bluuebunny/crossref_metadata_2025 huggingface dataset:
GENERATING ALBEDO BOUNDARY CONDITIONS FOR NODAL DIFFUSION CODES USING SERPENT 2
Authors: Valtavirta Ville , Kuopanportti Jaakko , Rintala Antti
Published: February 2021

We use the Serpent Monte Carlo code to produce total and partial albedo boundary conditions that can be used to model the Loviisa NPP VVER-440 core with the nodal neutronics tools of Fortum. The albedo generation process is described in detail. The dependence of the generated albedos on boron content and water density is investigated and a clear distinction is noted in water density dependence between regions containing mostly water and those containing mostly structural materials. The Serpent generated albedos are currently used in production calculations for modeling the Loviisa reactors at Fortum.
GPU ACCELERATION OF DOPPLER BROADENING FOR NEUTRON TRANSPORT CALCULATIONS1
Authors: Burke Paul E. , Remley Kyle E. , Griesheimer David P.
Published: February 2021

In radiation transport calculations, the effects of material temperature on neutron/nucleus interactions must be taken into account through Doppler broadening adjustments to the microscopic cross section data. Historically, Monte Carlo transport simulations have accounted for this temperature dependence by interpolating among precalculated Doppler broadened cross sections at a variety of temperatures. More recently, there has been much interest in on-the-fly Doppler broadening methods, where reference data is broadened on-demand during particle transport to any temperature. Unfortunately, Doppler broadening operations are expensive on traditional central processing unit (CPU) architectures, making on-the-fly Doppler broadening unaffordable without approximations or complex data preprocessing. This work considers the use of graphics processing unit (GPU)s, which excel at parallel data processing, for on-the-fly Doppler broadening in continuous-energy Monte Carlo simulations. Two methods are considered for the broadening operations – a GPU implementation of the standard SIGMA1 algorithm and a novel vectorized algorithm that leverages the convolution properties of the broadening operation in an attempt to expose additional parallelism. Numerical results demonstrate that similar cross section lookup throughput is obtained for on-the-fly broadening on a GPU as cross section lookup throughput with precomputed data on a CPU, implying that offloading Doppler broadening operations to a GPU may enable on-the-fly temperature treatment of cross sections without a noticeable reduction in cross section processing performance in Monte Carlo transport codes.
CONTINUOUS ENERGY COMET SOLUTION TO A SMALL MODULAR ADVANCED HIGH-TEMPERATURE REACTOR BENCHMARK PROBLEM (SmAHTR)
Authors: Rahnema Farzad , Zhang Dingkang
Published: February 2021

The continuous energy coarse mesh transport (COMET) method is a hybrid stochasticdeterministic solver that provides transport solutions to heterogeneous reactor cores. In this paper, COMET is tested against continuous energy Monte Carlo in solving the recently developed stylized Small Modular Advanced High-Temperature Reactor (SmAHTR) Benchmark Problems based on the Oak Ridge National Laboratory pre-conceptual design (core configurations). These problems are well-suited to test the performance of advanced neutronics tools because of their unique neutronics characteristics such as the multiple heterogeneities. The COMET solutions for the three benchmark problems were found to agree very well with the continuous energy Monte Carlo reference solutions. The discrepancy in the core eigenvalue (k-eff) varied from 40 pcm to 51 pcm. The average and maximum relative differences in the pin fission densities were in the range of 0.20% to 0.21% and 0.77% to 0.94%, respectively. It was also found that COMET was more than 2,000 times fast than MCNP. It can be concluded that COMET can model the SmAHTR core configuration with high fidelity and significantly high computational efficiency.
A HYBRID MONTE-CARLO-DETERMINISTIC METHOD FOR AP1000 EX-CORE DETECTOR RESPONSE SIMULATION
Authors: Zheng Qi , Shen Wei , Li Xuesong , Hao Tengfei , He Qingming , Li Jie , Liu Zhouyu
Published: February 2021

