bluuebunny/crossref_metadata_2025 huggingface dataset:
RECENT CAPABILITY AND PERFORMANCE ENHANCEMENTS OF THE WHOLE-CORE TRANSPORT CODE nTRACER
Authors: Choi Namjae , Park Hansol , Lee Han Gyu , Jae Seungug , Jeon Sori , Joo Han Gyu
Published: February 2021

The whole-core transport code nTRACER has made many advances in recent years. Several innovative cross section treatment methods were developed, a new axial transport solver was introduced for stabilizing the 2D/1D scheme, and substantial computational enhancements were achieved using NVIDIA CUDA and Intel Math Kernel Library (MKL). In addition, gamma transport solver was implemented to predict the power distributions more physically, and the flexibility of the restart calculation was improved using an offline processing code nTIG (nTRACER Input Generator). This paper is the compilation of the recent progresses in nTRACER developments.
3D CORE CALCULATION BASED ON THE METHOD OF DYNAMIC HOMOGENIZATION
Authors: Galia Antonio , Zmijarevic Igor , Sanchez Richard
Published: February 2021

The classical two-step calculation scheme has been extensively used to perform three-dimensional deterministic core calculations thanks to its fast results. On the other hand, direct 3D transport calculations and 2D/1D Fusion methods, mostly based on the method of characteristics, have recently been applied showing a prohibitive computational time for routine design purposes as well as in the context of multiphysics and core depletion calculations, due to current machine capabilities.

The Dynamic Homogenization method is here proposed as an alternative technique that may lie between the classical and the direct approaches in terms of precision and performance. In this work, the method is applied to the NEA ”PWR MOX/UO2 Core Benchmark” for a 3D configuration. Comparison of pin power relative errors and computational cost against the two-step and direct approaches are presented.
NUMERICAL BENCHMARK OF STRONGLY TO LOOSELY COUPLED ASSEMBLIES USING THE TRANSIENT FISSION MATRIX METHOD
Authors: Routsonis K. , Blaise P. , Tommasi J.
Published: February 2021

Advances in computational methods have given rise to the study and simulation of different aspects of reactor behavior. As such, topics associated with high computational costs become feasible candidates for further investigation and one of them is reactor space-time kinetics (STK). Until recently, STK simulation and point kinetics approximation were limited to deterministic codes, with Monte Carlo codes being too costly to start with. However, recent developments in this area have allowed the use of certain methods in stochastic codes. One such technique is based on the Transient Fission Matrix (TFM) model, a hybrid method that uses a system response obtained through Monte Carlo and stored in fission and time matrices as input for deterministic calculations. The result enables a view of the STK in terms of neutron propagation probability and propagation time across the system. The TFM method was applied to a simple coupled core configuration to generate a numerical benchmark. The Serpent 2 Monte Carlo code was used for the stochastic part of the calculation. The configuration consists of two fuel assemblies placed in a light water tank, with a water blade of varying width between them. TFM, flux and fission results were obtained for varying water blade widths, ranging between 0 cm and 20 cm. The data is then used to analyze the behavior of the system, as well as the effects of the coupling between the two assemblies. As the assemblies move further apart, the system slowly transitions from two tightly coupled assemblies that essentially form a single core, to two almost independent cores. This study enables to produce a benchmark for future calculations and predefine an innovative way of designing high dominant ratio configurations, required for tackling Monte Carlo residual problems. An actual experimental program could be led in ad hoc zero power reactor (ZPR), such as the KUCA reactor of Kyoto University.
DEVELOPMENT OF MPACT FOR FULL-CORE SIMULATIONS OF MAGNOX GAS-COOLED NUCLEAR REACTORS
Authors: Ade Brian J. , Luciano Nicholas P. , Conant Andrew J. , Gentry Cole A. , Stimpson Shane G. , Collins Benjamin S. , Seog Kim Kang , Mills Robert
Published: February 2021

The MPACT code, jointly developed by Oak Ridge National Laboratory and University of Michigan, is designed to perform high-fidelity light water reactor (LWR) analysis using wholecore pin-resolved neutron transport calculations on modern parallel-computing hardware. MPACT uses the subgroup method for resonance self-shielding, while the primary neutron transport solver uses a 2D/1D method that is based on the method of characteristics (MoC) for the x-y planes coupled with a 1D diffusion or transport solver in the axial dimension. Additional geometry capabilities are currently being developed in MPACT to support hexagonal-pitched lattices, as well as interstitial geometry (i.e., control rods at the corner of four adjacent pin cells). In this research, the MPACT method is tested on gas-cooled reactors by applying MPACT to full-core MAGNOX reactor test problems. MAGNOX test problems were chosen due to the availability of high-quality reactor design and validation data (available through an ongoing collaboration with the National Nuclear Laboratory in the United Kingdom) and the existence of a relatively complex axial power shape that is expected to challenge the MPACT method. MPACT’s convergence for partial- and full-core problems will be tested and verified. MPACT will be compared with high-fidelity continuous-energy Monte Carlo simulations to verify core reactivity, power distributions, and performance of the available cross section data libraries and energy group structures.
MATRIX EXPONENTIAL METHODS FOR PARALLEL COMPUTING OF ISOTOPIC DEPLETION AND SPECIES TRANSPORT FOR MOLTEN SALT REACTOR ANALYSIS
Authors: Taylor Zack , Collins Benjamin , Maldonado Ivan
Published: February 2021

