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The paper presents different methods for evaluating sensitivities and uncertainties of the effective delayed neutron fraction (βeff) with respect to microscopic nuclear data in the framework of the SCALE 6.2 code system. The prompt k ratio method is used to calculate βeff approximately. Sensitivities of βeff to microscopic reaction data are calculated with deterministic linear perturbation theory from sensitivities of the multiplication factor, and by performing linear regression analysis on results obtained from random sampling. The latter approach is also applied for estimating sensitivities of kinetic parameters in six delayed neutron groups, as obtained by adjoint flux weighting in lattice calculations, to nuclear reactions. Finally, it is demonstrated that the uncertainties of βeff strongly depend on the choice of the covariance data set; in particular, the use of SCALE 6.2 covariance data for the average number of delayed neutrons per fission in some cases leads to very high βeff uncertainties.

Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphite-moderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. This work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent and KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.

The OECD Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) established the Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV) in 2014 to bridge the gap between advanced, multi-physics simulation capabilities and the relatively low availability of dedicated, high-fidelity experimental data and benchmarks specifically for multi-physics modelling and simulation tools. The EGMPEBV was mandated to establish mechanisms for the certification of experimental data and benchmark models and to establish the processes and procedures for the validation of multi-physics modelling and simulation tools. The EGMPEBV oversees three task forces, covering (1) experimental data qualification and benchmark evaluation, (2) validation guidelines and needs and (3) example application of validation experiments. These have generated numerous reports surveying the state-of-the-art in multi-physics validation, challenge areas and recommendations for the evaluation of multi-physics benchmarks, while in parallel developing the specifications for multi-physics benchmarks. Three benchmark specifications are in active development, including a reactivity compensation scenario in the Rostov Unit 2 VVER-1000, multi-cycle depletion of the TVA Watts Bar Unit 1 and study of pellet cladding mechanical interaction within ramp tests performed at the Studsvik R2 reactor. We provide an overview of the recent progress in these areas and a summary of the future activities of the EGMPEBV in establishing international multi-physics benchmarks.

The paper presents solutions for the neutron transport sub-exercises regarding pin cells, fuel assemblies, and supercells of the OECD/NEA UAM-SFR Benchmark. For this, the GRS XSUSA methodology is applied with a recently introduced approach performing linear regression analysis on the output samples, denoted as XSUSA(LR). The main purpose for this is to calculate sensitivity profiles and the main contributions to the total output uncertainties via the first-order uncertainty propagation formula, i.e. by multiplying the corresponding total/partial nuclear data covariance matrices with the total/partial sensitivity vectors. Confidence intervals were estimated by applying Jackknife resampling. Implicit effects are considered. All output uncertainties are compared to the corresponding values calculated with deterministic linear perturbation theory using TSUNAMI from SCALE 6.2.3. Various optimizations were applied with respect to nuclear data and problem geometry. The output uncertainties are significant – around 1.5% for multiplication factors and 5% for Doppler and sodium void reactivity effects.

International audience; High-fidelity, multi-physics modeling and simulation (MetS) tools are being developed and utilized for a variety of applications in nuclear science and technology and show great promise in their abilities to reproduce observed phenomena for many applications. Even with the increasing fidelity and sophistication of coupled multi-physics MetS tools, the underpinning models and data still need to be validated against experiments that may require a more complex array of validation data because of the great breadth of the time, energy and spatial domains of the physical phenomena that are being simulated. The expert group on Multi-Physics Experimental Data, Benchmarks and Validation (MPEBV) of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) was formed to address the challenges with the validation of such tools. The work of the MPEBV expert group is shared among three task forces to fulfill its mandate and specific exercises are being developed to demonstrate validation principles for common industrial challenges. This paper describes the overall mission of the group, the specific objectives of the task forces, the linkages among the task forces, and the development of a validation exercise that focuses on a specific reactor challenge problem.