The ex-core detector-response calculation is a typical deep-penetration problem, which is challenging for the Monte Carlo method. The response of the ex-core detector is an important parameter for the safe operation of the nuclear power plants. Meanwhile, evaluation of the ex-core detector response during each step of fuel-loading is used to guide the fuel-loading sequence. The response can also be used to reconstruct core-power distribution for online monitoring of long-term power. The detector used for the ex-core response is the source-range detector which is sensitive to thermal neutrons. For a Monte Carlo shielding calculation of the above detector response, the thermal flux under 0.625eV is needed, which is too small to be tallied by traditional Monte Carlo simulations. In practice, the tally results are close to zero in the detector region under direct Monte Carlo calculation. Even if the number of particles is increased to a significant amount, the statistical variance is still very large. The high variance along with a significant calculation time leads to a small Figure Of Merit (FOM). In order to solve this problem and to improve the tally efficiency of the ex-core detector response, a hybrid Monte-Carlo-deterministic method is employed in this study, and an in-house hybrid Monte-Carlo-deterministic particle transport code, NECP-MCX, is developed in this paper. The method takes the space-energy-dependent adjoint fluxes to generate importance parameters for the mesh-based weight window in the Monte Carlo calculation. Simultaneously, the mesh-based source biasing is performed with the consistent importance parameters to make the starting weight of neutrons matching with the survival weight of the weight windows. As the mesh used in the hybrid Monte-Carlo-deterministic method is superimposed, the mesh of the weight window will not be affected by the complex geometry model. The adjoint flux is obtained by the efficient SN method with the multi-group cross-section data. The whole toolset is convenient to use with single set of the modelling data for both Monte Carlo and deterministic simulations. Compared with the direct Monte Carlo simulation, the hybrid Monte-Carlo-deterministic method has a higher efficiency for a typical deep-penetration problem such as the AP1000 ex-core detector-response simulation.
AN IMPROVED DOUBLE HETEROGENEITY MODEL FOR PEBBLE-BED REACTORS IN DRAGON-5
Authors: Chambon Richard , She Ding , Hébert Alain
Published: February 2021

In pebble-bed reactors, the fuel is contained in small grains, which are included in a graphite matrix. Some burnable poison particles may also be present. In this work, an additional ’double heterogeneity’ model is introduced in the DRAGON5 lattice code. The model is based on the legacy work of She (INET) and has been improved to overcome intrinsic limitations. It is based on a simplified physical model whereas the two already existing models were based on the collision probability analysis or on renewal theory. The advantage of this new model is its simplicity to implement. The theory shows that the correction suggested in the original model should not be arbitrary, but a constant equivalent particle fraction of 0.63. Numerical comparison between the models is generally good. However it does not support the theory of a constant equivalent fraction. Additional work is needed to reduce the discrepancy between the models in some cases.
NUMERICAL OPTIMIZATION OF A MULTIPHYSICS CALCULATION SCHEME
Authors: Cattaneo Paolo , Lenain Roland , Merle Elsa , Patricot Cyril , Schneider Didier
Published: February 2021

This work concerns the numerical optimization of a multiphysics calculation scheme. The considered application is a 5x5 Pressurized Water Reactor (PWR) assemblies mini-core surrounded by radial and axial reflectors. The scenario adopted for the analysis is steady-state nominal conditions and fission products set to the equilibrium concentration. The neutronics is modelled at the pin-cell scale and the thermal-hydraulics at the subchannel level. Depending on the scenario, the damped fixed-point algorithm might not be sufficiently robust or efficient enough. For this reason, a technique based on the partial convergence of the solvers is tested. In every multiphysic iteration, a maximum number of iterations is imposed for both the neutronics and the thermal-hydraulics solvers. In combination with that, the solver restarts from the results of the last calculation. In this way, if the method is convergent, the initialization progresses towards the fixed-point solution. The results show that such a technique improves both the robustness and the speed of the algorithm. Within this approach, the range of relaxation factors that makes the algorithm converge is significantly broadened and the importance of this parameter on the global performance is reduced. The computing time also decreases by a factor between 10 and 20. Furthermore, this gain does not strongly depend on the exact imposed maximum number of iterations. Some preliminary observations are also reported in respect with the application of such a technique to the Anderson acceleration method.
NEUTRONIC MODELING OF A FAST CORE WITH MODERATING MATERIALS USING APOLLO3R CODE
Authors: Zajaczkowski M. , Palau J.-M. , Pascal V. , Jean Saint C.de
Published: February 2021

In order to improve passive safety of Sodium-cooled Fast Reactors The French Alternative Energies and Atomic Energy Commission (CEA) has proposed a new core design called CADOR - an SFR core with enhanced Doppler reactivity feedback. One of its most important design features is the introduction of solid moderating materials inside each fuel assemblies to slightly decrease the average neutron energy. The article focuses on development and validation of a neutronics calculation scheme able to produce accurate results in case of CADOR and other fast cores with moderating materials. The study uses two different fuel assembly models moderated by metallic beryllium and zirconium hydride (ZrH2) respectively The study includes discussion of neutron scattering treatment and different ways of spatial homogenization and energy condensations. The results indicate that the accurate scattering treatment leads to much better estimation of Doppler constant, especially in case of ZrH2 moderated core. By using combined deterministic-Monte Carlo calculation scheme we are able to quantify the biases on global reactivity, reactivity feedbacks and control rod worth. We demonstrate that spatial homogenization plays a more important role in case of moderated CADOR assemblies and thus preserving certain level of heterogeneity within fuel assemblies can lower the calculation bias significantly.
NEUTRONIC MODELING STRATEGIES FOR A LIQUID FUEL TRANSIENT CALCULATION
Authors: Blanco J.A. , Rubiolo P. , Dumonteil E.
Published: February 2021

Framework

• A detailed and highly flexible numerical tool to study criticality accidents has been developed

• The tool implements a Multi-Physics coupling using neutronics, thermal-hydraulics and thermal-mechanics models based on Open FOAM and SERPENT codes

