scholarly journals Use of erbium as a burnable absorber for the VVER reactor core life extension

2020 ◽  
Vol 2020 (3) ◽  
pp. 62-71
Author(s):  
Saleh Hayel Alassaf ◽  
Vladimir Igorevich Savander ◽  
Abyelhamd Abdelnaby Hassan
Keyword(s):  
Reactor Core ◽  
Vver Reactor ◽  
Life Extension ◽  
Burnable Absorber

scholarly journals Use of erbium as a burnable absorber for the VVER reactor core life extension

2020 ◽  
Vol 6 (4) ◽  
pp. 275-279
Author(s):  
Saleh H. Alassaf ◽  
Vladimir I. Savander ◽  
Ahmed A. Hassan
Keyword(s):  
Spectral Component ◽  
Liquid Radioactive Waste ◽  
Reactor Core ◽  
Weight Fraction ◽  
Life Extension ◽  
Coolant Temperature ◽  
Weight Content ◽  
Burnable Absorber ◽  
Reactivity Coefficient ◽  
Fuel Elements

The paper presents the results of a computational and theoretical analysis concerned with the use of erbium as a burnable absorber in VVER-type reactors. Partial refueling options for the reactor life extension to 18 and 24 months is considered, the refueling ratio being equal to three for the 18-month life and to two for the 24-month life. Erbium is expected to be present in all fuel elements in the FA with the same weight content. The influence of the erbium weight content on such neutronic characteristics of the reactor and fuel as burn-up, reactivity coefficients, residual volume of “liquid" control, and amounts of the liquid radioactive waste (LRW) formed was assessed. The calculations were performed using a simplified model of refueling without FA reshuffling. An infinite array of polycells consisting of FAs with different in-core times was considered. The escape of neutrons from the core was taken into account by selecting the critical value K∞ at the end of life. Erbium does not burn up in full for the lifetime which affects the fuel burn-up as compared with the liquid excessive reactivity compensation system. The reduction is 0.7% per 0.1% of the erbium weight load in the fuel elements. This, however, also reduces the maximum content of the boron absorber in the coolant and the LRW accumulation in the ratio of 5% per 0.1% of the erbium weight load. Erbium influences the spectral component of the coolant temperature reactivity coefficient which turns out to be negative even with its minor weight fraction in fuel elements, and a reduction in the boron absorber fraction leads to a positive value of the density reactivity coefficient. As a result, the overall coolant temperature reactivity coefficient has a negative value throughout the lifetime.

scholarly journals Design and Calibration of a Hall Effect System for Measurement of Six-Degree-of-Freedom Motion within a Stacked Column

Sensors ◽  
2021 ◽  
Vol 21 (11) ◽  
pp. 3740
Author(s):  
Olafur Oddbjornsson ◽  
Panos Kloukinas ◽  
Tansu Gokce ◽  
Kate Bourne ◽  
Tony Horseman ◽  
...  
Keyword(s):  
Hall Effect ◽  
Degrees Of Freedom ◽  
Reactor Core ◽  
Degree Of Freedom ◽  
Life Extension ◽  
Scale Model ◽  
Design Development ◽  
Seismic Safety ◽  
Six Degrees Of Freedom ◽  
Six Degree Of Freedom

This paper presents the design, development and evaluation of a unique non-contact instrumentation system that can accurately measure the interface displacement between two rigid components in six degrees of freedom. The system was developed to allow measurement of the relative displacements between interfaces within a stacked column of brick-like components, with an accuracy of 0.05 mm and 0.1 degrees. The columns comprised up to 14 components, with each component being a scale model of a graphite brick within an Advanced Gas-cooled Reactor core. A set of 585 of these columns makes up the Multi Layer Array, which was designed to investigate the response of the reactor core to seismic inputs, with excitation levels up to 1 g from 0 to 100 Hz. The nature of the application required a compact and robust design capable of accurately recording fully coupled motion in all six degrees of freedom during dynamic testing. The novel design implemented 12 Hall effect sensors with a calibration procedure based on system identification techniques. The measurement uncertainty was ±0.050 mm for displacement and ±0.052 degrees for rotation, and the system can tolerate loss of data from two sensors with the uncertainly increasing to only 0.061 mm in translation and 0.088 degrees in rotation. The system has been deployed in a research programme that has enabled EDF to present seismic safety cases to the Office for Nuclear Regulation, resulting in life extension approvals for several reactors. The measurement system developed could be readily applied to other situations where the imposed level of stress at the interface causes negligible material strain, and accurate non-contact six-degree-of-freedom interface measurement is required.

scholarly journals Investigation of Coolant Local Hydrodynamics in the Mixed Core of the VVER Reactor

2020 ◽  
Vol 63 (2) ◽  
pp. 151-162
Author(s):  
S. M. Dmitriev ◽  
A. V. Gerasimov ◽  
A. A. Dobrov ◽  
D. V. Doronkov ◽  
A. N. Pronin ◽  
...  
Keyword(s):  
Cross Sections ◽  
Experimental Studies ◽  
Reactor Core ◽  
Cross Flow ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Computational Hydrodynamics ◽  
Cross Section Area ◽  
Fuel Assemblies ◽  
Flow Pressure

