Neutronic Point Of View Research Articles

Challenges towards an acceleration in stellarator reactors engineering: The dual coolant lithium–lead breeding blanket helical-axis advanced stellarator case

Under the ambitious EUROfusion mission of “bringing the stellarator line to technological maturity”, the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a “quasi-toroidal” BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context.

Annular fuel for VVER-1000 reactor: Moderation impact on neutronic behavior

The impacts of moderation on the neutronic behavior of the VVER-1000 reactor assembly have been investigated by using annular fuel and compared with the conventional one. The study revealed that annular fuel shows a softer neutron spectrum, a better control rod worth and 16% less fissile content compared to the conventional fuel. After 60 MWd/kg burnup, annular fuel contains 6.45% and 255.41% more Pu-239 and Pu-241 respectively than the conventional fuel which can be a valuable source of fissile materials for the MOX fuel. Additionally, annular fuel shows a higher degree of symmetry in power peaking factor (PPF), higher scattering rate, 3.41% less tritium production at the beginning of cycle, and better shutdown margin without significant changes in axial & radial flux. From the neutronics point of view, annular fuel can be a propitious replacement of conventional fuel for the VVER-1000 reactor.

Development of a coupled neutronics-CFD system for 540 MWE Indian Pressurized heavy water reactors (IPHWRs) and verification with slow online refuelling transient

A loosely coupled reactor is subjected to local perturbations due to reactivity device movement that affects the core neutronics in a local region of the reactor core. It is a challenge to predict the fuel-coolant characteristics like power, enthalpy, temperature etc., from a purely neutronics point of view hence a coupled reactor physics-CFD code is a must for the reactor. Additionally, the continual change in core composition due to on-power refueling along with coolant feedback throughout change in reactor power and/or reactivity device movement makes it clear that development of a coupled code is not sufficient in itself. The coupled code should capture the burnup effect of the fuel and the change in fuel heat transfer characteristics due to fission gas accumulation in the plenum gap. In this work a cross-section library with small step cross-section data for fuel and coolant has been developed. Simple interpolation gives a reliable data for neutronics computations. The coupled code discussed in this paper is an attempt to capture fuel characteristics accurately using a small step cross-section data and explicit computation of the effect of plenum gas in the fuel. The neutronics calculation is carried out in modified TRIVENI code and the CFD computation is carried out in Openfoam. Recent trends in reactor physics is towards high fidelity computing using Monte Carlo for pin-resolved neutron transport coupled with thermalhydraulics code. However, these high fidelity codes are computer resource intensive where as our code can be run on meagre computing resources. There is scope for improvement in not so well understood phenomena like plenum gas behaviour on the fuel characteristics and this work is highlighting this aspect of PHWR fuel.

Searching for optimal accident tolerant fuel for the VVER-1200 reactor from the neutronic point of view

ABSTRACT In an attempt to find an optimum fuel, which can withstand the harsh operation condition in the VVER-1200 reactor, different fuel types were investigated in this work. Uranium carbide and uranium nitride are proposed as alternative fuels to UO2. MCNPX 2.7 with cross-section library ENDF/B-VII was used to design 3D model of VVER-1200 to investigate the neutronic properties of the suggested fuel types. The output results of the burnup of the proposed fuels have been analyzed to determine which of them achieves the longest fuel cycle. It was found that UO2 and U12C achieved criticality till 35 GWD/MTHM, while U14N and U15N achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. The traditional method of reactivity management has been tested on the proposed fuel to determine to what extent it is valid for them. To ensure the operation safety of the proposed fuel, some safety factors such as moderator temperature coefficient (MTC‎), fuel temperature coefficient (FTC‎), and effective delayed neutron fraction (βeff) have been calculated. The distribution of the thermal neutron flux and normalized power has been displayed to give us a good insight view of the VVER-1200 assembly to determine which of the proposed fuels is optimum compared to UO2.