• Two neutronics models: Quasi-Static Monte Carlo and SPN

Objective:

In this work a system composed by a 2D square liquid fuel cavity filled with a fuel molten salt has been used to:

• Investigate the performance of the tool’s thermal-hydraulics and neutronics solvers coupling numerical scheme

• Evaluate possible strategies for the implementation of the Quasi-Static (QS) method with the Monte Carlo (MC) neutronics code

• Compare the QS-MC approach precision and computational cost against the Simplified P3 (SP3) method
HIGH FIDELITY AND REDUCED ORDER SOLUTIONS TO AN SMR-LEVEL PROGRESSION PROBLEM WITH THE KRAKEN COMPUTATIONAL FRAMEWORK
Authors: Valtavirta Ville , Lauranto Unna , Hovi Ville , Peltonen Jussi , Rintala Antti , Tuominen Riku , Leppänen Jaakko
Published: February 2021

Kraken is a new computational reactor analysis framework under development at VTT Technical Research Center of Finland Ltd. The framework builds heavily on the new generation of Finnish simulation codes such as the Serpent Monte Carlo code, the Ants nodal neutronics solver and the FINIX fuel behavior module.

This paper describes the application of Kraken to its first realistic full core problem, a small modular reactor (SMR) core in its fresh state in order to evaluate control rod worths, shutdown margins and reactivity coefficients for the core.
3-D COUPLED SIMULATION OF A VVER 1000 WITH PARCS/ATHLET
Authors: Henry Romain , Périn Yann , Velkov Kiril , Nikonov Sergei Pavlovich
Published: February 2021

A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems.

Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations, neither on the geometry nor in the nuclear data.

Since the whole core SERPENT models for production of such cross-section libraries would be computationally costly (and the standard 2-D approach may introduce unnecessary large approximations), 3-D models of each assembly type in infinite radial lattice configurations have been created.

These cross-sections are then used to evaluate effective multiplication factors for different core configurations with the diffusion code PARCS. The results are compared with the reference SERPENT calculations. In the next step, a thermal-hydraulic model with the system code ATHLET applying an assembly-wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transient test calculations.

This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, a simple exercise with coupled code system PARCS/ATHLET is presented and analysed.
A HIGH-FIDELITY SIMULATION OF THE C5G7 BENCHMARK BY USING THE PARALLEL ENTER CODE
Authors: Ruan Zhenglin , Guo Haibing
Published: February 2021

In simulation of advanced nuclear reactors, requirements like high precision, high efficiency and convenient to multi-physics coupling are putting forward. The deterministic transport method has the advantage of high efficiency, capable of obtaining detailed flux distribution and efficient in multi-physics coupling, but its accuracy is limited by the homogenized reaction cross-section data and core modelling exactness. The traditional two-steps homogenization strategy may introduce substantial deviation during the assembly calculation. It is possible to conduct a whole core deterministic transport simulation pin-by-pin to achieve higher accuracy, which eliminates the assembly homogenization process. The C5G7 benchmarks were proposed to test the ability of a modern deterministic transport code in analyzing whole core reactor problems without spatial homogenization. Different deterministic code that developed by different methods were applied to the benchmark simulation and some of them solved the benchmark accurately. However, there still exist some drawbacks in the given calculation processes which carried out by some other deterministic transport codes and we could find that the fuel pin cell in the assembly were not exactly geometrically modelled owing to the limit of the code. Consequently, the calculation precision could be improved by utilizing a high-fidelity geometry modelling. In this paper, the C5G7 benchmarks with different control rod position and different configuration were calculated by the finite element SN neutron transport code ENTER [1], and the results were presented after massively parallel computation on TIANHE-II supercomputer. By introducing a large scale high-fidelity unstructured meshes, high fidelity distributions of power and neutron flux were gained and compared with the results from other codes, excellent consistency were observed. To sum up, the ENTER code can meet those new requirements in simulation of advanced nuclear reactors and more works and researches will be implemented for a further improvement.
NEW REFERENCE APOLLO3® CALCULATION SCHEME FOR LIGHT WATER REACTORS – ANALYSIS OF THE BEAVRS BENCHMARK
Authors: Vidal Jean-François , Frölicher K. , Archier P. , Hébert A. , Buiron L. , Palau J-M. , Pastoris S. , Raynaud D.
Published: February 2021

In the past few years, developments in the APOLLO3® deterministic code have mainly been devoted to Fast Reactor applications. In this paper, we investigate the possibility of using some of these methods to build an accurate two-step calculation scheme for commercial Pressurized Water Reactors, with application to the BEAVRS benchmark at hot zero power conditions of cycle 1. Our objective is to assess the performances of the best “standard” calculation currently possible with APOLLO3® and to have a starting point for the development of improved transport solvers and innovative calculation schemes.

