scholarly journals Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core

2018 ◽  
Vol 4 (1) ◽  
pp. 7-11
Author(s):  
Pavel A. Alekseev ◽  
Aleksei D. Krotov ◽  
Mikhail K. Ovcharenko ◽  
Vladimir A. Linnik

The paper investigates the possibility for reducing the radial power peaking factor kr inside the core of a water-cooled water-moderated thermionic converter reactor (TCR). Due to a highly nonuniform power density, the TCR generates less electric power and the temperature increases in components of the thermionic fuel elements, leading so to a shorter reactor life. A TCR with an intermediate neutron spectrum has its thermionic fuel elements (TFE) arranged inside the core in concentric circles, this providing for a nonuniform TFE spacing and reduces kr. The water-cooled water-moderated TCR under consideration has a much larger number of TFEs arranged in a hexagonal lattice with a uniform pitch. Power density flattening in a core with a uniform-pitch lattice can be achieved, e.g., through using different fuel enrichment in core or using additional in-core structures. The former requires different TFE types to be taken into account and developed while the latter may cause degradation of the reactor neutronic parameters; all this will affect the design’s economic efficiency. It is proposed that the core should be split into sections with each section having its own uniform lattice pitch which increases in the direction from the center to the periphery leading so to the radial power density factor decreasing to 1.06. The number of the sections the core is split into depends on the lattice pitch, the TFE type and size, the reflector thickness, and the reactor design constraints. The best lattice spacing options for each section can be selected using the procedure based on a genetic algorithm technology which allows finding solutions that satisfy to a number of conditions. This approach does not require the reactor dimensions to be increased, different TFE types to be taken into account and developed, or extra structures to be installed at the core center.

2012 ◽  
Vol 27 (1) ◽  
pp. 75-83
Author(s):  
Milan Pesic

In 1958, the experimental RB reactor was designed as a heavy water critical assembly with natural uranium metal rods. It was the first nuclear fission critical facility at the Boris Kidric (now Vinca) Institute of Nuclear Sciences in the former Yugoslavia. The first non-reflected, unshielded core was assembled in an aluminium tank, at a distance of around 4 m from all adjacent surfaces, so as to achieve as low as possible neutron back reflection to the core. The 2% enriched uranium metal and 80% enriched uranium dioxide (dispersed in aluminum) fuel elements (known as slugs) were obtained from the USSR in 1960 and 1976, respectively. The so-called ?clean? cores of the RB reactor were assembled from a single type of fuel elements. The ?mixed? cores of the RB reactor, assembled from two or three types of different fuel elements, were also positioned in heavy water. Both types of cores can be composed as square lattices with different pitches, covering a range of 7 cm to 24 cm. A radial heavy water reflector of various thicknesses usually surrounds the cores. Up to 2006, four sets of clean cores (44 core configurations) have been accepted as criticality benchmarks and included into the OECD ICSBEP Handbook. The RB mixed core 39/1978 was made of 31 natural uranium metal rods positioned in heavy water, in a lattice with a pitch of 8?2 cm and 78


2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Giovanni Laranjo Stefani ◽  
Frederico Antônio Genezini ◽  
Thiago Augusto Santos ◽  
João Manoel de Losada Moreira

In this work a parametric study was carried to increase the production of radioisotopes in the IEA-R1 research reactor. The changes proposed to implement in the IEA-R1 reactor core were the substitution of graphite reflectors by beryllium reflectors, the removal of 4 fuel elements to reduce the core size and make available 4 additional locations to be occupied by radioisotope irradiation devices. The key variable analyzed is the thermal neutron flux in the irradiation devices.  The proposed configuration with 20 fuel elements in an approximately cylindrical geometry provided higher average neutron flux (average increment of 12.9 %) allowing higher radioisotope production capability. In addition, it provided 4 more positions to install  irradiation devices which allow a larger number of simultaneous irradiations practically doubling the capacity of radioisotope production in the IEA-R1 reactor. The insertion of Be reflector elements in the core has to be studied carefully since it tends to promote strong neutron flux redistribution in the core. A verification of design and safety parameters of the proposed  core was carried out. The annual fuel consumption will increase about 17 % and more storage space for spent fuel will be required.   


2017 ◽  
Vol 39 (4) ◽  
pp. 55-60
Author(s):  
A. A. Avramenko ◽  
N. P. Dmitrenko ◽  
М. M. Kovetskaya ◽  
Yu. Yu. Kovetskaya

Heat and mass transfer in a model of the core of a nuclear reactor with spherical fuel elements and a helium coolant was studied. The effect of permeability of the pebble bed zone and geometric parameters on the temperature distribution of the coolant in the reactor core is analyzed.  


Author(s):  
J. J. Grudzinski ◽  
C. Grandy

The reactivity of a fast spectrum nuclear reactor core is sensitive to changes in the fuel position. The core is formed by a hexagonal array of fuel assemblies which contain the fuel elements. The main structural components of the assemblies are thinwall hexagonal ducts. The fuel elements represent negligible stiffness in the fuel assembly compared to the ducts such that the ducts determine the location of the fuel. Thermal gradients across the fuel assembly cross sections create differential thermal expansion which causes the assemblies to bow. This bowing is proportional to the power to flow ratio such that it can become an important part of the reactivity change during reactor transients such as during reactor start-up, transient overpower (TOP), and unprotected loss of flow without scram (ULOF). In addition to these short term transients, thermal and fast neutron flux gradients within the core cause the assembly ducts to swell and bow over time due to irradiation creep and swelling. These latter effects produce permanent bowing of the ducts which change the reactivity over time and more importantly affect the mechanical forces required to refuel the core as the bowing is greater that the duct-to-duct clearance. Understanding these bowing responses is important to the understanding of both the transient behavior of a fast reactor as well as the refueling loads. Through proper design of the core restraint system, the bowing response can be engineered to provide negative feedback during the above mentioned transients such that it becomes part of the inherent safety of a fast reactor. Similarly, the opposing effects of creep and swelling can be manipulated to reduce the permanent core bowing deformations. We provide a discussion of the key features of analyzing and designing a core restraint system and provide a brief survey of the past work.


2016 ◽  
Vol 18 (1) ◽  
pp. 29 ◽  
Author(s):  
Tukiran Surbakti ◽  
Tagor Malem Sembiring

Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini akan dibahas analisis keselamatan uji bahan bakar mini U7Mo-Al dan U6Zr-Al ditinjau dari segi neutronik. Perhitungan dilakukan dengan menggunakan program komputer WIMSD-5B dan Batan-3DIFF. Hasil analisis menunjukkan bahwa kedua bahan bakar uji dapat diiradiasi dengan derajat bakar < 70 % selama 12 siklus operasi tanpa melampaui batas keselamatan neutronik. Kerapatan panas bahan bakar uji U7Mo-Al lebih besar dari bahan bakar U6Zr-Al.  Kata kunci: Bahan bakar mini, analisis neutronik, teras reaktor,  analisis keselamatan


2017 ◽  
Vol 2017 (4) ◽  
pp. 27-37
Author(s):  
Pavel Aleksandrovich Alekseev ◽  
Aleksei Dmitrievich Krotov ◽  
Mikhail Karpovich Ovcharenko ◽  
Vladimir Alekseevich Linnik

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi

<p>Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the  typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO<sub>2</sub> and MOX fuels. The  typical PWR-1000 core  is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are  UO<sub>2</sub> fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO<sub>2</sub> and MOX fuel are negative, so that the reactor is operated safely.</p>


1981 ◽  
Vol 103 (4) ◽  
pp. 627-636 ◽  
Author(s):  
B. M. Ma

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.


Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.


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