bluuebunny/crossref_metadata_2025 huggingface dataset:
IMPLEMENTATION AND COMPARISON OF LATTICE MODULE AND DISCONTINUITY FACTORS MODULE IN PROTEUS TOOLSETS
Authors: Ni Kan , Hou Jason , Avramova Maria
Published: February 2021

State-of-the-art core nodal diffusion calculation involves the use of assembly discontinuity factor (ADF) to improve computational accuracy by introducing degree of freedom describing the relationship between interfacial discontinuities in nodal calculation [1]. The form of ADF known as the Flux based ADF (FDF) generated from flux information is recommended in the conventional two-level core calculation scheme. The multi-group cross-sections were generated using SCALE 6.2 NEWT and verified with KENO-VI [2]. A lattice module has been created for the high-fidelity neutron transport code MOCEX [3] to generate the group constants and side-independent ADFs. This new capability is verified against the reference code SCALE 6.2 NEWT under both serial and parallel modes. The implementation of ADF is performed in this work and further verified by comparing core keff. The calculation results show that the newly implemented ADF module consistently improved the accuracy of the PROTEUS-NODAL (NODAL) diffusion solver, which will become an affordable candidate for the following research of High-to-Low (Hi2Lo) transport-to-diffusion informing scheme [4].
ON THE FLY PREDICTION OF TH-DEPENDENT SPATIAL MACROSCOPIC CROSS-SECTIONS USING FFT
Authors: Terlizzi S. , Kotlyar D.
Published: February 2021

Monte Carlo (MC) codes can accurately model neutron transport in nuclear reactors. However, the efficient inclusion of thermal-hydraulic (TH) feedback within the MC calculation sequence is still an open problem, particularly when burnup’s time-evolution must be included in the analysis. For this reason, deterministic codes, leveraging the use of macroscopic cross-sections generated with higher order methods from 2D lattice calculations, are still widely used to perform reduced-order multiphysics analyses. However, traditional cross-sections generation procedures typically decompose the large core problem into multiple assembly-level problems; thus not having the ability to capture inter-nodal effects. Moreover, the pre-generation procedure requires additional pre-computational time to perturb/branch the problem for various operational conditions (e.g. fuel temperature), which, again, is decoupled from the core. In this paper, we propose a new method leveraging the use of Fourier transfer functions to predict the cross-sections distribution due to a variation in TH conditions. The method was tested against a 3D BWR unit-cell problem with realistic density profile and axial fuel heterogeneity. The method was able to compute the mono-energetic cross-sections distribution with maximum error lower than 2%. Insights on the influence of the statistics used to generate the cross-sections on the accuracy of the results is also provided.
DEMONSTRATION OF A LINEAR PROLONGATION CMFD METHOD ON MOC
Authors: Li Jin , Xu Yunlin , Wang Dean , Shen Qicang , Kochunas Brendan , Downar Thomas
Published: February 2021

Coarse Mesh Finite Difference (CMFD) method is a very effective method to accelerate the iterations for neutron transport calculation. But it can degrade and even fail when the optical thickness of the mesh becomes large. Therefore several methods, including partial current-based CMFD (pCMFD) and optimally diffusive CMFD (odCMFD), have been proposed to stabilize the conventional CMFD method. Recently, a category of “higherorder” prolongation CMFD (hpCMFD) methods was proposed to use both the local and neighboring coarse mesh fluxes to update the fine cell flux, which can solve the fine cell scalar flux discontinuity problem between the fine cells at the bounary of the coarse mesh. One of the hpCMFD methods, refered as lpCMFD, was proposed to use a linear prolongation to update the fine cell scalar fluxes.

Method of Characteristics (MOC) is a very popular method to solve neutron transport equations. In this paper, lpCMFD is applied on the MOC code MPACT for a variety of fine meshes. A track-based centroids calculation method is introduced to find the centroids coordinates for random shapes of fine cells. And the numerical results of a 2D C5G7 problem are provided to demonstrate the stability and efficiency of lpCMFD method on MOC. It shows that lpCMFD can stabilize the CMFD iterations in MOC method effectively and lpCMFD method performs better than odCMFD on reducing the outer MOC iterations.
IMPLEMENTATION OF QUADRATIC AXIAL TRIAL FUNCTIONS IN THE HIGH-FIDELITY TRANSPORT CODE PROTEUS-MOC
Authors: Zhang Guangchun , Yang Won Sik
Published: February 2021

