Search

According to the Swiss waste management policy, spent nuclear fuel is planned to be disposed in a deep underground repository. Prior to final disposal, spent nuclear fuel is stored at reactor sites and in a centralized dry storage facility. Since the operation dates of the final repository are unknown, extended periods of interim storage have to be considered. A research program to investigate the fuel rod integrity during long-term dry storage has recently been launched at Paul Scherrer Institute. In the context of the project, fuel rod performance simulations will be carried out with the code Falcon. Until now, an extensive literature survey concerning current trends in dry storage modelling has been written and first developments of Falcon’s capabilities towards dry storage modeling have been made. In this work, Falcon has been modified by implementing a long-term cladding creep model. The original and upgraded versions of Falcon have been used to simulate a demonstration case consisting of the base irradiation, wet storage, drying and dry storage. Results obtained with both versions have shown an important difference in the cladding hoop creep which justifies implementation of the new creep model.

This paper presents an improved uncertainty quantification technique for the validation of CASMO-5 on the spent fuel reactivity worth experiments of the LWR-PROTEUS Phase II program. In the program, eleven spent fuel samples manufactured from rods irradiated in a Swiss PWR (discharge burnups of 20–120 MWd/kg) were measured in the PROTEUS research reactor. In this work, both irradiation and reactivity worth measurement steps were modeled with CASMO-5 and the uncertainty on the code prediction were calculated using SHARK-X. For the first time in this work, we propagated the nuclear data uncertainties coming from cross-sections and fission yields for all samples in both fuel irradiations and the reactivity worth experiment models. We found that the fission yield and cross section uncertainties have similar contributions to the uncertainty of the reactivity worth prediction and that the reactivity worth probability distribution is non-normal because of the non-normal distribution of the perturbed fission yield data produced by the GEF code. Propagating input uncertainties through fuel irradiation as well as considering fission yield uncertainty led to larger computational uncertainty than previously reported. However, the observed trends with respect to exposure, fuel type, and moderating conditions are similar. In particular, a linear regression analysis showed that the predictions of reactivity with exposure by CASMO-5 are very accurate for various moderating conditions and for very high burnup. Finally, we estimated the effects of the irradiation history specifications on the relative reactivity worth bias and its uncertainty using two independent irradiation histories. While the bias uncertainty coming from uncertain cross sections was similar when considering either irradiation histories, we observed differences in the bias value. The irradiation history specification is a significant source of modeling bias and should be further investigated.

This paper presents the results of a series of studies, using detailed pin power distributions derived from Monte Carlo simulation results, to assess the effects of local dryout during normal operation on the cladding temperatures, heat fluxes and potentially on cladding oxidation for modern boiling water reactor (BWR) fuel. The development of a new three-dimensional LWR fuel thermo-mechanics solver, OpenFoam Fuel BEhAviour Tool (OFFBEAT), was initiated for these studies. Three-dimensional simulations of heat transfer in an axial section of a fuel pin have been conducted using OFFBEAT assuming different local dryout conditions. The results have shown that local power effects and, in particular, strong power gradients across the fuel pin can contribute significantly towards increased heat fluxes in the cladding, reducing the safety margin to dryout. Further, the studies have shown that, under the conservative assumption that the fuel pellet is not centred, but is instead touching the cladding on one side, the safety margin is further decreased. Investigations of possible neutronics feedback effects following dryout, taking into account the detailed intra-pin temperature distribution before and after dryout, have confirmed that there is no mitigating neutronics effect under dryout conditions. Finally, a basic study of the cladding oxidation under dryout conditions has shown that days to weeks of continuous dryout, or alternatively weeks of repeated dryout and rewetting would be required to obtain oxide layers thicknesses that would affect the mechanical integrity of the cladding. This work is an initial step towards developing a comprehensive multi-physics capability for simulations and studies of local three-dimensional coupled effects during dryout, and on this basis strengthen our understanding of the potential impact on the cladding integrity.

The present study provides a detailed analysis of the calculations of 50 isotopic compositions for the Post Irradiation Examination sample GU1 from the ARIANE program. Two types of approach are performed: a lattice (2 dimensional) and a full core (3 dimensional) calculations. In the case of lattice calculations, the effect oflocation is also studied. Different code versions (for CASMO5 and SNF) are used in order to contribute to the validation work for specific calculation schemes and to assess the impact of nuclear data libraries as well. The effects of nuclear data are also quantified for cross sections, spectra and fission yields, together with their partial contributions. It is concluded that the consideration of the effect of the mass balance, nuclear data uncertainties, as well as uncertainties from operational conditions and manufacturing tolerances help in obtaining consistency between calculated and measured isotopic concentrations. This study is then complemented with decay heat calculations for the assembly of interest, consistent with the PIE study. It indicates that, without considering geometrical deformation due to the long storage time, nuclear data are still the main source of uncertainties. Expanded uncertainties (including biases, experimental and calculation uncertainties) are finally proposed for both isotopic compositions and decay heat.