Neutronic and Thermalhydraulic Evaluation of Annular Fuel in PWR with Three Different Nanofluids

In pressurized nuclear reactors, one of the fuel design principles is the fuel centerline temperature limit. Since the thermal conductivity of UO2 is not high, the temperature increases rapidly from the fuel surface to the fuel center. To overcome this limitation, the use of annular type fuels instead of solid fuels is one of the improvement efforts. In this study, the effects of using annular fuel in a typical PWR are investigated. At the same time, the effects of adding nanoparticles (〖Al〗_2 O_3,Cu and TiO_2) to the coolant water in the case of solid and annular fuels are also investigated. For both cases, neutronic and thermalhydraulic calculations are made and compared. In neutronic point of view, it has been shown that the use of annular fuel does not cause a significant change in the k_eff value but the presence nanoparticles reduces the k_eff. This reduction is maximum for Cu+water and minimum for 〖Al〗_2 O_3+water nanofluids. In thermalhydraulic point of view, it has been shown that the annular fuel positively affects the fuel temperature limits and the addition of nanoparticles could provide an additional contribution to this, provided that it is in very small amounts (φ

Neutronic analysis of silicon carbide Cladding-ATF fuel combinations in small modular reactors

With more attention being paid to the development of small modular reactors (SMR), the intrinsic safety issues have reached a critical point. Accident-tolerant fuel (ATF), such as the U3Si2, UO2-BeO, silicon carbide cladding etc. can maintain or improve fuel performance under normal reactor operating conditions and can maintain core integrity for a long period after the design-basis accident scenarios, providing a sufficient time margin to take measures. Therefore, applying the ATF to the SMR also enables the possibility of operating the SMR more safely. Most neutronic analyses of the ATF, however, have focused on the large-scale commercial pressurized water reactor. In this study, we adopted a given ATF design (the combination of U3Si2 and silicon carbide cladding) and compared it with other combinations of fuel and three different types of silicon carbide claddings in one single-fuel assembly and the full core of the mPower. We evaluated and analyzed the aspect of neutronic physics with Serpent-2 and OpenMOC computational codes. The results for the full core showed that the core life of U3Si2-SiC could be increased by 3.1% which feasibly could be used in the SMR from the neutronic point of view. The U3Si2, however, had a lower effect on neutron absorption than the UO2 when using the burnable poison rod in the full core. The neutronic penalties of different types of silicon carbide cladding were not significant in the SMR. Overall, the silicon carbide cladding-ATF fuel combinations were feasible for use in the mPower SMR.

Assessment of shutdown dose rates in the ITER Collective Thomson Scattering system and in equatorial port plug 12

The ITER Collective Thomson Scattering (CTS) system will be the main diagnostic responsible for measuring the velocity distribution function of fusion-born alpha particles in the plasma. As the CTS diagnostic is integrated in the equatorial port plug 12 (drawer 3), with direct apertures to the port interspace where maintenance hands-on operation will be carried out, it is essential to assess the shutdown dose rates (SDDR) in these maintenance areas. In this work, the D1S-UNED3.1.4 Monte-Carlo transport code, based on the implementation of the direct-one-step methodology in MCNP5 v1.60, was used to estimate the dose rate level 12 days (106 s) after shutdown in the port interspace. The results show that the CTS system does not contribute significantly to the SDDR in the area where hands-on maintenance is foreseen with contribution to dose rates less than 1 µSv/h. This is consistent with previous estimates, although with the most recent model of the CTS design there is a slight increase of the SDDR values. This deviation can be attributed to design changes and improved shielding modelling and/or most importantly, to statistical fluctuations of the D1S simulations. From a neutronics point of view, the increase in the SDDR falls within the range of the statistical fluctuations, and the design is still compliant with the radiation safety ALARA principle aiming at minimizing radiation doses, and there is no requirement for further design optimizations.

Searching for managing the reactivity and increasing the fuel cycle life in the PWR by an untraditional method

This study aims to find an economic method to decrease the excess reactivity and increase the fuel cycle length. The neutronic feasibility to replace the 238U with 232Th and the traditional assembly with Blanket-Seed assembly was analyzed. MCNPX code version 2.7 was used to design a three-dimension model of PWR to simulate the neutronic characteristics of the investigated fuel types and designs. The radial thermal neutron flux and the radial thermal power distribution through the assembly have been studied. The radial thermal neutron flux and thermal power distribution provide an insight view about the proposed fuel types and designs. The effect of the proposed fuel types and design on the infinity multiplication factor (Kinf), the breeding fissile concentration, the reactor-grade plutonium (rgPu), the minor actinides (MAs), conversion ratio (CR) and the concentration of the most radioactive fission products have been analyzed and compared with that of the standard fuel. From the neutronic point of view, the results showed the economic feasibility of using 232Th in regulating the initial excess reactivity.

