Optimization of spent nuclear fuel storage

Legislativní a bezpečnostní limity jsou v současné době dosahovány pomocí neutronového absorbátoru umístěného v obalových souborech. Díky svým mechanickým, chemickým a hlavně neutronickým vlastnostem je používána směs oceli a neutronového absorbátoru v podobě bóru. Výsledná slitina je používána pro transportní i skladovací komponenty. Bór je možné přidávat přímo do obalových souborů nebo zvlášť do plátů mezi trubky. Nicméně, se zvyšujícím se obohacením paliva je stále složitější bór ve slitině mechanicky zpracovávat, zároveň však dochází ke zmenšování bezpečnostních rezerv. Představené unikátní řešení je založeno na speciálních fixních neutronových absorbátorech vložených přímo do palivového souboru. Neutronový absorbátor je trvale spojen ve speciálně navržených pouzdrech. Prezentované řešení efektivněji snižuje reaktivitu systému než standardně používané absorbční pláty vložené mezi soubory i s použitím neutronové pasti. Nový přístup umožňuje výrazně změnit konstrukci obalových souborů (kontejnerů na vyhořelé palivo). Například snížením obsahu bóru ve slitině nebo zmenšením rozteče mezi palivovými soubory, což umožňuje snížit tloušťku materiálu kontejneru a tím i jeho celkovou hmotnost. Navíc je možné použít absorbátory i na zvětšení kapacity bazénu použitého paliva. Absorbátory umožňují výrobu kompaktnější mříže skladování s větší kapacitou.ObhájenoLegislative and criticality safety margins are commonly achieved by placing neutron absorbers in the cask basket design. Currently, boron content in steel or aluminium alloy is exclusively used in spent nuclear fuel transport and storage facilities absorber components as the absorber material. The reason is that the mechanical and chemical properties of light boron nuclei can be added directly to basket tube materials or placed in extra sheets between the tubes, and boron is a very good neutron absorber. Nevertheless, with higher fuel enrichment and a limit on boron content in alloys, criticality safety criteria are not easily met. A unique solution presented in this dissertation thesis is based on special fixed neutron absorbers placed directly within the fuel assembly. A neutron absorber, permanently connected in specially designed tubes, decreases system reactivity more efficiently than absorber sheets between the assemblies. This solution is more efficient than absorber tubes even with a neutron flux trap. Hence, it allows significant basket design changes (e.g., lowering boron content in steel or decreasing fuel assembly pitch in the basket resulting in lower cask wall diameter and total cask mass). It is even possible to use absorbers to increase the capacity of the spent fuel pool. Absorbers allow safe storage in a more compact rack with increased capacity

Persistent homology approach for human presence detection from 60 GHz OTFS transmissions

Orthogonal Time Frequency Space (OTFS) is a new, promising modulation waveform candidate for the next generation integrated sensing and communication (ISAC) systems, providing the environment-awareness capabilities together with high speed wireless data communications. This paper presents original results of OTFS-based person monitoring measurements in the 60 GHz millimeter-wave frequency band under realistic conditions, without the assumption of an integer ratio between the actual delays and Doppler shifts of the reflected components and the corresponding resolution of the OTFS grid. As the main contribution of the paper, we propose the use of the persistent homology technique as a method for processing of gathered delay-Doppler responses. We highlight the advantages of persistence homology approach over the standard constant false alarm rate target detector for selected scenarios

Radioisotopes production and neutron transmutations and its application at research reactors

katedra jaderných reaktor

Optimization of spent nuclear fuel storage

Legislativní a bezpečnostní limity jsou v současné době dosahovány pomocí neutronového absorbátoru umístěného v obalových souborech. Díky svým mechanickým, chemickým a hlavně neutronickým vlastnostem je používána směs oceli a neutronového absorbátoru v podobě bóru. Výsledná slitina je používána pro transportní i skladovací komponenty. Bór je možné přidávat přímo do obalových souborů nebo zvlášť do plátů mezi trubky. Nicméně, se zvyšujícím se obohacením paliva je stále složitější bór ve slitině mechanicky zpracovávat, zároveň však dochází ke zmenšování bezpečnostních rezerv. Představené unikátní řešení je založeno na speciálních fixních neutronových absorbátorech vložených přímo do palivového souboru. Neutronový absorbátor je trvale spojen ve speciálně navržených pouzdrech. Prezentované řešení efektivněji snižuje reaktivitu systému než standardně používané absorbční pláty vložené mezi soubory i s použitím neutronové pasti. Nový přístup umožňuje výrazně změnit konstrukci obalových souborů (kontejnerů na vyhořelé palivo). Například snížením obsahu bóru ve slitině nebo zmenšením rozteče mezi palivovými soubory, což umožňuje snížit tloušťku materiálu kontejneru a tím i jeho celkovou hmotnost. Navíc je možné použít absorbátory i na zvětšení kapacity bazénu použitého paliva. Absorbátory umožňují výrobu kompaktnější mříže skladování s větší kapacitou.ObhájenoLegislative and criticality safety margins are commonly achieved by placing neutron absorbers in the cask basket design. Currently, boron content in steel or aluminium alloy is exclusively used in spent nuclear fuel transport and storage facilities absorber components as the absorber material. The reason is that the mechanical and chemical properties of light boron nuclei can be added directly to basket tube materials or placed in extra sheets between the tubes, and boron is a very good neutron absorber. Nevertheless, with higher fuel enrichment and a limit on boron content in alloys, criticality safety criteria are not easily met. A unique solution presented in this dissertation thesis is based on special fixed neutron absorbers placed directly within the fuel assembly. A neutron absorber, permanently connected in specially designed tubes, decreases system reactivity more efficiently than absorber sheets between the assemblies. This solution is more efficient than absorber tubes even with a neutron flux trap. Hence, it allows significant basket design changes (e.g., lowering boron content in steel or decreasing fuel assembly pitch in the basket resulting in lower cask wall diameter and total cask mass). It is even possible to use absorbers to increase the capacity of the spent fuel pool. Absorbers allow safe storage in a more compact rack with increased capacity

