Burnup Reactivity Research Articles

A Neutronics and Burnup Analysis of the Accelerator-Driven Transmutation System with Different Cross Section Libraries

The difference of the neutronics and burnup characteristics of the Accelerator-Driven System (ADS) caused by the different nuclear data libraries, JENDL-3.2, ENDF/B-VI and JEF-2.2, was analyzed. It is important to know the current accuracy of the cross section data of the Minor Actinides (Np, Am, Cm) to improve the reliability of the analysis and to guarantee the subcriticality of the fuel region of ADS. The analysis was performed with benchmark problems of the OECD/NEA ADS transmutor. An ATRAS code system and a MVP code were used for the analysis. Group cross section sets were newly created from JENDL-3.2, ENDF/B-VI and JEF-2.2, respectively. To make the difference of the MA cross section data clearly, the group cross sections generated from JENDL-3.2 were used for cooling and structural nuclides of ENDF/B-VI and JEF-2.2. From the results of the analysis, JENDL-3.2 and JEF-2.2 gives close k-effective values and similar trends of burnup reactivity swing. However, ENDF/B-VI gives different k-effective values and burnup changes from other two libraries

Lumped Group Constants of FP Nuclides for Fast Reactor Shielding Calculation Based on JENDL-3.2

Fission Products were not considered in conventional fast shielding analyses that were predominantly developed in clean core experiments. However, in power reactors with high burn-up, the accumulation of FP nuclides affects the neutron balance mainly due to the absorption reaction and it cannot be neglected when calculating the neutron flux of a high burn-up reactor. In this study, the lumped group constants of FP nuclides used for a fast reactor shielding calculation were computed with the JENDL-3.2 library and compiled to the JSDJ2/JFTJ2 set. The effect of considering FP nuclides on the neutron flux calculations was evaluated in the JOYO experimental fast reactor. These tests showed conventional calculations that ignored FP nuclides overestimated neutron flux by about 2%. The effect on reaction rate calculation of the spent fuel in IVS (in-vessel storage rack) is a maximum of 10%.

A comparative study to investigate burnup in research reactor fuel using two independent experimental methods

A comparative study to investigate burnup in research reactor fuel using two independent experimental methods

Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead–bismuth, Pb–Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.

A comparative study to investigate burnup in research reactor fuel using two independent experimental methods

Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the first method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity difference technique was used. At the same location in the same core configuration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group cross-sections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of different fuel elements using these methods were found to be in good agreement.

Blanket Design Studies of a Lead-Bismuth Eutectic-Cooled Accelerator Transmutation of Waste System

The results of blanket design studies for a lead-bismuth eutectic (LBE)-cooled accelerator transmutation of waste system are presented. These studies focused primarily on achieving two important and somewhat contradictory performance objectives: First, maximizing discharge burnup, so as to minimize the number of successive recycle stages and associated recycle losses, and second, minimizing burnup reactivity loss over an operating cycle, to minimize reduction of source multiplication with burnup. The blanket is assumed to be fueled with a nonuranium metallic dispersion fuel; pyrochemical techniques are used for recycle of residual transuranic (TRU) actinides in this fuel after irradiation. The key system objective of high-discharge burnup is shown to be achievable in a configuration with comparatively high power density and relatively low burnup reactivity loss. System design and operating characteristics that satisfy these goals while meeting key thermal-hydraulic and materials-related design constraints have been preliminarily developed. Results of the performance evaluations indicate that an average discharge burnup of ~27% is achieved with a ~3.5-yr fuel residence time. Reactivity loss over the half-year cycle is 5.3%Δk. The peak fast fluence value at discharge, the TRU fraction in the charged fuel, and the peak coolant velocity are well within the assumed design limits. Owing to its use of nonuranium fuel, this proposed LBE-cooled system can consume light water reactor-discharge TRUs at the maximum rate achievable per unit of fission energy produced (~1.0 g/MWd).

Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B 4C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.