The main goal of this work is to perform pin-by-pin calculations of Swiss LWR fuel assemblies with neutron transport deterministic methods. At Paul Scherrer Institut (PSI), LWR calculations are performed with the core management system CMSYS, which is based on the Studsvik suite of codes. CMSYS includes models for all the Swiss reactors validated against a database of experimental information. Moreover, PSI has improved the pin power calculations by developing models of Swiss fuel assemblies for the Monte Carlo code MCNP, with the isotopic compositions obtained from the In-Core Fuel Management data of the Studsvik suite of codes, by using the SNF code. A step forward is to use a neutron code based on fast deterministic neutron transport methods. The method used in this work is based on a planar Method of Characteristics in which the axial coupling is solved by 1D SP3 method. The neutron code used is nTRACER. Thus, the methodology of this work develops nTRACER models of Swiss PWR fuel assemblies, in which the fuel of each pin and axial level is modelled with the isotopic composition obtained from SNF. This methodology was applied to 2D and 3D calculations of a Swiss PWR fuel assembly. However, this method has two main limitations. First, the cross sections libraries of nTRACER lack some of the isotopes obtained by SNF. Fortunately, this work proves that the missing isotopes do not have a strong effect on keff and the power distribution. Second, the 3D models require high computational memory resources, that is, more than 260 Gb. Thus, the nTRACER code was modified, so now it uses only 8 Gb, without any loss of accuracy. Finally, the keff and power results are compared with Monte Carlo calculations obtained by Serpent.

The effect of nuclear data (fission yields, cross sections and emitted spectra) is quantified for spent nuclear fuel assemblies from a realistic boiling water reactor operated over 25 cycles. Nominal calculations are performed with the CASMO5, SIMULATE-3 and SNF codes and the ENDF/B-VII.0 nuclear data library. The uncertainties are calculated with the same codes, using a Monte Carlo propagation method, and the ENDF/B-VII.1 covariance matrices. The conclusions are that (1) the nuclear data have a non-negligible impact for spent fuel quantities (e.g., decay heat, neutron emission or isotopic contents); (2) the importance of varying all data together is demonstrated, showing an under- or overestimation of uncertainties if fission yields are sampled separately from the other nuclear data; and finally (3) the importance of considering the full irradiation history (multi-cycle assembly life) is also demonstrated, showing also an under- or overestimation of uncertainties when performing the nuclear data sampling for a single reactor cycle.

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS2M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources.

This research aims at validating the SIMULATE-3K code and complementing the results with the quantification of nuclear data uncertainties against the Special Power Excursion Reactor Test III (SPERT-III) experiments. To that aim, the SHARK-X methodology, under development at PSI, for the propagation of nuclear data uncertainties in CASMO5 2-D lattice calculations to 3-D core transient simulations is applied for the analysis of a SPERT-III super-prompt critical test conducted at cold startup conditions. Concerning transient results, both total power and reactivity show a good agreement with the measurements at the initial phase of power excursion, while a slight discrepancy is obtained at the final phase of the transient. The estimated uncertainties regarding both steady-state parameters such as k-eff and static reactivity worth, as well as dynamical quantities such as power pulse width and enthalpies are presented. Results show non-negligible sensitivity upon the employed nuclear data library. The uncertainty quantification results show relatively small biases for k-eff and reactivity. The uncertainty in peak power is around 3%, while it is negligible for the time to peak power and the pulse width. The time evolution of the standard deviation and skewness of the total power showed special shapes with relatively high maximum values. In addition, the uncertainty due to nuclear data in the two important safety parameters, i.e. maximum nodal fuel temperature and enthalpy reaches maximum value around 2% and 10%, respectively.

In view of massive operations with LWR spent nuclear fuel for the forthcoming final geological disposal, as expected, e.g., in Switzerland, needs are growing for exhaustive and enhanced criticality safety assessments to allow optimization of the spent nuclear fuel transportation, storage and disposal strategies. At the Paul Scherrer Institute (PSI) the current base methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practices, on establishing an upper subcritical limit (USL) for the effective neutron multiplication factor ( k eff ) , based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark k eff values. Recently, new calculation capabilities have been developed at PSI in particular to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries. The computation tool NUSS for random sampling the ACE formatted ND libraries was developed for that purpose. Thereby, this paper concerns how these new calculation capacities could improve the present PSI CSE methodology and the prospects for its upgrading towards burnup credit applications are consequently discussed.