PROTEUS-MOC is a pin-resolved high-fidelity transport code, in which the axial variation of angular flux is represented in terms of orthogonal polynomials. Currently, PROTEUS-MOC employs linear functions and requires relatively fine axial meshes to achieve high accuracy, which increases the number of axial meshes and hence the memory requirement. In this study, aiming to reduce the memory requirement and potentially the computational time by allowing larger axial meshes, we have extended the PROTEUS-MOC transport solution method to quadratic trial functions. Preliminary tests for the performance of quadratic trial functions have been performed using the 3-D C5G7 benchmark problem. Test results showed that for the same axial mesh configuration with relatively large sizes, the quadratic approximation yields about 2 to 5 times more accurate pin powers than the linear approximation, depending on the degree of axial variation of angular fluxes. The quadratic approximation also allows the use of about 3 times coarser axial meshes than the linear approximation for comparable pin power accuracy, which consequently reduces the memory requirement by about 2 times. The memory reduction is not proportional because of the increased number of coefficients in each element from 2 to 3. However, the quadratic approximation did not reduce the computational time as expected because of the deteriorated performance of the pCMFD acceleration scheme due to large axial mesh sizes.
APPLICATION OF THE SPH EQUIVALENCE TECHNIQUE TO THE CABRI FULL CORE IN NON-FUNDAMENTAL MODE
Authors: Hébert Alain , Taforeau Julien , Ingremeau Jean-Jacques
Published: February 2021

We developed a SPH equivalence technique in non-fundamental mode condition between a CABRI full-core model solved with the method of characteristics (MOC) in 2D and a simplified full-core model solved with the simplified P3 (SP3) method, linear anisotropic sources and discretized with Raviart-Thomas finite elements over a pure Cartesian mesh. The MOC and SP3 calculations are performed with DRAGON5 and DONJON5 codes, respectively. A three-parameter database is generated by DRAGON5 and is interpolated in DONJON5 as a function of the core condition. An objective function is set as the root mean square (RMS) error (MOC-SP3 discrepancy) on absorption distribution and leakage rates defined over the macro-geometry in DONJON5. Our algorithm is a quasi-Newtonian gradient search based on the Limited memory Broyden-Fletcher-Goldfarb-Shanno (LBFGS) method. Numerical results are presented with Hafnium bars withdrawn or inserted.
THE DOUNREAY FAST REACTOR VAULT CHARACTERISATION USING THE MONTE CARLO N-PARTICLE CODE AS A TOOL TO SUPPORT DECOMMISSIONING PLANNING
Authors: Baldioli Vittoria , Gunn Richard , MacDonald Sanders , Mackay Aly
Published: February 2021

A combined empirical and theoretical characterisation approach was taken in order to investigate the activity distribution and inventory in the Dounreay Fast Reactor (DFR) vault, so as to aid with the decommissioning and waste management strategy of the cooling circuits located within the shielded vault. The DFR vault is a no-man access, shielded void, which surrounds the reactor core and the graphite shield.

Empirical data was collected through various Non-Destructive Assay methods, utilising all available accesses into the reactor vault. This collection of information was then used to make a first assumption about the activity distribution within the vault; this was then included into a MCNP model of the vault as the source terms.

The model was run and a dose rate map of the vault obtained, and compared to the empirical dose rate data previously gathered in the accessible penetrations in the vault; the model was altered until a satisfactory agreement was obtained between theoretical and measured data.

Once this was obtained then, the model was used to provide an estimate of the activity inventory within the vault, and specifically within the various reactor components.
SERPENT/SUBCHANFLOW COUPLED BURNUP CALCULATIONS FOR VVER FUEL ASSEMBLIES
Authors: Ferraro Diego , García Manuel , Imke Uwe , Valtavirta Ville , Tuominen Riku , Leppänen Jaakko , Sanchez-Espinoza Víctor
Published: February 2021

The continuous improvement in nuclear industry safety standards and reactor designers’ and operators’ commercial goals represent a push for the development of highly accurate methodologies in reactor physics. This fact, combined with the availability of vast computational resources, allowed the development of a wide range of coupled state-of-the-art neutronic-thermal-hydraulic calculation tools worldwide during last decade. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes, suitable for high-fidelity calculations for PWR and VVER reactors. This work presents the results for a pin-by-pin coupled burnup calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel thermal-hydraulics code SUBCHANFLOW (SCF, developed by KIT, Germany) for two different VVER-type fuel assembly types. For such purpose, a recently refurbished master-slave coupling scheme is considered, which provides several new features such as burnup and transient calculations capabilities for square and hexagonal geometries. Main aspects of this coupling are presented for this burnup case, showing some of the capabilities now available. On top of that, the obtained global results are compared with available published data from a similar high-fidelity approach for the same FA design, showing a good agreement. Finally, a brief analysis of the main resources requirement and main bottlenecks identification are also included. The results presented here provide valuable insights and pave the way to tackle the final goals of the McSAFE project, which includes full-core pin-by-pin depletion calculation within a fully coupled MC-TH approach.
PRELIMINARY INVESTIGATIONS OF TRANSPORT IN HETEROGENEOUS RANDOM MEDIA
Authors: Larmier C. , Mazzolo A. , Zoia A. , Lemaire S. , Riz D.
Published: February 2021

Random media emerge in several applications in reactor physics and safety analysis. Most often, models of stochastic media assume spatial homogeneity, whereas real-world complex materials, such as fuel chunks resulting from core degradation, typically display apparent heterogeneities. In a series of previous works, we have shown that stochastic tessellations can be successfully used in order to describe the material properties of several classes of random media. In this paper we extend these results to the case of heterogeneous random media by using Voronoi tessellations with space-dependent seed distributions, allowing for spatial gradients.
csg2csg: A TOOL TO ASSIST VALIDATION & VERIFICATION
Authors: Davis Andrew , Lilley Steven , Shimwell Jonthan
Published: February 2021

