Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO₂ and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO₂ and MOX as a function of burn-up and irradiation temperature is proposed.     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.     

Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO₂ LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO₂ of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.     

Core concept of compound process fuel cycle

This paper presents fast reactor core concept and its feasibility as a part of newly proposed compound process fuel cycle in which spent fuels of light water reactor are multi-recycled without conventional reprocessing but with only pyrochemical processing, fuel re-fabrication and reloading to the fast reactor core. Results of the core survey analyses in order to find out the feasibility of this concept, taking example for BWR MOX spent fuel of 60 GWd/t burn-up, show that four times recycling of LWR spent fuel with the burn-up of more than 300 GWd/t can be achieved without increasing MA content. Such benefits will be expected in this concept as reduction of fuel cycle cost due to simplified reprocessing procedure, reduction of environmental impacts due to reduced high level waste, efficient utilization of nuclear fuel resources, enhancement of nuclear non-proliferation, and suppression of LWR spent fuel pile-up.     

A study of the generation of 232U in UO2 and MOX fuels

To clarify the generation pathway of ²³²U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of ²³²U generated through various pathways were evaluated for UO₂ and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in ²³²U concentrations in UO₂ and MOX fuels mainly arise from differences in the initial compositions of ²³⁴U, ²³⁵U, and ²³⁶U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of ²³⁰Th, ²³¹Pa, ²³⁵U, and ²³⁶U, as well as the (n,2n) cross sections of ²³⁷Np and ²³⁸U, have a large effect on the generation of ²³²U. Additional investigation showed that ²³²U concentration is strongly affected not only by time after irradiation but also by time before irradiation.     

Dissolution kinetics of Indian PHWR natural UO2 fuel pellets in nitric acid – Effect of initial acidity and temperature

Understanding the dissolution behaviour of plutonium rich MOX fuel is one of the major challenges in fast reactor spent fuel reprocessing. As an initial step, kinetics for the dissolution of Indian PHWR natural UO₂ sintered pellets in nitric acid was studied experimentally. In this paper we have reported the effect of initial nitric acid concentration and temperature on dissolution rate of sintered UO₂ pellets. The shrinking core model was used to correlate the experimental results. The apparent activation energy estimated from the temperature effect of the chemical reaction was found to be about 90.5 kJ/mol. The estimated apparent activation energy indicates that the dissolution rate of UO₂ pellets was controlled by chemical reaction under the experimental conditions.     

MgO-pyrochlore composite as an inert matrix fuel: Neutronic and thermal characteristics

Inert matrix fuels are an important component of advanced nuclear fuel cycles, as they provide a means of utilizing plutonium and reducing the inventory of ‘minor’ actinides. We examine the neutronic and thermal characteristics of MgO-pyrochlore (A₂B₂O₇: La₂Zr₂O₇, Nd₂Zr₂O₇ and Y₂Sn₂O₇) composites as inert matrix fuels in boiling water reactors. By incorporating plutonium with resonance nuclides, such as Am, Np and Er, in the A-site of pyrochlore, the k_infvs. burn-up curves are shown to be similar to those of UO₂, although the Doppler coefficients are less negative than UO₂. The Pu depletion rates are 88-90% (²³⁹Pu) and 54-58% (total Pu) when the inert matrix fuels experience a burn-up equivalent of 45 GWd/tHM UO₂. Because of the high thermal conductivity of MgO, the center-line temperatures of the MgO-pyrochlore composites at 44.0 kW/m are lower than those of UO₂ pellets. After burn-up, the A-site cation composition is 15-35 at.% lower than that of the B-site cations in pyrochlore (e.g., A_1.84B_2.17O_7.00) due to the fission of Pu in the A-site and the presence of fission product elements in the A- and B-sites of the pyrochlore structure.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU^®1 design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.     

