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As a part of the validation of the two-dimensional lattice code CASMO5, isotopic predictions from the most recent version are compared to measured actinide and fission products concentrations from mixed-oxide fuel samples taken from irradiated pressurized water reactor fuel assemblies in the ARIANE program. The calculations are performed with a sin...
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... geometric layout of the fuel assemblies and the axial samples positions in the fuel pins are shown in Figure 2. The fuel assemblies are a PWR type (14x14) with 179 fuel pins distributed in three plutonium content zones (Table 2), 16 guide tubes for the control rods and 1 instrumentation tube. The design parameters of the fuel assemblies and the fuel material compositions are provided in the ARIANE program documentation [5,6]. The detailed irradiation history of the samples is provided as well. The data include the elapsed time of each power step, moderator temperature (ºC), moderator density (g/cc), boron (ppm), fuel center and fuel surface temperatures (ºC), the samples power (W/cm) and the sample burnup (MWd/tHM), The provided fuel center and fuel surface temperatures (ºC) are calculated with MICROLUX code, [5,6] 3. CALCULATION ...
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Citations
... The goal of this, so-called "MxN", simulation (or multi-segment option) is to compute the sample exposure by performing burnup calculations for an entire, 3 × 3-assembly system. Examples of such approach can be found in reference [15] for the ARIANE BM1 sample, for pin power calculations of a full PWR core [16], and for bowing effects in references [17,18]. ...
In this paper the isotopic compositions from 8 Boiling Water Reactor samples are analyzed following different irradiation assumptions as well as different simulation tools. These samples are part of a proprietary experimental program by a Spanish consortium, and they were obtained from a GE14 assembly irradiated in Sweden. Calculated nuclide concentrations are compared with measured ones providing biases for a selection of isotopes and samples; calculated uncertainties are also provided. Finally, the decay heat from one the sample segment is calculated and compared among the different simulation assumptions. It is shown that depending on the considered nuclear data library and modeling, different contributors affect the calculated quantities, indicating a certain level of prediction power.
... The atomic densities were varied by +0.1% to +30% of their original atomic densities in order to include all potential deviations and to verify the linearity of the model (which has been confirmed on the neutron emission rates in this work). Indeed, Ref. [12] indicates that CASMO5 predictions of isotopic densities may differ by about 1-20% from experimental measurements, for the nuclides selected in this study. Therefore, the range of 0.1-30% was set to wrap these values, with a margin. ...
... This can be explained by the impact of the composition perturbation in one assembly on the neutron emission rates of the surrounding ones. As an example, Ref. [12] says that CASMO5 predictions of the 235 U atomic density may deviate up to +13% from experimental data. Such a deviation in the composition of assembly n°3 leads to a 10% increase in the fast neutron flux at the RPV, which is significant given the global uncertainty of fluence estimations (that may be about 12% [13]). ...
... The sensitivity coefficients of the fast neutron flux incident on the RPV, at azimuthal position 0°, to fuel composition perturbations were verified to be linear and could be evaluated for different assemblies and isotopes. By combining these coefficients with the data provided in Ref. [12], it has been shown that fuel composition perturbations can have a significant impact on the fast neutron flux estimates (up to 10% bias on the flux due to 13% on the 235 U atomic density). It was observed that the estimated sensitivity coefficients are related to the type and the total exposure of the perturbed assembly as well as to the position of the perturbed assembly within the core. ...
The operation of many French nuclear 900 MWe Pressurized Water Reactors (PWR) is being extended beyond their design lifetime. In the perspective of further extension, the evaluation of reactor pressure vessels (RPV) aging is of great importance. The embrittlement of the vessel is mainly induced by fast neutron irradiation (neutrons with energies above 1 MeV). Consequently, the fast neutron fluence is one of the quantities used by safety authorities to characterize the structural integrity of the RPV. Estimates of this quantity are based on evaluations of the fast neutron flux that can be performed by a two-step approach. The first step estimates the full core fission neutron source term and the second step models the transport of neutrons from the core to the RPV using the fission neutron distribution determined in the previous step. Such fast neutron flux evaluations are subject to bias and uncertainties, which must be precisely analyzed in order to enhance nuclear safety. Especially, a lack of knowledge in fuel composition can have a significant impact on the quantification of the fast neutron flux, since it may impact both steps of the calculation. Therefore, this study proposes to discuss the sensitivity of fast neutron flux assessments to perturbations of nuclear fuel composition. The analysis has been performed using a deterministic code (CASMO5) to compute the fission neutron source term and a Monte-Carlo model (MCNP6) to compute the neutron attenuation.