Comparison of the Burnup Characteristics and Radiotoxicity Hazards of Rock-like Oxide Fuel with Different Types of Additives

Burnup characteristics of rock-like oxide (PuO2-ZrO2: ROX) fueled LWR have been studied with an additive ThO2, UO2 or Er2O3. A ROX fueled LWR core has reactivity coefficients and burnup reactivity swing problems, which can be mitigated by these additives. From the burnup calculations of the ROX fueled LWR, it was found that the additive ThO2 causes a few percent decrease in Pu transmutation in ROX fuel, while the additive UO2 or Er2O3 decreases it largely. The production of 241 Am, which is a parent of long-lived 237Np, is smaller in ThO2 added fuel than that in UO2 or Er2O3 added fuel. However, the productions of 243Am and 244Cm are more in ThO2 added fuel than those in UO2 or Er2O3 added fuel. The long-lived fission products (LLFPs) production in ThO2 added fuel is nearly the same as that in UO2 added fuel, and smaller than that in Er2 O3 added fuel. As a result of these Pu transmutation and minor actinides (MAs) and LLFPs productions, ingestion radiotoxicity hazard index of ThO2 added ROX spent fuel is lower than that of UO2 or Er2O3 added ROX spent fuel, except at the 105 years cooling. From these calculations, ThO2 is recommended to be used as the additive in ROX fuel.

Benchmark Calculations for Standard and DUPIC CANDU Fuel Lattices Compared with the MCNP-4B Code

Cell-code benchmark calculations have been performed for the standard CANDU and DUPIC CANDU fuel lattices compared with the MCNP-4B code. To consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The lattice codes WIMS-AECL and HELIOS were then benchmarked by the MCNP code for the major physics parameters such as burnup reactivity, coolant void reactivity, fuel temperature coefficient, etc. The calculations have shown that the physics parameters estimated by the lattice codes are consistent with those by MCNP. However, there is a tendency that the error increases slightly when the fuel burnup is high. This study has shown that the WIMS-AECL produces reliable results for CANDU fuel analysis. However, it is recommended that the cross-section library be updated to be used for the high-burnup fuels even though the current results are generally acceptable. This study has also shown that the HELIOS code has the potential to be used for CANDU fuel lattice analysis in the future.

Parametric Studies on Plutonium Transmutation Using Uranium-Free Fuels in Light Water Reactors

To compare the once-through use of U-free fuels for plutonium burnup in light water reactors (LWRs), plutonium transmutation, minor actinide (MA) and long-life fission product (LLFP) buildup and radiotoxicity hazards were compared for PuO2 + ZrO2 (rock-like oxide: ROX) and PuO2 + ThO2 (thorium oxide: TOX) fuels, loaded in a soft-to-hard neutron spectrum LWR core (a moderator-to-fuel volume ratio Vm/Vf is from 0.5 to 3.0). For better understanding and proper improvement of the reactivity coefficient problem of ROX, the fuel temperature coefficient, the void coefficient, and the delayed neutron fraction were also studied. A mixed-oxide (MOX)-fueled LWR was considered for reference purposes.From the result of the cell burnup calculation, ROX fuel transmutes 90% of net initially loaded weapons-grade Pu, and 2.5% of initially loaded Pu is converted to MAs when Vm/Vf is 2.0 and discharge burnup in effective full-power days is equivalent to that of 33 GWd/t in MOX fuel. Reactor-grade Pu-based ROX fuel transmutes 80% of net initially loaded Pu, and 6.7% of initially loaded Pu converts to MAs with the same condition as the weapons-grade Pu ROX fuel. TOX fuel also has a good Pu transmutation capability, but the 233U production amount is approximately a half of the fissile Pu transmutation amount. The MA production amount in TOX fuel is lower than that in MOX and ROX fuels. The LLFP production amount in ROX fuel is lower than that in MOX and TOX fuels. The radiotoxicity hazard of ROX spent fuel is lower compared to that in TOX and MOX spent fuels.The thermal neutron energy region is important in ROX fuel for fuel temperature coefficient and void coefficient problems. From these calculations, 15 to 20% 232Th-added ROX fuel seems the best to use as a once-through Pu-burning fuel compared to TOX and MOX fuels in conventional LWRs, because of its higher Pu transmutation, lower radiotoxicity hazard.