Matrix exponential methods have long been utilized for isotopic depletion in nuclear fuel calculations. In this paper we discuss the development of such methods in addition to species transport for liquid fueled molten salt reactors (MSRs). Conventional nuclear reactors work with fixed fuel assemblies in which fission products and fissile material do not transport throughout the core. Liquid fueled molten salt reactors work in a much different way, allowing for material to transport throughout the primary reactor loop. Because of this, fission product transport must be taken into account. The set of partial differential equations that apply are discretized into systems of first order ordinary differential equations (ODEs). The exact solution to the set of ODEs is herein being estimated using the matrix exponential method known as the Chebychev Rational Approximation Method (CRAM).
UTILIZING A REDUCED-ORDER MODEL AND PHYSICAL PROGRAMMING FOR PRELIMINARY REACTOR DESIGN OPTIMIZATION
Authors: Stewart Ryan , Palmer Todd S.
Published: February 2021

Reactor core design is inherently a multi-objective problem which spans a large design space, and potentially larger objective space. This process relies on high-fidelity models to probe the design space, and sophisticated computer codes to calculate the important physics occurring in the reactor. In the past, the design space has been reduced by individuals with extensive knowledge of reactor core design; however, this approach is not always available. In this paper, we utilize a set of high-fidelity models to generate a reduced-order model, and couple this with a genetic algorithm to quickly and effectively optimize a preliminary design for a prototypical sodium fast reactor. We also examine augmenting the genetic algorithm with physical programming to generate the fitness function(s) that evaluates the degree to which a core has been optimized. Physical programming is used in two variations of multi-objective optimization and is compared with a traditional weighting scheme to examine the solutions present on the Pareto front.

Optimization on the reduced-order model produces a set of solutions on the Pareto front for a designer to examine. The uncertainty for the objective functions examined in the reduced-order model is less than 7% for the given designs, and improves as additional data points are employed. Utilizing a reduced-order model can significantly reduce the computation time and storage to perform preliminary optimization. Physical programming was shown to reduce the objective space when compared with a traditional weighting scheme. It also provides an intuitive and computationally efficient way to produce a Pareto front that meets the designer’s objectives.
DESIGN OF A CODE-AGNOSTIC DRIVER APPLICATION FOR HIGH-FIDELITY COUPLED NEUTRONICS AND THERMAL-HYDRAULIC SIMULATIONS
Authors: Romano Paul K. , Hamilton Steven P. , Rahaman Ronald O. , Novak April , Merzari Elia , Harper Sterling M. , Shriwise Patrick C.
Published: February 2021

While the literature has numerous examples of Monte Carlo and computational fluid dynamics (CFD) coupling, most are hard-wired codes intended primarily for research rather than as standalone, general-purpose codes. In this work, we describe an open source application, ENRICO, that allows coupled neutronic and thermal-hydraulic simulations between multiple codes that can be chosen at runtime (as opposed to a coupling between two specific codes). In particular, we outline the class hierarchy in ENRICO and show how it enables a clean separation between the logic and data required for a coupled simulation (which is agnostic to the individual solvers used) from the logic/data required for individual physics solvers. ENRICO also allows coupling between high-order (and generally computationally expensive) solvers to low-order “surrogate” solvers; for example, Nek5000 can be swapped out with a subchannel solver.

ENRICO has been designed for use on distributed-memory computing environments. The transfer of solution fields between solvers is performed in memory rather than through file I/O.We describe the process topology among the different solvers and how it is leveraged to carry out solution transfers. We present results for a coupled simulation of a single light-water reactor fuel assembly using Monte Carlo neutron transport and CFD.
ON THE EFFECT OF FINE-SCALE RADIAL VOID DISTRIBUTIONS ON ASSEMBLY CRITICALITY AND PIN POWER DISTRIBUTION
Authors: Price Dean , Gacek Andrew , Kozlowski Tomasz , Radaideh Majdi I.
Published: February 2021

The assumption that void fraction, and by extension coolant density, is uniform in the radial direction is a common approximation used in lattice physics simulations. In this study, models without uniform radial void fraction are used and lattice criticality and pin powers are investigated in two ways. One way uses hypothetical models that reflect extreme radial void distributions; modifications such as uniform radial pin enrichment and the removal of gadolinium rods are included in these models as well. Experimentally-determined boiling water reactor radial void distributions are also replicated in neutronics models using Serpent 2. It is found in the hypothetical models that the presence of gadolinium rods has a large effect on the interaction between lattice criticality and radial void distribution. It was also found that considering experimental radial void fraction distributions had the largest effect on the pin power of the rods containing gadolinium. Furthermore, it is observed that considering realistic radial void distributions, in general, decreased lattice criticality. The reason can be attributed that to the passive negative-feedback design of light water reactors. These are useful findings because calculation of more accurate peaking factors can lead to efficient and yet safer reactor operation.
REVIEW ON THE NEUTRON SOURCE STRENGTH OF HEU METAL BURST REACTOR
Authors: Liu Xiaobo
Published: February 2021

There were two similar series of burst waiting time experiments carried out on Godiva-II and Caliban respectively and analyzed for neutron initiation study, unfortunately both the strength of spontaneous fission neutron source given in the published papers are not consistent with the spontaneous fission data of U-235 and U-238.