... The atomic densities were varied by +0.1% to +30% of their original atomic densities in order to include all potential deviations and to verify the linearity of the model (which has been confirmed on the neutron emission rates in this work). Indeed, Ref. [12] indicates that CASMO5 predictions of isotopic densities may differ by about 1-20% from experimental measurements, for the nuclides selected in this study. Therefore, the range of 0.1-30% was set to wrap these values, with a margin. ...
... This can be explained by the impact of the composition perturbation in one assembly on the neutron emission rates of the surrounding ones. As an example, Ref. [12] says that CASMO5 predictions of the 235 U atomic density may deviate up to +13% from experimental data. Such a deviation in the composition of assembly n°3 leads to a 10% increase in the fast neutron flux at the RPV, which is significant given the global uncertainty of fluence estimations (that may be about 12% [13]). ...
... The sensitivity coefficients of the fast neutron flux incident on the RPV, at azimuthal position 0°, to fuel composition perturbations were verified to be linear and could be evaluated for different assemblies and isotopes. By combining these coefficients with the data provided in Ref. [12], it has been shown that fuel composition perturbations can have a significant impact on the fast neutron flux estimates (up to 10% bias on the flux due to 13% on the 235 U atomic density). It was observed that the estimated sensitivity coefficients are related to the type and the total exposure of the perturbed assembly as well as to the position of the perturbed assembly within the core. ...
The operation of many nuclear Pressurized Water Reactors (PWR) is being extended beyond their design lifetime threshold. In the perspective of a possible further lifetime extension, satisfying safety requirements is a priority. The characterization of the structural integrity of the reactor pressure vessel (RPV) is an important issue as it is a guiding parameter that influences the reactor lifetime. The embrittlement of the RPV material is primarily induced by the bombardment of fast neutrons (with energies greater than 1 MeV). Consequently, the fast neutron fluence is one of the quantities used by safety authorities to characterize the structural integrity of the RPV. As reactors are aging, the global bias and uncertainty of fast neutron fluence calculations are analyzed precisely in order enhance nuclear safety assessments. Fast neutron fluence estimates are based on fast neutron flux assessments that can be carried out using a two-step approach. The first step is to compute the full core fission neutron source term and the second step consists in the transport of neutrons from the core to the RPV, using the fission neutron distribution determined in the previous step. Such a two-step methodology implies approximations on the fission spectrum modeling. Therefore, this article discusses the sensitivity of neutron flux at the RPV to the fission spectrum modeling. To do this, neutron flux predictions from decoupled schemes that model the fission spectrum at the scale of a fuel assembly or a reactor core have been compared to those of a reference Monte-Carlo fundamental mode calculation. The analysis was performed using the MCNP6 code.
... The present study contributes to such validation efforts for the CASMO5 transport code [1][2][3][4]. The calculated nuclide inventory for the relocated Post Irradiation Examination sample GU3 from the ARIANE program [5] will be presented and compared to measurements, and calculated uncertainties as well as biases will be proposed. ...
This study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources (nuclear data, operating conditions and manufacturing tolerances) are also provided, and are combined with biases into expanded uncertainties. This study is similar to a previous one on the GU1 sample and fit in the framework of code validation, as well as in the estimation of code predictive power for spent fuel characterization.
The Mixed Oxide samples (MOX) ARIANE Post Irradiation Examination samples BM1 and BM3 have been analyzed in this work, based on various two- and three-dimensional models. Calculated and measured nuclide inventories are compared based on CASMO5, SIMULATE and SNF simulations, and calculated values for the decay heat of the assembly containing the samples are also provided. For uncertainty propagation, the covariance information from three different nuclear data libraries are used. Uncertainties from manufacturing tolerances and operating conditions are also considered. The results from these two samples are compared with the ones from two UO 2 samples, namely GU1 and GU3, also from the ARIANE program, applying the same calculation scheme and uncertainty assumptions. It is shown that a two-dimensional assembly model provides better agreement with the measurements than a two-dimensional single pin model, and that the full core three-dimensional model provides similar results compared to the assembly model, although no ¹⁴⁸ Nd normalization is applied for the full core model. For the MOX assembly decay heat, as expected, heavy actinides have a higher contribution compared to the cases with the UO 2 samples; additionally, decay heat uncertainties are moderately smaller in the case of the MOX assembly.
