Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

Thermal–hydraulic modeling of water/Al2O3 nanofluid as the coolant in annular fuels for a typical VVER-1000 core

In this paper, a thermal–hydraulic analysis of nanofluid as the coolant is performed in a typical VVER-1000 reactor with internally and externally cooled annular fuel. The fuel assembly for annular case with 8 × 8 arrays is considered for annular pin configuration. The considered nanofluid is a mixture composed of water and particles of Al₂O₃ with various volume percentages. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Moreover, the temperature distributions of the fuel, clad and coolant are described for water/Al₂O₃ nanofluid in solid fuel and annular fuel. It is observed that as the concentration of Al₂O₃ nanoparticles increases, due to higher heat transfer coefficient of Al₂O₃ nanofluid, the temperature of the coolant is increased and the central fuel temperature is reduced. Thus, it improves margin from peak fuel temperature to melting. Finally, it is illustrated the use of the annular fuel instead of solid fuel in core of the reactor, security and efficiency of the nuclear power plant will be increased.     

Source efficiency as function of fuel and coolant in accelerator-driven systems

We have studied the efficiency of spallation neutron sources for different combinations of coolant and fuel in 80 MW_th, sub-critical, cores. It has been found that the proton source efficiency, ψ^∗, is reduced by 10% when switching coolant from helium to lead–bismuth eutectic. Substituting MOX fuel with an americium based fuel, results in another 10% reduction of ψ^∗. The relatively high source efficiencies found for prototype accelerator-driven systems, using standard MOX fuel and helium coolant, may thus be difficult to achieve in future systems dedicated to the transmutation of higher actinides. Our results are in agreement with previous investigations of the dependence of the source efficiency on the selection of coolant.     

Core physics analysis of 100% MOX core in IRIS

International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO₂ fuel with less than 5% ²³⁵U, and the analysis has been previously completed confirming good performance for that case. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance and safety parameters, such as critical boron concentration, peaking factors, discharge burnup, reactivity coefficients, shut-down margin, etc. In addition, the basis to perform load follow maneuvers via the Westinghouse innovative strategy MSHIM has been established. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the operational and safety limits considered thus confirming viability of this option from the reactor physics standpoint.     

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.     

Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.     

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.     

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.     

Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.     

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.     

Effect of fuel burnup on the effectiveness of ultrasonic detection of leaky fuel elements in VVER-1000 fuel assemblies

The effect of fuel burnup in VVER-1000 fuel elements on the utilization effectiveness of ultrasonic detection of leaky fuel elements is examined. It is determined that the limitations of this method are due to the interaction of fuel-element cladding with the fuel pellets. Threshold for fuel burnup in VVER-1000 fuel elements with E-110 alloy cladding, determining the application limits of ultrasonic detection of leaky fuel elements in fuel assemblies, is determined.     

Estimation of oxygen potential for U-Pu-O mixed oxide fuel

A thermochemical analysis of the estimation of oxygen potential for mixed oxide fuel is presented. The MOX fuel is treated as a pseudo-quaternary solid solution of UO₂-U_cO_d-PuO₂-Pu_aO_b with b = 1.5a and d = 2.25c. The hypo-stoichiometric fuel is viewed as UO₂-(1/a)Pu_aO_b-PuO₂ pseudo-ternary system and the hyper-stoichiometric fuel is viewed as UO₂-(1/c)U_cO_d-PuO₂ pseudo-ternary system and the graphical representation is visualized as a ‘diamond plot’. The oxygen potential for MOX is determined either by PuO₂-Pu_aO_b equilibria or by the UO₂-U_cO_d equilibria subject to the constraint that the oxygen potential is equal both in Pu-O subsystem and in U-O subsystem. The choice of stoichiometric parameters a and c affects the relative contribution of configurational entropy to the oxygen potential which determines the shape of the ‘S’-type curve near exact stoichiometry. From the analysis of literature data, a = 2 and c = 8 was derived at for the ‘diamond plot’ model. Oxygen potentials for MOX fuel with 21, 28% and 44% plutonia was estimated using the ‘diamond plot’ model. A simplified nomograph for the oxygen potential of MOX is presented.     

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.     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

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.     

On the influence of spatial discretization in LWR-burnup calculations with HELIOS 1.9 – Part II: Mixed oxide (MOX) fuel

Cell and burnup calculations are the basis for all deterministic static and transient 3D full core calculations for different operational states of the reactor. The arising differences in the integral transport solution (neutron flux and k_inf) for different discretization strategies during the burnup of mixed oxide (MOX) fuel due to different spatial discretization are demonstrated. The influence of different discretization strategies on the calculation of homogenized few group cross-sections is investigated. The influence of the discretization strategies on the calculation time is evaluated.     

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.     

VVER nuclear fuel burnup with different absorbers

A computational study is performed of the fuel burnup in VVER-1000 using different absorbers in open and closed fuel cycles. It is shown that mixtures of plutonium isotopes (energy and others) can give the same effect as gadolinium, which is currently used. Fuel burnup increases. When neptunium, americium, and curium isotopes are used as a consumable absorber in a closed fuel cycle, the accompanying effect is elimination of long-lived α-emitting radionuclides which have accumulated in long-term repositories.