Gamma-ray spectroscopy on irradiated MTR fuel elements

The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

A Design Study on the FBR Metal Fuel and Core for Commercial Applications

A design methodology for the FBR metal fuel and core is developed. It consists of the fuel integrity prediction by a mechanistic analysis code, and the core neutronic/thermal-hydraulic design in which the effect of the characteristic fuel behavior and fuel cycle technologies are properly considered. Based upon this methodology, a fuel and core design study for reactors of various output scale (150–1500 MWe) is conducted, and the feasibility of metal fuel FBRs for the future commercial applications is reviewed. The results show that the averaged burnup of as high as 90–150 MWd/t is achievable at the refueling interval of 1–2 years with 3–4 batches, regardless of the output scale. The maximum allowable cladding temperature is assumed to be 650°C to avoid the liquid phase formation, which leads to the achievable core outlet temperature of ∼510°C based on the current hotspot factors estimation. It is found that high breeding ratio of ≥1.2 can be enabled with relatively small blanket amount, owing to the very high internal conversion capability. Another advantage is that it is possible to significantly reduce the burnup reactivity loss at the slight expense of the burnup, in which case the core excess reactivity becomes as small as a few dollars.

A pan-shape transuranic burner core with a low sodium void worth

A 1575 MWt transuranic (TRU) burner reactor core with a low sodium void worth has been developed by devising a pan-shaped active core design. The core consists of two types of fule subassemblies that differ in the height of the fueled regions. This strategy has allowed an extreme "pancaking" of the inner core region while the radial dimension increase is limited by placing longer fuels in the outer core region. The fuel cycle analysis has been performed in the equilibrium cycle consisting of external feed fuel with reprocessed typical PWR spent fuel and fissile makeup with recycled TRU elements. The neutronic performance characteristics obtained from the equilibrium cycle analysis show that it would work safely as well as economically as measured in terms of burnup reactivity swing peak power density Doppler coefficient TRU burning and sodium void worth. The core has relatively low double power peaks in both the inner and outer cores without enrichment zoning and this enables to make an active core volume smaller. The developed TRU burner core has been subjected to an extensive parametric study on the reprocessing schemes. Investigations are given for sodium void worth transmutation capability burnup reactivity swing and minor actinide (MA) contents. Through this series of study a variant of the TRU burner core that is aimed at preferentially burning MA has been determined. This MA burner core uses U-235 as well as the homogeneously recycled TRU elements. This MA burner core consumes 140 kg of MA per year without penalizing the sodium void reactivity observed in the TRU burner core. Finally the combined introduction of developed TRU and MA burner cores shows the functional effectiveness of reducing PWR discharged TRU inventory.

Application of modified explicit higher-order perturbation (MHP) method to PWR core analysis

A PWR core analysis was performed by using the Modified Explicit Higher-Order Perturbation (MHP) method. A new perturbation calculation code with two energy groups in a two-dimensional x- y geometry based on the MHP method TWOPEC was developed to evaluate the change both in the neutron flux distribution and the reactivity caused by a perturbation. The results were compared with those by the direct eigenvalue calculations to examine the capability of the MHP method. It was demonstrated that both the neutron flux distribution and the reactivity can be accurately evaluated by the MHP method for a string and local perturbation caused by the control rod insertion when one uses large numbers of expansion modes and perturbation orders. The MHP method was applied to the pseudo burnup calculations that regarded the change in cross sections caused by the fuel burnup as the perturbation. It was found that one can also obtain satisfactory results not only for the burnup reactivity but also for the control rod worth at each burnup step.

A Design Study on the FBR Metal Fuel and Core for Commercial Applications.

A design methodology for the FBR metal fuel and core is developed. It consists of the fuel integrity prediction by a mechanistic analysis code, and the core neutronic/thermal-hydraulic design in which the effect of the characteristic fuel behavior and fuel cycle technologies are properly considered. Based upon this methodology, a fuel and core design study for reactors of various output scale (150–1500 MWe) is conducted, and the feasibility of metal fuel FBRs for the future commercial applications is reviewed. The results show that the averaged burnup of as high as 90–150 MWd/t is achievable at the refueling interval of 1–2 years with 3–4 batches, regardless of the output scale. The maximum allowable cladding temperature is assumed to be 650°C to avoid the liquid phase formation, which leads to the achievable core outlet temperature of ∼510°C based on the current hotspot factors estimation. It is found that high breeding ratio of ≥1.2 can be enabled with relatively small blanket amount, owing to the very high internal conversion capability. Another advantage is that it is possible to significantly reduce the burnup reactivity loss at the slight expense of the burnup, in which case the core excess reactivity becomes as small as a few dollars.