The csg2csg tool is a minimal dependency Python3 based program to facilitate the comparison of radiation transport Monte Carlo codes. The csg2csg can currently parse MCNP and OpenMC geometry descriptions, and it can subsequently export MCNP, FLUKA, PHITS, OpenMC and Serpent geometry files. This translation is achieved by reading the input geometry into a class based system of objects corresponding to surfaces, cells, materials and so on. The objects are then translated into a ‘generic’ form. This form once generalised can be converted to any of the supported code formats and then radiation transport calculations can be performed. The code has been tested on several geometries; most notably a complex JET model and a 70k cell ITER C-Model geometry.
ANALYSIS OF THE OECD/NEA SFR BENCHMARK WITH ANTS REDUCED-ORDER NODAL DIFFUSION SOLVER AND THE SERPENT MONTE CARLO CODE
Authors: Szogradi Marton
Published: February 2021

In order to meet modern industrial and scientific demands the Kraken multi-physics platform’s development was recently launched at VTT Technical Research Centre of Finland. The neutronic solver of the framework consists of two calculation chains, providing full core solutions by the Serpent high fidelity code (1) and the AFEN/FENM-based reduced-order diffusion solver called Ants (2) capable of handling square and hexagonal geometries in steady-state. Present work introduces the simulation of a large 3600 MWth Sodium-cooled Fast Reactor (SFR) described within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of OECD. Full-core 3D results were obtained by Serpent for carbide- and oxide-fuel cores, moreover group constants were generated for Ants utilizing 2D super-cell and single assembly infinite lattice models of Serpent. The continuous-energy Monte Carlo method provided the reference results for the verification of the reduced-order method. Implementing the spatially homogenized properties, 3D solutions were obtained by Ants as well for both core configurations. Comparison was made between the various core designs and codes based on reactivity feedbacks (Doppler constant, sodium voiding, control rod worth) considering power distributions. Regarding reactivity sensitivity on geometry, axial fuel- and radial core expansion coefficients were evaluated as well.
COUNTERING NEUTRON CLUSTERING IN MONTE CARLO WITH A NEUTRON SOURCE INJECTION
Authors: Cosgrove P. , Kowalski M.A. , Shwageraus E. , Parks G.T.
Published: February 2021

Neutron clustering is a recently identified problem with Monte Carlo eigenvalue calculations which can produce significantly erroneous results. Previous work by Sutton & Mittal (2017) considered neutron clustering as a problem of maintaining ‘genetic diversity’ within the neutron population. This paper proposes reducing the extent of neutron clustering by replacing fission neutrons in the source bank with uncorrelated neutrons, sampled from a uniform source distribution – effectively adding new neutron genealogies to the population. The efficacy of the method is demonstrated on a number of simple problems, showing improved behaviour of the Shannon entropy and neutron centre-of-mass. Although currently limited in scope, this paper intends to provide a route to reducing clustering effects in more general problems.
A FULL REFERENCE APOLLO3® DETERMINISTIC SCHEME FOR THE JHR MATERIAL TESTING REACTOR
Authors: Lebreton Matthieu , Politello Julien , Vidal Jean-François , Rimpault Gérald
Published: February 2021

JHR is a new material testing reactor under construction at CEA Cadarache. Its high flux core contains 37 fuel assemblies loaded along concentric rings into alveolus of an aluminum matrix. For the operation of the reactor, twenty-seven of these fuel assemblies hovnst hafnium rods in their center while the other ones but also the beryllium radial reflector can accommodate experimental devices. In order to accurately predict its operating core characteristics but also its irradiation performance, a recently developed scheme based on the APOLLO3® platform is being developed which uses the sub-group method for spatial self-shielding, the 2D method of characteristics and the 3D unstructured conform MINARET Sn transport solver.

A 2D model of JHR has been built and optimized for calculating, at the lattice step, the self-shielded and condensed cross sections thanks to the sub-group method and the method of characteristics. Results are benchmarked against a TRIPOLI-4® stochastic reference calculation. A more refined spatial mesh gives better results on fission rates and reactivity compared to the ones of the former APOLLO2 scheme.

The classical 2-step calculations use the hypothesis of infinite lattice configuration, which is reasonable for the assemblies close to the center but not for peripheral ones. Hence, a new approach is being set up taking into account the surrounding of each assembly. The newly 3-step scheme uses the Sn solver MINARET and gives better results than the traditional 2-step scheme. This approach will be applied to a 3D modelling of the heterogeneous JHR core configurations incorporating experimental devices and enabling burn up calculations.
WHOLE CORE COUPLING METHODOLOGIES WITHIN WIMS
Authors: Tollit Brendan , Charles Alan , Poole William , Cox Andrew , Hosking Glynn , Lindley Ben , Smith Peter , Smethurst Andy , Lavarenne Jean
Published: February 2021

The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation.