In this paper, the discrepancies of spontaneous fission data of U-235 and U-238 from different references are reviewed and clarified and the correct strength of spontaneous fission neutron source for Godiva-II and Caliban are calculated respectively, then the multiplication equivalence method of arbitrary source is introduced and the equivalent fundamental-mode(EFM) co-efficiencies of spontaneous fission neutron source for Godiva-II and Caliban are calculated with MC software, lastly the strength of neutron source for these two HEU metal burst reactor in EFM are deduced. The EFM neutron strength of this two burst reactors are much less than proposed values used for burst waiting investigation in published papers respectively, and with the new data, the inconsistency of burst waiting time to the published results is presented and discussed.
ADAPTIVE TIME-STEP CONTROL FOR THE MODAL METHOD TO INTEGRATE THE MULTIGROUP NEUTRON DIFFUSION EQUATION
Authors: Carreño A. , Vidal-Ferràndiz A. , Ginestar D. , Verdú G.
Published: February 2021

The distribution of the power inside a reactor core can be described by the time dependent multigroup neutron diffusion equation. One of the approaches to integrate this time-dependent equation is the modal method, that assumes that the solution can be described by the sum of amplitude function multiplied by shape functions of modes. These shape functions can be computed by solving a _-modes problems. The modal method has a great interest when the distribution of the power cannot be well approximated by only one shape function, mainly, when local perturbations are applied during the transient. Usually, the shape functions of the modal methods are updated for the time-dependent equations with a constant time-step size to obtain accurate results. In this work, we propose a modal methodology with an adaptive control time-step to update the eigenfunctions associated with the modes. This algorithm improves efficiency because of time is not spent solving the systems to a level of accuracy beyond relevance and reduces the step size if they detect a numerical instability. Step size controllers require an error estimation. Different error estimations are considered and analyzed in a benchmark problem with a out of phase local perturbation.
PREDICTOR-CORRECTOR QUASI-STATIC METHOD FOR TIGHTLY-COUPLED REACTOR MULTIPHYSICS TRANSIENT CALCULATIONS
Authors: Kooreman Gabriel , Griesheimer David P.
Published: February 2021

Several methods have recently been developed to solve multiphysics transients using the Improved Quasi-Static method and its derivatives. In order to address some perceived drawbacks of these methods, we have developed a new method for solving multiphysics transient calculations using the Predictor-Corrector Quasi-Static method. Our method involves computing reactivity feedback parameters during each transport timestep in order to enable reactivity feedback on the small timescale used by the point kinetics phase of the calculation. The advantage of this approach is that the transport solver does not need to store local flux information between time steps, potentially making it more appropriate for use with a Monte Carlo solver. We have demonstrated the accuracy of our method by solving a simple model problem that exhibits difficult multiphysics behavior. Additionally, we have compared our results against another recently published multiphysics coupling scheme, confirming that our approach does not negatively affect the accuracy of the transient solution.
VALIDATION OF THE FEMAXI/SIMMER COUPLING BY USING THE E9 CABRI-2 EXPERIMENT
Authors: Gianfelici Simone , Perez-Martin Sara , Pfrang Werner , Schulenberg Thomas , Rineiski Andrei
Published: February 2021

Accidents prevention and mitigation of the related consequences are mandatory for the safety assessment of the Generation IV Sodium-cooled Fast Reactors (SFRs). In order to reach these goals, the accurate evaluation of the most challenging transients, which may lead to severe accidents, is fundamental. In particular, an accurate evaluation of the degradation of the thermo-mechanical properties of the fuel pins due to irradiation, affecting the course of accidental transient conditions, is necessary. In our studies, the reference calculation route for the analysis of a severe accident in SFRs is based on the sequential use of the SAS-SFR and the SIMMER-III/IV mechanistic codes for the initiation (up to hexcan failure of a fuel sub-assembly) and the late (with core melting) phases of the scenario, respectively. Nevertheless, the module devoted to the SAS-SFR/SIMMER data transfer shows some limitations when innovative pins and core designs are considered. Having this in mind, the present paper proposes an innovative approach based on the coupling of a fuel performance code FEMAXI and the SIMMER code. When applied to severe accident studies of new reactors designs, such an approach is expected to improve the consistency of the SIMMER results as well as to extend its range of application. The FEMAXI/SIMMER coupling is described in the paper. Further, the results of the code validation against the CABRI-2 E9 experimental test, performed at the CABRI reactor using an OPHELIE-6 annular fuel pin irradiated in the PHENIX reactor, are shown and discussed. In the E9 test, a slow power ramp non-failure transient, corresponding to control rod withdrawal conditions, was investigated. New FEMAXI/SIMMER results are compared also to stand-alone SAS-SFR and SIMMER evaluations.
EXTENDED MSM METHOD TO ESTIMATE THE REACTIVITY OF A SUB-CRITICAL CORE DRIVEN BY AN ACCELERATOR BASED NEUTRON SOURCE
Authors: Billebaud A. , Lecouey J.-L. , Chevret T. , Chabod S. , Doligez X. , Kochetkov A. , Krása A. , Lecolley F.-R. , Lehaut G. , Marie N. , Messaoudi N. , Uyttenhove W. , Vittiglio G. , Wagemans J. , Bécares V. , Villamarin D. , Mellier F.
Published: February 2021

