bluuebunny/crossref_metadata_2025 huggingface dataset:
IMPROVED ANISOTROPIC LINEAR SOURCE FORMULATION FOR MULTIPHYSICS PROBLEMS
Authors: Herring Nicholas F. , Fitzgerald Andrew , Kochunas Brendan , Downar Thomas
Published: February 2021

This work seeks to extend an existing formulation of the method of characteristics with linear source approximation for problems with dynamic cross sections. The previous formulation eliminated cross section dependence of precomputed coefficients for systems with an isotropic source. The method is extended to include a formulation for spatially flat anisotropic scattering that eliminates cross section dependence of precomputed coefficients without adding additional operations; increasing efficiency in multiphysics simulations where cross sections can be subject to change.

The new formulation is implemented in the MPACT code and tested on two problems: 3D transport assembly calculations using MPACT’s 2D/1D method and a 3D assembly with T/H feedback using MPACT’s 2D/1D method coupled with COBRA-TF. This work demonstrates that the new linear source formulation allows for the number of mesh elements to be significantly reduced while maintaining accuracy, leading to shorter run-times for 3D cases with fixed cross sections, and substantial reduction of memory usage for 3D cases with fixed cross sections. The multiphysics calculations show similar runtimes for the same accuracy with significant reduction of memory. For similar accuracy, the method proved effective in reducingmemory usage by, on average, 30% for 3D problems and 21% for multiphysics problems.
IMPROVEMENT TO NEM SP3 MODELLING AND SIMULATION
Authors: Xu Yuchao , Hou Jason , Ivanov Kostadin N.
Published: February 2021

Accurate reactor core steady state safety analysis requires coupling between thermal-hydraulics and three dimensional multigroup pin by pin neutronics. Concerning the neutronics modeling, the Nodal Expansion Method (NEM) code is developed at North Carolina State University in the framework of high fidelity multiphysics coupling with CTF. NEM includes a simplified third-order Spherical Harmonic (SP3) solver. In this work, the solver has been improved by incorporating higher order scattering matrix library. The boundary conditions were corrected with one dimensionalP3theory and a consistent coupling coupling between zeroth- and second-order flux moments was established. Two methods for generating second order discontinuity factors (DFs) has ben developed, one based on the Generalized Equivalence Theory (GET) and one based on Parial Current Equivalence Theory (PCET). DFs were generated with three lattice sizes: single pin, 2 pins and assembly level. These developments were tested using the C5G7 benchmark. The results of the SP3 solver improvement, by usingP2andP3scattering cross sections, show a 50% decrease in the eigenvalue (keff) prediction error compared to the reference transport solution. The GET DFs are applied in the C5G7 core pin by pin calculation and are compared with PCET DFs. The results show that PCET have a better performance in global results (eigenvalue). Concerning the different lattice sizes studies, the results show that DFs generated in smalll colorsets can improve local solutions. However, in order to reveal strong global trends, DFs should be generated in a larger corloset representative of the whole core. For the core calculations, DFs generated with the three colorsets together with an additional mixed type DFs were tested. For the mixed type, DFs generated from assembly size lattice were used for the internal interfaces and DFs generated from 2 pins size lattice were used for the assemblies boundary interfaces. These mixed DFs outperformed all the other configurations indicating that they manage to accomplish a satisfying compromise between global and local trends.
A COMPARISON OF STOCHASTIC MESH CELL VOLUME COMPUTATION STRATEGIES FOR THE RANDOM RAY METHOD OF NEUTRAL PARTICLE TRANSPORT
Authors: Tramm John R. , Siegel Andrew R. , Lund Amanda L. , Romano Paul K.
Published: February 2021

The random ray method is a recently developed neutron transport method that can be used to perform efficient full-core, general-purpose, high-fidelity 3D simulations of nuclear reactors. While Tramm et al. have so far documented the new random ray algorithm in several publications, one critical detail has not yet been published: how to best determine the volume of each source region (or cell) of the simulation. As the “true” analytical constructive solid geometry cell volumes are typically not known a priori they must be computed by the application at runtime, which is not straightforward in TRRM as different rays are used each power iteration such that the sampled volume of each cell also changes between iterations. In the present study, we analyze two different on-the-fly stochastic methods for computing the cell volumes and quantify their impacts on the accuracy of scalar flux estimates. We find that the “na¨ıve” stochastic volume estimator (which arises naturally from the derivation of the Method of Characteristics), is highly biased and can result in over 1,000 pcm error in eigenvalue. Conversely, we find that the “simulation averaged” estimator is unbiased and is therefore equivalent to the use of analytical cell volumes even when using a coarse ray density. Thus, the new simulation averaged method is a critical (and as yet undocumented) component of the TRRM algorithm, and is therefore vital information for those in the reactor physics community working to implement random ray solvers of their own.
BENCHMARKING OF THE SERPENT 2 MONTE CARLO CODE FOR FUSION NEUTRONICS APPLICATIONS
Authors: Valentine A. , Colling B. , Worrall R. , Leppänen J.
Published: February 2021

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed.

