Void coefficient analysis of an LWR lattice experiment with high-enrichment MOX fuel

A series of physics experiments performed between 1985 and 1990 in the PROTEUS reactor at the Paul Scherrer Institute in Switzerland included the investigation of a Pu-fuelled light water reactor (LWR) lattice with a moderator-to-fuel volume ratio of 2.07 and an effective enrichment of about 8%. The analysis of the measurements in this test lattice and in the tighter light water high conversion reactor (LWHCR) lattices investigated previously permits the determination of the k_∞ void coefficient of the LWR lattice for cases of partial and total voiding. A comparison of the measured changes of k_∞ with values calculated using the cell codes WIMS and KAPER4 shows a satisfactory prediction of the partial void coefficient in the range from 0–54% voidage. Discrepancies increase up to twice the estimated experimental error in cases of further voiding to 100% void. The total void coefficient (0–100% void) results from large compensating effects from individual reaction rate ratios. Its accurate prediction by cell calculations appears to be fortuitous. Improved nuclear data and refined calculational methods are thus required for a more accurate calculation of the void coefficient in high-enrichment MOX-LWRs.     

Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.     

A methodology to predict a fission gas release ratio of MOX fuel with heterogeneous microstructure

The objective of this study is to formulate a methodology to predict a fission gas release ratio of MIMAS MOX. An irradiated MIMAS MOX fuel with plutonium rich agglomerates was subjected to elemental analyses by electron probe micro analysis and secondary ion mass spectrometry in order to investigate xenon distribution. The results of the elemental analyses showed that the plutonium rich agglomerates at the periphery of the fuel pellet sample retained a high concentration of xenon as gas bubbles. Then, the results were used as reference data for modification of models in a fuel rod analysis code, FEMAXI-7. Using the modified FEMAXI-7, we applied an approach to prediction of fission gas release ratio of MOX fuel with plutonium rich agglomerates. In the approach, two separated analyses using FEMAXI-7 were performed for the plutonium rich agglomerates and the matrix. Fission gas release ratios obtained from the two analyses were processed through weighted-average with burnup ratios of the plutonium rich agglomerates and the matrix. Finally, the fission gas release ratios were compared with results of rod puncture tests. As a result of the comparison, it was confirmed that the proposed approach could well predict fission gas release ratio of MOX fuel with plutonium rich agglomerates.     

Transport of MOX fuel: continuous challenge

Abstract

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. This paper focuses on the casks used to transport the fresh and used mix oxide (MOX) fuel. To transport the fresh MOX boiling water reactor and pressurised water reactors fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimise the loading in terms of integrated dose and also of course capacity. Mixed oxide used fuel has now its dedicated cask: the TN 112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the Electricité de France nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.     

我国压水堆装载MOX燃料对核燃料循环情景影响初步分析

为了满足持续增长的国家能源需求,核电将有更大规模的发展。本文对我国未来的核电发展和核燃料循环进行了情景研究,预测了2050年前核电对天然铀资源和燃料制造能力的需求情况,核电站产生的乏燃料量,分离钚产生量。乏燃料后处理能力作为我国核燃料循环体系的重要组成部分,将对我国核燃料循环情景产生重要影响。本文对后处理规模和分离钚的利用进行了假设,研究了两种情景模式下后处理和分离钚利用对我国铀资源需求和核废物产生的影响。     

Restructure behavior analysis in fast breeder reactor MOX fuel by X-ray CT

Both high- and low-density MOX fuel pellets of uranium and plutonium oxides were irradiated in the experimental fast reactor JOYO. After irradiation, these fuel pellets were examined by X-ray CT and their irradiation behavior was evaluated for formation of the central void. In particular, the central void size and temperature of fuel pellets at the beginning and end of irradiation were analyzed. The central voids in the low-density fuel pellets were bigger than those of the high-density fuel pellets at the same linear heating rate (LHR), and the threshold LHR and temperature at which the central voids were formed were lower than those of the high-density fuel pellets. It was understood from these results that the irradiation behaviors of high- and low-density fuel pellets were different.     

A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

Utilization of Mixed Uranium–Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse ‘AVVER-1000LEUandMOXAssemblyComputationalBenchmark’ and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group ‘JEFF31GX’ cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k_∞). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium–Gadolinium (UGD) pin, fission rate distributions in UGD, UO₂ and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.     

