Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO₂ assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of ²³⁵U, well reproduce those of ²³⁹Pu and ²⁴⁰Pu, yet underestimate those of ²³⁶U, ²³⁷Nd, ²³⁸Pu, ²⁴¹Pu, and ²⁴²Pu. The inventories of minor actinides are underestimated by the present analysis except for ²⁴¹Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO₂ assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO₂ assemblies were well reproduced by the burnup calculations of MVP-BURN.     

Analytical investigations of irradiated inert matrix fuel

Yttria stabilised zirconia (YSZ) inert matrix fuel (IMF) fabricated at PSI and irradiated 3 years in the Halden Material Test Reactor (HBWR) since 2000, has been examined by Electron Probe Microanalysis (EPMA) and Secondary Ion Mass Spectroscopy (SIMS) after irradiation and compared with data gained for the unirradiated material. The examined pellet cross-section was estimated to have an equivalent burn-up of 22 MW d kg⁻¹. EPMA measurements demonstrate that the burn-up was rather flat over more than the half pellet radius. A Pu consumption of about 2.5 wt% has been measured with a higher rate in the fuel border zone. The high fuel temperature is responsible for a certain homogenisation of the mineral phases in the fuel centre region whereas the border zone has remained rather with an as-fabricated phase distribution. The central part was also characterised by a dense porosity distribution as well as a temperature and relocation driven depletion of the volatile fission products Xe and Cs. In addition, SIMS has been realised on the same specimen in order to determine the semi-quantitative distribution of different isotopes in the pellet.     

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.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

Thermoelastic stresses in pellet fragments and conditions for fragments formation

A finite element code CHICAM in which thermal and elastic effects are analysed together has been used to calculate the thermoelastic stress fields in whole and hollow pellets in a radial temperature field of approximately parabolic shape. In addition, the residual thermoelastic stresses were calculated for whole and hollow pellets containing a crack and for various geometries of pellet fragments. The residual stress fields in the pellet fragments differ very considerably from the axial symmetrical stress fields of the whole and hollow pellets. The stress intensity factor K_I is given as a function of the crack length for the whole pellet and for the half pellet showing that during the first rise to power a fuel pellet will fracture into several pie-shaped fragments not necessarily of the same angular width. The calculations show that the residual stress fields in fractured pellets cannot be simulated by a stress and strain analysis with axial symmetry.     

Pellet compliance model based on out-of-pile simulation

The thermal and mechanical behavior of fuel rods is significantly influenced by the extent of their relocation and by compliance of the cracked pellets. Movement of the cracked pellet pieces towards the cladding results in softer pellets with crack voids which accommodate some fraction of the thermoelastic pellet deformation and make the pellet more compliant under the restraint of the cladding. It is difficult to model such a pellet compliance independently of experimental observations because the cracked pellet behavior is uncertain by nature.Electrically heated simulation of pellet-cladding mechanical interaction (PCMI) facilitates much quicker and more flexible experimentation than actual in-pile tests. Testing apparatus consists of the simulated fuel rod with hollow UO₂ pellets and a tungsten rod in the center, and a diameter measuring device including three pairs of diameter sensors. Test parameters include the pellet-cladding gap and the cladding thickness. Results show that rods with a smaller gap have a larger increasing rate of cladding diameter. This suggests that a group of cracked pellet pieces induced by thermal stress has an apparent compliance which increases with pellet-cladding gap. Results also show more sensitivity to cladding thickness than those calculated assuming pellets having intrinsic stiffness. This also suggests the compliant nature of cracked pellets.Such a compliant nature can almost be described by reducing the elasticity of the pellet. A simple pellet compliance model was obtained by fitting calculations with measurements to describe a cracked pellet as a uniform axisymmetric body with apparent elasticity.     

A tritium permeation model for conceptual fusion reactor designs

A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10^?3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×10⁴ s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.     