The Modified Source Multiplication method is used to determine an unknown reactivity level of a reactor from a known one if one has access to the detector counting for both levels when the reactor is fed by a constant neutron source like an Am-Be source. When available, an accelerator driven source, in continuous mode, can be useful as its intensity can be tunable and then adapted to the experimental conditions. However, in that case, the MSM technique must be extended to account for an external source whose intensity, energy and angular distributions can vary from one measurement to another. In this paper, this Modified Multi-Source Multiplication (MMSM) method is applied to measurements done during the FREYA project in the GUINEVERE facility, operated with the GENEPI-3C accelerator providing a mixture of (D,T) and (D,D) neutrons. The monitoring of these sources through the detection of the associated charged particles allows the calculation of the MMSM factors and the estimate of the reactivity values. The results are compared in different configurations with the reactivity obtained with an Am-Be source or in dynamic measurements performed with GENEPI-3C. Their excellent agreement shows the possibility of using such accelerator-based neutron sources for MSM measurements when they are correctly monitored. This is of great interest for deep sub-critical level characterization for which detector count rates per source neutrons might be low.
FACILITIES FOR NANO MATERIALS EXAMINATION AT THE PULSTAR REACTOR
Authors: Hawari A. I. , Liu M. , Cai Q.
Published: February 2021

The PULSTAR is a 1-MWth nuclear research reactor located at North Carolina State University. It is fueled by uranium dioxide assemblies enriched to 4% or 6% in U-235 and is currently under licensing for operation at 2-MWth power. The PULSTAR is a center for irradiation testing and pre/post irradiation examination of materials. Among its unique capabilities are positron annihilation spectroscopy (PAS) and neutron powder diffraction (NPD) facilities. The PAS facility provides an intense positron beam reaching 6 × 108 e+/s, which drives two spectrometers; the e+-PAS and the Ps-PAS, used for studies of defects in thin film materials. A Na-22 bulk PAS system is also operational, which is used for studying millimeter scale materials. All spectrometers are capable of performing Doppler Broadening Spectroscopy (DBS) and Positron Annihilation Lifetime Spectroscopy (PALS). To date, the PAS systems have been used to characterize various materials (unirradiated and irradiated) that include graphite, soft matter, and metal-organic frameworks (MOF). The NPD facility uses a double focusing single crystal silicon rotating monochromator producing neutron beams with different energies. A position sensitive detection bank covers a scattering angle of 5°-125°. The facility is used in the examination of unirradiated and irradiated materials including graphitic materials, magnetic materials without rare-earth elements, and anode materials used in lithium batteries. The collected diffraction patterns can be processed to produce atomic pair distribution functions. The PAS and NPD facilities are available through user programs including the US DOE’s Nuclear Science User Facilities (NSUF) and the US NSF’s Research Triangle Nanotechnology Network (RTNN).
DOSIMETRY MODELING AND EXPERIMENTAL VALIDATION FOR THE PETALE PROGRAM IN THE CROCUS REACTOR
Authors: Laureau A. , Lamirand V. , Gruel A. , Frajtag P. , Pautz A.
Published: February 2021

The PETALE experimental program in the CROCUS reactor intends to provide integral measurements on reactivity worth and dosimetry measurement to constrain nuclear data relative to stainless steel heavy reflectors. The experimental setup consists in eight successive plates of pure iron, pure nickel, pure chromium, or nuclear-grade stainless steel set at the close periphery of the core. The plates are interleaved with up to nine dosimeters that consist of thin activation foils with different possible materials to be sensitive to different ranges of the neutron spectrum. A precise measurement with a good estimation of the uncertainties and correlations is required, especially when comparing reaction rates, e.g. transmission measurement and/or spectral indices.

The present work focuses on the validation of the dosimetry technics developed in preparation of this experimental program. Different aspects are discussed: monitors, efficiency calibration, self-absorption correction, self-shielding and nuclear data uncertainties. The different sources of uncertainties for the experiment-calculation comparisons are characterized, taking into account all the correlation between the different dosimeters. These correlations are a mandatory element for the aimed Bayesian assimilation in order to avoir overfitting when considering dosimeter providing a similar information.
INVESTIGATION OF APPROPRIATE LADDER NUMBER ON PROBABILITY TABLE GENERATION
Authors: Tada Kenichi
Published: February 2021