Spent mixed oxide fuel rejuvenation in fusion breeders

A fusion breeder is presented for the rejuvenation of spent nuclear fuel. A (D, T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing ten rows in radial direction, covers the cylindrical fusion plasma chamber. The first three fuel rod rows contain Canadian deuterium uranium (CANDU) reactor spent nuclear fuel which was used down to a total enrichment grade of 0.418%. The following seven fuel rod rows contain light water reactor (LWR) spent nuclear fuel, which was used down to a total enrichment grade of 2.17%. This allows a certain degree of fission power flattening. Fissile zone is cooled with pressurised helium gas with volume ration of V coolant/ V fuel=2 in the fissile zone. Spent fuel rejuvenation occurs through the neutron capture reaction in 238U. The new fissile material increases the nuclear quality of the spent fuel which can be described as the cumulative fissile fuel enrichment (CFFE) grade of the nuclear fuel which is the sum of the isotopic ratios of all fissile material ( 235U+ 239Pu+ 241Pu) in the mixed oxide (MOX) fuel. Under a first-wall fusion neutron current load of 10 14 (14.1-MeV n/cm 2 s), corresponding to 2.25 MW/m 2 and by a plant factor of 100%, the CANDU spent fuel can achieve an enrichment degree of 1% after ∼7 months, suitable for reutilization in a CANDU reactor. LWR spent fuel requires >15 months to reach an enrichment grade ∼3.5%, suitable for reutilization in a LWR. A longer rejuvenation period (up to 48 months) increases the fissile fuel enrichment levels of the spent fuel reactor to much higher degrees (>3% for CANDU spent fuel and over 5% for LWR spent fuel), opening possibilities an increased burn-up in critical reactors and a re-utilization in multiple cycles.

A Liquid-Metal Reactor for Burning Minor Actinides of Spent Light Water Reactor Fuel—II: Nuclear Data Uncertainty Analysis

A comprehensive sensitivity and uncertainty analysis was performed on a 1200-MW(thermal) minor actinide burner designed for a low burnup reactivity swing, negative Doppler constant, and low sodium void worth. Sensitivities of the performance parameters were generated using depletion perturbation methods for the constrained closed fuel cycle of the reactor. The uncertainty analysis was performed using the sensitivity and covariance data taken from ENDF/B-V and other published sources. The uncertainty analysis of a liquid-metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. The relative uncertainties in the burnup swing, Doppler constant, and void worth were conservatively estimated to be 220, 120, and 59%, respectively. An analysis was performed to prioritize the minor actinide reactions for reducing the uncertainties.

A Liquid-Metal Reactor for Burning Minor Actinides of Spent Light Water Reactor Fuel—I: Neutronics Design Study

A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics.

Fission Product Model for BWR Analysis with Improved Accuracy in High Burnup

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation.

Experience with uranium-erbium fuel at the Ignalinsk Atimic Power Plant

Experience with uranium-erbium fuel assemblies at the Ignalinsk Atomic Power Plant confirms the possibility of removing additional absorbers and considerably increasing the depth of fuel burnup, while maintaining permissible values of the steam reactivity and other characteristics important for reactor safety. Conversion of the reactor to uranium-erbium fuel permits not only the maintenance of αϕ within specified limits but also compensation for its growth on account of replacement of the control rods. The effectiveness of uranium-erbium fuel up to degrees of burnup close to the design value (∼20 MW·day/kg) has been demonstrated. The good agreement of the theoretical predictions and the actual characteristics offers hope that the complete conversion of reactors to uranium-erbium fuel and appropriate removal of additional absorbers will yield high fuel burnup and considerable improvement in the economic performance of atomic power plants. Other means of increasing the fuel burnup in RBMK-1500 reactors might be to increase the fuel enrichment and erbium content, to use zirconium spacers, and to reduce the operational reserve of reactivity.