Simple FFH pilot experiment model based on DTT-like machine

The global need for energy in the world is constantly increasing. Critical fission reactors have proved great efficiency in the energy production, but the fear of nuclear wastes and accidents due to an uncontrolled chain reaction makes these unpleasant to public. More safe fusion reactors, on the opposite, have low efficiency. Hybrid reactors capable of using the advantages of both are studied, but not yet developed. In this paper, a simple fusion–fission pilot experiment model has been developed. A Tokamak with the same characteristics of DTT (Divertor Tokamak Test facility) has been considered as a reference machine for the fusion component. The fusion system has been coupled with a relatively simple low-power fission blanket configured into three different modes by using different fuels and materials. This model could be useful in order to investigate the properties of the fusion–fission hybrid coupling from a neutronic point of view.

Evolution of the importance of neutron-induced reactions along the cycle of an LFR

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. Several LFR concepts are now in design phase, such as MYRRHA and ALFRED, and accurate nuclear data are required for the neutronic and safety assessment of the fast reactor designs. In this work, an assessment of the evolution of the importance of neutron-induced reactions along the cycle of a reference LFR design (i.e., ALFRED) with the state-of-the-art JEFF-3.3 nuclear data library is performed. Sensitivity analyses have been carried out with MCNP6 code in order to identify the most relevant isotopes and reactions from the neutronic point of view at BoL, BoC and EoC. Furthermore, an uncertainty quantification has been performed with the SUMMON system to study the evolution of uncertainties in the keff along the reactor cycle. The results from this work provide an exhaustive picture on the influence of nuclear data on core criticality performance, identifying key quantities and nuclear data needs relevant to achieve an improved safety level for LFR.

ENHANCED NUCLEAR SAFETY OF SPENT FUEL

Criticality safety analysis of highly enriched nuclear fuels requires sophisticated calculation techniques like burnup credit in order to show that fuel reactivity is below legislative limits. However, new fuel types are continuously introduced and the increase of the boron concentration in the absorbing tubes of spent fuel pool racks or spent fuel casks is necessary. Borated steel or aluminum alloys may not be economically reasonable and other types of absorber components can be feasible. Absorber tubes with neutron traps between them are very effective from the neutronics point of view; the only better solution is to place absorbers directly within the fuel assembly. This study shows a concept for enhanced nuclear safety of spent fuel. The analysis is dedicated to the selection and optimization of the most suitable materials for this concept.

Neutronic study for the dual use of nanofluid as a primary coolant and neutron absorber in VVER-1000 nuclear power reactor

The use of nanofluid as a primary coolant and reactivity controller in VVER-1000 reactor to enhance the heat transfer is the goal of this study. In this study, we have studied using nanofluid in the reactor for two cases: first, only using nanofluid as primary coolant, second, using nanofluid as a primary coolant and reactivity controller. The simulation has been done in case of normal operation of the reactor with several concentration for the different types of nanofluid. Their effect on the neutronic field has been investigated to obtain the optimum type of nanofluid from the neutronics point of view. It can be concluded that among the four types of nanofluids Al2CO3, CuO, Cu and Ag2O for the first case Ag2O is the optimum type, however, Cu is the optimum type for the second case where the boron is absent.

Calculation of fuel burnup and radionuclide inventory for the HEU and potential LEU fuels in the IRT research reactor

Abstract The study of core conversion of the IRT research reactor from the HEU to LEU fuels has been conducted in this paper. Fuel burnup, buildup of actinides and fission product inventory for the IRT research reactor with UO2-Al dispersion fuel for the various different uranium enrichments and the uranium density such as: (IRT-3M 90% 1.1 g/cm3, IRT-3M 36% 2.51 g/cm3, IRT-4M 19.75% 3.25 g/cm3, IRT-4M 19.75% 3.48 g/cm3, IRT-4M 19.75% 3.71 g/cm3) have been carried out using the GETERA code. The multiplication factor (keff) of the IRT-3M and IRT-4M reactors were calculated using the MCNP4C code. The results showed, after 200 days of the reactor operation time that the total consumptions of 235U for the five cases were 73.85, 73.13, 72.66, 72.80 and 73.21 g respectively. The amounts of 239Pu produced in the core were: 0.43, 3.39, 5.90, 6.13 and 6.35 g respectively. The total radioactivities of fuel assembly were: 2.78E+15, 2.83E+15, 2.86E+15, 2.88E+15 and 2.89E+15 Bq respectively. Finally, from the neutronic view point, the LEU fuel was validated to be a suitable candidate for the fuel conversion of the IRT in the future.