The probability table method is widely used for continuous energy Monte Carlo calculation codes to treat the self-shielding effect in the unresolved resonance region. The ladder method is used to calculate the probability table. This method generates pseudo resonance structures using random numbers based on the averaged resonance parameters. The probability table affects the ladder number,i.e., the number of pseudo resonance structures. Since the higher the ladder number the longer it takes to create a cross-section library, reducing the value of the ladder number is an effective way to reduce the time spent in the creation of a cross-section library. However, the exact value of the most appropriate ladder number to use has not been found. Currently, users have to use a nuclear data processing system to manually set the ladder number to a large enough value in order to obtain a sufficiently converged probability table. In this study, the probability tables for all nuclides prepared in JENDL-4.0 are used to find the optimal ladder number. The results show that the probability table differences are small enough when the ladder number is 100. The impact of these differences on neutronics calculations is found to be small.
EXPERIMENT OF UNIQUE COMBINATION NUMBER DUE TO THE THIRD-ORDER NEUTRON-CORRELATION
Authors: Endo Tomohiro , Imai Sho , Watanabe Kenichi , Yamamoto Akio , Sakon Atsushi , Hashimoto Kengo , Yamanaka Masao , Pyeon Cheol Ho
Published: February 2021

From zero-power reactor noise measurement, the second- and third-order neutron correlation factors Y and y3 can be evaluated by analyzing mean, variance, the third-order central moment of neutron count data. Theoretically, it is expected that the neutron-correlation ratio y3/Y2 converges to the unique combination number “3” at a near-critical state in an arbitrary system without depending on the fissile material and the neutron-energy spectrum of core, as the neutron counting gate width T increases sufficiently. Thus, the information about the difference between y3/Y2 and “3” has the potential to judge whether a target unknown system is critical or not and to roughly guess the absolute value of subcriticality. In this study, the detector dead-time effect on y3/Y2 is theoretically investigated based on the heuristic method using the single-, pair-, and trio-detection probabilities with the fundamental mode approximation. As a result, it is clarified that the saturation value of y3/Y2 converges to “3” independent of the dead time, when a target system is a critical state. For validation, actual experimental results are presented for a non-multiplication system driven by 252Cf spontaneous source, and shallow and deep subcritical systems at Japanese experimental facilities (UTR-KINKI and KUCA) under the shutdown state. Consequently, it is demonstrated that y3/Y2 shows a significant difference from “3” in the non-multiplication system. In the case of subcritical systems driven by inherent neutron sources, it is confirmed that the ratios y3/Y2 are close to the unique combination number “3,” and the slight difference from “3” is measurable by the long-time reactor noise measurement for the deep subcritical system.
EXPERIMENTAL VALIDATION OF NUCLEAR REACTION DATA FOR INVENTORY SIMULATIONS ON MOLYBDENUM
Authors: Gilbert M.R. , Packer L.W. , Stainer T.
Published: February 2021

Molybdenum is being considered as a potential material for future nuclear fusion experiments and power plants. It has good thermo-mechanical properties and can be easily fabricated, making it attractive as an alternative first wall material to tungsten, which is the current leading candidate. Unfortunately, according to simulations, fusion-neutron irradiated Mo may become too activated during reactor operation to maintain the objective of fusion of avoiding the generation of long-term, higher-level radioactive waste.

However, these simulated predictions rely heavily on having high-quality, accurate nuclear reaction data. For example, reliable simulations of the radiological response of isotopically-tailored Mo would be required by reactor designers and eventually nuclear regulators to assess if it is a viable low-activation fusion solution (in comparison to pure Mo). In recent years, UKAEA has developed benchmarks to test and validate the FISPACT-II inventory code and the input international nuclear data libraries against experimental measurements. This paper presents nuclear code prediction comparisons to new data acquired from γ-spectroscopy measurements of Mo irradiated in the ASP 14 MeV facility at AWE in the UK.

Results demonstrate that FISPACT-II predictions for Mo are remarkably accurate; particularly for activity generated from the shorter-lived radionuclides explored by these laboratory experiments, such as91Mo and97Nb, etc., and their metastable isomeric states.
EXPERIMENTAL STUDY ON NEUTRON CORRELATION ANALYSIS FOR A SUBCRITICAL SYSTEM DRIVEN BY A PULSED SPALLATION NEUTRON SOURCE IN KUCA
Authors: Nakajima Kunihiro , Takahashi Kazuki , Sakon Atsushi , Hohara Sin-ya , Sano Tadafumi , Yamanaka Masao , Pyeon Cheol Ho , Hashimoto Kengo
Published: February 2021

The Feynman-α and the Rossi-α methods have been frequently employed to determine the subcriticality of subcritical reactor systems driven by Poisson source such as Am-Be neutron source. In actual accelerator-driven systems (ADS), a spallation device will be applied as an intense neutron source. This device will be probably operated in a pulse mode and it is impossible to apply a conventional analysis method to determine the subcriticality in any ADS. In previous theoretical studies, some advanced formulae of neutron correlation analysis for spallation neutron source have been presented. However, the experimental study has been hardly reported to date. The major objectives of this study are to examine experimentally an applicability of these complicated formulae to a subcritical reactor system driven by an actual pulsed spallation neutron source and to determine the prompt-neutron decay constant α of the system. To achieve these goals, we constructed an ADS core at the Kyoto University Critical Assembly (KUCA). The core was composed of highly-enriched uranium fuel assemblies surrounded by many polyethylene reflector assemblies. We carried out a series of the Feynman-α and the Rossi-α analyses for the system driven by pulsed spallation source. As a result, the prompt-neutron decay constants were experimentally obtained by using a fitting formula. The prompt-neutron decay constants determined by Feynman-α and Rossi-α analyses agrees with each other within a statistical error range of least-squares fitting.
APPLICATION OF ATHERMAL RECOMBINATION CORRECTED DPA MODEL IN NUCLEAR DATA PROCESSING CODE NECP-ATLAS
Authors: Yin Wen , Zu Tiejun , Cao Liangzhi
Published: February 2021