At the lattice level, we show that the subgroup method using the REL383 energy mesh, associated with a MOC flux calculation, provides accurate results on different clusters of 3x3 cells with UOX and MOX fuel, including a heterogeneity at the center (guide-tube full of water or with common absorbers Ag-In-Cd or B4C inserted, and mixed uranium-gadolinium oxide fuel). These good results have been confirmed on BEAVRS assembly, rods in and rods out.

At the core level, 20-group 3D calculations with the MINARET Sn solver have been performed at the cell level to analyze BEAVRS Hot Zero Power results (reactivity, power map, and control rods worths). Results are rather satisfactory, considering the low computing cost, but the power map prediction needs to be improved.
OVERVIEW OF THE OECD-NEA EXPERT GROUP ON MULTI-PHYSICS EXPERIMENTAL DATA, BENCHMARKS AND VALIDATION
Authors: Valentine Timothy , Avramova Maria , Fleming Michael , Hursin Mathieu , Ivanov Kostadin , Petruzzi Alessandro , Rohatgi Upendra , Velkov Kiril
Published: February 2021

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.
APPLICATION OF THE JSIR2S CODE PACKAGE FOR SHUTDOWN DOSE RATE CALCULATIONS ON JET
Authors: Ambrožič Klemen , Vilarri Rosaria , Batistoni Paola , Snoj Luka
Published: February 2021

In this paper we present a computational exercise for shut-down dose rate calculations for the JET tokamak using the in-house developed JSIR2S code package as part of its validation. The computation is performed in two parts: neutron transport and transport of secondary gamma radiation. In order to calculate neutron activation reaction rates with sufficiently low variance, hybrid variance reduction techniques using the ADVANTG code have been utilized. Probability based sampling of secondary source particles was performed. Calculated gamma dose rates after shut down are compared with dose rate measurements performed on site using ionization chambers. The C/E agreement for 1st octant is between 0.8 to 1 while statistically meaningfull results for the 2nd octant are yet to be obtained.
STUDY OF CAVITATION IN LIQUID SODIUM AND SIMULATION OF DYNAMIC CORE DEFORMATIONS
Authors: Houbar Sofiane , Gerschenfeld Antoine , Cyril Patricot
Published: February 2021

Due to the high confinement of the closed fuel assemblies in Sodium-cooled Fast Reactors (SFR), the liquid sodium filling the space between these assemblies may vaporize in case of an important mechanical excitation (cavitation process). That phenomenon would therefore induce a specific force feedback to the assemblies leading to a specific neutronic response of the core. Neutronic of SFR is indeed sensible to mechanical deformation. Hence the necessity to have a good prediction of the displacement of these mechanical structures in the framework of Fluid-Structure Interaction (FSI).

To study this kind of transient at a reactor scale, we propose in this paper a coupling procedure between the fluid and the structures. The fluid behaviour is investigated by Merkle’s three-equations model and simulated within a coarse mesh. The structures are modelled by a mass-spring system subjected to the fluid forces. In order to stabilize the numerical code coupling, a relaxation process is added and some results of this computational work are presented.
COUPLED MODELLING OF THE EBR-II SHRT-45R INCLUDING PHOTON HEAT DEPOSITION
Authors: Valerio Domenico , Abrate Nicolò , Dulla Sandra , Nallo Giuseppe Francesco , Piero Ravetto
Published: February 2021

The Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code has been developed during the last years at Politecnico di Torino, implementing a full-core coupled neutronic/thermal-hydraulics model for steady-state and transient analysis of liquid-metal cooled fast breeder reactor (LMFBR). In the framework of the validation activities for the code, an analysis of the sodium-cooled reactor EBR-II, previously carried out in the frame of a IAEA Coordinated Research Project, is performed with FRENETIC including the most recent physics models. In particular, photon transport and heat deposition are taken into account, a feature which has been proved in previous studies to be relevant to the correct study of the EBR-II core. To this purpose, a set of nuclear data for photons has been generated by means of the Monte Carlo code Serpent-2, and it is demonstrated that the code is able to take into account the photon heat deposition in the EBR-II.
ANALYSIS OF MODIFIED PASSIVE SAFETY SYSTEM IN FAST RECTORS TRANSIENTS
Authors: Oggioni Carlo , Keckler Chris , Fratoni Massimiliano
Published: February 2021

The Autonomous Reactivity Control (ARC) system is a passive safety system aiming to provide an additional negative reactivity feedback during reactor transient scenarios. This paper shows how the performance of the ARC system can be enhanced by introducing a hydraulic diode that allows for different engagement and disengagement speeds of system. The benefits of the proposed system is assessed in a reference soidum-cooled fast reactor (SFR) during multiple postulated transient scenarios. The reactor and plant dynamic response are evaluated using SAS4A/SASSYS-1, whereas for the internal ARC system fluid dynamics, the SAM code is adopted. The two codes are externally coupled using a custom driver script which coordinates data exchange and restart calculations at each time step, with Picard iterations used to converge each time step. All the transients analyzed in this work show that the enhanced ARC system is effective in reducing peak temperatures and in reducing the oscillatory behavior encountered in some cases with the standard ARC system.
TRANSIENT CALCULATIONS OF SPERT III EXPERIMENTS
Authors: Pautz Andreas , Zwermann Winfried
Published: February 2021

Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET. Homogenised 2-group cross sections for DYN3D are thereby generated with the Monte Carlo neutron transport code Serpent 2 in combination with the ENDF/B-VII.1 cross section library. Results in terms of maximum power, energy release, and reactivity compensation are in good agreement with the experimental values. The time-dependent contributions to the reactivity feedback are investigated for both a cold-startup test and a hot-standby test. These findings prove the suitability of the combined application of the simulation codes to predict the reactor dynamic behaviour in the event of prompt-critical and super-prompt critical transients even for small reactor cores. Furthermore, static core characteristics of the SPERT III E-core reactor at cold-startup condition are analysed with using a static DYN3D model, a detailed Serpent reference model, and a simplified Serpent model consistent with the DYN3D model. The critical control rod position and the excess reactivities of both the control rods and the transient rod obtained with the Serpent reference model are consistent with the experimental values. For the same parameters, the DYN3D model is in good agreement with the Serpent simplified model.
VALIDATION OF AXIAL VOID PROFILE MEASURED BY NEUTRON NOISE TECHNIQUES IN CROCUS
Authors: Hursin Mathieu , Pakari Oskari , Perret Gregory , Frajtag Pavel , Lamirand Vincent , Pázsit Imre , Dykin Victor , Por Gabor , Nylén Henrik , Pautz Andreas
Published: February 2021

Recently a joint project has been carried out between the Paul Scherrer Institut, the Ecole Polytechnique Federale de Lausanne and swissnuclear, an industrial partner, in order to determine the axial void distribution in a channel installed in the reflector of the zero power research reactor CROCUS, using neutron noise techniques. The main objective of the present paper is to report on the validation of the results against an alternative measurement technique using gamma-ray attenuation and simulations with the TRACE code. For the gamma-ray attenuation experiments, the channel used in CROCUS is installed out of the core in a Plexiglass water tank. The source and detector are fixed and the channel is moved axially to keep the geometry of the source/detector arrangement untouched. This is key to measure the void effect by gamma attenuation due to the low contrast of this technique. The paper compares the experimental results obtained with both techniques, with the outcomes of simulations carried out with the TRACE code. Even though the quantitative void fraction estimations are not consistent, the trends obtained with the simulation and experimental techniques are the same. The discrepancies between the various experimental techniques and the simulation outcomes are related to the heterogeneous distribution of the water-air mixture in the radial sections of the channel.
FIRST STEPS TO COUPLED HYDRAULIC AND MECHANICAL CALCULATIONS WITHIN A PARAMETER STUDY TO DEFINE POSSIBLE CORE DESIGNS FOR THE CONVERSION OF FRM II
Authors: Shehu Kaltrina , Bojanowski Cezary , Bergeron Aurelien , Petry Winfried , Reiter Christian
Published: February 2021

The Forschungs-Neutronenquelle Heinz Meier-Leibnitz (FRM II) is actively participating in the worldwide efforts on developing high-density uranium fuels in order to reduce the enrichment of fuels used in high flux research reactors. This work is part of a parameter study to define possible compatible FRM II core designs for conversion. As a first step, a code-to-code verification is performed and experimental data is used for validation. The Gambill experiment was performed in the early 1960’s in support of the HFIR program and provides results regarding the heat transfer coefficient and friction factors of water flowing through an electrically heated thin rectangular channel. A comparison is made between the Gambill Test and the results simulated by Ansys CFX and STAR-CCM+.
CONCEPTUAL FUEL ELEMENT DESIGN CANDIDATES FOR CONVERSION OF HIGH FLUX ISOTOPE REACTOR WITH LOW-ENRICHED URANIUM SILICIDE DISPERSION FUEL*
Authors: Chandler D. , Betzler B. R. , Bae J. W. , Cook D. H. , Ilas G.
Published: February 2021

Engineering design studies are underway to assess the feasibility of converting the High Flux Isotope Reactor (HFIR) to operate with low-enriched uranium silicide dispersion (LEU3Si2-Al) fuel. These studies are supported by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. A systematic approach employing neutronic and thermal-hydraulic analyses has been performed with the ORNL Shift and HFIR Steady State Heat Transfer Code tools, respectively, to predict reactor performance and thermal safety margins for proposed LEU3Si2-Al fuel designs. The design process was initiated by generating an optimized design with fabrication features identified from previous studies that result in excellent performance and safety metrics. The approach continued by substituting a single fabrication feature anticipated to be difficult to manufacture with another feature expected to perform an analogous function to that of the removed feature. Four conceptual fuel element design candidates, with various fabrication features, for conversion of HFIR with 4.8 gU/cm3 LEU3Si2-Al fuel have been generated and shown to meet pre-defined performance and safety metrics. Results to date indicate that HFIR could convert with the subject fuel system and meet performance and safety requirements if, among other considerations, fabrication of the specific design features are demonstrated and qualification of the fuel is complete under HFIR-specific conditions.
IMPACT OF VARIOUS SOURCE OF COVARIANCE INFORMATION ON INTEGRAL PARAMETERS UNCERTAINTY DURING DEPLETION CALCULATIONS WITH CASMO-5
Authors: Hursin Mathieu , Rochman Dimitri , Vasiliev Alexander , Ferroukhi Hakim , Pautz Andreas
Published: February 2021