A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.
A SEMI-ANALYTIC EIGENVALUE EXTENSION TO THE DOPPLER SLAB ANALYTIC BENCHMARK1
Authors: Remley Kyle E. , Griesheimer David P.
Published: February 2021

Advancement in multiphysics simulation has motivated interest in availability of analytic and semi-analytic benchmark solutions. These solutions are sought because they can be used to assess the accuracy of complicated numerical schemes necessary to simulate coupled physics systems. While there exist analytic solutions for fixed-source problems, benchmark-quality eigenvalue solutions are of interest because eigenvalue problems more closely align with analyses undertaken with coupled solvers. This paper extends a fixed-source benchmark, the Doppler Slab benchmark, to the eigenvalue case. A novel solution for this benchmark is derived. Numerical implementation of the benchmark is demonstrated through verification of numerical computation of the power reactivity coefficient.
EFFECT OF THE UNIFORM FISSION SOURCE METHOD ON LOCAL POWER VARIANCE IN FULL CORE SERPENT CALCULATION
Authors: Bilodid Yurii , Leppänen Jaakko
Published: February 2021

One of challenges of the Monte Carlo full core simulations is to obtain acceptable statistical variance of local parameters throughout the whole reactor core at a reasonable computation cost. The statistical variance tends to be larger in low-power regions. To tackle this problem, the Uniform-Fission-Site method was implemented in Monte Carlo code MC21 and its effectiveness was demonstrated on NEA Monte Carlo performance benchmark. The very similar method is also implemented in Monte Carlo code Serpent under the name Uniform Fission Source (UFS) method.

In this work the effect of UFS method implemented in Serpent is studied on the BEAVRS benchmark which is based on a real PWR core with relatively flat radial power distribution and also on 3x3 PWR mini-core simulated with thermo-hydraulic and thermo-mechanic feedbacks. It is shown that the application of the Uniform Fission Source method has no significant effect on radial power variance but equalizes axial distribution of variance of local power.
ANALYSIS OF DOSE RATES AROUND THE SLOVENIAN SILO-TYPE LILW REPOSITORY USING ADVANTG
Authors: Kotnik Domen , Kos Bor , Čalič Dušan , Snoj Luka
Published: February 2021

The ADVANTG code was used to analyze dose rates from the proposed Slovenian silo-type low and intermediate level waste (LILW) repository. Detailed calculations of dose rates are challenging as gamma-sources are located in thick concrete containers and the effect of backscattering in the air has to be considered. Compared to analog MCNP simulations maximum relative speed-up of up to 9500 was achieved by using ADVANTG. We have analyzed the influence of different repository configurations on the final annual γ-dose rates. For different measurement positions, which were located from 10 m to 150 m from the outer edge of the repository, γ-dose rates were calculated. The results show that for all configurations of the open repository where the waste from different sources is deposited, except using the waste from Krško NPP, the calculated annual γ-dose rates for all measurement positions were lower than the prescribed limit values for professionals of 20 mSv/year. To ensure that the γ-dose rates are lower than the prescribed limit values even for the case with the waste from Krško NPP an additional analysis was made where the concrete casks, in which the waste is stored, were modelled in more detail and the quantity of additional grout on top of the waste was varied.
EFFECT OF ENERGY DEPOSITION MODELLING IN COUPLED STEADY STATE MONTE CARLO NEUTRONICS/THERMAL HYDRAULICS CALCULATIONS
Authors: Tuominen Riku , Valtavirta Ville , García Manuel , Ferraro Diego , Leppänen Jaakko
Published: February 2021

In coupled calculations with Monte Carlo neutronics and thermal hydraulics the Monte Carlo code is used to produce a power distribution which in practice means tallying the energy deposition. Usually the energy deposition is estimated by making a simple approximation that energy is deposited only in fission reactions. The goal of this work is to study how the accuracy of energy deposition modelling affects the results of steady state coupled calculations. For this task an internal coupling between Monte Carlo transport code Serpent 2 and subchannel code SUBCHANFLOW is used along with a recently implemented energy deposition treatment of Serpent 2. The new treatment offers four energy deposition modes each of which offers a different combination of accuracy and required computational time. As a test case, a 3D PWR fuel assembly is modelled with different energy deposition modes. The resulting effective multiplication factors are within 30 pcm. Differences of up to 100K are observed in the fuel temperatures.
RMC/ANSYS MULTI-PHYSICS COUPLING SOLUTIONS FOR HEAT PIPE COOLED REACTORS ANALYSES
Authors: Ma Yugao , Liu Minyun , Chen Erhui , Xie Biheng , Chai Xiaoming , Huang Shanfang , Wang Kan , Yu Hongxing
Published: February 2021