The Displacement per Atom (DPA) is an important factor used to quantify the irradiation damage of materials. The radiation damage energy production cross section obtained by nuclear data processing code is an essential nuclear data used to calculate the DPA of materials. Nuclear data processing code, such as NJOY, adopts NRT-DPA model to calculate the radiation damage energy production cross section. However, the NRT-DPA model has several well-known limitation. Especially, the number of radiation defects produced in energetic cascades in metals is only ~1/3 the NRT-DPA model prediction. To improve the precision accuracy of DPA calculation, the state-of-the-art Athermal Recombination Corrected DPA (ARC-DPA) model is adopted in heat production and radiation damage energy production cross section module of nuclear data processing code NECP-Atlas. ARC-DPA model can be obtained by simply multiplying with the new efficiency function. The parameters in the new efficiency function are material constants that can be determined for a given material from Molecular Dynamics (MD) simulations or experiments. The material constants of some materials are obtained by MD simulations firstly in this paper. Numerical results show that the radiation damage cross section produced by nuclear data processing code NECP-Atlas adopting NRT-DPA model can agree well with the NJOY code. What’s more, the results from NECP-Atlas adopting ARC-DPA model can provide more physically realistic descriptions of primary defect and agree well with the results from MD simulation numerically.
INTERDIFFUSION OF Mo AND W UNDER IN-PILE IRRADIATION CONDITIONS
Authors: Ivanov A.S. , Churin V.A.
Published: February 2021

The study has established that reactor radiation leads to a significant increase of interdiffusion coefficients in a Mo-W alloy compared to thermally activated diffusion at temperatures of ~1500°С. Furthermore, a weakening of their concentration dependence is observed. The approximation procedure developed in this work allowed approximating experimental data for each sample using variation of only one parameter. It has been demonstrated that the interaction of molybdenum and tungsten nuclei with fast neutrons as well as with hard gamma quanta may cause the observed diffusion effects.
NEW VALIDATION OF SI CROSS-SECTION USING SILICA SAND
Authors: Czakoj Tomáš , Košťál Michal , Losa Evžen , Rypar Vojtěch
Published: February 2021

Silicon cross-sections, being important for a criticality safety of final spent fuel disposals, were recently reevaluated within the IAEA INDEN project. Similarly, the thermal Scattering Law matrix for silicon dioxide, which is also important for criticality safety, was also reevaluated in the ENDF/B-VIII.0 library. Due to these reasons, a series of validation experiments with silica sand were performed at the LR-0 reactor. This paper describes these validation experiments, which used two different amounts of silica sand placed in the core. The first part of validation was carried out as critical experiments in order to benefit from the suitability of integral experiments for validation. The second part -fast neutron spectrum measurement in the sand -was performed in order to obtain knowledge of its characteristics and its agreement with a calculation. The results showed significant improvement of the Thermal Scattering Law matrix for silicon dioxide available in the ENDF/B-VIII.0 library. They also showed that the new INDEN evaluation of silicon cross-sections, together with its description in the ENDF/B-VIII.0 gives disagreement rate closest to the experiments carried out without silica sand insertions. The spectrum measurement showed that the calculations of fast neutron spectra in the sand show only slight differences between different evaluations of silicon cross-sections. However, the fast neutron spectrum is not dependent on the Thermal Scattering Law. The calculated spectra show relatively good agreement with the measurement.
FEATURES OF THE SOURCE TERM CALCULATION CAPABILITY OF THE NECP-X CODE
Authors: Wen Xingjian , Liu Zhouyu , Huang Kai , Cao Liangzhi
Published: February 2021

The source term calculation capability is developed for the high-fidelity neutronics code NECP-X. Generally, a full activation library is used, but the memory requirement is unacceptable for the high-fidelity calculation. In order to minimize the memory requirement during the calculation with very strict conditions, a new generalized activation chain compressed method is proposed based on the influence qualification. One basic compression element is a reaction channel or an isotope, and the influence of every compression element to the final results are qualified. To enlarge the range of application of the new compressed library, an effective method to determine representative problems, which utilizes the neutron spectra and neutron flux, is developed and analyzed. Based on the ENDF-VII.0, EAF-2010 evaluated nuclear library and the influence qualification-based activation library compression method, a new compressed activation library is generated. The VERA-3A problem and the KAIST problem are used to assess the accuracy and the efficiency of the new activation library. 85 measurements of decay heat from decay heat measurement facilities GE-Morris and CLAB are used to validate the decay heat calculation in NECP-X. The results show good accuracy of NECP-X in predicting radiation source term of the spent nuclear fuel and significant memory saving when using new compressed activation library.
INTRODUCTION OF OSCAR-4 AT THE HIGH FLUX REACTOR (PETTEN)
Authors: Erasmus B. , Hendriks J.A. , Hogenbirk A. , van der Marck S.C. , Asquith N.L.
Published: February 2021