This paper describes the effect of input uncertainties on a set of integral parameters (kinf, nuclide compositions) associated with the validation of CASMO-5 against PIE data. The nuclear data under consideration are the cross-sections, fission spectrum and neutron multiplicities and fission yields. Various sources of covariance information are considered, novel ones (ENDFB-VIII.0, JEFF-3.3) as well as more widely distributed ones (JENDL-4.0, ENDF/B-VII.1, Scale 6.1 and Scale 6.2). All possible nuclide reaction pairs (cross sections, fission spectrum and averaged number of neutron per fission) have been perturbed, e.g. all isotopes available in both the respective covariance libraries and the CASMO-5 library. The evolution of the uncertainty estimates with exposure is complemented with sensitivity analysis to determine the main contributors to the uncertainty. The Pearson coefficient defined between the model output and a given input is used in this work to assess the part of the variance in the model output coming from the considered input uncertainty. It is a very promising measure of sensitivity as it is computationally cheap even though it assumes linearity of the output with respect to input perturbations. The evolution of the uncertainty with exposure, both in terms of trends and magnitude are however very different. Sensitivity analysis allows determining why the trends and magnitudes are different.
BASIC STUDY ON A PSEUDO TREND PHENOMENON OF THE FEYNMAN-α ANALYSIS WITH BUNCHING METHOD
Authors: Hohara Sin-ya , Sakon Atsushi , Endo Tomohiro , Sano Tadafumi , Nakajima Kunihiro , Takahashi Kazuki , Hashimoto Kengo
Published: February 2021

In these years, reactor noise analysis methods have been studied to apply for the Debris’ criticality management at the Fukushima Daiichi NPP, Japan. The Feynman-α analysis with bunching method is one of the candidate techniques, however the bunching method itself has never been validated in detail. This synthesis technique is useful to reduce a time required for the experiment, however it is known that a non-physical trend unrelated to the state of a nuclear reactor is generated by the multiple use of time series data, and this phenomenon (we call “pseudo trend phenomenon”) has never been systematically studied in detail. In this study, Poisson events, whose statistical characteristics were clarified, were employed to investigate the pseudo trend phenomenon of the bunching method. The time-sequence count data for various statistical parameters were generated by the Monte Carlo time series simulator. Comparing the two results obtained by applying the conventional bunching method and the moving-bunching method for the same Poisson event time series, and it was found that the same pseudo trend component appears in both results of the bunching method and the moving bunching method. In addition, it was also found that the fluctuation width of the pseudo trend component is smaller than the statistical fluctuation range.
ON-THE-FLY INTERPOLATION OF CONTINUOUS TEMPERATURE-DEPENDENT THERMAL NEUTRON SCATTERING DATA IN RMC CODE
Authors: Zheng Lei , Feng Zhiyuan , Wang Kan
Published: February 2021

Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future.
CRITICAL EXPERIMENT OF THORIUM LOADED THERMAL CORES AT KUCA (1) A NEW CRITICAL EXPERIMENT OF THORIUM LOADED CORE WITH HARDER NEUTRON SPECTRUM IN KUCA
Authors: Sano Tadafumi , Hori Jun-ichi , Lee Jeaong , Takahashi Yoshiyuki , Takahashi Kazuki , Unesaki Hironobu
Published: February 2021

In order to perform integral evaluation of 232Th capture cross section, a series of critical experiments for thorium-loaded and solid-moderated cores in KUCA had been carried out. In these experimental cores, H/235U nuclide ratio ranged about from 150 to 315, and 232Th/235U nuclide ratio ranged about from 13 to 19. In this study, a new critical experiment with Th loaded core in KUCA, which had about 70 of the H/235U ratio and 12.7 of 232Th/235U ratio, was carried out. As results, the excess reactivity was 0.086 ± 0.003 (% dk/k) and the keff was 1.0009 ± 0.0003, where the effective delayed neutron fraction was 7.656E-3. The keff was also calculated by MVP3.0 with different nuclear libraries. The respective calculations with JENDL-4.0, JENDL-3.3 and ENDF/B-VII.0 lead to 1.0056 ± 0.0086 (%), 1.0048 ± 0.0085 (%) and 1.0056 ± 0.0086 (%).On the other hand, the further MVP3.0 calculations, where only the 232Th cross sections were taken from JENDL-4.0, JENDL-3.3 or ENDF/B-VII.0 but all other nuclides were done from JENDL-4.0, were carried out to examine an impact of the difference of 232Th cross section among these nuclear libraries to the keff. The keff calculated with respective 232Th cross sections from JENDL-3.3 and ENDF/B-VII.0 was 1.0038 ± 0.0086 (%) and 1.0040 ± 0.0086 (%).
RE-EVALUATION OF THE TSL FOR YTTRIUM HYDRIDE
Authors: Zerkle Michael L. , Holmes Jesse C. , Wormald Jonathan L.
Published: February 2021

