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The aim of this paper is to assess the reliability and accuracy of the PSI standard method, used in many previous works, for the quantification of ND uncertainties in the SPERT-III RIA transient, by quantifying the discrepancy between the actual inserted reactivity and the original static reactivity worth and their associated uncertainties. The assessment has shown that the inherent S3K neutron source renormalization scheme, introduced before starting the transient, alters the original static reactivity worth of the transient CR and reduces the associated uncertainty due to the ND perturbation. In order to overcome these limitations, two additional methods have been developed based on CR adjustment. The comparative study performed between the three methods has showed clearly the high sensitivity of the obtained results to the selected approach and pointed out the importance of using the right procedure in order to simulate correctly the effect of ND uncertainties on the overall parameters in a RIA transient. This study has proven that the approach that allows matching the original static reactivity worth and starting the transient from criticality is the most reliable method since it conservatively preserves the effect of the ND uncertainties on the inserted reactivity during a RIA transient.

Best estimate plus uncertainty for the safety assessment of nuclear power plant transient requires, among others, estimating the probability density function (PDF) of physical model parameters in t...

In this work, we explore the use of an iterative Bayesian Monte Carlo (iBMC) method for nuclear data evaluation within a TALYS Evaluated Nuclear Data Library (TENDL) framework. The goal is to probe the model and parameter space of the TALYS code system to find the optimal model and parameter sets that reproduces selected experimental data. The method involves the simultaneous variation of many nuclear reaction models as well as their parameters included in the TALYS code. The ‘best’ model set with its parameter set was obtained by comparing model calculations with selected experimental data. Three experimental data types were used: (1) reaction cross sections, (2) residual production cross sections, and (3) the elastic angular distributions. To improve our fit to experimental data, we update our ‘best’ parameter set—the file that maximizes the likelihood function—in an iterative fashion. Convergence was determined by monitoring the evolution of the maximum likelihood estimate (MLE) values and was considered reached when the relative change in the MLE for the last two iterations was within 5%. Once the final ‘best’ file is identified, we infer parameter uncertainties and covariance information to this file by varying model parameters around this file. In this way, we ensured that the parameter distributions are centered on our evaluation. The proposed method was applied to the evaluation of p+$$^{59}$$ 59 Co between 1 and 100 MeV. Finally, the adjusted files were compared with experimental data from the EXFOR database as well as with evaluations from the TENDL-2019, JENDL/He-2007 and JENDL-4.0/HE nuclear data libraries.

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.

In this paper, the impact of the thermal scattering data for H in H20 is estimated on criticality benchmarks, based on the variations of the CAB model parameters. The Total Monte Carlo method for uncertainty propagation is applied for 63 keff criticality cases, sensitive to H in H20. It is found that their impact is of a few tenth of pcm, up to 300 pcm maximum, and showing highly non-linear distributions. In a second step, an adjustment is proposed for these thermal scattering data, leading to a better agreement between calculated and experimental keff values, following an increase of scattering contribution. This work falls into the global approach of combining advanced theoretical modelling of nuclear data, followed by possible adjustment in order to improve the performances of a nuclear data library.

Recent advances in the development of methods and tools for enhanced accuracy reactor simulations have allowed moving towards time-dependent neutron transport solution without spatial cross-sections homogenization approximation. Although the full core reactor dynamic problems are still too much challenging for the modern codes, reduced order problems can be solved already today. Examples of the codes allowing such advanced transient calculations are, among several others, the SUHAM-TD and nTracer codes, which are discussed in this work. In order to perform an intermediate verification of the SUHAM-TD code, which is under development at the Kurchatov Institute, and to perform an independent verification of the nTracer code of the Seoul National University by its user team at the Paul Scherrer Institute, performances of these two codes have been assessed using the ongoing OECD/NEA C5G7-TD Phase-I benchmark exercises. In particular, two cases of positive reactivity insertion were selected for detailed comparison, being the most relevant from the point of view of future practical applications.

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.

This paper illustrates the capability of a global sensitivity analysis (GSA) framework applied to the TRACE thermal-hydraulics (TH) system code in the context of selected flooding experimen...

