The European Journal of Physics N (EPJ-N)
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Nuclear data uncertainty analysis for the Po-210 production in MYRRHA
No full textMYRRHA is a multi-purpose research reactor able to operate in sub-critical and critical modes and currently in the design phase at SCK•CEN. The choice of LBE was driven by its chemical stability, low melting temperature, high boiling point, low chemical reactivity with water and air and a good neutronic performance. As a drawback, the neutron capture in 209Bi results in the production of 210Po, a highly radiotoxic alpha emitter with relatively short half-life (≈138 days). The 210Po production represents a major safety concern that has to be addressed for the reactor licensing. In this work we used the ALEPH-2 burnup code to accurately calculate the 210Po production in a MYRRHA operating cycle. The impact of using different nuclear data libraries was evaluated and the reliability of the results was determined by quantifying the uncertainty of the 210Po concentration. The uncertainty quantification was carried out sampling the currently available nuclear data covariance matrices with the SANDY code. Also, estimates of the sensitivity profiles were obtained with a linear regression approach. The activation yield of the 209Bi neutron capture reaction was assessed as the largest nuclear data source of uncertainty, however the lack of covariances for such data represent a capital drawback for the 210Po content prediction
A methodology for performing sensitivity analysis in dynamic fuel cycle simulation studies applied to a PWR fleet simulated with the CLASS tool
No full textFuel cycle simulators are used worldwide to provide scientific assessment to fuel cycle future strategies. Those tools help understanding the fuel cycle physics and determining the most impacting drivers at the cycle scale. A standard scenario calculation is usually based on a set of operational assumptions, such as reactor Burn-Up, deployment history, cooling time, etc. Scenario output is then the evolution of isotopes mass in the facilities that composes the nuclear fleet. The increase of computing capacities and the use of neutron data fast predictors provide new opportunities in nuclear scenario studies. Indeed, a very high number of calculations is possible, which allows testing a high number of operational assumptions combinations. The global sensitivity analysis (GSA) formalism is specifically well adapted for this kind of problem. In this new framework, a scenario study is based on the sampling of operational data, which become input variables. A first result of a scenario study is the highlight of relations between operational input data and outputs. Input variable subspace that satisfy optimization criteria on an output, such as plutonium incineration or stabilization, can also be determined. In this paper, a focus is made on the methodology based on GSA. This innovative methodology is presented and applied to a simple fleet simulation composed of a PWR-UOx fuel and a PWR-MOx fuel. Calculations are done with the fuel cycle simulator CLASS developed by the CNRS/IN2P3 in collaboration with IRSN. The design of experiment is built from five fuel cycle input sampled variables. Sensitivity indices have been calculated on plutonium and minor actinide (MA) production. It shows that the PWR-UOx Burn-Up and the fraction of PWR-MOx fuel are the most important input variables that explain the plutonium production. For the MA production, main drivers depend strongly on isotopes. Sensitivity analysis also reveals input variable subspace responsible of simulation crash, what led to an important improvement of the model algorithms. An equilibrium condition on the plutonium mass in the stockpile used for building MOx fuel has been applied. The solution is represented as a subspace in the PWR-UOx Burn-Up and PWR-MOx fraction input space. For instance, achieving a plutonium equilibrium in a stockpile fed by a PWR-UOx that operates at 40 GWd/t requires a PWR-MOx fraction between 9 and 14%. This study also provides data related to plutonium incineration induced by the utilization of the MOx
Influence of nuclear structure data on fission observables
No full textThe simulation of the de-excitation of nuclei requires some models and data in order to construct the nuclear level scheme and the associated transition intensities. The aim of this work is to focus on nuclear structure data used at low energy where electromagnetic transitions can be measured. The RIPL3 database linked to the FIFRELIN Monte Carlo code contains such data and their influence on fission observables is reviewed
Characterization of fresh EMPIrE and SEMPER FIDELIS U(Mo)/Al fuel plates made with PVD-coated U(Mo) particles
No full textThe HERACLES group and the US-DOE work jointly to develop dispersed U(Mo)/Al as LEU fuel for conversion of high performance nuclear research reactors. Within this frame, two irradiation programs are in progress. In the first, EMPIrE, mini-plates are tested in the ATR reactor (USA) and in the second, SEMPER FIDELIS, full-size plates are irradiated in BR2 (Belgium). In both experiments, U(Mo)/Al plates with optimized microstructure are tested under high duty conditions. This paper focuses on analyses made at CEA Cadarache on seven fresh plates made of atomized particles, with or without Mo homogenization, and with ZrN coating. Five EMPIrE mini-plates and two SEMPER FIDELIS full-size plates were examined by optical microscopy (OM), scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). A particular attention is paid to the integrity of the ZrN coating (thickness, cracks…) and to the U(Mo) particles microstructure
A method for phenomenological and chemical kinetics study of autocatalytic reactive dissolution by optical microscopy. The case of uranium dioxide dissolution in nitric acid media
No full textDissolution is a milestone of the head-end of hydrometallurgical processes, as the stabilization rates of the chemical elements determine the process performance and hold-up. This study aims at better understanding the chemical and physico-chemical phenomena of uranium dioxide dissolution reactions in nitric acid media in the Purex process, which separates the reusable materials and the final wastes of the spent nuclear fuels. It has been documented that the attack of sintering-manufactured uranium dioxide solids occurs through preferential attack sites, which leads to the development of cracks in the solids. Optical microscopy observations show that in some cases, the development of these cracks leads to the solid cleavage. It is shown here that the dissolution of the detached fragments is much slower than the process of the complete cleavage of the solid, and occurs with no disturbing phenomena, like gas bubbling. This fact has motivated the measurement of dissolution kinetics using optical microscopy and image processing. By further discriminating between external resistance and chemical reaction, the “true” chemical kinetics of the reaction have been measured, and the highly autocatalytic nature of the reaction confirmed. Based on these results, the constants of the chemical reactions kinetic laws have also been evaluated