Yttrium hydride (YHx) is of interest as a high-temperature moderator material in advanced nuclear reactor systems because of its superior ability to retain hydrogen at elevated temperatures. Thermal neutron scattering laws (TSL) for hydrogen bound in yttrium hydride (H-YH2) and yttrium bound in yttrium hydride (Y-YH2) were previously evaluated by Naval Nuclear Laboratory using the ab initio approach and released in ENDF/B-VIII.0. In that work, density functional theory, incorporating the generalized gradient approximation (GGA) for the exchange-correlation energy, was used to simulate the face-centered cubic structure of YH2 and calculate the interatomic Hellmann-Feynman forces for a 2×2×2 supercell containing 96 atoms. Lattice dynamics calculations using PHONON were used to determine the phonon density of states. The calculated phonon density of states for H and Y in YH2 were then used to prepare H-YH2 and Y-YH2 TSL evaluations, in the incoherent approximation, using the LEAPR module of NJOY. In addition elastic scattering was assumed to be incoherent for both H and Y. While the incoherent elastic scattering approximation is appropriate for H-YH2, it introduces an undesirable approximation for Y-YH2. In this work, we re-evaluate the TSL for Y-YH2 using FLASSH (Full Law Analysis Scattering System Hub). Y-YH2 is evaluated using the FLASSH generalized coherent elastic scattering capability in order to capture the Bragg peaks associated with the YH2 crystal structure which were neglected in the prior NJOY-based evaluation due to limitations in LEAPR. An experimental approach to validate the Y-YH2 TSL using neutron transmission measurements is discussed.
DATA ASSIMILATION APPLIED TO PRESSURISED WATER REACTORS
Authors: Dixon J.R , Lindley B.A. , Taylor T. , Parks G.T.
Published: February 2021

Best estimate plus uncertainty is the leading methodology to validate existing safety margins. It remains a challenge to develop and license these approaches, in part due to the high dimensionality of system codes. Uncertainty quantification is an active area of research to develop appropriate methods for propagating uncertainties, offering greater scientific reason, dimensionality reduction and minimising reliance on expert judgement. Inverse uncertainty quantification is required to infer a best estimate back on the input parameters and reduce the uncertainties, but it is challenging to capture the full covariance and sensitivity matrices. Bayesian inverse strategies remain attractive due to their predictive modelling and reduced uncertainty capabilities, leading to dramatic model improvements and validation of experiments. This paper uses state-of-the-art data assimilation techniques to obtain a best estimate of parameters critical to plant safety. Data assimilation can combine computational, benchmark and experimental measurements, propagate sparse covariance and sensitivity matrices, treat non-linear applications and accommodate discrepancies. The methodology is further demonstrated through application to hot zero power tests in a pressurised water reactor (PWR) performed using the BEAVRS benchmark with Latin hypercube sampling of reactor parameters to determine responses. WIMS 11 (dv23) and PANTHER (V.5:6:4) are used as the coupled neutronics and thermal-hydraulics codes; both are used extensively to model PWRs. Results demonstrate updated best estimate parameters and reduced uncertainties, with comparisons between posterior distributions generated using maximum entropy principle and cost functional minimisation techniques illustrated in recent conferences. Future work will improve the Bayesian inverse framework with the introduction of higher-order sensitivities.
THERMAL SCATTERING LAW ENDF LIBRARIES FOR LIQUID FLIBE
Authors: Zhu Y. , Manring C. A. , Hawari A. I.
Published: February 2021

Several advanced nuclear reactor concepts have been proposed in the past few years where FLiBe molten salt represents a major constituent of the core. In this case, neutrons produced in fission slow down and moderate in FLiBe (a eutectic with a mixture of 2:1 ratio of LiF and BeF2) until they reach low energies (i.e, below 1 eV). At that stage, the thermalization process becomes dominant and the neutrons achieve a quasi-equilibrium energy state that is dependent on the temperature of the moderator. In neutronic simulations, the description of neutron thermalization is captured using the thermal scattering law (TSL), i.e., S(α,β), of the material in which low energy neutrons are interacting. S(α,β) defines the energy-momentum phase space that is available for an incoming low energy neutron. In addition, it is directly proportional to the double differential thermal neutron scattering cross section. In this work, the TSL of molten salt FLiBe is developed based on a generalized density of excitation states (GDOS) derived from atomic trajectories generated using classical molecular dynamics (MD) simulations that were performed with the LAMMPS code. The MD simulations utilized a Born-Mayer type atomic potential function that was verified to reproduce the properties of FLiBe including density and viscosity. The FLASSH code was used to evaluate the TSL’s ENDF File 7 in a temperature range extending from 773 K to 1673 K. In addition, ACE type cross section libraries are produced and tested with the objective of contributing the data to the National Nuclear Data Center for inclusion in the ENDF/B-VIII database.
IMPACT OF NUCLEAR DATA LIBRARIES ON PREDICTED FAST REACTOR PERFORMANCE
Authors: Nelson A.G. , Ramey K.M. , Heidet F.
Published: February 2021

