The OECD Halden reactor project fuels testing programme: methods, selected results and plans

The fuels testing programme conducted in the Halden reactor (heavy boiling water reactor (HBWR)) is aimed at providing data for a mechanistic understanding of phenomena, which may affect fuel performance and safety parameters. The investigations focus on implications of high burnup and address thermal property changes, fission gas release as influenced by power level and operation mode, fuel swelling, and pellet–clad interaction. Relevant burnup levels (>50 MWd kg⁻¹ U) are provided through long-term irradiation in the HBWR and through utilisation of re-instrumented fuel segments from commercial light water reactors (LWR). Both urania and MOX fuels are being studied regarding thermal behaviour, conductivity degradation, and aspects of fission gas release. Experiments are also conducted to assess the cladding creep behaviour at different stress levels and to establish the overpressure below which the combination of fuel swelling and cladding creep does not cause increasing fuel temperatures. Clad elongation measurements provide information on the strain during a power increase, the relaxation behaviour and the extent of a possible ratcheting effect during consecutive start-ups. Investigations foreseen in the programme period 2000–2002 include the behaviour of MOX and Gd-bearing fuel and other variants developed in conjunction with burnup extension programmes. Some LWR-irradiated fuel segments will undergo a burnup increase in the HBWR to exposures not yet achieved in LWRs, while others will be re-instrumented and tested for shorter durations.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

Fuel burnup performance of an OTTO refueling pebble bed reactor with burnable poison loading

Fuel burnup performance has been analyzed for a pebble bed reactor with a once-through-then-out (OTTO) refueling scheme and compared with a reference multi-pass scheme. A new fuel pebble was designed by adding spherical B₄C particles into its free fuel zone for controlling the infinite multiplication factor during burnup, and then reducing the axial power peak of the OTTO scheme. The objective is to maximize the fuel burnup performance of the OTTO scheme while keeping the power peak under a limit and ensuring the core criticality. Numerical calculations were performed based on the 400 MWt pebble bed modular reactor (PBMR) using the MVP code. For the fuel pebble of the PBMR containing 9 g uranium with 9.6 wt% ²³⁵U enrichment, 1600 B₄C particles with a radius of 70 μm are determined to flatten the k_∞ curve in the early burnup stage. The dependences of the neutronic properties of the core with the OTTO scheme on target fuel burnup show that the maximum target burnup of 74 GWd/t can be achieved so that the power peak is reduced to about 10.80 W/cm³ which is approximate that of the multi-pass scheme (10.85 W/cm³). This target burnup is about 22% less than that of the multi-pass scheme (95 GWd/t), i.e. the fuel utilization efficiency of the OTTO scheme is about 22% lower, which could be compensated by the construction and operation cost of the fuel handling system. This result also suggests that further investigations of the fuel burnup performance and other properties are needed in both neutronic and thermal hydraulic viewpoints to find out the optimal core performance.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

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.     

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.     

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other hand, in order to accommodate solid fission product swelling and to control fuel clad mechanical interaction of the stiffer fuel, the fuel smear density is reduced to 70%. In addition, plenum height is increased to accommodate for fission gases.     

The effect of aluminum corrosion on the advanced neutron Source Reactor fuel design

The corrosion of fuel plates can have a significant impact on the performance of reactors with aluminum fuel cladding. Increased fuel temperatures result when the oxide imposes an additional thermal resistance between the fuel plate and the coolant, while spallation of the oxide film may reduce cladding integrity. Characterization of oxide growth over the fuel cycle is therefore a necessary element in the design of such reactors. This paper describes the impact of fuel plate oxide formation on the Advanced Neutron Source Reactor (ANSR) core design.An oxide layer continually grows on the fuel plate surface throughout the fuel cycle. Although the cycle is short in duration (14 days), even a thin layer (< 25 μm) of low conductivity aluminum oxide causes a significant increase in fuel temperatures due to high ANSR power densities. The growth rate of the oxide is determined by many factors: impurities in the D₂O, coolant and oxide temperatures, heat flux, etc. Since several of these factors vary during the fuel cycle, prediction of oxide thicknesses must be time dependent. This is presently accomplished by using a series of core power density distributions calculated at different times within the fuel cycle in concert with an empirical oxidation model.Two specific thermal limits are imposed on the ANSR core design which are affected by oxide growth. A fuel centerline temperature limit of 400°C is established by fuel swelling behavior, while the oxide temperature drop is limited to 119°C to avoid oxide spallation. This paper discusses these limits and the constraints they impose on the core design. A recently developed oxide growth correlation is used in combination with thermal hydraulic analysis to show that fuel loading design can be tailored to minimize thermal limitations on the ANSR imposed by oxide growth. A power shape is presented which ideally causes a uniform oxide film to form over the entire plate surface, improving operating margins by several percent. In reality, this ideal shape cannot be obtained due to various aspects of the core: control rod movement, the time dependent nature of fuel burnup, the fuel/moderator relationship, etc. The design process is, therefore, iterative between thermal hydraulic and neutronic analyses. Results of additional calculations are presented which describe the performance of these more realistic fuel loadings and compare them to the ideal case.     