The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy Reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric Design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the-fly temperature treatment of cross sections was adopted in RMC code to solve the memory problems and to speed up simulations. This paper proposed a geometric updating strategy and reactivity feedback methods for the geometric deformation of the solid-state core. The neutronic and thermal-mechanical coupling platform is developed to analyze and further to optimize the heat pipe cooled reactor design. The present coupling codes analyze a 2D central cross-section model for MEGAPOWER heat pipe cooled reactor. The thermal-mechanical feedback reveals that the solid-state reactor has a negative reactivity feedback (~1.5 pcm/K) while it has a deterioration in heat transfer due to the expansion.
COUPLED TRANSIENT ANALYSIS OF A CORE WITH FUEL ASSEMBLY BOWING WITH A HYBRID CTF/DYN3D MODEL
Authors: Périn Y , Travleev A , Zilly M
Published: February 2021

Fuel assembly bowing is a known phenomenon observed in many PWR reactors all over the world. The phenomenon is relevant to safety as it can lead to increased water gaps between assemblies which results in higher pin peaking factors. The goal of the present study is to assess the effect of assembly bowing not only for stead-state nominal conditions but also during a transient. The selected transient is the loss of one reactor coolant pump as it can be limiting especially regarding the Departure from Nucleate Boiling (DNB) safety criterion. This study focuses on an extreme case where the bowing is simulated in the core hot assembly by keeping the water gap constant over the whole core active length. The resulting cross-sections and form functions obtained from a 2d infinite lattice model are used in the nodal diffusion code DYN3D applying its pin-by-pin reconstruction method. For the transient simulation, DYN3D is coupled with the thermal-hydraulics subchannel code CTF on the SALOME platform. Several modelling options are compared: nominal geometry for neutronics and thermal-hydraulics (TH); mixed: neutronics with increased water gap, TH with nominal geometry; and increased water gap for both neutronics and TH. The results confirm that the increased water gap should be considered in both models in order to reduce the conservatism.
MULTIPHYSICS MODELISATION OF AN UNPROTECTED LOSS OF FLOW TRANSIENT IN A SODIUM COOLED FAST REACTORS USING A NEUTRONIC-THERMAL-HYDRAULIC COUPLING SCHEME
Authors: Pascal V. , Gorsse Y. , Alpy N. , Ammar K. , Anderhuber M. , Baudron AM. , Campioni G. , Junquera A. Canas , Gerschenfeld A. , Patricot C. , Sciora P.
Published: February 2021

Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.
PRELIMINARY POWER TRANSIENT ANALYSIS OF THE SUPER FR WITH AXIALLY HETEROGENEOUS CORE
Authors: Okui Tsutomu , Yamaji Akifumi
Published: February 2021

The Super FR is one of the SuperCritical Water cooled Reactor (SCWR) concepts with once-through direct cycle plant system. Recently, new design concept of axially heterogeneous core has been proposed, which consists of multiple layers of MOX and blanket fuels. To clarify the safety performance during power transient, safety analyses have been conducted for uncontrolled control rod (CR) withdrawal and CR ejection at full power. RELAP/SCDAPSIM code was used for the safety analysis. The results show that the peak cladding surface temperature (PCST) is high in the upper MOX fuel layer. It is also shown that axial temperature gradient of cladding greatly increases in a short period. Suppressing such large temperature gradient may be a design issue for the axially heterogeneous core from the viewpoint of ensuring fuel integrity.
AN IRSN CONTRIBUTION TO THE UAM PROJECT: THERMAL-HYDRAULIC AND NEUTRONIC UNCERTAINTIES PROPAGATION IN A ROD EJECTION, FIRST RESULTS
Authors: Sargeni A. , Ivanov E.
Published: February 2021

The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark intended to an elaboration of the methodology of uncertainty propagation. The considered case studied a full PWR core behavior in fast (~0.1 sec) rod ejection transient. According to the benchmark, the core represented a Hot Zero Power state. Authors used brute-force sampling propagating nuclear data and thermo-fluid uncertainties using 3D computational IRSN chain HEMERA. It couples the reactor physics code CRONOS and thermal-hydraulic core code FLICA4. The nuclear data uncertainties were represented in a form of cross sections standard deviations (in percentage of the mean cross sections values) supplied by the UAM team. In addition to the original benchmark, the study includes a case with an increased power peak by supplementary rod ejection, i.e. with higher reactivity. Both the results are similar to what we obtained in the mini-core rod ejection: the power standard deviation follows, in percentage of the mean power, the mean power curve. We split the variance with a direct calculation: once the cross sections are modified and the thermal-hydraulics inputs are kept constant, another time the contrary. The results show that uncertainties dues to nuclear data dominate over ones due to the thermal-flow area. Furthermore, the major contributors in peak-of-power variance lie in a fast group of cross sections.
MSRE TRANSIENT BENCHMARKS USING SAM
Authors: Fei T. , Hua T. , Feng B. , Heidet F. , Hu R.
Published: February 2021