This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions.

The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code.
A MULTIPHYSICS SIMULATION SUITE FOR SODIUM COOLED FAST REACTORS
Authors: Dawn William C. , Palmtag Scott
Published: February 2021

A simulation suite has been developed to model reactor power distribution and multiphysics feedback effects in Sodium-cooled Fast Reactors (SFRs). This suite is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified P3 (SP3) neutron transport equations. In the FEM implementation, two-dimensional triangular elements and three-dimensional wedge elements are selected. Wedge elements are selected for their natural description of hexagonal geometry common to fast reactors.

Thermal feedback effects within fast reactors are modeled within the simulation suite. A thermal hydraulic model is developed, modeling both axial heat convection and radial heat conduction within fuel assemblies. A thermal expansion model is included and is demonstrated to significantly affect reactivity.

This simulation suite has been employed to model the Advanced Burner Reactor (ABR) benchmark, specifically the MET-1000. It has been demonstrated that these models sufficiently describe the multiphysics feedback phenomena and can be used to estimate multiphysics reactivity feedback coefficients.
PBMR-400 BENCHMARK SOLUTION OF EXERCISE 1 AND 2 USING THE MOOSE BASED APPLICATIONS: MAMMOTH, PRONGHORN
Authors: Balestra Paolo , Schunert Sebastian , Carlsen Robert W , Novak April J , DeHart Mark D , Martineau Richard C
Published: February 2021

High temperature gas cooled reactors (HTGR) are a candidate for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and online refueling. Pebble bed reactors pose unique challenges to analysts and reactor designers such as continuous burnup distribution depending on pebble motion and recirculation, radiative heat transfer across a variety of gas-filled gaps, and long design basis transients such as pressurized and depressurized loss of forced circulation. Modeling and simulation is essential for both the PBR’s safety case and design process. In order to verify and validate the new generation codes the Nuclear Energy Agency (NEA) Data bank provide a set of benchmarks data together with solutions calculated by the participants using the state of the art codes of that time. An important milestone to test the new PBR simulation codes is the OECD NEA PBMR-400 benchmark which includes thermal hydraulic and neutron kinetic standalone exercises as well as coupled exercises and transients scenarios. In this work, the reactor multiphysics code MAMMOTH and the thermal hydraulics code Pronghorn, both developed by the Idaho National Laboratory (INL) within the multiphysics object-oriented simulation environment (MOOSE), have been used to solve Phase 1 exercises 1 and 2 of the PBMR-400 benchmark. The steady state results are in agreement with the other participants’ solutions demonstrating the adequacy of MAMMOTH and Pronghorn for simulating PBRs.
IMPROVEMENTS TO THE MODELING OF THE TREAT REACTOR AND EXPERIMENTS
Authors: Ortensi J , Wang Y , Labouré V , Gleicher FN , Schunert S , DeHart MD
Published: February 2021

This paper summarizes the latest improvements and lessons learned from the modeling and simulation of the transient test reactor at Idaho National Laboratory using the MAMMOTH reactor physics application. MAMMOTH is a MOOSE-based, Finite Element Method application that specializes in the analysis of the spatial dynamics behavior of nuclear reactors. Since early 2018 several transient tests have been conducted at TREAT, thus providing the opportunity to apply and benchmark modern modeling and simulation tools. MAMMOTH was used to provide predictions of the power coupling factor between the core and the experiment for various experiments. Even though the power coupling factor predictions agree very well with the experimental data, within the bounds of the experimental uncertainty, one shortcoming was the underprediction of the total energy deposited in the core and experiment. Determination of the sources for this discrepancy is ongoing, but several key problems have been identified and resolved, thus providing valuable insights for future research. This paper discusses several of these lessons learned. First, the heat capacity data for the TREAT fuel has some significant problems due to limitations of the measurement techniques used circa 1960s. The sensitivity of the peak power and the total energy deposition to various representations of the heat capacity is approximately 5%. Second, the effects of the biological shield and thermal column on the modeling of the core are non-negligible, since they affect the mean generation time and the effective reflection of neutrons back into the core, which is suspected to be important during the core heat up. Matching the reactor period resolves the fact that the reduced spatial domain used in the MAMMOTH model underpredicts the mean generation time. The neutron reflection from these regions is marginally improved with the use of an albedo boundary condition. Third, modeling of the control rod movement with a multi-scheme method is introduced and its current limitations are exposed. Fourth, we explore the effects of using a homogenized model with Superhomogenization equivalence and how that differs from fully heterogeneous simulations. Finally, the energy condensation effects for this graphite core are significant. Solutions with 10 and 26 energy groups show the benefits of using a finer coarse group structure.
DEVELOPMENT AND VERIFICATION OF T-TRACE/PANTHER COUPLED CODE
Authors: Abarca A. , Avramova M. , Ivanov K. , Verdebout S. , De Meyer D. , Schneidesch C. R.
Published: February 2021

Multi-physics coupled simulations have become increasingly important during the last two decades being one of the major field of application in the nuclear technology. The nuclear reactors themselves are complex systems whose responses are driven by interactions between neutron kinetics, thermal-hydraulics, heat transfer, mechanics and chemistry. Probably, in a nuclear system, the most complex and important feedback effect takes place between the core neutron kinetics and thermal-hydraulics. The development of coupled thermal-hydraulic -neutron kinetics codes is a recurrent field of research for the nuclear industry.