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.     

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.     

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO₂ fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.     

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were ?1.0%Δk for the diffusion calculations (JENDL-3.2), ?0.3%Δk for the transport calculations (JENDL-3.3), and 0.2%Δk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was ?2.41 ± 0.08%Δk/kk’, which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error.     

Use of Thorium-Plutonium MOX in the inner pins of CANDU fuel bundles

This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters.The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO₂ fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO₂. This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case.     

Specific low temperature release of Xe from irradiated MOX fuel

A particular low temperature behaviour of the ¹³¹Xe isotope was observed during release studies of fission gases from MOX fuel samples irradiated at 44.5 GWd/tHM. A reproducible release peak, representing 2.7% of the total release of the only ¹³¹Xe, was observed at ∼1000 K, the rest of the release curve being essentially identical for all the other xenon isotopes. The integral isotopic composition of the different xenon isotopes is in very good agreement with the inventory calculated using ORIGEN-2. The presence of this particular release is explained by the relation between the thermal diffusion and decay properties of the various iodine radioisotopes decaying all into xenon.     

Benchmark on deterministic 2-D MOX fuel assembly transport calculationswithout spatial homogenization

An important issue of deterministic transport methods for whole core calculations concerns the accuracy of homogenization techniques. A direct calculation for whole core heterogeneous geometries was not feasible in the past due to the limited capability of computers. With modern computational abilities, direct whole core heterogeneous calculations are becoming feasible. This paper explores a recent OECD/NEA benchmark problem proposed to test the accuracy of modern deterministic transport methods when applied to reactor core problems without spatial homogenization.

For this work a two-dimensional configuration was investigated and an accurate Monte Carlo reference solution was obtained. Twenty participants submitted solutions for the two-dimensional configuration and all of the participant solutions were compared to a reference Monte Carlo solution. Overall all the results submitted by the participants agreed well with the reference solution. A majority of the participants obtained solutions that were more than acceptable for typical reactor calculations and the remaining errors in the participant solutions can be attributed to the high order space-angle approximation necessary to solve this particular benchmark problem. It is important to note that the high order space-angle approximation needed for this benchmark is not necessary typical for all such whole-core heterogeneous problems.     

MOX燃料在轻水堆核电站中的应用

目前MOX燃料已成为一种可用于轻水堆核电站成熟的核燃料。简要介绍了国外该领域的发展状况以及MOX燃料对反应堆性能的主要影响和应对措施。探讨了MOX燃料在国内压水堆核电站中的应用问题。     

Integral measurements with a plutonium inert matrix fuel rod in a heterogeneous light water reactor lattice

As part of PSI's validatory efforts for neutronics calculations pertaining to inert matrix fuel (IMF) deployment in light water reactors, first-of-their-kind integral measurements have been carried out at the PROTEUS facility employing a specially fabricated Pu-Er-Zr IMF rod in a heterogeneous (boiling water reactor) test lattice. Analogous experiments have also been conducted with MOX and “dummy” IMF rods, providing the basis for a systematic comparison of experimental and calculational results.     

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically ¹³⁷Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.     

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens’ theory and reported thermal conductivities of unirradiated (U, Pu) O₂ and irradiated UO₂ pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.     

Upgrading of X-ray CT technology for analyses of irradiated FBR MOX fuel

The X-ray CT technology previously developed by JAEA was upgraded. The shape of the X-ray source beam was changed from a circular shape to an elliptical one and the collimator slit width was decreased from 0.3 to 0.1 mm. The X-ray detector was improved by changing a CdWO₄ scintillator to a highly sensitive silicon semiconductor detector. The analysis code of X-ray CT image was revised with respect to the number of points by using two kinds of experimental results and taking into account the effects of crack existence and deviation of the central void position from the radial center of a fuel pellet. As a result, high resolution X-ray CT images could be obtained on the transverse cross section of irradiated fuel assemblies. The error of the dimensional measurement was improved from ±0.1 to ±0.03 mm by upgrading the instrument and revising the analysis code of X-ray CT image. The discriminating accuracy of density difference could be increased, and the low density region (undisturbed region) and high density region (equi-axial and columnar regions) in the X-ray CT image on the cross section of irradiated fuel could be discriminated from each other. The reliability of fuel performance analysis improves because a large number of PIE data can be collected, compared with the conventional destructive PIE.