The development of analysis models for the Swiss reactors is a key objective of the STARS project at the Paul Scherrer Institut (PSI). Within this context there is a need for the development of computational fluid dynamics (CFD) models of the Swiss reactors in support of future high fidelity investigations of steady-state and transient scenarios. This article presents initial results for the CFD analysis of a Siemens KWU PWR with a focus on the flow behaviour and heat transfer in the gap between the core shroud and core barrel. Temperatures and densities in this region of the reactor are important, for example, for accurate estimations of fast neutron fluence and activation in the steel structures of the core shroud, core barrel and reactor pressure vessel. The flow behaviour in this region may also be relevant for better understanding of ex-core detector responses. The flow conditions in the core bypass region were found to be in the transition-to-turbulence regime, with vortex shedding taking place downstream of the core formers as a result of flow instabilities. The non-stationary nature of the flow presented a challenge in terms of obtaining a solution within a reasonable time period. Two approaches were proposed to address this challenge: time-averaging of the flow-field information before solving the conjugate heat transfer problem; time-averaging of surface heat fluxes in order to derive detailed surface heat transfer coefficients. Both approaches yielded similar results with similar computational effort. Several characteristics and features of the core bypass flow are discussed. Updated Monte Carlo simulation results show that the influence of the core bypass temperatures on the neutron fluence predictions is non-negligible. This highlights the importance of including accurate bypass temperatures in future Monte Carlo simulations focused on ex-core regions.

Nonlinear and skewed distributions for safety-related parameters of a realistic Pressurized Water Reactor (PWR) are observed due to the current limited knowledge of the inelastic scattering of neutrons on 238 U. The current knowledge of this cross section, about 20% uncertainty, does not lead to normally distributed core parameters, thus increasing the probability of rare events. In this paper, it is demonstrated how the simulations of 14 years of PWR operation with realistic history and fuel loading pattern induce non-Normal distributions for 3-dimensional peaking quantities and why a reduction of the 238 U inelastic cross section uncertainty by a factor 2 is desirable.

At the Paul Scherrer Institut, the development of the SHARK-X methodology to propagate nuclear data uncertainties in CASMO-5 2-D assembly depletion calculations started few years ago. Before that, the CMSYS platform had been established to serve as framework for the continuous development and validation of core models based on CASMO/SIMULATE for all the Swiss operating reactors and cycles. Recently, a first attempt to integrate SHARK-X within CMSYS was initiated in order to complement the core analyses with nuclear data uncertainty quantifications. A proof-of-principle is presented in this paper on the basis of real operated UO2 and MOX cycles of a Swiss PWR plant. In this framework, the nuclear data uncertainties related to cross-sections, neutron multiplicity and fission spectrum, provided via SCALE-6.2b4 based variance-covariance matrices, are propagated using the stochastic sampling method in all types of core calculations, including cycle depletion, validation against flux measurements and safety calculations typically required for reload licensing. For each calculation type and core design, the statistical convergence of relevant quantities of interest is studied in order to determine the required optimal number of samples. Thereafter, the uncertainties of selected quantities of interest are presented and compared between both core designs. For instance, the estimated uncertainties are shown for critical boron concentration as well as 3-D power/burnup distributions for cycle depletion analyses. Regarding the validation of the core models against in-core detector measurements, the uncertainties in biases between calculated and experimental 3-D reaction rates are also evaluated. Finally, for calculations of safety parameters related to core design and licensing, uncertainties in reactivity coefficients, control rod worths and kinetic parameters are presented.

Fuel assembly bowing effects on isotopic concentrations during depletion are investigated for a range of full-scale fuel assembly bow models for Simplified Pressurized Water Reactor (PWR) assemblies and full cores. First, two-dimensional (2D) 3 × 3 assembly arrays with a central fuel assembly displaced are simulated using both CASMO-5 and SERPENT-2. Investigations are carried out on burn-up evolutions and horizontal re-distributions of nuclide concentrations resulting from assembly bowing effects. Longitudinal distributions of the fuel content changes are also illustrated by simulating a three-dimensional (3D) assembly bowed in C-shape with SERPENT-2. Progressive deformations, varied bowing magnitudes and orientations are considered as influential factors for studying the impact of bowing. Finally, by implementing a measured bowing map from an operating reactor, simulations for 2D PWR full cores are performed with CASMO-5 to reveal the bowing impacts in terms of assembly locations.

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

This work is a contribution towards the development of computational methods that allow investigation and prediction of reactor noise response due to in-core perturbations such as vibrations of reactor internals and fluctuations in thermal-hydraulic parameters. Further improvements are made to the modelling approach to allow for vibrations of fuel assemblies at higher vibrational modes. Modelling of such realistic noise sources, and assessment of the induced noise response are presented in this paper. Results show that the simulation of individual noise sources and their combinations affect the neutron noise phenomenology. The neutron noise induced radially and, in particular, axially, reflects the imposed vibrational patterns associated to the higher modes of the fuel assemblies. The results provide confidence in the application of the developed methodology based on commercial time-domain codes for modelling of realistic noise scenarios and numerical noise analyses of the PWR cores.