This contribution, developed in the Consortium for Nuclear Power (CNP) framework, has the objective of develop a dynamic coupling, using TCP/IP based socket communication, between the thermal-hydraulic system code T-TRACE, Tractebel-ENGIE version of the latest US NRC TRACE release, and the multi-group 3-D nodal diffusion and core physics code PANTHER, developed and maintained by EDF Energy (UK). As a first step of the development, a fully temporally explicit coupling scheme has been developed between TRACE and PANTHER based on a boundary conditions exchange at the core level at each temporal iteration. The OECD TMI MSLB benchmark has been selected as verification scenario for testing the ongoing developing T-TRACE/PANTHER coupled code. The developed coupled code is benchmarked code-to-code against TRACE/PARCS and T-RELAP5/PANTHER.
THE LiCore POWER PLANT SIMULATOR OF THE MOLTEN SALT FAST REACTOR
Authors: Laureau A. , Rosier E. , Merle E. , Beils S. , Bruneau O. , Blanchon J.C. , Gathmann R. , Heuer D. , Passelaigue F. , Vaiana F. , Zanini A.
Published: February 2021

Molten salt reactors as liquid-fuelled reactors are flexible in terms of operation or design choices, but they are very different in terms of design, operation and safety approach compared to solid-fuelled reactors. Such reactors call for a new definition of their operating procedures and safety approach. Dedicated developments and studies have been performed in the frame of the European SAMOFAR project of Horizon2020 and in parallel in France involving CNRS, CORYS and Framatome to develop a system code called LiCore adapted to such reactors, corresponding to a basic-principle power plant simulator. The neutronic model LiCore, at the centre of the simulator, corresponds to an improved point-kinetics model to take into account the specificities of a MSR, notably the circulation of the delayed neutron precursors out of the core. Coupled to a simple piston model for the fuel motion in the core, this code can perform calculations faster than real time to simulate the behaviour of the fuel circuit. Transient calculations performed with LiCore are presented, together with comparisons first to a simple point-kinetics model and then to 3D calculations with the TFM-OpenFOAM coupled code. Finally, the LiCore code has recently been integrated in the ALICES platform, the integrated simulation toolset designed by CORYS for the development, maintenance and operation of major simulator such as power plant simulators.
3-D TRANSIENT COUPLED SIMULATION OF SUPERPHENIX WITH PARCS/ATHLET
Authors: Bousquet Jeremy , Henry Romain , Seubert Armin
Published: February 2021

Most safety criteria for Sodium cooled Fast Reactors (SFR) are local core parameters. Thus, application of 3-D neutron kinetic and thermal-hydraulic coupled codes including detailed modelling of core expansion effects is mandatory for best estimate evaluations of safety margins. A recently published benchmark based on measurements performed at the Superph´enix (SPX) reactor offers the opportunity to validate codes and methods for SFR safety assessment. In this paper, the SPX core is modelled in ATHLET and PARCS. Explicit models for axial and radial core expansion effects for 3-D coupled calculations were recently implemented in PARCS. In ATHLET, axial thermal expansion of structures (strongback, vessel, control rod drive line) influencing the relative position of the control rod in the active core is modelled. The newly implemented models are tested on a transient initiated by a reactivity insertion of -50 pcm. A point kinetic simulation is also performed to compare with the 3D solution. Both ATHLET-point kinetic model and ATHLET-PARCS simulations deliver similar power responses during the transient but with an offset. By analysing the different feedbacks in the point kinetic model, it can be concluded that models of expansion of the different structures are well implemented. In the future further analysis of different transients of the benchmark are planned.
BENCHMARK EVALUATION OF REACTIVITY EFFECTS AND REACTIVITY COEFFICIENTS IN THE MOLTEN SALT REACTOR EXPERIMENT
Authors: Shen Dan , Fratoni Massimiliano
Published: February 2021