Assessing neutron emission of LWR spent fuel is necessary for the back-end of the fuel cycle, such as the dimensioning of transport and storage casks of spent fuel. Although core and depletion codes can calculate the isotopic composition of the discharged fuel and therefore infer its neutron source, accurate measured neutron emission values remain rare mainly because of the difficulty to prepare, handle and characterize spent fuel. Measured neutron emission values are, however, extremely relevant to code validation, as neutrons emitted by LWR spent fuel mainly originates from spontaneous fissions of minor actinides (e.g., 242Cm, 244Cm and 252Cf) that are produced only after a large number of neutron captures in the reactor core. This paper reports on neutron emission measurements of selected LWR-PROTEUS spent fuel samples and their comparisons with a core and depletion calculation chains based on CASMO-5, SIMULATE-3 and the SNF codes. The measured LWR-PROTEUS samples are comprised of 11 samples irradiated in a Swiss PWR. The samples are UO2 or MOX and have discharge burn-ups ranging from 20 to 120 GWd/t. We measured the 40-cm long samples in a hot-cell of the Paul Scherrer Institut using a measurement station made of polyethylene and a BF3 detector. We repeated the measurements several times and in different conditions to ensure the accuracy and reproducibility of the results. We derived ratios of neutron rates emitted by the different samples and absolute neutron emission rates by comparison with a reference 252Cf source, which we re-calibrated for this exercise. The experimental uncertainty (1σ) on the absolute neutron emission varies from 3% to 4%. We compared a subset of the measured values to the calculation predictions and showed an agreement within less than 7% for all but one sample.

The impact of the H in H2O thermal scattering data are calculated for burnup quantities, considering models of a UO2 pincell with DRAGON and SERPENT. The Total Monte Carlo method is applied, where the CAB model parameters are randomly varied to produce sampled (random) LEAPR input files for NJOY. A large number of burnup calculations is then performed, based on the random thermal scattering data. It is found that the impact on k∞ is relatively small (less than 35 pcm), as for nuclide inventory (less than 1% at 50 MWd/kgU) and for decay heat (less than 0.4%). It is also observed that the calculated probability density functions indicate strong non-linear effects.

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 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.

Some of the KWU pre-KONVOI PWRs operating across Europe saw a systematic increase in the neutron noise levels over several cycles in the last decade, and subsequently, core internals’ movements, especially vibrations offuel assemblies with specific designs were identified as one of the plausible causes. Therefore, it is important to develop computational methods that can allow to investigate and predict the reactor noise response to fuel assemblies vibrations. To this aim, the 3D nodal reactor dynamics code SIMULATE- 3K is used at PSI with a special module called the ‘assembly vibration model’ that imitates time-dependent motions of fuel assemblies by dynamically modifying the water-gaps surrounding the laterally moving fuel assemblies. The varying water-gaps are represented by the variation in the corresponding two-group macroscopic cross sections generated using the lattice code CASMO-5 in 2D. The studies conducted so far to assess the methodology for full core noise simulations were based on assuming vibrations of a clamped-free cluster of fuel assemblies that are unsupported from both ends. However, as this represents a non-physical movement, further developments were made at PSI to allow simulating more realistic movements of fuel assemblies such as the cantilevered mode vibration. The updated methodology, along with evaluations of the simulated noise response to realistic vibration modes, is presented in this paper. Results show that, as expected, the radial and axial neutron noise behaviour follow the vibration pattern of the imposed time dependent axial functions corresponding to the natural oscillation modes of the fuel assemblies, thereby providing confidence in the application of the developed methodology for numerical neutron noise analyses of the PWR cores

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 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.

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.

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions. Keywords: Accident Management, Core Exit Temperature, Integral Test Facility, Peak Cladding Temperature, PKL, Primärkreislauf-Versuchsanlage (Primary Coolant Loop Test Facility), PWR, Station Blackout, TRACE, Validation

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.