Basic design of the TEPLATOR core - construction

The study shows the base optimization of the TEPLATOR core. One of the most difficult challenges for this concept is dealing with irradiated fuel assemblies. Because spent nuclear fuel has insufficient reactivity, the main aim of this study is to investigate various effects on TEPLATOR operation from the perspective of the core design. The analysis was executed by Serpent and TRITON code and shows the influence of individual components of the TEPLATOR CORE. The crucial role plays the choice of a suitable moderator; it determines the construction fundamentals of the core. Based on this choice an ideal fuel pitch, a dimension of a reflector, and parameters of cooling were arranged. The construction with or without fuel channels was dealt with. After consideration of all these effects, the first core of this kind was designed. The first DEMO is designed with 50 MW of thermal power and 55 spent fuel assemblies of VVER-440 type in the core, heavy water as both moderator and coolant. More is described in the article

Ex-core neutron flux monitoring system in graphite prism for gen. IV reactors

Nowadays Gen. IV reactors are attracting more attention, and so the same attention should be given to the safety and security requirements of these reactors. One of the most critical safety criteria is connected to the control of the chain reaction and thus to the neutron flux monitoring system. The neutron flux monitoring system is an integral part of the instrumentation and control system (I&C) on reactor that has to be able to provide a full monitoring range information throughout the whole operation cycle. Neutron flux can be in general monitored via ex-core or in-core instrumentation. The measurement can be conducted by fission chambers, selfpowered neutron detectors (SPND), proportional counters, ionization chambers and other detectors such as plastic organic scintillators etc. The accurate positioning of neutron flux detectors has to be determined for each type of Gen. IV reactor with respect to all technical and physical aspects. This paper is aimed on ex-core measurement of neutron flux of Gen IV graphite moderated reactors, while discussing the international standards, safety requirements, technical requirements, quality requirements etc. This research will be applied to the molten salt modular reactor. The specification of the neutron flux monitoring system together with possible positioning is discussed with respect to the modular reactor dimensions and power. In the final part, first steps towards to the ex-core measurements for Gen IV are made. In order to be able to determine the correct position of the detector, the model of the reactor, where preliminary test will be made is created in Serpent code and this model is validated

Fuel pebble optimization

The pebble bed reactor is randomly filled with thousands of spheres with fuel and moderator elements known as pebbles. Each fuel pebble contains thousands of coated TRISO particles stochastically embedded in the graphite matrix. Compared to PWR reactors with a regular lattice of fuel pellets in the regularly placed fuel assemblies, the fuel lattice is not easily modelled in the calculation codes. This study shows the fuel pebble models and basic optimization from the neutronics point of view. It analyzes the influence of all possible spherical crystalline structure arrangements during burnup. It also analyses the basic influences such as moderator, fuel power, enrichment and overcoat effect

VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6

Geometry models for Monte Carlo transport codes have been using standard constructive solid geometry (CSG). The standard approach is using analytical equations for defining surfaces from which spatial cells are constructed. Both union and intersection operators are available, therefore, arbitrary 3-D geometry can be modeled with CSG. However, this approach can be quite time consuming and possibly error prone for complex models. Monte Carlo transport codes are continuously developed, one of the paths is using CAD-based mesh geometry. MCNP6 features unstructured meshes (UM) created with Abaqus/CAE as geometry description. Attila4MC package for creation of UM geometry from CAD model can be used for MCNP6 models. VVER-1000 fuel assembly model in UM geometry was created for TVSA-T.mod.2 fuel type. This TVSA fuel type is exclusively operated at Temelin NPP in Czechia. Creation of the model consists of deleting small assembly parts that can be neglected for transport calculations, choosing the size of tetrahedron mesh cells and verification of the model. Basic validation of the model was performed, initially for criticality calculations. In the future, the model will be used for criticality safety analyses, preparation of boundary conditions for diffusion codes and radiation shielding analyzes of spent fuel transport and storage facilities

Feasibility of using erbium as burnable poison in VVER-1000 fuel assembly

This study shows the feasibility of using Erbium as a burnable absorber (BA) in the VVER-1000 reactor. Nowadays, Gd2O3 is commonly used BA in the form of additive compounds in the nuclear fuel for light water reactors. This analysis compares standard gadolinium absorber with a variant of absorber made of erbium in several forms. It is focused on the reactivity decrease and relative power of fresh fuel assemblies. The main aim of this study is to investigate neutronics properties of Erbium as the burnable neutron absorber and the possibility of the use in VVER-1000. Base on the Serpent calculations, a new type of fuel assemblies with Erbium absorbers are designed