In recent years, there has been renewed interest in Molten Salter Reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, the System Analysis Module (SAM) was enhanced to include MSR-specific modeling features including a delayed neutron precursor drift model and a modified point kinetics model. This paper discusses the validation of these features using the experiments conducted in the Molten Salt Reactor Experiment (MSRE). These experiments include the pump start-up and coast-down tests at zero power and a thermal convection test. For the zero power tests, the change in pump speeds induces flow rate changes in the core that impact the precursor concentrations. This introduces a neutron imbalance and requires the adjustment of the control rods to counter-balance this effect. SAM was used to evaluate the precursor concentration in the core as a function of time, and the resulting changes in reactivity were evaluated through the modified point kinetics equation. The results show good agreement with the experimental data. It should be noted that the pump performance curve used in this analysis was re-constructed based on the initial water test data of the fuel pump. The steady-state pump curve is assumed to be applicable to transient flow operations. The thermal convection test was conducted by shutting off the pumps, reducing the inlet core temperature for 360 minutes, and allowing the power to be adjusted by the inherent feedbacks of the system. The power level during this transient was evaluated by SAM as a function of time.
MULTI-PHYSICS TRANSIENT SIMULATIONS WITH TRIPOLI-4®
Authors: Faucher Margaux , Mancusi Davide , Zoia Andrea
Published: February 2021

In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback. For this purpose, the Monte Carlo code was extended for multi-physics capabilities and coupled to the thermal-hydraulics subchannel code SUBCHANFLOW. As a test case for the verification of transient simulation capabilities, a 3x3-assembly mini-core benchmark based on the TMI-1 reactor is considered with a pin-by-pin description. Two reactivity excursion scenarios initiated by control-rod movement are simulated starting from a critical state and compared to analogous simulations performed using the Serpent 2 Monte-Carlo code. The time evolution of the neutron power, fuel temperature, coolant temperature and coolant density are analysed to assess the multi-physics capabilities of TRIPOLI-4. The stabilizing e_ects of thermal-hydraulics on the neutron power appear to be well taken into account. The computational requirements for massively parallel calculations are also discussed.
REACTOR NOISE EXPERIMENTS CONSIDERING HIGH FREQUENCIES IN THE IPEN/MB-01 REACTOR
Authors: dos Santos Adimir , dos Santos Diogo Feliciano
Published: February 2021

Reactor noise experiments at high frequencies (up to 100 kHz) in subcritical configurations have been performed at the IPEN/MB-01 research reactor facility. The core configuration considered a short version of the IPEN/MB-01 core in a 24 × 26 rectangular array of fuel rods. The subcritical configurations considered the control banks totally withdrawn and the moderator poisoned with H3BO3. The boron concentrations were: 286.8 ± 10 and 578.6 ± 10 ppm. The pulses of two 3He detectors in the reflector region were summed and inserted into the IPEN/MB-01 Correlator and the APSD (Auto Power Spectral Density) was inferred through a mathematical model. The analysis reveals that the APSD in this frequency range is best described by a four-mode decay model. According to the two-region two-group (thermal and fast) kinetic model developed in this work, the first two decay modes describe the thermal group and the other two describe the fast group. After a tedious and severe analyses of the least-square fit of the experimental data, it was concluded that the kinetic behavior of the thermal and fast neutrons can be considered uncoupled. The analysis of the experimental data is still in progress and only the thermal group of the case of 286.8 ppm of boron was analyzed and some parameters could be inferred. The most important one so far is the prompt neutron generation time in the core region that could be inferred with a good level of accuracy.
REACTOR PERFORMANCE IMPROVEMENT OPTIONS TO SUSTAIN HIGH FLUX ISOTOPE REACTOR LEADERSHIP INTO THE FUTURE*
Authors: Chandler D. , Betzler B. R. , Cook D. H.
Published: February 2021

The mission of the Neutron Sciences Directorate (NScD) at the U.S. Department of Energy’s Oak Ridge National Laboratory (ORNL) is the undertaking of high-impact research into the structure and properties of materials across the spectrum of biology, chemistry, physics, materials science, and engineering. NScD operates two world-leading neutron scattering facilities: the High Flux Isotope Reactor (HFIR) and the Spallation Neutron Source. HFIR achieved full power in 1966, and over a half century later, it continues to serve a variety of national missions. HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world to support scientific missions including cold and thermal neutron scattering, isotope production, and materials irradiation research. To sustain leadership in neutron sciences into the future, ORNL is exploring areas in which HFIR can be improved to enhance its performance. Many improvement areas are being explored including upgrading the cold source and neutron scattering facilities. The improvement areas discussed herein include replacing the reactor pressure vessel, upgrading the neutron reflector, and ensuring that reactor performance is maintained or enhanced after converting from high-enriched uranium to low-enriched uranium fuel.
MEASUREMENT OF INTEGRAL CROSS SECTIONS OF SELECTED DOSIMETRY REACTIONS IN LR-0 REACTOR
Authors: Kostal Michal , Czakoj Tomas , Losa Evzen , Schulc Martin , Juříček Vlastimil , Šimon Jan , Rypar Vojtěch , Mareček Martin , Capote Roberto , Trkov Andrej
Published: February 2021