A set of benchmarks based on the experimental data from the Molten Salt Reactor Experiment (MSRE) is being compiled as part of International Reactor Physics Experiments Evaluation Reactor Physics Experiments Evaluation Project (IRPhEP). The initial benchmark that will be available in the 2019 edition of the IRPhEP handbook covers the first zero-power criticality experiment. Follow up benchmarks are under development based on the series of control rod calibration experiments performed at the MSRE, which consisted in progressive addition of a small amount (85g) of 235U in the salt followed by the insertion of the control rods acts to compensate for the excess reactivity insertion. Multiple reactivity effects and coefficients measurements are included in the benchmark: differential worth of a control rod, reactivity equivalent of 235U addition, control rod bank worth, reactivity effect of fuel circulation, isothermal temperature coefficient and fuel temperature coefficient. An uncertainty of 2% is attributed to the reported reactivity measurements from experimenters and it was believed that the uncertainty of reactor period measurement contributed the most of the experimental uncertainty. An additional 2% uncertainty was added to all reactivity measurements to represent the uncertainty for the correction factor applied to pull all the measurements on the same uranium concentration and this uncertainty was reasonably inferred by evaluating this factor on the MSRE benchmark model. The calculated reactivity equivalent of 235U additions (0.2228±0.0014, represented as the change of reactivity over the relative change of 235U mass in loop) matches well with the experiment value (0.223±0.006). Most of other calculations, including the control rod bank worth, reactivity effects of fuel circulation and isothermal and fuel temperature coefficients fall within one standard deviation from the experimental values as well.
RESONANCE CALCULATION BASED ON THE GLOBAL-LOCAL METHOD FOR THE COMPLEX GEOMETRY PROBLEMS
Authors: Cao Lu , Liu Zhouyu , He Qingming , Cao Liangzhi
Published: February 2021

With the development of nuclear reactor, new complicated designs introduce some challenges for the resonance method in the NECP-X code, which is called as the Global-Local method. There are two limitations of the present NECP-X. One is the geometry modeling limitation, and the other one is that the Global-Local resonance method can only deal with the cylindrical fuel rods in the current version. Therefore, some fuels like plate, annular fuels cannot be calculated in the NECP-X code because of these limitations. To overcome above issues, the constructive solid geometry (CSG) method is developed to model the complex problem, and the capability of constructing and solving the equivalent 1D plate and 1D annual fuel pin based on the framework of Global-Local method, is developed. A set of tests are calculated, including multi-annular fuel problems and plate fuel problems. The results show good accuracy of the new developed resonance self-shielding method.
DEVELOPMENT OF DECAY HEAT MODEL FOR RAST-K
Authors: Dzianisau Siarhei , Park Jinsu , Choi Sooyoung , Cherezov Alexey , Du Xianan , Lee Deokjung
Published: February 2021

Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest source of power in the core after fission. Being such a strong contributor to reactor power, it should be accurately determined at any time of reactor operation. Currently, there are two main approaches for DH estimation used in reactor simulation codes. One approach is based on careful inventorying of all produced target nuclides and their individual contributions to total power. Alternatively, the other popular approach is based on collapsing all target fission products into a small number of groups similar to delayed neutron estimation techniques. However, the last (multigroup) method currently has limitations when used in some transient scenarios such as transients occurred in fresh fuel.

In this study, the multigroup method was further developed for reducing limitations while retaining the advantage in computation speed. Then, it was implemented into Reactor Analysis code for Steady state and Transient (RAST-K) and tested against other codes. As a result, the improved method was found capable of determining DH power at all tested stages of reactor operation under any tested operation scenario. In particular, the test simulations using the improved method showed better results in those scenarios that were under accuracy limitations of the original multigroup method. Overall, the quality of transient calculations in RAST-K was improved when using the newly implemented DH module.
ANALYTIC DISCRETE-ORDINATES SOLUTION FOR TIME-DEPENDENT TRANSPORT IN FINITE MEDIA
Authors: Densmore Jeffery D. , Kooreman Gabriel
Published: February 2021

We present an extension of the Analytic Discrete-Ordinates method to time-dependent transport in finite media. The application of this technique to time-dependent transport is primarily accomplished through the use of a Laplace transform. In the case of finite media, a system of equations arises from enforcing boundary conditions. Instead of directly solving this system, we construct a solution in terms of a Neumann series. We then show that terms can be neglected when numerically evaluating the inverse Laplace transform such that the series reduces to a finite sum. With this extension, we use convergence acceleration to generate a high-precision benchmark.
THE SOURCE JERK INTEGRALMETHOD FOR SUB-CRITICALITY MEASUREMENTS IN ADS
Authors: Kochetkov A. , Billebaud A. , Chabod S. , Krása A. , Lecolley F.-R. , Lecouey J.-L. , Lehaut G. , Marie N. , Messaoudi N. , Vittiglio G. , Wagemans J.
Published: February 2021

Three sub-critical (SC) core configurations were investigated in the VENUS-F zero power reactor coupled with the GENEPI-3C accelerator. The SC10 and SC12 were a mock-up of a MYRRHA start-up core and SC11 represented a more complex MYRRHA core loaded with various types of in-pile-sections. The sub-criticality of 11 variants of these VENUS-F cores was changed in several steps from -6$ down to -30$ using the safety and control rods. Their sub-criticalities were determined with the Source Jerk Integral (SJI) method using 11 fission chambers located all over the reactor. For the data analysis, the 8-group delayed neutron parameters from the JEFF-3.1.2 evaluated nuclear data library were used. Reliability and reproducibility of the experimental results were tested by repeating the measurements, swapping the detectors and varying the accelerator beam intensity, thus changing the detector count rates and verifying the validity of the dead time corrections. The obtained results are compared with MCNP calculations.
AXIAL FLUX PROFILE IN THE ADVANCED TEST REACTOR
Authors: Manwaring Nathan , Reichenberger Michael
Published: February 2021