With the goal to enhance the capability to perform best-estimate simulations of Light Water Reactors (LWRs) transients, with strong coupling between core neutronics and plant thermal-hydraulic, a coupling between TRACE and SIMULATE-3K (TS3K) was developed in collaboration between PSI and Studsvik for analyses involving interactions between system and core. In order to verify the coupling scheme and the coupled code capabilities to simulate complex transients, the OECD/NEA Oskarshmn-2 (O-2) Stability benchmark was modeled with the coupled code TS3K. The main goal of this paper is to present TS3K analyses of the Oskarshamn-2 stability event, noting that this constitutes the first reported assessment of this code system for a BWR stability problem. A systematic analysis is carried out using different time-space discretization schemes in order to identify an optimized methodology to simulate correctly the O-2 stability event. In this context, the TS3K results are compared to the available benchmark data both for steady-state and transient conditions. The results show that using a refined model in space and time, the TS3K model can successfully capture the entire behavior of the transient qualitatively, i.e. onset of the instability with growing oscillation amplitudes, as well as quantitatively, i.e. Decay Ratio and resonance frequency. In addition, the results are compared also to those obtained using TRACE/PARCS and S3K stand-alone, which allows a systematic comparison between different codes.

The paper deals with the development of a TRACE/PARCS plenum to plenum model for the Swiss Boiling Water Reactor (BWR) Leibstadt (KKL), for cycle 19. It summarizes a verification of the model against an existing SIMULATE-3 model and a validation against Travelling Incore Probes (TIP) at various points of the cycle 19. First, the TRACE/PARCS agreement with SIMULATE-3, for a simplified BWR reactor is analysed. The discrepancies are less than 300 pcm in terms of keff, 6.5% axially and 2% radially in terms of nodal power prediction across the reactor cycle. The discrepancies are due to the mapping of the cross sections when considering the complex default branch/history structure of CASMO-4; and to the modelling of the vessel in TRACE, the solution of the heat conduction in the fuel and the evaluation of the bypass flow. When the KKL cycle 19 model is considered, the TRACE/PARCS agreement with SIMULATE-3 is satisfactory, in terms of keff with discrepancies less than 700 pcm across the reactor cycle; and in terms of Relative Power Fraction (RPF) as well, with discrepancies less than 9% axially and 2% radially, across the reactor cycle. Finally, when comparing to TIP measurements, the TRACE/PARCS methodology and models showed some similar discrepancies than SIMULATE-3, as long as the detector model is not used. When using its detector model to predict the TIP response, TRACE/PARCS shows larger discrepancies with TIP measurements.

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, k e f f , of the canister would comply with the imposed criticality safety criterion.

International audience; This paper presents an application of the Backward-Forward Monte Carlo (BFMC) method using measured critical boron concentrations for a specific PWR cycle. The considered prior nuclear data are the fission cross sections, $\overline{\nu}_{p}$ for$^{235}$U and$^{239}$Pu and the capture cross section of$^{238}$U. The posterior nuclear data exhibit cross-isotope correlations, moderate changes for the average quantities and reduced uncertainties. This work is the first one considering the BFMC method and an integral system mostly sensitive to thermal neutrons. It contributes to show the impact of integral experimental data for the evaluation of nuclear data and their covariance matrices, leading to cross-isotope correlations and a nuclear data uncertainty reduction.

International audience; In this paper, we present three Monte Carlo methods to include integral benchmark information into the nuclear data evaluation procedure: BMC, BFMC and Mocaba. They allow to provide posterior nuclear data and their covariance information in a Bayesian sense. Different examples will be presented, based on 14 integral quantities with fast neutron spectra (k$_{eff}$ and spectral indices). Updated nuclear data for$^{235}$U,$^{238}$U and$^{239}$Pu are considered and the posterior nuclear data are tested with MCNP simulations. One of the noticeable outcomes is the reduction of uncertainties for integral quantities, obtained from the reduction of the nuclear data uncertainties and from the rise of correlations between cross sections of different isotopes. Finally, the posterior nuclear data are tested on an independent set of benchmarks, showing the limit of the adjustment methods and the necessity for selecting well representative systems.

Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

The CASMO/SIMULATE/MCNP/FISPACT-II calculation route has been established at the Paul Scherrer Institute (PSI) for reactor dosimetry and activation studies. Furthermore, the in-house tool NUSS is in use at PSI for nuclear data (ND) related uncertainties quantifications with Monte Carlo neutron transport calculations. The use of randomly sampled ACE-formatted ND files not only allows propagation of the ND uncertainties, but also can serve for assessing the applicability of different types of experimental data for validation of calculation predictions of parameters of interest. In the present work an application of the PSI calculation scheme for analysis of activation reaction rates and the fast neutron fluence (FNF), throughout the Swiss pressurised water reactor (PWR) and simplified containment building models, is demonstrated. As particular examples of potentially available experimental data, two kinds of the neutron flux monitors are considered: a) the reactor pressure vessel scraping samples and detectors placed in the dosimetry channels, mounted at the core barrel and designed for validation of FNF calculations, and b) the ex-vessel dosimeters, specifically used by the Swiss waste management organisation (NAGRA) for validation of bio-shield activation predictions. The calculations were done with the ENDF/B-VII.1 library. The obtained results demonstrate importance of the ND uncertainties for the dosimetry evaluations. The assessment of the applicability of the selected experimental information for validation of the bio-shield irradiation calculations was done based on evaluation of the ND-related Pearson correlation coefficients.

This paper describes the effect of input uncertainties on a set of integral parameters (kinf, nuclide compositions) associated with the validation of CASMO-5 against PIE data. The nuclear data under consideration are the cross-sections, fission spectrum and neutron multiplicities and fission yields. Various sources of covariance information are considered, novel ones (ENDFB-VIII.0, JEFF-3.3) as well as more widely distributed ones (JENDL-4.0, ENDF/B-VII.1, Scale 6.1 and Scale 6.2). All possible nuclide reaction pairs (cross sections, fission spectrum and averaged number of neutron per fission) have been perturbed, e.g. all isotopes available in both the respective covariance libraries and the CASMO-5 library. The evolution of the uncertainty estimates with exposure is complemented with sensitivity analysis to determine the main contributors to the uncertainty. The Pearson coefficient defined between the model output and a given input is used in this work to assess the part of the variance in the model output coming from the considered input uncertainty. It is a very promising measure of sensitivity as it is computationally cheap even though it assumes linearity of the output with respect to input perturbations. The evolution of the uncertainty with exposure, both in terms of trends and magnitude are however very different. Sensitivity analysis allows determining why the trends and magnitudes are different.

The effects of the nuclear data uncertainties are quantified for specific core parameters from a real Boiling Water Reactor operated over 25 consecutive cycles. Nominal calculations are performed with the CASMO5 and SIMULATE-3 codes and the ENDF/B-VII.0 nuclear data libraries. The uncertainties are calculated using a Monte Carlo propagation method, and the ENDF/B-VII.1 covariance matrices. The conclusions are that the nuclear data uncertainties (1) have a limited impact on core parameters (axial power, linear heat generation rate, and comparison with in-core measurements), but (2) are larger for keff compared to the expectations from reactor experience. Additionally, the importance of simultaneously varying all input data in order to accurately calculate uncertainties is demonstrated.

In this study, an optimization of the canister loading for used nuclear fuels is performed, solely based on their decay heat values. By taking into account the individual irradiation history for each fuel assembly, as well as with a number of assumptions on the number of used fuels to be disposed, this study indicates that a canister filling fraction of 94% is achievable for both PWR and BWR fuel, thus lowering the number of required canisters compared to a previous study. This work emphasized the need for a precise fuel accounting, including the reprocessed ones, as well as considering the individual irradiation history for all fuel assemblies. Finally, it is demonstrated that a higher filling fraction can be reached by delaying the transfer of used nuclear fuels from storage casks to disposal canisters.

In this paper the isotopic content of the U1 sample is re-analyzed taking into account the full irradiation environment with a 3-dimensional core simulator. It is found that by accounting for the neighboring assemblies as well as the various axial segments, the agreement between the measured and calculated isotopic concentrations is greatly improved. Additionally, the sample burnup is impacted (reduced) compared to the previous studies. Varying nuclear data and code versions, an estimation of uncertainties is proposed. Finally, remarks about the U1 relevance for the estimation of the biases for assembly averaged quantities are presented.

In this paper, the possibilities and advantages of loading mixed used nuclear fuel (UNF) assemblies into canisters for final repository are presented from a criticality aspect within the burnup credit approach (BUC). UNF coming from a Swiss reactor are taken into account at the pin-by-pin level and two canister loading patterns are studied: mixed and homogeneous (i.e. mixing UNF with various burnup values and enrichments, or not). The advantages of the mixed cases are presented in terms of criticality, together with the impact of assembly axial and radial rotations. The main outcome of the mixed loading is that a variety of UNF assemblies can be safely loaded together, including those which were previously not allowed in the homogeneous case.