This paper is dedicated to the strength assessment of powered axle of shunting locomotive MDD-04, manufactured for Swiss market by Bulgarian Rolling stock manufacturer Express Service Ltd. Like many other components of the newly built locomotives, the axles are also subject to strict approval procedures. Based on geometrical characteristics of the axle combined with loads acting on the axle, as defined in European standard EN 13104:2009+A2:2012, strength assessment according to calculation method defined in this standard was performed. In this way obtained safety factor values were not satisfying, so it was decided to use Finite Elements Method for validation of first calculation results. Since the results obtained with first calculation method did not meet the requirements, and with the FEM they were satisfactory, an attempt was made to approve the design only with FEM. Although the results were satisfactory, the notified body demanded a change in the original design to be made, because of approval procedures prescribed in EN 13104. The results of the new calculations were more satisfying, so the powered axle could be approved by notified body. The new design of the axle was released for commissioning and use in new produced locomotives.

The paper is dedicated to comparative analysis of the results of the static strength calculation of the cantilever and the tests that have been carried out. Strength calculations have been made using the Finite Elements Method in the Department of Railway Engineering at the Technical University – Sofia. Three different types of Finite Elements (solid, shell and beam) have been used for calculations. The tests of the cantilever have been carried out in the Testing laboratory of Department of Railway Engineering. It was found that the stress results are very similar, especially in the areas with maximal values. This proves that a suitable calculation model with a relatively small number of finite elements has been developed. This allows solving a wide range of problems concerning the improvement of the cantilever with similar construction.

The model implemented into coupled neutron-thermohydraulic code system ATHLET + QUABOX/CUBBOX for RBMK-1000 vapor reactivity coefficient measurements computer modeling is described. The results of ATHLET + QUABOX/CUBBOX modeling of Kursk NPP unit 1 vapor reactivity coefficient measurements are presented. Reasonable agreement between measurements and modeling results was shown.

The development of multi-physics multi-scale coupled methodologies for Light Water Reactor (LWR) analysis requires comprehensive validation and verification procedures, which include well-established benchmarks developed in international cooperation. The Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) has provided such framework, and over the years a number of LWR benchmarks have been developed and successfully conducted. The first set of NEA/OECD benchmarks that permits testing of the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions have been completed and documented. These benchmarks provided a validation basis for the new generation of coupled “best-estimate” codes. The above mentioned OECD/NEA LWR benchmark activities have also stimulated follow up developments and benchmarks to test these developments. The models utilized have been improved when moving from one benchmark to the next and this created a need to validate them using high-quality experimental data. Second set of the NEA/OECD benchmarks have been initiated by the Expert Group on Uncertainty Analysis in Modelling (EGUAM) at the Nuclear Science Committee (NSC), NEA/OECD to address the current trends in the development of LWR multi-physics and multi-scale modeling and simulation. These benchmarks include the following common features, which address some of the issues identified in the first set of OECD/NEA benchmarks: (a) utilization of high-quality experimental data; (b) refined local scale modeling in addition to global predictions; (c) more detailed comparisons and analysis; (d) including uncertainty and sensitivity analysis of modeling predictions. The paper presents each of these new benchmarks by providing description and discussion of comparative analysis of obtained results. Special attention is devoted to uncertainty propagation in LWR multi-physics and multi-scale simulations for design and safety evaluations.

One of the most interesting technologies recognized as promising for the future of nuclear energy is the Accelerator Driven System (ADS). The purpose of the ADS is to function as an element of an integrated nuclear power enterprise comprising of conventional and advanced power reactors for energy production, and for reducing the radiotoxicity of the nuclear waste produced by these power reactors before entombment in a geologic repository ( Stanculescu, 2000 ). The use of new types of coolants such as molten lead or lead–bismuth eutectic alloy (LBE) represents another particular feature in the design concept of these reactors, since it permits one to take advantage of higher Heavy Liquid Metals (HLM) boiling temperature compared to water, leading to an improvement in thermal efficiency. Furthermore, it allows the reactor to be operated at a lower pressure, thus reducing the probability of a Loss-Of-Coolant Accident (LOCA) and, consequently, increasing reactor safety. The proposed use of relatively new types of coolants, especially in combination with new fuel types and cladding materials, demands specific attention to the thermal–hydraulics and core mechanics in normal and abnormal conditions. For that reason, a detailed analysis using system codes is necessary. In this paper, a new version of the ATHLET system code is tested, in which the physical properties of liquid metals like sodium, lead and LBE are implemented. The object of this study is the spallation loop of the MYRRHA facility, in which LBE is circulated by forced convection to remove the heat deposited by a proton beam. A detailed nodalization is set up for performing thermal–hydraulic calculations for both nominal conditions and accident scenarios in order to have a good characterization of the entire loop. The start-up transient has verified the correct removal of the heat generated in the target by the foreseen heat exchanger. During accidental transients, it was noted that the level of LBE in the beam line changes in agreement with preliminary studies on the target device. The pump failure test represents the most dangerous scenario, as the temperature in the target area will reach a very high value within a few seconds after the blockage of the pump impeller. The results obtained have subsequently been compared to the ones achieved with previous numerical simulations which were performed using a version of RELAP5/mod 3.3 system code that was modified at the University of Pisa to account for LBE properties. The comparison has confirmed the capability of the new version of the ATHLET code for the analysis of hydraulic circuits cooled by liquid metals, although both codes use different heat transfer correlations.