Since 2005 the nodal diffusion based code system, OSCAR-3, was used for reactor support calculations of operational cycles of the High Flux Reactor in Petten, The Netherlands. OSCAR uses a two-step deterministic calculation, in which homogenized cross sections are generated in lattice environments using neutron transport simulations, and then passed to a nodal diffusion core simulator to model the full reactor. Limitations in OSCAR-3 led to the need for improved modelling capabilities and better physics models for components present in the reactor core. OSCAR-4 offers improvements over OSCAR-3 in its approach to homogenization, and the new version of the diffusion core simulator allows for better modelling of movable components such as control rods. Fuel inventories calculated using OSCAR-4 can also easily be exported to MCNP, which allows the calculation of individual plate powers and local reaction rates amongst others. For these reasons OSCAR-4 is currently being introduced as a core support tool at the High Flux Reactor. In this work the steps that were followed to validate the reactor models are presented, and include results of validation calculations from both OSCAR-4 and MCNP6 over multiple reactor cycles. In addition differences in cross section library evaluations and their impact on the results are presented for the MCNP model.
ALTERNATIVE METHOD OF NON-DESTRUCTIVE TESTING FOR NUCLEAR POWER PLANT
Authors: Surin V.I. , Alwaheba A.I. , Beketov V.G , Gazal Abu
Published: February 2021

This study presents the results of electrophysical non-destructive testing of collectors welded joints to the body of steam generators PGV 1000Mof the reactor WWER-1000. The studied steam generators located in the Resource Сenter of National Research Nuclear University MEPhI at the site of reactors construction factory “Atommash” in Volgodonskcity-Russia. The primary loop collector was welded to the body of steam generators PGV 1000Mand the technical condition of welded joints was investigated by electrophysical non-destructive testing (EPhT). In steam generatorsmay appear corrosion and cracks due to corrosive environment and effect of stresses however,EPhTcan detect early stages of cracks.

The study highlights the theoretical part of electrophysical diagnostics and control method, using modern mathematical signals analysis, and issues related to the practical implementation of EPhT in the factory. The novel alternative method of non-destructive testing can be used in diagnostic oil and gas pipelines, elements of complex technical structures and industrial equipment. It can be used in many industries where it is necessary to process control of the entire structure or its individual parts under operating conditions.

Based on control results, potentiograms were constructed for various structural levels of the diagnostic signal. The results of controlled welded joints in steam generators showed their satisfactory condition.
AUTOMATED FUEL MANAGEMENT OPTIMIZATION FOR FAST REACTORS
Authors: Jarrett Michael , Heidet Florent
Published: February 2021

The Versatile Test Reactor (VTR) is expected to operate in a persistent non-equilibrium state due to inter-cycle variations in experimental loading. The goal of planning and optimizing the fuel loading for this mode of operation can differ from equilibrium cycle optimization. In this work, a general algorithm for optimizing a core reload of a fast reactor with respect to some objective function is developed. The objective function used in this work is a preliminary model that is defined to capture most of the core parameters expected to be of interest, but elements could be added or subtracted as needed for different types of problems. The optimization method is a discrete evolutionary algorithm. Instead of using diffusion or transport to evaluate each potential core configuration that is considered during the execution of the optimization method, the necessary inputs to the objective function (k-effective and assembly power distribution) are evaluated approximately by treating the reloaded configuration as a small change to the previous configuration, for which a diffusion or transport solution has already been calculated. This approximate calculation facilitates evaluation of the objective function for several hundred potential configurations without a neutron transport solution, which would be a significant bottleneck in the optimization method. In the results, the evolutionary algorithm demonstrates good responsiveness to the tuning of the parameters of the objective function.
BOUNDARY CONDITION MODELING EFFECT ON THE SPENT FUEL CHARACTERIZATION AND FINAL DECAY HEAT PREDICTION FROM A PWR ASSEMBLY
Authors: Hernandez-Solis Augusto , Ambrožič Klemen , Čalič Dušan , Fiorito Luca , Kos Bor , Kromar Marjan , Schillebeeckx Peter , Stankovskiy Alexey , Žerovnik Gašper
Published: February 2021

In this paper, two main exercises have been carried out to describe the effect that varying an albedo boundary condition has in the computation of observables such as decay heat, neutron emission rate and nuclide inventory from a PWR fuel assembly (or a configuration of assemblies) during a depletion scenario. The SERPENT2 code was then employed to emphasize the importance of modeling a proper boundary condition for such purposes. Moreover, the effect of taking into account more than a single fuel-pin region for depletion studies while varying the type of boundary condition, was also accounted for. The first exercise has the main objective of comparing in a single fuel assembly the albedo variations ranging from 1.1 up to full vacuum conditions. By comparing to the reference assembly (considered to be the case of full reflective conditions), relative differences up to +17% were observed in decay heat and up to almost -30% in neutron emissions. Also, a clear dependence on the albedo was detected if more than one depletable zone was considered while computing the integral value of observables of interest. Regarding the second exercise, where a 3 × 3 configuration of fuel assemblies is being now considered with a reflector section in the middle, a negligible effect on the observables was observed for the single fuel pin zone case; instead, an effect in the 244Cm computation when analyzing two fuel pin-zones produced a change in the neutron emission rate during cooling time up to 2.5% (while comparing it to the reference single assembly case).
SIMULATION OF BREED AND BURN FUEL CYCLE OPERATION OF MOLTEN SALT REACTOR IN BATCH-WISE REFUELING MODE
Authors: Raffuzzi Valeria , Krepel Jiri
Published: February 2021