Utilization of rock-like oxide fuel in the phase-out scenario

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by nuclear material balance (NMB) analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the NMB code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels, and an ability to estimate waste and repository. Three scenario groups of once-through Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. Using two full-MOX or full-ROX reactors the Pu amount is reduced to about one-half and the isotopic vector of Pu deteriorated for being used as a nuclear weapon, especially in terms of spontaneous fission neutron generation. Effects of ROX reactors are more significant than MOX reactors in terms of both reduction in the Pu amount and deterioration of the isotopic vector. Repository footprint and potential radiotoxicity are not reduced by the MOX and ROX reactors because the heat and toxicity of MOX and ROX spent fuels are considerably high.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

Modeling and criticality evaluation of the voloxidation process

Voloxidation is a necessary process in the dry reprocessing of spent nuclear fuels. The criticality evaluation plays a considerable role in the design of voloxidation apparatus. As conservative results are always preferred in a criticality evaluation, an optimized model was built in consideration of both the geometry of voloxidation apparatus and the occurring forms of evaluation material. The criticality evaluation of fresh UO₂ fuel and PWR spent fuel were then performed by employing Monte Carlo techniques, respectively. It is demonstrated that there is no criticality risk concerning the voloxidation process dealing with fresh UO₂ or PWR spent fuel if water does not intrude into the cell. However, if water intrudes and mixes with the fuels, the subcritical mass limit is 40.1 ± 0.1 kg for fresh UO₂ and 19,155 ± 50 kg for spent fuel. The contributions of ¹H and ²³⁵U were analyzed quantitatively by the TSUNAMI code to clarify the competition between ¹H moderation effect and its dilution effect on the concentration of ²³⁵U.     

A study on the burn-up characteristics of inert matrix fuels

Proper disposal of minor actinides (MA), long-lived fission products (LLFPs), and transuranium element (TRU) plays a key role in the sustainable development of fission nuclear power. Adoption of inert matrix fuels (IMFs) can effectively reduce the amount of ²³⁷Np and Np element in the spent fuel of present-day commercial power reactors. In order to study the burn-up characteristics of IMFs caused by the unique composition, burn-up calculations and MA accumulation of two typical IMFs, PuO₂ + ZrO₂ + MgO and PuO₂ + ThO₂, are performed in this paper. Results indicate that k_inf at beginning of life (BOL) and reactivity drop with burn-up for PuO₂ + ZrO₂ + MgO are much larger than those of PuO₂ + ThO₂ IMF. The yields of ²³⁷Np and Np element in IMFs are two orders smaller than those of UO₂ and mixed oxide (MOX) fuels. For the same PuO₂ volume fraction and a certain burn-up, the masses of ²³⁷Np, Np element, and ²⁴¹Am for PuO₂ + ZrO₂ + MgO are smaller than those of PuO₂ + ThO₂; however, the mass of total MA is larger. IMF has high destruction efficiencies of TRU and plutonium (Pu). The results and conclusion provide basic data for the ongoing IMF design and application study.     

VALMOX: validation of nuclear data for high burn-up MOX fuels

VALMOX, an acronym for validation of nuclear data for high burn-up MOX fuels, is one of the projects of the cluster evolutionary fuel concepts: high burn-up and MOX fuels (EVOL). It covers 30 months, from October 2001 to March 2004.It considers the evaluation of the actinide inventory of MOX fuel at high burn-up (typically 60 GWd/t) in light water reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file.So far, the major effort was placed on the evaluation of MOX fuel irradiations in pressurised water reactors, and first results will be presented and compared.     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.     

Releases of Cesium and Poorly Volatile Elements from UO2 and MOX Fuels under Severe Accident Conditions

Radionuclide release from fuel under severe accident conditions has been investigated in the VEGA program at the Japan Atomic Energy Agency. In this program, three types of fuel, two UO₂ fuels irradiated at PWR and BWR and a MOX fuel irradiated at the ATR Fugen, were heated up to about 3130K in helium atmosphere at 0.1 MPa. Comparison of experimental data and evaluation with computer code analyses showed that Cs release is essentially identical among the three fuels. The Cs release from fuel may differ below about 1770K due to a difference in migration to grain boundaries during irradiation. The difference was not also observed for releases of poorly volatile elements, namely, U, Pu, Sr and Mo between UO₂ and MOX fuels. The release rate of Pu became slightly higher than that of U at 3130 K. The release rate of Sr increased at 3130 K, while that of Mo was quite low at temperatures above 2310 K.