 25 ms, sensitivity studies are performed to assess the effects of specific modelling options and assumptions applied in the REA 3D kinetic analysis and to compare the sensitivities between UO2 and MOX cores. On the basis of all these studies, it is found that the MOX core pulse width is usually around 5 to 10 ms lower than for UO2 cores. For rod worths close to 2.0$, corresponding to around 500 pcm above prompt criticality and which was obtained for many of the investigated MOX cycles particularly at end-of-cycle conditions, a representative MOX pulse width magnitude is found to be around 15 ms. To shortly address the applicability of these results, i.e. obtained for very specific MOX cores, additional calculations were performed for hypothetical cases with increased MOX core fractions and with higher MOX assembly Pu enrichments. And it was found that even for these types of cores, the pulse width would, although being slightly reduced, remain in the range 10–15 ms for rod reactivities up to 500 pcm above prompt criticality.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

Analysis of the SPERT-III E-Core Using ANCK Code with the Chord Weighting Method

ANCK/MIDAC is a three-dimensional (3D) code system that has been developed, which couples the 3D nodal kinetic code ANCK and the 3D drift flux thermal-and-hydraulics (T/H) code MIDAC. The adequacy of the ANCK feedback model was confirmed by analysis of the SPERT-III E-core reactivity accident tests. The chord weighting method was adopted in the effective fuel temperature (Teff) calculation model of ANCK and ANCK/MIDAC since it provides results that are in good agreement with those of a Monte Carlo calculation. The adequacy of the Teff model was also confirmed. The calculation results for hot startup and hot standby conditions using the ANCK code showed excellent agreement for the main parameters such as peak power, peak time, and behavior of reactor power. It was also confirmed that the ANCK feedback model with the chord weighting method was applicable to the reactivity-initiated accident (RIA) calculation for the case of 4.8 wt% UO2 core.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

BWR核設計コードによるMOX臨界試験EPICUREおよびMISTRALの解析

For the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores.

Alternative Applications of Homogeneous Thoria-Urania Fuel in Light Water Reactors to Enhance the Economics of the Thorium Fuel Cycle

Since the thorium-based fuel has many incentives including the reduction of plutonium generation and long-lived radiotoxic isotope production, the research on the use of thorium as a nuclear fuel for nuclear power reactors has been performed and will last for a long time. Focus is on the fuel economics of the thorium-based cycle for light water reactors (LWRs). Analyses show that the neutronic behavior of a mixed thorium and uranium dioxide (Th + U)O2 core in a pressurized water reactor (PWR) will not be significantly different from that of a UO2 core. This implies that homogeneous (Th + U)O2 fuel can be used in PWRs instead of the current UO2 fuel without any significant mechanical modification of the fuel design and without any change in the nuclear design limits. However, homogeneous (Th + U)O2 has not shown any economic advantage over UO2 fuel when current fuel management strategies are used. Thus, alternative applications of homogeneous (Th + U)O2 fuel in LWRs have been investigated to enhance the economics of the thorium fuel cycle. Specifically, thorium-uranium fuel with a 235U enrichment significantly <19.5 wt%, mixed cores of both duplex (Th + U)O2 and UO2 fuel assemblies, and use of homogeneous thorium-uranium fuel in small-to-medium PWRs with a 5-yr cycle length have been investigated. The proposed alternatives result in far better fuel economics than the homogeneous thorium-uranium fuel cycle. However, the proposed alternatives do not show the economic merit of thorium-based fuel options for existing LWRs as compared to the UO2 fuel option. However, the inclusion of spent-fuel disposal costs in the fuel cost estimate makes (Th + U)O2 fuel competitive with UO2 fuel. In the case of a spent-fuel disposal cost higher than 700 US$/kg HM, the long-lived core with better economic potential than the UO2-fueled core may be realized with the homogeneous (Th + U)O2 fuel.

Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs

An assessment is made of the potential for Th-based fuel to minimize Pu and minor actinide (MA) production in pressurized water reactors (PWRs). Destruction rates and residual amounts of Pu and MA in the fuel used for transmutation are examined. In particular, sensitivity of these two parameters to the fuel lattice hydrogen to heavy metal (H/HM) ratio and to the fuel composition was systematically investigated. All burnup calculations were performed using CASMO4, the fuel assembly burnup code. The results indicate that up to 1000 kg of reactor-grade Pu can be burned in Th-based fuel assemblies per gigawatt (electric) year. Up to 75% of initial Pu can be destroyed per passage through reactor core. Addition of MA to the fuel mixture degrades the burning efficiency. The theoretically achievable limit for total transuranium (TRU) destruction per passage through the core is 50%. Efficient MA and Pu destruction in Th-based fuel generally requires a higher degree of neutron moderation and, therefore, higher fuel lattice H/HM ratio than typically used in the current generation of PWRs. Reactivity coefficients evaluation demonstrated the feasibility of designing a Th-Pu-MA fueled core with negative Doppler and moderator temperature coefficients. Introduction of TRU-containing fuels to a PWR core inevitably leads to lower control material worths and smaller delayed-neutron yields than with conventional UO2 cores. Therefore, a major challenge associated with the introduction of Th-TRU fuels to PWRs will be the design of the whole core and reactor control features to ensure safe reactor operation.

Benchmarking MCNP and WIMS/RFSP against Measurement Data—I: Deuterium Critical Assembly

Benchmark calculations have been performed for the conventional Canadian deuterium uranium (CANDU) core analysis code RFSP and the Monte Carlo code MCNP-4B using experimental data from the deuterium critical assembly. The benchmark calculation was carried out for the effective multiplication factor (keff), void reactivity, local power peaking factor (LPPF), and power distribution of a uniform core with 1.2 wt% UO2 and two-region cores with PuO2-UO2 fuels. The RFSP calculation was performed with two energy groups, using lattice parameters generated by WIMS-AECL with the ENDF/B-V cross-section library. The RFSP calculation has shown that the root-mean-square (rms) errors of the keff and the void reactivity are within 0.6% δk and 0.3% δ(1/k), respectively. The MCNP simulation was performed using a fully heterogeneous core model that explicitly describes the individual fuel rod and channel. The simulation showed an excellent agreement for the keff against the measurement, while the rms error of the void reactivity was 0.4% δ(1/k). The LPPF and core power distribution estimated by both codes matched those of the measurements within 4 and 9%, respectively. Conclusively, the physics analysis by the RFSP code in conjunction with the WIMS-AECL produces credible results for the light water–cooled and heavy water–moderated system. In addition, the MCNP-4B code has proved its potential as a computational benchmarking tool for the heavy water–moderated system.

380. MOX燃料炉物理試験MISTRALの解析

Nuclear Power Engineering Corporation (NUPEC) has been performing conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program, MISTRAL, that was devoted to measuring the core physics parameters of such advanced cores. The program consists of one reference UO2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice.NUPEC has been analyzing the experimental results with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. This indicates that these applied analysis methods give the same accuracy for the UO2 core and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores.

Rock-Like Oxide Fuels and Their Burning in LWRs

Research on the plutonium rock-like oxide (ROX) fuels and their once-through burning in light water reactors has been performed to establish an option for utilizing and disposing effectively the excess plutonium. The ROX fuel is a sort of the inert matrix fuels and consists of mineral-like compounds such as yttria stabilized zirconia, spinel and corundum. A particle-dispersed fuel was devised to reduce damage by heavy fission fragments. Some preliminary results on swelling, fractional gas release and microstructure change for five ROX fuels were obtained from the irradiation test and successive post-irradiation examinations. Inherent disadvantages of the Pu-ROX fuel cores could be improved by adding 238U or 232Th as resonant materials, and all improved cores showed a nearly the same characteristics as the conventional UO2 core during transient conditions. The threshold enthalpy of the ROX fuel rod failure was found to be comparable to the fresh UO2 rod by pulse-irradiation tests simulating reactivity initiated accident conditions.