The nuclear data evaluation process inherently yields a nuclear data set designed to produce accurate results for the neutron energy spectra corresponding to a specific benchmark suite of experiments. When studying reactors with spectral conditions outside of, or not well represented by, the experimental database used to evaluate the nuclear data, care should be given to the relevance of the nuclear data used. In such cases, larger biases or uncertainties may be present than in a reactor with well-represented spectra. The motivation of this work is to understand the magnitude of differences between recent nuclear data libraries to provide estimates for expected variability in criticality and power distribution results for sodiumcooled, steel-reflected, metal-fueled fast reactor designs.

This work was specifically performed by creating a 3D OpenMC model of a sodium-cooled, steel-reflected, metal-fueled fast reactor similar to the FASTER design but without a thermal test region. This OpenMC model was used to compare the differences in eigenvalues, reactivity coefficients, and the spatial and energetic effects on flux and power distributions between the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 nuclear data libraries. These investigations have revealed that reactivity differences between the above libraries can vary by nearly 900 pcm and the fine-group fluxes can vary by up to 18% in individual groups. Results also show a strong variation in the flux and power distributions near the fuel/reflector interface due to the high variability in the 56Fe cross sections in the libraries examined. This indicates that core design efforts of a sodium-cooled, steel-reflected, metalfueled reactor will require the application of relatively large nuclear data uncertainties and/or the development of a representative benchmark-quality experiment.
IMPROVED THERMAL SCATTERING DATA PREPARATION
Authors: Trainer Amelia , Forget Benoit , Conlin Jeremy
Published: February 2021

Convenient access to accurate nuclear data, particularly data describing low-energy neutrons, is crucial for trustworthy simulations of thermal nuclear systems. Obtaining the scattering kernel for thermal neutrons (i.e., neutrons with energy ~1 eV or less) can be a difficult problem, since the neutron energy is not sufficient to break molecular bonds, and thus the neutrons must often interact with a much larger structure. The “scattering law” S(α; β), which is a function of unitless momentum α and energy β transfer, is used to relate the material’s phonon frequency distribution to the scattering kernel. LEAPR (a module of NJOY) and GASKET are two nuclear data processing codes that can be used to prepare the scattering law and use different approaches to approximate the same equations. LEAPR uses the “phonon expansion method” which involves iterative convolution. Iteratively solving convolution integrals is an expensive calculation to perform (to ease this calculation, LEAPR uses trapezoidal integration for the convolution). GASKET uses a more direct approach that, while avoiding the iterative convolutions, can become numerically unstable for some α; β combinations. When both methods are properly converged, they tend to agree quite well. The agreement and departure from agreement is presented here.
KRITZ-1-Mk CRITICAL MEASUREMENTS AT TEMPERATURES FROM 20 °C TO 250 °C
Authors: Mennerdahl Dennis
Published: February 2021

Benchmarks are needed to validate methods to account for temperature-dependence of nuclear data. An evaluation of 37 KRITZ-1-Mk critical water height measurements, together with associated iso-reactivity temperature effects and coefficients, is released with the 2019 Handbook of the International Reactor Physics Experiment Evaluation Project (IRPhEP). The KRITZ zero-power research reactor, operated between 1969 and 1975 in Studsvik (Sweden), was contained in a pressure vessel, allowing full size fuel assemblies or fuel rods in light water at temperatures up to 250 °C without boiling. Preliminary results were published in 1971 and 1972 for four series of altogether 37 measurements with Marviken (Boiling Heavy Water Reactor) UO2 fuel rods, each containing a 235U isotopic mass fraction of 1.35 %. Temperature was the predictor variable, while critical water height was the response variable. Each series was characterized by the fuel rod lattice design and by the soluble boron concentration in water. The KRITZ measurements were focused on temperature-dependence (differences). High measurement correlations reduced the ?k uncertainties, typically from 195 pcm to 40 pcm for a large temperature change. Thermal expansion of fuel and reactor components was not measured. Detailed and simple benchmarks include estimated thermal expansion as a simplification. Benchmark calculation results using JEFF-3.3 nuclear data reduce the large biases observed for older libraries but a remarkable positive temperature trend is observed for series 4. In 2019, Studsvik Nuclear released information on KRITZ-1-Mk and on other KRITZ-1 and KRITZ-2 critical measurements with Boiling Water Reactor fuel assemblies and fuel rod clusters.