The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty quantification is being developed by the International Atomic Energy Agency and expected to be released in January 2020; a preliminary version IRDFF-1.05 was released in 2014 and is being tested in this work. Validation of the cross-section evaluations proposed for this library is a high priority task. The validation can be realized using integral cross sections measured in standard and/or reference neutron benchmark fields. Integral quantities feature significantly lower uncertainties than differential nuclear data. If the neutron spectrum where the cross section is measured is well characterized, then the Spectrum Averaged Cross Section can be used for validating of existing evaluations.
NUCLEAR DATA UNCERTAINTY QUANTIFICATION FOR THE DECAY HEAT OF PWR MOX FUELS USING DATA ASSIMILATION OF ELEMENTARY FISSION BURSTS
Authors: Vallet V. , Huyghe J. , Vaglio-Gaudard C. , Lecarpentier D. , Reynard-Carette C.
Published: February 2021

Currently there is no integral experimental data for code validation regarding the decay heat of MOX fuels, excepted fission burst experiments (for fission products contributions at short cooling times) or post-irradiated experiments on nuclide inventories (restricted number of nuclide of interest for decay heat). The uncertainty quantification mainly relies on uncertainty propagation of nuclear data covariances. In the recent years, the transposition method, based on the data assimilation theory, was used in order to transpose the experiment-to-calculation discrepancies at a given set of parameters (cooling time, fuel burnup) to another set of parameters. As an example, this method was used on the CLAB experiments and the experiment-to-calculation discrepancies at 13 years were transposed to an UOX fuel between 5 and 27 years and for burnups from 10 to 50 GWd/t. The purpose of this paper is to study to what extent the transposition method could be used for MOX fuels. In particular, the Dickens fission burst experiment of 239Pu was considered for MOX fuels at short cooling times (< 1h30) and low burnup (< 10 GWd/t). The impact of fission yields (FY) correlations was also discussed. As a conclusion, the efficiency of the transposition process is limited by the experimental uncertainties larger than nuclear data uncertainties, and by the fact that fission burst experiments would only be representative to the FY contribution of the decay heat uncertainty of an irradiated reactor fuel. Nevertheless, this method strengthens the decay heat uncertainties at very short cooling times, previously based only on nuclear data covariance propagation through computation.
ON THE VALIDATION OF VESTA 2.2.0 USING THE ARIANE-GU3 SAMPLE
Authors: Ichou R. , Dechenaux B.
Published: February 2021

The validation of the VESTA 2.2.0 Monte Carlo depletion code has been initiated using the Spent Fuel Isotopic Composition Database (SFCOMPO). The work presented in this paper is limited to one fuel sample, the GU3 PWR-UOX sample from the ARIANE program, which has a reported burn up of 52.5 MWd.kgHM-1. The chemical analyses of the studied fuel sample were performed by 2 independent laboratories at the end of irradiation and cooling time. US and European evaluated nuclear data libraries, namely ENDF/B-VII.1 and JEFF-3.2, but also the more recent ENDF/B-VIII.0 and JEFF-3.3 are used for the VESTA 2.2.0 calculations. The isotopic concentration results are compared to experimental data and the C/E agreement is analyzed in the light of the previous VESTA 2.1.5 validation results obtained using ENDF/B-VII.0 and JEFF-3.1 nuclear data libraries.
ENDF/B-VIII.0 CROSS SECTION TESTING FOR COPPER NUCLEAR CRITICALITY SAFETY APPLICATIONS
Authors: Shaw Alex , Rahnema Farzad , Holcomb Andrew , Bowen Doug
Published: February 2021