The OECD/NEA recently launched an international benchmark on a combined feedwater transient and stability event that occurred at the Swedish Oskarshamn-2 (O2) nuclear power plant (NPP). The primary benchmark objective is to assess advances in coupled neutronic/thermal–hydraulic codes for simulations of challenging transients including the appearance of unstable power oscillations. The Paul Scherrer Institut (PSI) is participating in this benchmark in order to enlarge the validation basis of its advanced stability analysis methodology currently under development for Swiss B WRs and based on the state-of-the-art SIMULATE-3K (S3K) code. This paper presents the development, optimization and validation of a S3K model for the O2 benchmark, including the analysis of the entire February 1999 event as well as additional stability tests considered as blind cases. With the optimized model, the S3K solution is compared to available benchmark event data both for at steady-state and transient conditions. For the latter, the qualitative as well as quantitative behavior of S3K results is compared and discussed in relation to the experimental observations. The modeling aspects found in this context to mostly affect the S3K ability to reproduce the event are also presented. As a result, the S3K model indicates that if the reactor scram had not been introduced, the observed diverging oscillations would reach a maximum amplitude before decaying back into a stable state. However, it is also found that the core could instead have evolved into limit-cycle oscillations if a stabilization of the feedwater flow and temperature had occurred just before the scram signal. Finally, the S3K model optimized for the event analysis is subsequently validated in a blind mode approach against several stability tests conducted before and after the event. Considering that the uncertainties in measured decay ratio values are not known and could be relatively large, the obtained results can on a qualitative basis be considered as satisfactory.

The present work proposes a Bayesian method to combine theoretical fission yields with a set of reference data. These two sources of information are merged using a Monte Carlo process, and leads to a so-called Bayesian Monte Carlo update. Examples are presented for the independent fission yields of four major actinides, using the GEF code as a source of theoretical calculations and an evaluated library of fission yields for the reference data. The impact of the updated fission yields and their covariances is shown for two distinct applications: a UO 2 pincell with burn-up up to 40 GWD/tHM and decay heat calculations of a thermal neutron pulse on 235 U and 239 Pu.

NUSS-RF is a tool for nuclear data uncertainty propagation through neutronics calculations with continuous-energy Monte Carlo codes and ACE-formatted nuclear data libraries. Many existing codes, in...

In this paper we are proposing a re-evaluation of the thermal-neutron induced capture cross section of 147Nd. A unique measurement exists from which this cross section was calculated in 1974. This original calculation is based on an assumed value for a specific gamma-ray fraction (called F2), taken from the neighboring nucleus 145Nd. With the availability of reaction codes such as TALYS, such fraction can nowadays be calculated using specific reaction models and parameters. The new value of F2 indicates a decrease of the thermal cross section by 45%, leading to 243 barns, instead of the 440 barns previously reported. This new cross section impacts the calculation of the number density for the well-known burn-up indicator 148Nd, but as shown, the change is close to the usual experimental uncertainties for the 148Nd number densities, thus having a limited impact on burn-up calculation.

This work is presenting a comparison of results for different methods of uncertainty propagation due to nuclear data for 330 criticality-safety benchmarks. Covariance information is propagated to keff using either Monte Carlo methods (NUSS: based on existing nuclear data covariances, and TMC: based on reaction model parameters) or sensitivity calculations from MCNP6 coupled with nuclear data covariances. We are showing that all three methods are globally equivalent for criticality calculations considering the two first moments of a distribution (average and standard deviation), but the Monte Carlo methods lead to actual probability distributions, where the third moment (skewness) should not be ignored for safety assessments.

Nuclear reactors are stochastic systems, in which several parameters fluctuate even during steady-state conditions. In addition, structural components like fuel assemblies, vibrate due to strong hydraulic forces. This stochasticity is the cause of the neutron population fluctuating behavior; a phenomenon referred as neutron noise. This paper systematically analyzes the capabilities of the Studsvik’s codes to model the fuel assembly vibration and to assess the neutron noise response. To this aim, the CASMO-5 code is utilized for generating cross sections by parametrizing the water-gap thickness of fuel segments. The fuel assembly vibration modelling is approximated by the dynamic modification of the water-gap widths between adjacent assemblies using the transient code SIMULATE-3K. The cross sections dependency on the water-gap width is analyzed and successfully compared against Serpent-2 reference results. Last, the utilized codes capability is verified qualitatively or/and quantitatively through a series of cases, at both lattice and nodal level.

This paper presents calculation studies towards validation of a methodology for estimations of the mass of tramp uranium in a boiling water reactor (BWR) core from water chemistry measurements. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of 235U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle. Thus, the use of the so-called “Pu-based model” for the determination of the tramp uranium mass (this means in particular using the yields for 239Pu fission) appears justified in the cases considered. On that basis, replacing the older “U-based model” by the “Pu-based model” is recommended.