The D-COM blind problem on fission gas release: The predictions of the TRANSURANUS and future codes

The concentration of retained xenon, the percentage of porosity and the UO₂ grain size have been measured as a function of radial position in the base irradiated rod AG11-8 and the transient tested rod AG11-10. In the base irradiation, densification of the fuel took place and slight grain growth occurred at the pellet centre. Gas release was not detected. During the transient test, 15–20% of the xenon inventory was released from the fuel grains. Gas release was accompanied in the central region of the fuel by an increase in the porosity from 4.7 to 6–8%. These findings are compared with the predictions made by the fuel performance code TRANSURANUS. The code predictions are in good agreement with the experimental observations. FUTURE was used to investigate the development of gas bubbles and the mechanisms controlling gas release in the rods during the base irradiation and the transient test. According to FUTURE fission gas will have accumulated on the grain boundaries during the base irradiation. The code indicates that variations in the fuel microstructure resulting from the base irradiation will have caused the level of gas release to vary along the fuel stack in rods AG11-9 and AG11-10 during the transient test. FUTURE also suggests that fission induced bubble re-solution became increasingly important for release during the latter stages of the transient test. Moreover, the code calculations imply that bubble migration could have played a significant role in the release process.     

Verification of FEMAXI-7 code by using irradiation test in Halden reactor for He-pressurization effect on FGR of BWR fuels under power transient

Power ramp test for He-pressurization effect on fission gas release (FGR) of about 42GWd/tUO₂ boiling water reactor (BWR) fuel rods was analyzed by the fuel performance code FEMAXI-7. The experimental data were obtained with the two rods, which were base irradiated in the Halden reactor for 12 years (IFA-409), then subjected to the power ramp tests (IFA-535) to investigate the He-pressurization effect. The FEMAXI-7 calculations were performed by inputting rod specifications and experimental conditions in both the baseand test irradiations. The results showed that the calculations reasonably followed the trends of measured cladding elongation and FGR during the power ramp test, depending on the pellet temperature and fission gas atoms diffusion rate. Based on the calculated results, the reason that no apparent He-pressurization effect was observed in the experiment was considered to be caused by insufficient gas communication during strong pellet–clad mechanical interaction (PCMI) and enhanced gap thermal conductance by the solid–solid contact due to gap closure.     

Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO₂ rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.     

Interaction of Impurity （Li, Be, B and C） and Hydrogen Isotope Pellet Injection with Reactor-relevant Plasmas

Based on the two-dimensional kinetic ablation theory of the hydrogen pellet ablation developed by Kuteev [B.V. Kuteev, Nuclear Fusion, 35 (1995) 431], an algorithm of erosion speed and ablation rate calculations for Li, Be, and B impurity pellets in reactor-relevant plasma has been derived. Results show compatibilities of lithium pellet injection used in α-particle diagnostics are positive in comparison with other solid impurity pellets (e.g. Be, B and C). Using the 2-D Kuteev lentil model, including kinetic effects, we find that currently existing pellet injection techniques will not meet core-fueling requirements for ITER-FEAT. A pressure as high as 254 MPa must be applied to a pellet accelerator with a 200 cm-long single-stage pneumatic gun, in order to accelerate a pellet with a radius rpo = 0.5 cm to a velocity of vpo, 24 × 105 cm/s penetrating 100 cm into the ITER plasma core. Comparisons of pellet velocity- and radius-dependent penetration depth between the Neutral Gas Shielding and the Kuteev's     

Burnup simulations and spent fuel characteristics of ZrO2 based inert matrix fuels

Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF’s) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO₂ based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.     