In the update from ENDF/B-VII.1 to ENDF/B-VIII.0, copper cross sections were significantly altered in the intermediate and fast spectrum of the ENDF-VIII.0 library. Performance of this ENDF data requires validation to determine whether recent evaluation has proven beneficial. To examine the performance of the new library, particularly new copper data, critical benchmarks from the ICSBEP handbook were chosen for their sensitivity to copper cross section changes and modeled using SCALE continuous energy Monte Carlo simulations. Selected benchmarks were modeled in ENDF-VII.1 and ENDF-VIII.0 to compute keff within a statistical uncertainty of 10 pcm and compared in reference to the benchmark experimental criticality. Due to spectrum choices in selection based on the changes to cross section data, the set of benchmarks consist of intermediately enriched uranium, highly enriched uranium, or plutonium systems. 11 separate benchmark evaluations containing 32 individual configurations highly sensitive to copper were selected, modelled, and compared to benchmark experimental criticality. This work demonstrates a significant decrease in the deviation between calculated and experimental criticality as a result of the ENDF-VIII.0 library; a decrease in absolute mean deviation from 522.5±39.3 to 249.6±39.3, and a decrease in root mean square deviation from 630.8±46.1 to 338.1±74.9. Additionally, the role of recently evaluated copper data in this improved agreement is presented, confirming the benefit of reaffirming cross section data.
THE TCA BENCHMARK FOR VALIDATION OF TEMPERATURE FEEDBACK CALCULATIONS
Authors: van der Marck Steven C. , Asquith Nicola L.
Published: February 2021

The TCA benchmark was investigated as a possible candidate for validation of temperature feedback calculations. This benchmark has low-enriched uranium fuel, light water moderation and reflection, and a temperature range of 20–60 °C. The use of three different nuclear data libraries was considered, viz. ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0. Since the results were not as good as hoped for, additional studies were performed to identify the cause(s) of discrepancies.

The benchmark values depend on a choice of delayed neutron data, so it was investigated whether this could be the cause of discrepancies. Also, an assessment was made based on critical configurations only, i.e. avoiding the use of delayed neutron data, whether the calculations exhibit a bias in relation to the benchmark in the results for the effect of temperature. Indications were found that such a bias exists.

It is concluded that the choice of delayed neutron data has a significant effect on the benchmark values themselves. The use of three major nuclear data libraries leads to a range of benchmark values for each configuration, rather than one value. Also, one has to take into account the possibility of a bias in the calculation of temperature effects. It is not clear at this point what the cause of the bias is.
DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA
Authors: Kim Kang Seog , Ade Brian J. , Luciano Nicholas P.
Published: February 2021

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnox reactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.
SFCOMPO DATABASE OF SPENT NUCLEAR FUEL ASSAY DATA – THE NEXT FRONTIER
Authors: Ilas Germina , Gauld Ian , Ortego Pedro , Tsuda Shuichi
Published: February 2021

SFCOMPO is the world’s largest database for measured spent nuclear fuel assay data. An international effort coordinated by the Nuclear Energy Agency (NEA) resulted in a significant expansion of the database and its release online in 2017 as a downloadable application. The SFCOMPO Technical Review Group (TRG) was recently formed under the direction of NEA’s Nuclear Science Committee/Working Party on Nuclear Criticality Safety and was mandated to maintain and further coordinate the development of SFCOMPO. This TRG is currently focused on (1) critical evaluation of the experimental assay data by independent experts and (2) development of benchmarks and benchmark models that can be applied to validate burnup codes. This will improve the quality and documentation of the experimental datasets and enable their use by the international community to support code validation for design and safety analysis of spent nuclear fuel transportation, storage, and repository applications. It follows the precedent and draws on the experience gained from similar NEA efforts in the International Reactor Physics Experiment Evaluation Project and the International Criticality Safety Benchmark Experiment Project. Ongoing SFCOMPO evaluations have served as a test bed to develop templates for documenting evaluations, develop review guidance, improve approaches for a global uncertainty analysis, and devise a strategy focused on providing practical information of highest value to the user community. The current effort, status, and associated challenges are discussed.
EVALUATION OF CRITICAL EXPERIMENTS IN THE UNIVERSITY OF WISCONSIN NUCLEAR REACTOR (UWNR) WITH UNCERTAINTY QUANTIFICATION
Authors: Park Young-Hui , Cheng Ye , Elzohery Rabab , Wilson Paul P.H. , Roberts Jeremy A. , DeHart Mark D.
Published: February 2021

An improved computational model of the University of Wisconsin Nuclear Reactor (UWNR) was developed to support the benchmark evaluation of recent data acquired during an experimental campaign conducted at UWNR. Previous efforts led to a scripted UWNR model for automated generation of MCNP6 and Serpent inputs. This capability was extended to SCALE/KENO. All three tools were used to evaluate a variety of zero-power, fresh-critical configurations, and the results agreed well. The MCNP6 model was extended to support shuffling the core configuration, which allows the modeling of burnup for evaluation of depleted critical configurations. The MCNP6 model successfully predicts core reactivity over time, after accounting for the initial reactivity bias. The inclusion of SCALE/KENO input generation enables sensitivity and uncertainty analyses using the TSUNAMI and Sampler modules of SCALE. A preliminary uncertainty analysis was performed with TSUNAMI for nuclear data uncertainties while direct perturbation calculations were performed using MCNP6 for geometry and material uncertainties, which helped to identify model parameters with the largest effect on the eigenvalue. A transient UWNR transport Model in Mammoth/Rattlesnake is under development to simulate the transient experiments. The existing MCNP6 and Serpent models are used to provide the CAD file for meshing and homogenized cross-sections. In conclusion, the evaluation of UWNR benchmark data provides increased confidence in various states of the UWNR computational model and will provide a unique model for use by other analysts.
COMPRESSION OF PINWISE NUCLIDE CONCENTRATIONS IN CMS5
Authors: Hykes Joshua
Published: February 2021