Influence of accident tolerant fuel cladding material on the European pressurised reactor core neutronic characteristics

Research on accident tolerant fuel became an important research area to improve safety of the nuclear reactors, especially following the Fukushima disaster. Numerous studies have demonstrated that silicon carbide is a promising candidate material that could potentially replace the zirconium cladding in LWR reactors, which is currently under active research by multiple research and industry organizations. Many of material testing of silicon carbide were already performed and the results are available. Also molybdenum is considered as fuel cladding. This paper is focused on the potential of using silicon carbide and molybdenum as a fuel cladding in the European Pressurized Reactor (EPR) core from the neutronic point of view.

Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

Abstract In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

In this paper, the SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding, and its behavior from the neutronic viewpoint was compared with the Zircaloy. To evaluate the average temperature of the fuel, cladding, and moderator used in the neutronic studies, a simplified thermal-hydraulic simulation of the core was performed using the RELAP5 code for the two cladding materials. Then, after that, a standard PWR fuel element was simulated, considering normal conditions of operation at Beginning Of Life (BOL) and full power, and using the average temperatures coming from the RELAP5 code calculation. The effects of burnable poison rods and temperature feedback on the core neutron multiplication factor have been evaluated changing the clad material at criticality steady state condition. Finally, the fuel element is analyzed during burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code.

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

This paper presents the results of a neutronics analysis related to the homogeneous recycling of different mixtures of transuranic elements (transuranics) (TRU) in pressurized water reactors (PWRs) loaded with mixed oxide (MOX) fuel using enriched uranium instead of depleted uranium (UenrO2-TRUO2, i.e., MOX-EU). It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely, the use of reprocessed uranium. From a neutronics point of view, it is possible to multirecycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one-third and two-thirds of the fleet. Recycling neptunium and americium with plutonium significantly decreases the decay heat of the waste stream between 100 to 1000 years compared to that of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235U, and recycling it makes a major difference in the natural uranium needs. For example, at equilibrium, a PWR fleet recycling its plutonium, neptunium, and americium in MOX-EU needs 28% more natural uranium than a reference UO2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19% less if the reprocessed uranium is recycled back in the reactors, i.e., a 47% difference. Reenriching the reprocessed uranium is not necessary.

ITER core CXRS diagnostic: Assessment of different optical designs with respect to neutronics criteria

The Charge Exchange Recombination Spectroscopy (CXRS) diagnostic aims to measure emission lines of impurity isotopes in the ITER plasma in order to quantify several parameters like the composition of the plasma (density of helium, deuterium or tritium), the ion temperature or rotation velocities. The core plasma CXRS shall be installed in one of the ITER Upper Port Plugs (UPP #3). Currently, four different optical layouts are being assessed with respect to the optical performance, engineering feasibility, cost, maintenance especially with respect to remote handling and the performance of their neutron radiation shielding.This work is devoted to the neutronic analysis performed in support for the design of ITER CXRS-core Diagnostic System, presently under development by the IC3 Consortium (FZJ, KIT, BME, Wigner RCP, TU/Eindhoven, FOM-DIFFER, CCFE, CIEMAT, Optimal Optik). Results of the neutronic analyses are presented showing the differences between the four different designs with respect to several nuclear responses such as neutron fluxes around the upper port plug and in the port interspace, maps of nuclear heating around the UPP including the toroidal/poloidal field coils as well as the vacuum vessel. Furthermore, radiation damage maps were calculated covering large areas of the upper port plug and of its environment. The results indicate the viability of one of the preferred designs from the neutronic point of view but also show the potential for improvements.

Neutronics characterization of an erbia fully poisoned PWR assembly by means of the APOLLO2 code

Recently, increasing demands on the reduction of fuel cycle costs have led to higher burnup fuel designs. According to the erbia-credit super high burnup fuel concept, developed by mixing low content of erbia to UO2 powder directly after reconversion process so that all fuel pins in a given fuel assembly are homogeneously doped, the present study aims to characterize, from a neutronic point of view, a 17 × 17 pressurized water reactor assembly enriched to 10.27 wt.% in 235 U with an erbia content of 1 at.% (i.e. 0.7 wt.%) by means of the deterministic neutronic code APOLLO2. For this purpose, a simplified thermal-hydraulic analysis was performed in order to evaluate the effects on fuel thermal conductivity of adding erbia to uranium oxide. The results obtained allow to conclude that an Er-doped assembly enriched to >5 wt.% in 235 U represents an advantageous solution for very long fuel cycles, and it is so suited for very high burnups.