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The present paper concerns a novel type of integral database generated in the course of the LWR-PROTEUS Phase I experiments, viz. the relative reactivity effects of removing individual fuel pins from a highly heterogeneous BWR assembly. The measurements reported were conducted in the central assembly of a 33 test-zone configuration of SVEA-96+ assemblies in PROTEUS, under full-density water moderation conditions. Calculations of the pin-removal reactivity worths have been carried out using two different LWR assembly codes, viz. CASMO-4 and BOXER. The discrepancies observed between experiment and calculation have been analysed in detail, on the basis of an extended reactivity decomposition methodology. This has allowed quantification of the different phase-space contributions (in terms of reaction rate types, energy groups and spatial regions) of a given calculated pin-removal reactivity worth, thus providing useful insights regarding the most important sources of error in each of the assembly codes investigated.[All rights reserved Elsevier].

Different foil activation techniques have been used for measuring spatial distributions of the 63Cu(n,)64Cu reaction within two pins of a SVEA-96 Optima2 boiling water reactor fuel assembly, at the critical facility PROTEUS. This reaction is of interest because its 1/v cross-section gives it a good representation of the 235U fission rate. Initially, radial capture rate profiles were measured with mechanically punched copper foils. More detailed profiles were then determined by using a 0.2 mm copper wire spiral (~200 m resolution), as well as 5-, 10-, and 20-ring UV-lithography, electroplating, and molding (UV-LIGA) foils (up to a 100 m resolution). For azimuthal measurements, apart from manually cut activation foils (into 8 sectors), 8- and 12-sector LIGA foils were used. The highly versatile LIGA foils have the additional advantage of being very easily separated into individual pieces after irradiation without the use of punches or other cutting tools. In order to account for the invasive character of the foil activation techniques, corrections to account for sample perturbations and for self-shielding effects were determined via simplified Monte Carlo (MCNP4C) modeling of the experimental setup. The final results from the various measurements of 63Cu(n,)64Cu within-pin distributions have been compared with MCNP computations employing a detailed model of the full SVEA Optima2 fuel assembly

The reaction-rate ratio C8/Ftot, neutron captures in 238U to total fissions, has been measured in 80 out of 96 fuel rods of a Westinghouse SVEA-96+ boiling water reactor fuel assembly. High-resolution gamma spectroscopy was performed on individual fuel rods, withdraw: from the SVEA-96+ assembly after irradiation at low power in the center of the LWR-PROTEUS reactor core. Absolute experimental errors of 1.7% and relative errors of 0.6% (for rod-to-rod ratios) were achieved. The experimental results were used as a database for validation of four different calculational tools: CASMO-4 and HELIOS as commercial assembly codes, the Paul Scherrer Institute in-house code BOXER, and the Monte Carlo transport code MCNPX. In general, on the level of a few percent, there is good agreement between experiment and calculations, the use of a recently proposed 239Np gamma-ray emission probability improving even further the agreement. However, the highly heterogeneous design of the SVEA-96+ assembly (both in terms of material compositions and neutron moderation conditions) causes some problems. Clear deviations from assembly mean values are found among the burnable absorber fuel rods that are grouped in clusters (direct neighbors), a unique feature of this assembly design. For these rods the codes overpredict C8/F tot by several percent, including MCNPX. Additional trends, not present in the results from the Monte Carlo calculation which generally shows the best overall agreement with experiment, are identified for the deterministic codes.

As part of a joint research programme between the Paul Scherrer Institute (PSI) and swissnuclear, with the co-operation of the Leibstadt nuclear power plant in Switzerland and fuel suppliers Westinghouse Sweden, measurements and calculations have been made of the axial and radial distributions of fission and 238U capture rates in the fuel rods of a Westinghouse SVEA-96 Optima2 boiling water reactor assembly. The measurements, made in the zero-energy research reactor PROTEUS at PSI, have been compared with calculations carried out using the Monte Carlo code MCNPX. The results reported are for the regions near the ends of the part-length fuel rods, which are a feature of SVEA-96 Optima2 assemblies. The sudden increase in moderation above the ends of the part-length rods leads to power peaking in the adjacent rods. Careful attention needs to be given to this phenomenon in the deployment of such fuel, the present paper providing experimental evidence for the ability of a stochastic code to predict such effects. [All rights reserved Elsevier].