The Molten Salt Reactor (MSR) is one of the most revolutionary Gen-IV reactors and it can be operated, especially with chloride salts, in the so-called breed and burn fuel cycle. In this type of fuel cycle the fissile isotopes from spent fuel do not need to be reprocessed, because the excess bred fuel covers the losses. The liquid phase of the MSR fuel assures its instant homogenization, and the reactor can be operated with batch-wise refueling thus reaching an equilibrium state. At the same time, the active core of the chloride fast MSR needs to be bulky to limit neutron leakage. In this study, the code Serpent 2 was coupled to the Python script BBP to simulate batch-wise operation of the breed and burn MSR fuel cycle. The script, previously developed for solid assemblies shuffling, was modified to simulate fuel homogenization after fertile material addition. Several fuel salts and fission products removal strategies were simulated and their impact was analyzed. Similarly, the influence of blanket volume was assessed in a two-fluid core layout. The results showed that the reactivity initially grows during the irradiation period and later decreases. The blanket has a large impact on the performance and it can be used to further increase the fuel burnup or to shrink the active core size. The breed and burn fuel cycle in MSR can reach high fuel utilization without fuel reprocessing and a multi-fluid layout can help to decrease the core size.
FUEL-CYCLE SCENARIO TO REDUCE RADIOACTIVE WASTE FROM LIGHT-WATER REACTOR
Authors: Wada Satoshi , Hiraiwa Kouji , Yoshioka Kenichi , Sugita Tsukasa , Kimura Rei
Published: February 2021

It is important to reduce the amount of trans-uranium (TRU) produced from the existing nuclear power plants to realize sustainable nuclear energy since the some TRU nuclides remain for a long time and have high radioactivity and radiotoxicity. One of the promising solutions is to transmute the TRU nuclides to those with lesser radioactivity and radiotoxicity in the existing nuclear reactors. In the current scheme, the TRU nuclides are transmuted in fast reactors and/or accelerator-driven-systems, however, this scenario seems unpromising in Japan: after the Fukushima Daiichi accident, it is required to reduce the production of TRU nuclides from the light-water reactors. In the previous studies, a concept of FORSETI was investigated, and a nuclear-fuel cycle simulation code ATRUNCYS was developed to study the low TRU production scenario. The FORSETI concept consists of two types of fuels: 1) UO2 fuels with high-assay low-enriched-uranium, and 2) MOX fuels with highly fissile concentrated plutonium reprocessed from the FORSETI-UO2 fuels. The current paper focuses on the following two scenarios: a) once-recycled scenario with the current fuel design, and b) once-recycled scenario with the FORSETI concept. The two scenarios were compared by using the ATRUNCYS code where the simulation studies showed that the amount, radioactivity, and radiotoxicity of resulting waste can be decreased in the FORSETI concept: In the case 1), the production of TRU nuclides decreased in the UO2 fuel; In the case 2), the fission rate increased and neutron-capture reactions of 240Pu and 241Pu decreased in the MOX fuels.
A NOVEL COMPUTATIONAL PLATFORM FOR THE PROPAGATION OF NUCLEAR DATA UNCERTAINTIES THROUGH THE FUEL CYCLE CODE ANICCA
Authors: Solis Augusto Hernandez , Rodriguez Ivan Merino , Fiorito Luca , Van den Eynde Gert
Published: February 2021

This paper presents the first results of a computational platform dedicated to the propagation of nuclear data covariances, all the way to fuel cycle scenario observables. Such platform, based on in-house codes developed at SCK•CEN in Belgium, both for the creation of the many-randomized nuclear data libraries based on ENDF format and for fuel cycle scenario-studies (known as SANDY and ANICCA, respectively), was employed for the uncertainty assessment of the time-dependent inventory computed from a mono-recycling of Plutonium scenario based on a PWR fleet. An essential part of the procedure that deals with the creation of input data libraries to ANICCA, has been carried out this time by the SERPENT2 code. Due to the fact that its neutron transport and depletion parallelized calculation in 72 cores for up to 1640 days and 60 MWd/kg-HM takes almost one hour, it is feasible to finish a total of 100 ANICCA runs based on randomized input libraries created from ENDF/B-VII.1 neutron-reaction covariances in about one week. Therefore, it is consider that the computation of the output population statistics can be inferred from 100 observables representing time-dependent mass inventories. To mention a few results from the aforementioned NEA/OECD benchmark scenario, it was found out that the relative standard deviation of the accumulated plutonium in the final disposal after 120 years was of 7%, while for curium it corresponded to 8%. Thus, sources of uncertainty arising from neutron-reaction covariances do have an impact in the final quantitative analysis of the fuel cycle output uncertainties.