In this work, a method is proposed for combining differential and integral benchmark experimental data within a Bayesian framework for nuclear data adjustments and multi-level uncertainty propagation, using the Total Monte Carlo method. First, input parameters to basic nuclear physics models implemented within the TALYS code, were randomly varied to produce a large set of random nuclear data files. Next, a probabilistic data assimilation was carried out by computing the likelihood function for each random nuclear data file based first on only differential experimental data and then on integral benchmark data. The individual likelihood functions from the two updates were then combined into a global likelihood function. The proposed method was applied for the adjustment of n+208Pb in the fast energy region below 20 MeV. The adjusted file was compared with available experimental data as well as evaluations from the major nuclear data libraries and found to compare favourably.

Shark-X is a set of Perl-based tools build around the deterministic neutron transport code CASMO-5, used to perform uncertainty quantification for lattice calculations at the Paul Scherrer Institut (PSI). While perturbing an input parameter, e.g. a cross section in a given energy range, the resulting perturbation to the effective cross section, e.g. multigroup cross section, can be expressed as the sum of an explicit and implicit component. The implicit term stems from changes in the neutron flux resulting from the cross section perturbation. This effect is neglected when using the standard perturbation theory methodology, even though its importance for uncertainty quantification is reported in the literature. The objectives of the paper are, in view of further enhancing Shark-X, to review a set of existing methods and to propose novel methodologies in order to accurately take the implicit effect into account in the resonance shielding treatment of CASMO-5; as well as to assess the importance of the new tool for typical LWR applications. Amongst the methods considered, those based on an equivalence in dilution allow perturbing the group-averaged cross sections without having to solve the slowing down problem in conjunction with the perturbed point-wise cross sections. However, such methods are found to show only limited improvements as compared to the standard perturbed self-shielded cross sections determined neglecting the implicit effect. As a result, a simpler approach based on the pre-determination of perturbation factors is adopted and implemented into Shark-X and CASMO-5. The new implementation is successfully tested and verified by comparing k-inf sensitivity coefficients computed with Shark-X with those from a Monte-Carlo reference solution, and this for a range of simple homogeneous mixtures as well as for LWR pin cells and lattices. Finally, the paper demonstrates through simple considerations, that the magnitude of the implicit effect is related to the energy group structure in the resolved resonance region, which might be specific to the code used to compute the sensitivity coefficients. The finer is the energy mesh; the smaller is also the impact of the implicit effect, and as such the necessity for a specific treatment.

A lot of research work has been carried out in fine tuning model parameters to reproduce experimental data for neutron induced reactions. This however is not the case for proton induced reactions where large deviations still exist between model calculations and experiments for some cross sections. In this work, we present a method for searching both the model and model parameter space in order to identify the 'best' nuclear reaction models with their parameter sets that reproduces carefully selected experimental data. Three sets of experimental data from EXFOR are used in this work: (1) cross sections of the target nucleus (2) cross sections of the residual nuclei and (3) angular distributions. Selected models and their parameters were varied simultaneously to produce a large set of random nuclear data files. The goodness of fit between our adjustments and experimental data was achieved by computing a global reduced chi square which took into consideration the above listed experimental data. The method has been applied for the adjustment of proton induced reactions on Co-59 between 1 to 100 MeV. The adjusted files obtained are compared with available experimental data and evaluations from other nuclear data libraries., Comment: 2019 International Conference On Nuclear Data for Science and Technology

This paper presents a Bayesian approach based on integral experiments to create correlations between different isotopes which do not appear with differential data. A simple Bayesian set of equations is presented with random nuclear data, similarly to the usual methods applied with differential data. As a consequence, updated nuclear data (cross sections, [see formula in PDF], fission neutron spectra and covariance matrices) are obtained, leading to better integral results. An example for 235U and 238U is proposed taking into account the Bigten criticality benchmark.

The JEFF library does not contain fission yield covariances, but simply best estimates and uncertainties. This situation is not unique as all libraries are facing this deficiency, firstly due to the lack of a defined format. An alternative approach is to provide a set of random fission yields, themselves reflecting covariance information. In this work, these random files are obtained combining the information from the JEFF library (fission yields and uncertainties) and the theoretical knowledge from the GEF code. Examples of this method are presented for the main actinides together with their impacts on simple burn-up and decay heat calculations.

International audience; The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞ , macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

This paper presents a Bayesian approach based on integral experiments to create correlations which do not appear with differential data. Some quantities such as the fission cross section (σ ), neutron multiplicity (ν p ), neutron spectra (χ ), etc. are usually neither modeled together nor measured in coincidence, thus there is no correlation matrices in evaluated nuclear data libraries. One can nevertheless use the information from integral experiments such as fast criticality-safety benchmarks to correlate such quantities for possible inclusion in nuclear data libraries. A simple Bayesian set of equations is presented with random nuclear data, similarly to the usual methods applied with differential data. An example for 239 Pu is proposed.