The characterization of radioactive waste: a critical review of techniques implemented or under development at CEA, France
No full textThis review paper describes the destructive and non-destructive measurements implemented or under development at CEA, in view to perform the most complete radioactive waste characterization. First, high-energy photon imaging (radiography, tomography) brings essential information on the waste packages, such as density, position and shape of the waste inside the container and in the possible binder, quality of coating and blocking matrices, presence of internal shields or structures, presence of cracks, voids, or other defects in the container or in the matrix, liquids or other forbidden materials, etc. Radiological assessment is then performed using a series of non-destructive techniques such as gamma-ray spectroscopy, which allows characterizing a wide range of radioactive and nuclear materials, passive neutron coincidence counting and active neutron interrogation with the differential die-away technique, or active photon interrogation with high-energy photons (photofission), to measure nuclear materials. Prompt gamma neutron activation analysis (PGNAA) can also be employed to detect toxic chemicals or elements which can greatly influence the above measurements, such as neutron moderators or absorbers. Digital auto-radiography can also be used to detect alpha and beta contaminated waste. These non-destructive assessments can be completed by gas measurements, to quantify the radioactive and radiolysis gas releases, and by destructive examinations such as coring homogeneous waste packages or cutting the heterogeneous ones, in view to perform visual examination and a series of physical, chemical, and radiochemical analyses on samples. These last allow for instance to check the mechanical and containment properties of the package envelop, or of the waste binder, to measure toxic chemicals, to assess the activity of long-lived radionuclides or pure beta emitters, to determine the isotopic composition of nuclear materials, etc
Checking, processing and verification of nuclear data covariances
No full textThe aim of this paper is to present the activities carried out by NEA Data Bank on checking, processing and verification of JEFF-3.3T4 covariances. A picture of the completeness and status of the JEFF-3.3T4 covariances is addressed. The verification of JEFF-3.3T4 covariances is performed with nuclear data sensitivity tool providing the keff uncertainty as a function of the contributing nuclide-reaction pairings including cross-reaction covariances. A total number of 4501 ICSBEP benchmarks is used in this analysis. This exercise is also extended to covariance libraries such as JENDL-4.0 updated files, ENDF/B-VII.1, SCALE-6.2rev8 and ENDF/B-VIII.0β5, allowing comparison of these results with both the experimental criticality benchmark and different methodologies of evaluation
Comments on the status of modern covariance data based on different fission and fusion reactor studies
No full textBoth the availability and the quality of covariance data improved over the last years and many recent cross-section evaluations, such as JENDL-4.0, ENDF/B-VII.1, JEFF-3.3, etc. include new covariance data compilations. However, several gaps and inconsistencies still persist. Although most modern nuclear data evaluations are based on similar (or even same) sets of experimental data, and the agreement in the results obtained using different cross-sections is reasonably good, larger discrepancies were observed among the corresponding covariance data. This suggests that the differences in the covariance matrix evaluations reflect more the differences in the (mathematical) approaches used and possibly in the interpretations of the experimental data, rather than the different nuclear experimental data used. Furthermore, “tuning” and adjustments are often used in the process of nuclear data evaluations. In principle, if adjustments or “tunings” are used in the evaluation of cross-section then the covariance matrices should reflect the cross-correlations introduced in this process. However, the presently available cross-section covariance matrices include practically no cross-material correlation terms, although some evidence indicate that tuning is present. Experience in using covariance matrices of different origin (such as JEFF, JENDL, ENDF, TENDL, SCALE, etc.) in sensitivity and uncertainty analysis of vast list of cases ranging from fission to fusion and from criticality, kinetics and shielding to adjustment applications are presented. The status of the available covariance and future needs in the areas including secondary angular and energy distributions is addressed
A stochastic method to propagate uncertainties along large cores deterministic calculations
No full textDeterministic uncertainty propagation methods are certainly powerful and time-sparing but their access to uncertainties related to the power map remains difficult due to a lack of numerical convergence. On the contrary, stochastic methods do not face such an issue and they enable a more rigorous access to uncertainty related to the PFNS. Our method combines an innovative transport calculation chain and a stochastic way of propagating uncertainties on nuclear data: first, our calculation scheme consists in the calculation of assembly self-shielded cross sections and a pin-by-pin flux calculation on the whole core. Validation was done and the required CPU time is suitable to allow numerous calculations. Then, we sample nuclear cross sections with consistent probability distribution functions with a correlated optimized Latin Hypercube Sampling. Finally, we deduce the power map uncertainties from the study of the output response functions. We performed our study on the system described in the framework of the OECD/NEA Expert Group in Uncertainty Analysis in Modelling. Results show the 238U inelastic scattering cross section, the 235U PFNS, the elastic scattering cross section of 1H and the 56Fe cross sections as major contributors to the total uncertainty on the power map: the power tilt between central and peripheral assemblies using COMAC-V2 covariance library amounts to 5.4% (1σ) (respectively 7.4% (1σ) using COMAC-V0)