This work demonstrates an approach to determine probability of perturbation of the axial profile of the thermal neutron flux in the Advanced Test Reactor. The axial flux profile is expected to follow a theoretical cosine shape, due to the minimal use of vertically-withdrawn shims. Reactivity is normally controlled by rotation of Outer Shim Control Cylinders, uniformly affecting neutron flux at all axial locations. The Advanced Test Reactor routinely accepts for irradiation experiments of a variety of designs. Among the analyses required by the safety basis approved by the United States Department of Energy is the characterization of a new experiment’s potential for perturbing the axial flux, which could exacerbate power peaking in the driver fuel. However, this perturbation can be more or less severe in different locations within the fuel. Therefore, the best characterization of axial flux perturbation requires knowledge of baseline axial flux. Such information is obtained by measuring decay in activated uranium flux wires irradiated at known positions in cooling channels in plate-type fuel elements. Due to variability in measured axial flux, it is not usually clear whether a given anomalous measurement is caused by an actual perturbation. Assuming normality in random measurement errors, the probability of an actual perturbation is quantified.
NEUTRON FIELD MOCK-UP DEVELOPMENT FOR THE FLUORIDE SALT REACTORS NEUTRONIC RESEARCH
Authors: Losa Evžen , Košťál Michal , Czakoj Tomáš , Šimon Jan , Burianová Nicola , Juříček Vlastimil , Rypar Vojtěch
Published: February 2021

Experimental work at the LR-0 reactor was carried out to determine the possibility of the mock-up neutron field creation for the fluoride salt-based reactors. Previous calculations and experiments have shown that the fast part of the molten salt reactor (MSR) spectrum is shaped by fluorine and even the Teflon material is suitable for neutronics in that energy range. Properly selected spectrum indices can describe the neutron field of the MSR in fast thermal and intermediate parts of the spectrum. Current research has focused on a deeper study of the possibility of using the filtered thermal and intermediate neutron spectrum of the experimental light water reactor for the physics of fluoride salt-based reactor. LiF-BeF2 (FLIBE) capsule and teflon cylinders are used as a spectrum filters in the LR-0 reactor. Measured results show acceptable C/E-1 agreement in the reaction rates and satisfactory agreement for usage of the FLIBE filtered neutron spectrum determined by the 181Ta(n, γ) monitor as a mock-up in thermal to intermediate energy in the fluoride high-temperature reactor (FHR). Concerning MSR, intermediate spectrum can be reproduced to some extent but not as good as in the case of FHR.
GENERATION AND INITIAL VALIDATION OF A NEW CASMO5 ENDF/B-VIII.0 NUCLEAR DATA LIBRARY
Authors: Ferrer Rodolfo , Rhodes Joel
Published: February 2021

A new nuclear data library for the CASMO5 advanced lattice physics code has been generated based on the recently-released ENDF/B-VIII.0 evaluation. The ENDF/B-VIII.0 evaluation represents the state-of-the-art in nuclear data and features new evaluations from the CIELO project for 1H, 16O, 56Fe, 235U, 238U and 239Pu. A description of the library generation procedure used to process these data into the CASMO5 586 energy group structure is provided. Initial validation of the new ENDF/B-VIII.0-based library, referred to as the E8R0 library, is also presented and involves the comparison of predicted k–eff and fission rate distributions to measurements from various critical experiments. The critical experiments used in the initial validation of the E8R0 library consist of the B&W 1810 series, B&W 1484 series, DIMPLE S06A/B, and TCA reflector experiment with iron plates. The results from the initial validation indicate that the new E8R0 library provides a satisfactory performance in terms of CASMO5 predicted k–eff and fission distributions.
REACTOR PHYSICS EXPERIMENT IN A GRAPHITE-MODERATION SYSTEM FOR HTGR
Authors: Fukaya Y. , Goto M. , Nakagawa S. , Nakajima K. , Takahashi K. , Sakon A. , Sano T. , Hashimoto K.
Published: February 2021

The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce a generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment to observe sub-criticality. To achieve the objectives, the reactor core of graphite-moderation system named B7/4”G2/8”p8EUNU+3/8”p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core is composed of the fuel assembly, driver fuel assembly, graphite reflector, and polyethylene reflector. The fuel assembly is composed of enriched uranium plate, natural uranium plate and graphite plates to realize the average fuel enrichment of HTTR and it’s spectrum. However, driver fuel assembly is necessary to achieve the criticality with the small-sized core. The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In this study, the overview of the criticality experiments is reported. The reactor configuration with graphite moderation system is rare case in the KUCA experiments, and this experiment is expected to contribute not only for an HTGR development but also for other types of a reactor in the graphite moderation system such as a molten salt reactor development.
COMPARISONS AMONG THREE COUPLING FORMS OF CYLINDRICAL NODAL EXPANSION METHODS
Authors: Wen Yutong , She Ding , Shi Lei , Zhao Jing
Published: February 2021