A methodology has been developed for assessing the profitability of passenger carriers in the event of a complete or partial replacement of the old rolling stock with new one. The evaluation has taken into account the energy costs, the cost of servicing the train stock, repair costs, the cost of infrastructure charges, the number of days for the movement of trains. A correction coefficient was introduced for bringing the analysed costs closer to the real ones reported in the accounting documents of BDZ-PP Ltd. It was found out that when the rolling stock is completely changed, company’s expenses are expected to decrease by BGN 82 million per year. The most significant share is due to a decrease in energy resources – 51%, followed by repair – 39%, train staff – 17% and infrastructure fees – 16%.

Incorporating full three-dimensional models of the reactor core into system transient codes allows for a "best-estimate" calculation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations on the development of coupled thermal-hydraulic and neutronics codes. Appropriate benchmarks have been developed in international cooperations led by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) that permit testing of the neutronics-thermal-hydraulics coupling and verification of the capability of the coupled codes to analyze complex transients with coupled core-plant interactions. Three such benchmarks are presented in this paper-the OECD/U.S. Nuclear Regulatory Commission (NRC) pressurized water reactor main steam line break benchmark, the OECD/NRC boiling water reactor turbine trip benchmark, and the OECD/U.S. Department of Energy/Commissariat a l'Energie Atomique V1000 coolant transient benchmark. To meet the objectives of the validation of best-estimate coupled codes, a systematic approach has been introduced to evaluate the analyzed transients employing a multilevel methodology. Since these benchmarks are based on both code-to-code and code-to-data comparisons, further guidance for presenting and evaluating results has been developed. During the course of the benchmark activities, a professional community has been established, which allowed our carrying out in-depth discussions of different aspects considered in the validation process of the coupled codes. This positive output has certainly advanced the state of the art in the area of coupling research.

The Organization for Economic Cooperation and Development (OECD) Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark has been calculated for all three exercises by the coupled code system ATHLET-QUABOX/CUBBOX developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). The results obtained are presented, and a detailed comparison with other solutions of the benchmark is discussed. An attempt is made to explain the differences observed in the solutions by the different modeling of physical processes in the codes. The sensitivity of results on modeling features is also investigated. In addition, the effect of different mapping schemes between fuel assemblies of the core loading and the thermal-fluid dynamics on the accuracy of three-dimensional (3-D) neutronics solutions is studied. The results for the MSLB transient are also evaluated to compare 3-D neutronics and point-kinetics solutions in view of integral and local parameters. Thus, the experiences with the coupled code system ATHLET-QUABOX/CUBBOX during the MSLB benchmark activity are summarized.

Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

This study presents uncertainty and sensitivity analyses of the delayed neutron fraction of light water reactor and sodium-cooled fast reactor fuel assemblies. For these analyses, the sampling-based XSUSA methodology is used to propagate cross section uncertainties in neutron transport and depletion calculations. Cross section data is varied according to the SCALE 6.1 covariance library. Since this library includes nu-bar uncertainties only for the total values, it has been supplemented by delayed nu-bar uncertainties from the covariance data of the JENDL-4.0 nuclear data library. The neutron transport and depletion calculations are performed with the TRITON/NEWT sequence of the SCALE 6.1 package. The evolution of the delayed neutron fraction uncertainty over burn-up is analysed without and with the consideration of delayed nu-bar uncertainties. Moreover, the main contributors to the result uncertainty are determined. In all cases, the delayed nu-bar uncertainties increase the delayed neutron fraction uncertainty. Depending on the fuel composition, the delayed nu-bar values of uranium and plutonium in fact give the main contributions to the delayed neutron fraction uncertainty for the LWR fuel assemblies. For the SFR case, the uncertainty of the scattering cross section of U-238 is the main contributor.