Thermochemical Stability, Radiation Testing, Fabrication, and Reprocessing of Mononitride Fuel

A thermodynamic analysis and experimental investigations have shown that mononitride fuel is thermochemically stable up to 1973–2073 K, at which temperature the equilibrium vapor pressure of nitrogen does not exceed 4.5·10^–7–2.1·10^–6 MPa. It is concluded on the basis of a generalization of the data from radiation testing of mononitride fuel with burnup up to 9–10% in fast and 16.8% in thermal reactors with lineal power density from 400 to 1300 W/cm that it should operate reliably in fuel elements with helium and liquid-metal sublayers. The requirement for the impurity (oxygen and carbon) content in it is formulated. When both oxygen and carbon impurities are present simultaneously in mononitride, the mass fraction of each should not exceed 0.15%. The methods for fabricating mononitride fuel are determined by the final product of the reprocessing of irradiated fuel. Consequently, methods for fabricating mixed nitride fuel from oxides and metals are now being developed.     

The analysis of the insertion of UO2 microspheres in the gap between pellet and clad as a new design solution to increase the burnup of the nuclear fuel

The insertion of UO₂ microspheres (eventually graphite coated) in the gap between pellet-clad is observed to decrease substantially the clad hoop plastic strain concomitantly with the elimination of the rim effect at high burnup if low enrichment is used for the microspheres. Taking into account the special features of the specialized finite element code ELFIN'90 for the behavior of fuel elements, it was possible to introduce this new type of material viewed as a granular media. The results of the new code version ELFIN'MS applications to a PHWR fuel for a power ramp irradiation history show that the hoop plastic strain is reduced by about three times in comparison to standard fuel, and that the ridge phenomenon disappears. To establish critical plastic strain limit for irradiated clad failure onset, quantitative evaluations of iodine chemisorbtion on graphite and at the surface of the irradiated zircaloy, are presented. The indications on technology procedure are also discussed. Therefore, the insertion of 2–3 layers of UO₂ microspheres of 100 μm diameter, graphite coated to retain corrosive fission products for clad and with the diameter greater than the design gap, can be considered a design solution to increase the burnup of nuclear fuel.     

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.     

燃料热导率降级对CAP1000大破口失水事故的影响分析

在高燃耗情况下,燃料芯块的热导率随燃耗降低,该现象被称之为热导率降级(TCD)现象。TCD现象影响失水事故(LOCA)前稳态工况的燃料平均温度和燃料储能,进而影响大破口LOCA过程中的包壳峰值温度(PCT)。本研究采用大破口LOCA分析程序WCOBRA/TRAC对CAP1000冷段双端剪切断裂事故进行了不同燃耗的敏感性分析,并获得了不同工况下的PCT。分析中采用美国核燃料研究所(NFI)修正的TCD模型对降级后的燃料热导率进行模拟,同时考虑了燃耗大于30GW·d/tU后FQ和FΔh峰值因子的降低。敏感性分析表明,考虑TCD和峰值因子降低的影响,PCT极限工况不再出现在低燃耗区间,而出现在燃耗为29GW·d/tU附近。与其他燃耗水平相比,该燃耗点的PCT第1峰值和第2峰值均处于最高水平。本研究结果可为高燃耗情况下非能动电厂大破口LOCA的分析评估提供参考。     

Statistical estimation of fast-reactor fuel-element lifetime

Conclusions On the basis of a statistical analysis, the main parameters having a significant influence on the theoretical determination of fuel-element lifetimes in the operation of power fast reactors in steady power conditions may be isolated. These include the creep and swelling of the fuel and shell materials, prolonged-plasticity lag, shell-material corrosion, gap contact conductivity, and the strain diagrams of the shell and fuel materials obtained for irradiated materials at the corresponding strain rates. By means of deeper investigation of these properties of the materials, it is possible to increase significantly the reliability of fuel-element lifetime predictions in designing fast reactors and to optimize the structure of fuel elements more correctly.Another important result of the statistical analysis is information regarding the increase in the number of unsealed fuel elements with fuel burnup. It is shown that for the given conditions of reactor operation the number of unsealed fuel elements rises almost exponentially, beginning with a burnup >7% of heavy atoms. The results of such calculations must obviously be taken into account in the cost-benefit analysis of projected new reactors and in choosing the optimal fuel burnup.Translated from Atomnaya Énergiya, Vol. 48, No. 1, pp. 16–19, January, 1980.     