Nodal expansion method (NEM) is a typical nodal method in solving neutron diffusion equation in coarse mesh spatial discretization. NEM has been extended to cylindrical geometry in previous studies. Cylindrical NEM can be realized by different forms, where the variables in the nodal coupling equations are classified as net current, flux or partial current at the nodal surface, respectively. The three types of coupling forms of NEM have been implemented in PANGU code for the high temperature gas-cooled reactor (HTGR) physics analysis. This paper derives the three types of coupling forms of NEM, and analyzes their performance in solving the HTGR model containing void region.
RECENT VALIDATION STUDY OF THE TECHNIQUE FOR EXPRESS EVALUATION OF BURNUP IN LEAKING FUEL ASSEMBLIES OF WWER POWER UNITS
Authors: Vilkhivskaya O.V. , Evdokimov I.A. , Likhanskii V.V. , Afanasieva E.Yu.
Published: February 2021

The present work continues the series of papers on the revision of the conventional technique for evaluation of leaking fuel burnup during reactor operation at nuclear power plants (NPPs). The focus was made on reduction of uncertainties in evaluation of leaking fuel burnup in modern fuel cycles at WWER-1000 power units. A set of models was proposed for express calculation of the build-up of caesium isotopes in fuel and to relate 134Cs/137Cs activity ratio with fuel burnup for each rod in the core. These models are based on routine neutronic calculations of pin-by-pin linear heat generation rates which are performed at NPPs for each particular fuel cycle with particular core loading pattern (however, these calculations do not provide data on caesium inventory in fuel). Previously, the proposed models have been validated against several practical cases. This latest validation study relied on the analysis of the most recent fuel cycles at two NPPs that reported spike-events and identified the leaking fuel assemblies (LFAs) after the reactor shutdown. The calculated 134Cs/137Cs activity ratios in the fuel of the LFAs were compared to the NPPs data on the activity measurements, and to the post-irradiation examination (PIE) data provided for one FA. A reasonable agreement between the model predictions and the experimental data on 134Cs/137Cs activity ratios in the fuel as a function of its burnup is shown for the advanced FA designs in modern fuel cycles.
ENHANCEMENT OF VALIDATION STUDIES FOR REACTOR DOSIMETRY AND ACTIVATION PREDICTIONS WITH THE NUCLEAR DATA SAMPLING METHODOLOGY
Authors: Vasiliev A. , Pecchia M. , Rochman D. , Ferroukhi H.
Published: February 2021

The CASMO/SIMULATE/MCNP/FISPACT-II calculation route has been established at the Paul Scherrer Institute (PSI) for reactor dosimetry and activation studies. Furthermore, the in-house tool NUSS is in use at PSI for nuclear data (ND) related uncertainties quantifications with Monte Carlo neutron transport calculations. The use of randomly sampled ACE-formatted ND files not only allows propagation of the ND uncertainties, but also can serve for assessing the applicability of different types of experimental data for validation of calculation predictions of parameters of interest. In the present work an application of the PSI calculation scheme for analysis of activation reaction rates and the fast neutron fluence (FNF), throughout the Swiss pressurised water reactor (PWR) and simplified containment building models, is demonstrated. As particular examples of potentially available experimental data, two kinds of the neutron flux monitors are considered: a) the reactor pressure vessel scraping samples and detectors placed in the dosimetry channels, mounted at the core barrel and designed for validation of FNF calculations, and b) the ex-vessel dosimeters, specifically used by the Swiss waste management organisation (NAGRA) for validation of bio-shield activation predictions. The calculations were done with the ENDF/B-VII.1 library. The obtained results demonstrate importance of the ND uncertainties for the dosimetry evaluations. The assessment of the applicability of the selected experimental information for validation of the bio-shield irradiation calculations was done based on evaluation of the ND-related Pearson correlation coefficients.
USE OF WIMS/PANTHER TO ANALYSE NITROGEN WORTH MEASUREMENTS AT THE UK FLEET OF AGRS
Authors: Birkett Eddie , Knight Martin
Published: February 2021

The UK fleet of Advanced Gas-Cooled Reactors (AGRs) are moderated with graphite and cooled with carbon dioxide. Each reactor consists of over 300 fuel channels with a cluster of UO2 fuel pins surrounded by graphite. Primary shutdown is achieved with control rods, which are at interstitial locations at six AGR stations, and in vacated fuel channels in the seventh. Nitrogen injection into the coolant is used for diverse shutdown or hold down.

The reactivity worth of nitrogen was measured during cold commissioning at each AGR during the 1970s and 1980s. This measurement involved inducing a flux transient through a known change in nitrogen pressure, with conversion into a reactivity worth using kinetic data. Noting the time elapsed since this work (over 40 years) it is sensible to review this validation and reconsider these measurements using a consistent source of kinetic data.

The lattice code WIMS and the reactor code PANTHER can be used to analyse the commissioning cores, in particular to determine the core-wide kinetic data. Furthermore the calculated worth of nitrogen can be compared with measurement. For the six interstitial AGR reactor designs nitrogen worth is under-predicted by about 10% and by about 5% for the inchannel design.

This consistent under-prediction is investigated further with SERPENT.