The paper is focused on the detailed thermo-hydraulic (TH) modelling of the Reactor Pressure Vessel (RPV) only with the help of a TH system code including the full scope of Nuclear Power Plant (NPP) modelling (primary and secondary sides, RPV, active core and all operational and safety systems). The idea of the pseudo-3D TH methodology of the GRS system code ATHLET [1] is based on the application of a set of parallel TH (PTH) channels with cross flow connections between them. The aim of the activities described in the paper is to validate the recent developments of the methodology based on a fine TH nodalization of ATHLET on different experiments and thus to meet the challenge of detailed modelling of neutron-kinetics (NK) in the active core and the TH processes in the whole RPV with a good accuracy applying only a system TH code. This methodology enables to boost the performance of coupled code safety analysis for production purposes and it also allows the efficient application of uncertainty and sensitivity (U&S) analyses. Primary concern of the validation is related to the methodology itself that includes the correct modelling of the mixing flow phenomena mainly in the coolant legs, downcomer, bottom and upper plenum of RPV, and active core. Another aim of the study is to find the limits of application and accuracy of the system code ATHLET by the simulation and prediction of local (core) safety related parameters with the enhanced nodalization and PTH channel application.The paper describes and discusses the comparisons done with the ATHLET system code using the new enhanced nodalization scheme with 3 types of measured data: data from two experiments (C1 and C3) on fluid mixing at Upper Plenum Test Facility (UPTF) [2] and a set of detailed neutron-physical/thermal-hydraulic measurements performed at Kalinin-3 NPP during an asymmetric transient.The performed studies are of great interest because they have shown acceptable accuracy and also some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the PTH channel methodology within a 1-D TH system code.Copyright © 2012 by ASME

Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

In the present contribution, an overview of the sampling based XSUSA method for sensitivity and uncertainty analysis with respect to nuclear data is given. The focus is on recent developments and applications of XSUSA. These applications include calculations for critical assemblies, fuel assembly depletion calculations, and steadystate as well as transient reactor core calculations. The analyses are partially performed in the framework of international benchmark working groups (UACSA – Uncertainty Analyses for Criticality Safety Assessment, UAM – Uncertainty Analysis in Modelling). It is demonstrated that particularly for full-scale reactor calculations the influence of the nuclear data uncertainties on the results can be substantial. For instance, for the radial fission rate distributions of mixed UO2 /MOX light water reactor cores, the 2σ uncertainties in the core centre and periphery can reach values exceeding 10%. For a fast transient, the resulting time behaviour of the reactor power was covered by a wide uncertainty band. Overall, the results confirm the necessity of adding systematic uncertainty analyses to best-estimate reactor calculations.

The paper demonstrates the influence of uncertainties in microscopic nuclear data on the results of reactor simulations, both for stationary and transient states. It gives an overview on the methods in use for uncertainty and sensitivity analyses with respect to nuclear data, and discusses the pros and cons. For full-scale reactor simulations, in particular with coupled neutron transport and thermo-hydraulics, random sampling provides a powerful means to propagate nuclear data uncertainties through the complete calculation sequence. Results of uncertainty analyses performed with the GRS XSUSA – “cross section (XS) Uncertainty and Sensitivity Analysis” methodology are shown for radial power distributions from steady-state PWR calculations and for the time evolution of the reactor power in the course of a control rod withdrawal from a PWR mini-core. In all cases, the output uncertainties are considerable. For the radial power distributions, relative 1σ uncertainties of up to 10% are observed, and for the power peak during the transient, the relative 1σ uncertainty reaches 20%. These large uncertainties strongly suggest to routinely accompany best-estimate simulations by uncertainty analyses with respect to nuclear data, in particular for systems beyond LWR for which much less operation experience is available.