The radiative neutron capture rates for isotopes of astrophysical interest are commonly calculated on the basis of the statistical Hauser Feshbach (HF) reaction model, leading to smooth and monotonically varying temperature-dependent Maxwellian-averaged cross sections (MACS). The HF approximation is known to be valid if the number of resonances in the compound system is relatively high. However, such a condition is hardly fulfilled for keV neutrons captured on light or exotic neutron-rich nuclei. For this reason, a different procedure is proposed here, based on the generation of statistical resonances. This novel technique, called the “High Fidelity Resonance” (HFR) method is shown to provide similar results as the HF approach for nuclei with a high level density but to deviate and be more realistic than HF predictions for light and neutron-rich nuclei or at relatively low sub-keV energies. The MACS derived with the HFR method are systematically compared with the traditional HF calculations for some 3300 neutron-rich nuclei and shown to give rise to significantly larger predictions with respect to the HF approach at energies of astrophysical relevance. For this reason, the HF approach should not be applied to light or neutron-rich nuclei. The Doppler broadening of the generated resonances is also studied and found to have a negligible impact on the calculated MACS., SCOPUS: ar.j, info:eu-repo/semantics/published

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

This paper presents results from Paul Scherrer Institut (PSI) on the three phases of the Peach Bottom Boiling Water Reactor Turbine Trip Benchmark. The first part of the paper presents the PSI analysis using RETRAN-3D of Phase 1, where the system pressure is predicted based on a predefined core power distribution. These calculations elucidate the importance of accurate modeling of the steam separator region and of nonequilibrium effects. In the second part, the CORETRAN results of Phase 2 are summarized and the core 3-D response to the pressurization transient prior to SCRAM is discussed. The CORETRAN results show a slight axial flux redistribution toward the top of the core, while radially a flux redistribution is observed toward core regions with assemblies that are initially moderately voided and where the axial power shape is increasingly top-peaked. The impact of the control rod configuration as well as the assembly coolant inventory dynamics on the 3-D flux redistribution is also discussed. The third part presents results of the Phase 3 calculation using RETRAN-3D, which is a culmination of the analytical work of Phases 1 and 2.

Causality analysis is a substantial tool for identifying cause-and-effect links between different components of a system and has been extensively used in various areas of science such as neuroscience, climatology, and econometrics. This analysis is carried out in terms of the renormalized partial directed coherence and the directed transfer function connectivity measures. Applying such analysis in the nuclear reactor field is of paramount importance since it can help in inferring cause-and-effect relationships between highly coupled processes, and consequently, it can assist on the safe and reliable operation of a nuclear power plant during the occurrence of possible disturbances or malfunctions. The effectiveness of the connectivity analysis is demonstrated through several simulated and measured test cases. Results show that the connectivity analysis is able to identify accurately the importance and central role of the activation signal when it is applied on a simple analytical model and a simulated nuclear reactor system. In addition, the application on more realistic and complex measured data sets of a Swiss boiling water reactor illustrates the capability of this analysis to indicate possible causes behind the observed anomalies or trends observed at certain conditions and, more importantly, allows a better understanding of the underlying interactions among different neutronic and thermal-hydraulic processes. Published under license by AIP Publishing.

At PSI, the reference approach for core safety analyses of all the Swiss operating reactors is based on a deterministic 2-D lattice depletion/3-D two-group nodal diffusion code platform. As a complement to this, a cycle check-up (CHUP) methodology is being established where, for instance, the continuous energy Monte Carlo transport solver MCNP can be plugged-in at any selected operating point of a given reactor/cycle to conduct in-depth examinations of the 3-D flux/power distributions. After briefly outlining the concepts of the CHUP methodology, its application with MCNP to obtain additional 3-D pin power information than provided by current state-of-the-art nodal core simulators is illustrated. More specifically, the pin power and local intra-pin power distributions in operated assemblies, representative of modern BWR fuel designs, are examined. Focus is given to fuel rods characterised by strongly heterogeneous spectral environments. The CHUP/MCNP methodology is applied to evaluate and compare pin and intra-pin power behaviour between different cycles and/or burnup steps, for normal conditions as well as when assuming local perturbations of the coolant and/or mechanical structures. Although only pin-average nuclide compositions are at this stage used, it is shown that already in this case, strong to very-strong azimuthal gradients take place both radially as well as axially and that these can become substantially more pronounced by very small local perturbations, in particular from geometrical changes related to assembly bowing.

The current standard for computational neutronic analysis of nuclear power plants (NPPs) is the so-called conventional approach, which relies on few-group, coarse-mesh diffusion calculations. The r...