Extension of the TRANSURANUS burn-up model

The validation range of the model in the TRANSURANUS fuel performance code for calculating the radial power density and burn-up in UO₂ fuel has been extended from 64 MWd/kgHM up to 102 MWd/kgHM, thereby improving also its precision. In addition, the first verification of calculations with post-irradiation examination data is reported for LWR-MOX fuel with a rod average burn-up up to 45 MWd/kgHM. The extension covers the inclusion of new isotopes in order to account for the production of ²³⁸Pu. The corresponding one-group cross-sections used in the equations rely on results obtained with ALEPH, a new Monte Carlo burn-up code. The experimental verification is based on electron probe microanalysis (EPMA) and on secondary ion mass spectrometry (SIMS) as well as radiochemical data of fuel irradiated in commercial power plants. The deviations are quantified in terms of frequency distributions of the relative errors. The relative errors on the burn-up distributions in both fuel types remain below 12%, corresponding to the experimental scatter.     

Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code

Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4–7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0-31 MWd/kgUO₂. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release.

In the irradiation period up to about 23 MWd/kgUO₂, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO₂ pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO₂ cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.     

Investigation of the influence of the initial state of powder,pore-forming additions,and U<Subscript>3</Subscript>O<Subscript>8</Subscript> on the microstructure and strength of fuel pellets

It is shown that the initial state of uranium dioxide powder has no effect on the density, microstructure, and strength of pellets. Pore-forming agents and U₃O₈ used in fabrication lower the pellet strength because their particles are not spherical. To increase pellet strength, it is recommended that U₃O₈ be subjected to special processing to spheroidize the particles before mixing for uranium dioxide powder.     

Elastic and viscoplastic pellets fragmentation modeling using an axisymmetrical 1D finite element code

During first rise to power in Power Water Reactor, fuel pellets crack because of thermal expansion. The phenomena of pellet cracking and fragments relocation have a major influence on rod behaviour and especially on the cladding behaviour in the case of pellet–cladding interaction.This article presents the modeling used to take into account the fragmented state of the pellet in the EDF fuel rod thermo-mechanical code, CYRANO3^®. The aim is to simulate more realistic stress and strain fields in the pellet.The investigated method consists in adding parameters in the 1D finite elements calculations in order to integrate the multi-dimensional fragmentation effects in the axisymmetrical 1D code CYRANO3^®. These parameters modify the material behaviour by describing the fuel as an anisotropic damaged material. The modeling accounts for the opening and closing of radial pellet cracks. It has been implemented in the code for elastic and viscoplastic fuel behaviours.     

Diffusion Behavior of Cesium,Strontium and Cobalt in Pellet Waste Package

A leaching model for a radioactive pellet package for nuclear power plant wastes was studied. Calculated results obtained in term of the model were compared with experimental ones, where volumetric averages of diffusion coefficients in the Na₂SO₄ pellet and in the cement-glass were used to solve the diffusion equation in the pellet package. Since Na₂SO₄ pellets dissolved quickly, the pellet part were simulated by saturated Na₂SO₄ solution. The leaching ratios were meaured with different volume ratios of cement-glass in the package to discuss the validity of the model. The cement glass diffusion coefficient and distribution factor, which were required to solve the equation, were also measured. The calculated results agreed with the experimental ones within a 20% error.     

Online integrated visual inspection and sorting system for fast reactor fuels

Mixed oxide (MOX) fuel for prototype fast breeder reactor (PFBR) is designed to have initial burn up of 100,000 MWD/T. The major differences from thermal reactor fuel are relatively smaller dimension with central hole and higher plutonium concentration (21% and 28% of PuO₂) MOX pellets which are loaded into 2.5 m long clad tubes with depleted UO₂ blanket pellets at either end of the MOX stack. The relatively smaller dimension of fuel pellets for PFBR results in large volume at fabrication and inspection. To ensure fast and accurate inspection and sorting of as sintered pellets with less radiation exposure to personnel an integrated on line pellet inspection system for remote visual inspection and sorting of pellets based on diameter has been developed. Details of the integrated pellet inspection system developed at Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur along with the results of the performance trials has been described in this paper.     

Artificial neural network modeling for fission gas release in LWR UO2 fuel under RIA conditions

A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO₂ fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.     

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.