Modern thermal-hydraulics (TH) systems analysis codes are increasingly being updated to include three-dimensional (3D) TH components towards capturing the complex flow phenomena in large components of light water reactors (LWRs). These are becoming increasingly popular for modelling reactor pressure vessels and cores since they potentially offer a more realistic representation of, for example, the effects of asymmetric loop flows and injection, which cannot be captured using traditional multi-channel models. Experience suggests, however, there may be unexpected side effects of upgrading from traditional multi-channel models. In this paper, we present comparative results demonstrating our experiences in switching to 3D components using the US-NRC system code TRACE. We present a series of studies in which 1D and 3D nodalisations are compared for representative PWR loss-of-coolant-accident (LOCA) scenarios. We see a clear trend for many of the LOCA scenarios considered. In particular, the peak cladding temperature (PCT) for 3D nodalisations is significantly lower than for traditional multi-channel models in many cases. Counter current flow limitation (CCFL) at the core upper tie plate appears to be one of the driving phenomena for many of the observed differences. This trend is, however, not universal; For a surge line break in the large scale test facility (LSTF), heterogeneities in the core upper tie plate lead to hot regions in the core, which are not captured in the multi-channel model. For large PWR cores, 3D nodalisations appear to represent the reality more closely. For smaller cross-sections, however, the validity of the TRACE’s CCFL prediction for 3D nodalisations is questionable. Specific guidelines are therefore needed to help TRACE users select the most appropriate 3D nodalisation for different geometries.

The role of dry storage in spent nuclear fuel management becomes more and more important. Originally intended to serve as a temporary solution for a few decades until final disposal, now dry storage period is to be extended to 100 years and beyond. It has to be proven for licensing that the fuel rod integrity during dry storage is ensured. Since it is difficult to provide experimental support, the licensing process has to rely largely on numerical simulations. This paper reviews the literature associated with the modeling of spent nuclear fuel under dry storage conditions. The main phenomena threatening the fuel rod integrity and the numerical models representing them are described here. Moreover, the most recent or currently ongoing experimental efforts that could support the modeling of nuclear fuel in dry storage in the future are discussed. Finally, this paper reviews approaches to dry storage modeling chosen by different countries. In general, these approaches are similar and can be described as a calculation chain consisting of neutronics–thermo-hydraulics–fuel performance computations where sensitivity and uncertainty studies are crucial elements.

The current research summarizes recent developments performed within the PSI BEPU methodology and the corresponding results on the SPERT-III RIA experiments. In this regard, three main studies have been carried out. First, an uncertainty breakdown study has been performed in order to identify the dominant nuclear data and to quantify their respective contributions to the overall uncertainties of the parameters of interest. On the other hand, the second study was devoted to assess the impact of considering the effect of cross section perturbation (limited for the time being to U-238 capture) in the resonance self-shielding calculation, known as the implicit effect, on different parameters’ uncertainties in SPERT-III RIA transient. Finally, the third study was conducted with the main goal to assess the impact of kinetic parameter uncertainties, such as delayed neutron yields and associated decay constants, on the overall uncertainties. The uncertainty quantification results of the first study, based on the ENDF/B-VII.1 library, showed that, for steady-state, the first dominant contributor to the overall uncertainties of the static reactivity worth and k-eff is U-235 nubar, followed by U-238 then U-235 capture cross sections. However, for the transient, the first dominant contributor to the overall uncertainties, in e.g. total power and reactivity, is the U-238 inelastic scattering. More interestingly, it was found that, the second dominant contribution to the overall power uncertainty, during the initial excursion phase, is due to the U-235 nubar, while during the power reversal phase, the U-238 capture becomes the second dominant contribution, and even the first one by the end of the transient. Moreover, results of the second study demonstrated a clear systematic decrease of the uncertainty of all parameters due to the inclusion of the implicit effect. In steady state, due to the impact of the implicit effect, the Doppler coefficient relative uncertainty could be reduced up to 30%, which is reflected in the transient, especially in the reversal phase of the power excursion where Doppler effect plays a key role, by a systematic decrease of the relative uncertainty of total power and other parameters. Finally, in the third study, the estimated uncertainties, regarding both steady-state and transient parameters, have shown systematic larger values due to kinetic parameter uncertainties compared to those due to the other nuclear data.

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.

International audience; Within the NURESAFE project, a main steam line break benchmark has been defined and solved by codes integrated into the European code platform NURESIM. The paper describes the results of the calculations for this benchmark. Six different solutions using different codes and code systems are provided for the comparison. The quantitative differences in the results are dominated by the differences in the secondary system parameters during the depressurization. The source of these differences comes mainly from the application of different models for the two-phase leak flow available in the system codes. The use of two different thermal hydraulic system codes influences the results more than expected when the benchmark was created. The codes integrated into the NURESIM platform showed their applicability to a challenging transient like a main steam line break. © 2017 Elsevier B.V.