Critical experiments involving fuel pin gamma-scans have been performed to characterise total fission reaction rate axial distributions in a Westinghouse SVEA-96+ assembly under full-density water moderation [these investigations are part of the co-operation between PSI and Unterausschuss Kernenergie (UAK) or association of the Swiss nuclear operators, viz. Kernkraftwerk Leibstadt AG, Kernkraftwerk Gosgen-Daniken AG, BKW FMB Energie AG and Nordostschweizerische Kraftwerke AG]. The measurements have covered an 80-cm long region of the SVEA-96+ assembly involving two important axial heterogeneities: (a) the enrichment boundary separating the lower axial region (average 235U enrichment of 4.02 wt.%, 14 burnable absorber fuel pins) from the upper axial region (average 235U enrichment of 3.70 wt.%, 16 burnable absorber fuel pins), and (b) the presence of spacers at about 12 cm below this enrichment boundary. The experimental results have been compared to HELIOS/PRESTO-2 calculated values at three different levels, by means of radial point-wise, axial point-wise and axial node-wise comparisons. The results show good agreement (typically within 2%) between HELIOS/PRESTO-2 nodal pin powers and corresponding experimental values, with the largest discrepancy of 6-9% being highly localised in a pin subject to strong flux gradients and spectral interactions. Further improvements would require explicit three-dimensional lattice calculations and the development,of three-dimensional pin power reconstruction methods

Non-destructive and destructive methods have been compared to validate their corresponding assessed accuracies in the measurement of 134Cs/137Cs and 154Eu/137Cs isotopic concentration ratios in four spent UO2 fuel samples with very high (52 and 71 GWd/t) and ultra-high (91 and 126 GWd/t) burnup values, and about 10 (in the first three samples) and 4 years (in the latter sample) cooling time. The non-destructive technique tested was high-resolution gamma spectrometry using a high-purity germanium detector (HPGe) and a special tomographic station for the handling of highly radioactive 400 mm spent fuel segments that included a tungsten collimator, lead filter (to enhance the signal to Compton background ratio and reduce the dead time) and paraffin wax (to reduce neutron damage). The non-destructive determination of these isotopic concentration ratios has been particularly challenging for these segments because of the need to properly derive non-Gaussian gamma-peak areas and subtract the background from perturbing capture gammas produced by the intrinsic high-intensity neutron emissions from the spent fuel. Additionally, the activity distribution within each pin was determined tomographically to correct appropriately for self-attenuation and geometrical effects. The ratios obtained non-destructively showed a 1 statistical error in the range 1.9-2.9%. The destructive technique used was a high-performance liquid chromatographic separation system, combined online to a multicollector inductively coupled plasma mass spectrometer (HPLC-MC-ICP-MS), for the analysis of dissolved fuel solutions. During the mass spectrometric analyses, special care was taken in the optimisation of the chromatographic separation for Eu and the interfering element Gd, as also in the mathematical correction of the 154Gd background from the 154Eu signal. The ratios obtained destructively are considerably more precise (1 statistical error in the range 0.4-0.8% for most of the samples, but up to 2.8% for one sample). The HPGe gamma spectrometry can achieve a high degree of accuracy (agreement with HPLC-MC-ICP-MS within a few percent), only by virtue of the optimised setup, and the refined measurement strategy and data treatment employed. [All rights reserved Elsevier].

Characterisation of the SVEA-96 Optima2 boiling water reactor assembly, in terms of the radial distributions of normalised total fission and 238U capture rates, is reported at its central elevation, i.e. at the 92-pin section, where the one-third part-length pins are replaced by water. Measurements performed in the PROTEUS facility are compared with MCNPX predictions. The calculation model included the measured locations of the SVEA-96 Optima2 assemblies and sub-assemblies, within the PROTEUS test zone. Predicted and experimental fission and 238U capture rates are found to agree, respectively, within 3.5% and 4% for all pins. Fission rates in the burnable-absorber UO2-Gd2O3 fuel pins have been predicted without bias using the ENDF/B-VI data library but show an average 1.4% under-prediction with the JEFF-3.1 data library. A slight overestimation of the total fission rate in the pins located at the periphery of the assemblies was observed and has been attributed to an inaccurate modelling of the pin positions. However, there was no systematic bias observed due to the absence of the one-third pins at the corners of the assembly.[All rights reserved Elsevier].