The response of carbide fuel elements to fast thermal transients using the UNCLE-T-BUBE code

The response of fuel elements to fast thermal transients have great implications to the safety of LMFBR's. In this article, fission gas swelling and release, and clad stress and strain are computed for a carbide fuel element during several fast thermal transients as a function of steady stae power and percent burnup. The computations are made with the UNCLE-T-BUBE code which allows for equilibrium and nonequilibrium fission gas bubbles. In some of the transients, the code UNCLE-T-BUBE predicts fuel-clad gap closure, attended with a high clad hoop stress, whereas UNCLE-T does not. It is also found that allowing for nonequilibrium fission gas bubbles strongly affects fuel swelling and clad strain but has negligible effect on gas release.     

Testing of experimental fuel elements for VVÉR-1000 reactors in MR to high fuel burnup

Conclusions The results of testing in a multipurpose reactor and post-irradiation examinations indicate satisfactory performance of the fuel element for the VVÉR-1000, which is designed for a 3-year run. In addition to the computational data, the experimental data were used to substantiate the performance of fuel elements when fuel burnup is increased and atomic power plants are switched from the VVÉR-1000 to a 3-year cycle (with an average burnup of 40 MW-day/kg).I. V. Kurchatov Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 72, No. 2, pp. 116–120. February, 1992.     

Symbiotic systems consisting of large-FBR and small water-cooled thorium reactors (WTR)

Small long life water-cooled thorium reactors (WTR; 30–300 MWth) have been investigated. For realizing thorium cycle of the reactors, a uranium–thorium mixture core is introduced to fast breeder reactors (FBR; 3000 MWth) to be a ²³³U producer. In the present study, two distinct metallic fuel pins, with natural uranium and thorium, are loaded into a large sodium-cooled FBR. The FBR itself is self-sustained by the plutonium produced in the uranium pins. Under the equilibrium burnup state, the FBR spent fuels are periodically discharged with a certain discharge rate and then separated. Some actinides are returned to the FBR core while ²³³U, which is discharged from the thorium pins, is utilized for the WTR fresh fuel. Fissile support capability is the main investigated parameter of the study. The system achieves higher support capability at higher burnup and lower power of the WTR, and shows that larger number of uranium pins is better for the FBR criticality while larger number of thorium pins and lower burnup give better support factor capability. For a symbiotic system consisting 3000 MWth FBR and 100 MWth WTRs, where discharged fuel burnup is 96 and 60 GWd/t for the FBR and WTRs, one FBR can support 5 WTRs.     

Effect of Fast Neutron Irradiation on the Properties of Boron Carbide Pellet

Boron carbide pellets were irradiated in the experimental fast reactor “JOYO” to ¹⁰B burnup of up to 170x10²⁶cap/m³, fluences of 2x10²⁶/m²(E>0.1MeV), and maximum temperatures of about 1,200°C. Post irradiation examinations were made of microstructural changes, helium release, swelling, and thermal conductivity.

Boron carbide pellets irradiated to high burnups developed extensive cracking. Helium release from the pellets was initially low, but enhanced helium release was observed at high burnups and high temperatures. The swelling linearly increased with burnup, and when boron carbide was irradiated at high temperatures, the swelling rate began to decrease corresponding to the beginning of enhanced helium release. The correlation between swelling and the helium release was studied and the swelling was interpreted in terms of accumulation of helium in the boron carbide pellet. The thermal conductivity of the boron carbide pellets decreased rapidly by neutron irradiation accompanied with loss of temperature dependence.     

FRAPCON-2程序改进和校验

通过改进FRAPCON-2程序中的燃料导热系数模型和裂变气体释放模型,使之能对高燃耗的燃料进行性能分析计算。并利用Halden堆IFA 597.3 ROD8的试验数据对程序进行了验证。结果表明,改进后的程序所计算出的参数（如燃料温度和裂变气体释放份额）均与实测值符合很好,对程序的改进是成功的。     

Analysis on split failure of cladding of high burnup BWR rods in reactivity-initiated accident conditions by RANNS code

A computer code RANNS was developed to analyze fuel rod behaviors in the reactivity-initiated accident (RIA) conditions. RANNS performs thermal and finite-element mechanical calculation for a single rod in axis-symmetric geometry, where fuel pellet consists of 36 equal-volume ring elements and cladding metallic wall consists of eight equal-thickness ring elements and one outer oxide element. The code can calculate temperature profile inside the rod, contact pressure generated by pellet–clad mechanical interaction (PCMI), stress–strain distribution and their interactions elaborately. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by the fuel performance code FEMAXI-6.In the present study, analysis was performed on the simulated RIA experiments in the “nuclear safety research reactor” (NSRR), FK-10 and FK-12, with high burnup BWR rods in a cold-start up condition, and stress–strain evolution in the PCMI process was calculated extensively. In the analysis, the pellet–clad bonding was assumed both in the heat conduction and in mechanical restraint. The calculated hoop strain increase was compared with the measured strain gauge data, and satisfactory agreement was obtained. Simulation calculations with broader power pulses anticipated in RIA of commercial BWR were carried out and the resulted cladding hoop stress was compared with the failure stress estimated by comparison of analysis with experimental data.