Pinwise nuclide number density (ND) data from the lattice physics code CASMO5 is compressed using a preconditioned truncated singular value decomposition (TSVD) to reduce storage requirements. Previously only assembly-average or single-pin NDs were optionally saved in the CMS5 few-group cross section library. However, backend analysis has prompted the desire to have pin-by-pin NDs available in the library for use by the SNF code. Adding this data set significantly increases the size of the library, particularly for lattices modeled in full assembly geometry (that is, not in half or octant symmetry). To reduce the required storage, the SVD is used to approximate the entire data with a reduced basis. In the four test cases, compression ratios of 2.7 to 8.5 were achieved for the PWR cases, with maximum errors less than 0.1%. However, the rodded BWR segment proved more difficult, with an average compression ratio of 1.6. One advantage of this technique is that the compression ratio is higher for full-symmetry cases, where the need for the compression is also highest.
SYSTEMATIC ANALYSIS OF MULTIVARIATE SCENARIOS USING ADVANCED CLUSTERING METHODS
Authors: Marc Ernoult , Sylvain David , Doligez Xavier , Jiali Liang , Nicolas Thiolliere , Lea Tillard
Published: February 2021

The continuous improvement of fuel cycle simulators in conjunction with the increase of computing capacities have led to a new scale of scenario studies. Taking into consideration multiple variable parameters and observing their effect on multiple evaluation criteria, these scenario studies regroup several thousands of trajectories paving the different possible values for multiple operational parameters. If global methods like sensitivity analysis allow extracting useful information from these groups of trajectories, they only provide average and global values.

In this work we present a new method to analyze these groups of trajectories while keeping some localization in the information. Based on principal component analysis, clustering method have been implemented in order to mathematically extract, from the ensemble of trajectories simulated for a scenario study, subgroups of trajectories that have similar behaviors. Typical trajectories, representative of these subgroups, are then determined. The application of this new method on a sample scenario for two different output, the total amount of transuranic elements within the fuel cycle and the number of time the MOX fuel could not be built during the simulated time, is presented. The comparison of the results between the two analyses shows that the method allows good clustering for continuous and regular outputs but struggle with discrete highly non-linear ones.
MAPPING OF SODIUM VOID EFFECT AND DOPPLER CONSTANT IN ESFR-SMART CORE WITH MONTE CARLO CODE SERPENT AND DETERMINISTIC CODE ERANOS
Authors: Krepel Jiri , Raffuzzi Valeria
Published: February 2021

The Sodium Fast Reactor is one of the most technologically developed Gen-IV reactors, which can close the nuclear fuel cycle. Its criticality safety directly depends on the sodium void effect and Doppler constant. Hence the knowledge of their local distribution is important. These coefficients can be mapped by deterministic or Monte Carlo codes, where the latter provide higher modeling accuracy, but are also strongly computer demanding and subject to stochastic noise issues. In this study, the void effect and Doppler constant have been enumerated for the ESFR core by Serpent2 and ERANOS2 codes, preserving a six-batch operation scheme. The Serpent code was coupled to the Python script BBP to simulate batch-wise operation in a radially infinite inner core configuration; the ERANOS code was applied to the whole core geometry and the batch-wise operation was simulated by the EQL3D routine. Sodium void effect and Doppler constant spatial maps with different levels of refinement were produced, as well as the time evolution of the integral coefficients during the transition from initial cycle to equilibrium cycle. Both codes indicate deterioration of these coefficients during the transition. The equilibrium cycle performance of the inner core zone from the ERANOS calculation was compared with Serpent results and they showed reasonable agreement. For very fine mapping, the Monte Carlo method employed was computationally very demanding and the enumerated effect was lower than the stochastic noise. In general, the Serpent model practically excludes modeling assumptions and produces reliable results for reasonably sized maps, which can be combined if needed with the high spatial resolution results obtained by ERANOS simulations.
VOID COEFFICIENT SENSITIVITY ANALYSIS FOR THE TRIGA MARK II REACTOR AT L.E.N.A. (UNIPV)
Authors: Portinari D. , Cammi A. , Lorenzi S. , Aufiero M. , Calzavara Y. , Bidaud A.
Published: February 2021