Computerised gamma-ray emission tomography has been applied to single PWR UO2 fuel rods, with pellet averaged burnups of 52, 71, 91 and 126 GWd/t respectively, for the determination of 134Cs, 137Cs and 154Eu internal radial distributions. State-of-the-art image reconstruction techniques, analytical and iterative, have been applied, evaluated and compared using test phantoms first and, in a second step, the actual measured data. Further, linear attenuation maps, previously derived on the same samples by means of gamma-ray transmission tomography, have been used to correct for density inhomogeneities. The final results have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (periphery-to-centre ratios of ~2.5 for 137Cs and ~3 for 134Cs). In addition, a difference in the relative activity distributions of 137Cs and 134Cs has been observed for all the samples. In contrast, the europium shows an almost flat distribution. [All rights reserved Elsevier].

A boiling water reactor SVEA-96+ fresh fuel lattice has been used as the basis for a benchmark study of the void reactivity coefficient at assembly level in the full voidage range. Results have been obtained using the deterministic codes CASMO-4, HELIOS, PHOENIX, BOXER and the probabilistic code MCNP4C, combined for almost all cases with different cross section libraries. A statistical analysis of the results obtained showed that the void reactivity coefficient tends to become less negative beyond 80% void and that the discrepancies between codes tend to increase from less than 15% at voidages lower than 40% to more than 25% at voidages higher than 70%. The void reactivity coefficient results and the corresponding differences between codes were isotopically decomposed to interpret discrepancies. The isotopic decomposition shows that the minimum observed in the void reactivity coefficient between 80% and 90% void is largely due to the decrease in the relative importance of the 157Gd(n, ) rate with increasing voidage, and that the fundamental discrepancies between codes or libraries are mainly governed by the different predictions of the 238U(n, ) variation with voidage. [All rights reserved Elsevier].

A methodology has been developed for the accurate assessment of localised reactivity perturbations in a BWR lattice embedded in a larger multiplying system, based on a full-system, unperturbed calculation, and on perturbed calculations on reduced-geometry models with reflective boundary conditions (typically, reflected-assembly calculations). Reflective reduced-geometry calculations are to be followed by a fast transferability correction for making the results representative of what full system computations would have produced. In this way, one can avoid the problem of having insufficient accuracy in the results (in spite of extremely lengthy iterations), particularly for cases of small reactivity effects. Furthermore, the factorization of reactivity effect transferability, a key feature of the developed methodology, provides valuable insight into the different effects contributing to a particular integral transferability factor, along with a quantification of the relative importance of these effects for each individually considered case. The initial investment, needed for realizing the relatively low required computational effort involved in the postcorrection procedure, is to obtain a limited number of adjoint equation solutions defined for the reference state at full system level. Application results are reported for the numerical analysis of fuel pin removal reactivity effects in LWR-PROTEUS. The latter is a programme of integral experiments, employing essentially a central LWR test zone driven critical by surrounding driver and buffer regions

An important source of uncertainty in boiling water reactor physics is associated with the precise characterisation of the moderation properties of the coolant and by-pass regions, with significant impact on reactor physics parameters such as the lattice neutron multiplication, the neutron migration area and the pin-by-pin power distribution. In this paper, the effects of certain relevant void-fraction uncertainties on reactor physics parameters have been studied for a BWR assembly of the type Westinghouse SVEA-96 using CASMO-4, HELIOS/PRESTO-2 and MCNP4C. The SVEA-96 geometry is characterised by the sub-division of the assembly into four different sub-bundles, by means of an inner by-pass with a cruciform shape. The study has covered: (a) the effects of different cross-section data libraries on the void coefficient of reactivity, for a wide range of void fractions; (b) the consideration of a water film inside the sub-bundle walls, and (c) the impact of partly inserted absorber blades producing very different void fractions in different sub-bundles. [All rights reserved Elsevier]