Sensitivity analysis studies the effect of a change in a given parameter to a response function of the system under investigation. In reactor physics, this usually translates into the study of how cross sections and fission spectrum modifications affect the value of the multiplication factor, the delayed neutron fraction or the void coefficient for example. Generalized Perturbation Theory provides a useful tool for the assessment of adjoint weighed functions such as keff and void coefficient sensitivities. In this work, the capability of SERPENT code to perform sensitivity calculation based on GPT is used to study the TRIGA Mark II research reactor installed at L.E.N.A. of University of Pavia. A general sensitivity analysis to the most important reactor’s cross sections has been performed in order to highlight the biggest reactivity contributions. Two numerically challenging tasks related to GPT calculation have been performed thanks to the relatively quick Monte Carlo approach allowed by this reactor: investigating the linearity of the reactivity injection caused by the flooding of the central channel, and calculating the fuel void coefficient sensitivity to the coolant density.
UNCERTAINTY ANALYSIS TO C5G7-TD BENCHMARK BASED ON THE COST METHOD
Authors: Cao Liangzhi , Sui Zhuojie , Wang Bo , Wan Chenghui , Liu Zhouyu
Published: February 2021

A method of Covariance-Oriented Sample Transformation (COST) has been proposed in our previous work to provide the converged uncertainty analysis results with a minimal sample size. The transient calculation of nuclear reactor is a key part of the reactor-physics simulation, so the accuracy and confidence of the neutron kinetics results have attracted much attention. In this paper, the Uncertainty Quantification (UQ) function of the high fidelity neutronics code NECP-X has been developed based on our home-developed uncertainty analysis code UNICORN, building a platform for the UQ of the transient calculation. Furthermore, the well-known space-time heterogeneous neutron kinetics benchmark C5G7 and its uncertainty propagation from the nuclear data to the interested key parameters of the core have been investigated. To address the problem of “the curse of dimensionality” caused by the large number of input parameters, the COST method has been applied to generate multivariate normal-distribution samples in uncertainty analysis. As a result, the law of the assembly/pin normalized power and their uncertainty with respect to time after introducing an instantaneous perturbation has been obtained. From the numerical results, it can be observed that the maximum relative uncertainties for the assembly normalized power can up to be about 1.65% and the value for the pin-wise power distributions can be about 2.71%.
MAPPER – A NOVEL CAPABILITY TO SUPPORT NUCLEAR MODEL VALIDATION AND MAPPING OF BIASES AND UNCERTAINTIES
Authors: Mertyurek U. , Abdel-Khalik H.S. , Marshall W.B.J.
Published: February 2021

This paper overviews the initial results of a new project at the Oak Ridge National Laboratory, supported via an internal seed funding program, to develop a novel computational capability for model validation: MAPPER. MAPPER will eliminate the need for empirical criteria such as the similarity indices often employed to identify applicable experiments for given application conditions. To achieve this, MAPPER uses an information-theoretic approach based on the Kullback-Leibler (KL) divergence principle to combine responses of available or planned experiments with application responses of interest. This is accomplished with a training set of samples generated using randomized experiment execution and application of high-fidelity analysis models. These samples are condensed using reduced order modeling techniques in the form of a joint probability distribution function (PDF) connecting each application response of interest with a new effective experimental response. MAPPER’s initial objective will be to support confirmation of criticality safety analysis of storage facilities which require known keff biases for safe operation. This paper reports some of the initial results obtained with MAPPER as applied to a set of critical experiments for which existing similarity-based methods have been shown to provide inaccurate estimates of the biases.
GENERALIZED SENSITIVITY ANALYSIS CAPABILITY WITH THE DIFFERENTIAL OPERATOR METHOD IN RMC CODE
Authors: Shi Guanlin , Jia Conglong , Wang Kan , Cheng Quan
Published: February 2021

Sensitivity analysis is an important way for us to know how the input parameters will affect the output of a system. Therefore, recently, there is an increased interest in developing sensitivity analysis methods in continuous-energy Monte Carlo Code due to the fact that Monte Carlo method can perform high-fidelity simulations of nuclear reactor. Previous studies mainly focused on developing sensitivity analysis method suitable for analyze eigenvalue. There are relatively few researches for performing sensitivity analysis of generalized response function by using continuous-energy Monte Carlo code. So, in this work, the differential operator method (DOM) has been investigated and implemented in continuous-energy Reactor Monte Carlo code (RMC) to perform sensitivity analysis of generalized response function in the form of ratios of reaction rate. The DOM implemented in RMC is based on the analog Monte Carlo transport mode and non-analog Monte Carlo transport mode. The correctness of the newly implemented method has been verified by comparing the results with those calculated by using the collision history-based method through the Jezeble and Flattop benchmark problems. In general, the results given by the DOM agree well with those obtained by the collision history-based method with an accuracy of 5%. Moreover, it is also shown that the non-analog Monte Carlo transport mode can obtain lower relative standard deviation of the sensitivity coefficients than the analog Monte Carlo transport mode.