The article presents the results of experimental studies of the local hydrodynamics of the coolant flow in the mixed core of the VVER reactor, consisting of the TVSA-T and TVSA-T mod.2 fuel assemblies. Modeling of the flow of the coolant flow in the fuel rod bundle was carried out on an aerodynamic test stand. The research was carried out on a model of a fragment of a mixed core of a VVER reactor consisting of one TVSA-T segment and two segments of the TVSA-T.mod2. The flow pressure fields were measured with a five-channel pneumometric probe. The flow pressure field was converted to the direction and value of the coolant velocity vector according to the dependencies obtained during calibration. To obtain a detailed data of the flow, a characteristic cross-section area of the model was selected, including the space cross flow between fuel assemblies and four rows of fuel rods of each of the TVSA fuel assemblies. In the framework of this study the analysis of the spatial distribution of the projections of the velocity of the coolant flow was fulfilled that has made it possible to pinpoint regularities that are intrinsic to the coolant flowing around spacing, mixing and combined spacing grates of the TVSA. Also, the values of the transverse flow of the coolant caused by the flow along hydraulically nonidentical grates were determined and their localization in the longitudinal and cross sections of the experimental model was revealed. Besides, the effect of accumulation of hydrodynamic flow disturbances in the longitudinal and cross sections of the model caused by the staggered arrangement of hydraulically non-identical grates was determined. The results of the study of the coolant cross flow between fuel assemblies interaction, i.e. between the adjacent TVSA-T and TVSA-T mod.2 fuel assemblies were adopted for practical use in the JSC of “Afrikantov OKB Mechanical Engineering” for assessing the heat engineering reliability of VVER reactor cores; also, they were included in the database for verification of computational hydrodynamics programs (CFD codes) and for detailed cell-based calculation of the reactor core.

A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations

2021 ◽  
Author(s):  
Jaakko Leppänen ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
Antti Rintala ◽  
Unna Lauranto
Keyword(s):  
Fuel Cycle ◽  
Natural Circulation ◽  
Reactor Core ◽  
District Heating ◽  
Research Centre ◽  
Computational Framework ◽  
Conceptual Design Phase ◽  
Burnable Absorber ◽  
Control Rods ◽  
Type Fuel

Abstract The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.

Modeling the Spectral History in the Depletion of a PWR Core

2010 ◽  
Author(s):  
I. Bilodid
Keyword(s):  
Cross Sections ◽  
Uranium Oxide ◽  
Reactor Core ◽  
Actual Distribution ◽  
History Effects ◽  
Fuel Assemblies ◽  
Burnable Absorber ◽  
Uranium Plutonium ◽  
Moderator Density ◽  
Local Parameters

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.

Development of Safety-Enhanced Fast Reactor by Using Minor Actinide Bearing Internal Blanket

2020 ◽  
Author(s):  
Sho Fuchita ◽  
Satoshi Takeda ◽  
Koji Fujimura ◽  
Toshikazu Takeda ◽  
Kazuhiro Fujimata
Keyword(s):  
Fast Reactor ◽  
Inner Core ◽  
Reactor Core ◽  
Minor Actinide ◽  
Core Damage ◽  
Burnable Absorber ◽  
Sensitivity Studies ◽  
Core Height ◽  
Gas Expansion ◽  
Control Rods

Abstract For a 750MWe sodium-cooled fast reactor core using MOX fuel, safety-enhancement measures have been studied to reduce the risk of core damage under unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents. As passive measures the followings are considered: 1) adoption of the axial heterogeneous core configuration with sodium plenum and Gas Expansion Modules (GEMs) to lower sodium void reactivity for ULOF, and 2) addition of minor actinides (MAs) as burnable absorber and fertile nuclides to the internal blanket in the inner core to reduce burnup reactivity for UTOP. In this study, configurations of the safety-enhanced core were optimized based on sensitivity studies as follows. Firstly, effects of 1) above on the sodium void reactivity were evaluated by changing the inner core height, B-10 content of the upper shield, GEMs, and standby position of the backup control rods, which are the dominant factors of core behavior in the event of ULOF. Secondly, the effects of 2) above on the burnup reactivity were evaluated by changing the MA content in the internal blanket and the burnup period, which are the dominant factors of UTOP. Finally, by utilizing sensitivity analysis results, the safety-enhanced core which satisfies the provisional design goals has been developed. This core has negative sodium void reactivity and burnup reactivity less than 1 $.

scholarly journals Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent

2017 ◽  
Vol 781 ◽  
pp. 012029 ◽  
Author(s):  
K I Usheva ◽  
S A Kuten ◽  
A A Khruschinsky ◽  
L F Babichev
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Vver Reactor ◽  
Monte Carlo Code

scholarly journals Advanced Test Reactor: A National Scientific User Facility

2008 ◽  
Author(s):  
Clifford J. Stanley ◽  
Frances M. Marshall
Keyword(s):  
Neutron Flux ◽  
National Laboratory ◽  
Reactor Core ◽  
Nuclear Research ◽  
Reactor Power ◽  
Department Of Energy ◽  
Life Extension ◽  
Test Environment ◽  
User Facility ◽  
Test Reactor

This presentation and associated paper provides an overview of the research and irradiation capabilities of the Advanced Test Reactor (ATR) located at the U.S. Department of Energy Idaho National Laboratory (INL). The ATR which has been designated by DOE as a National Scientific User Facility (NSUF) is operated by Battelle Energy Alliance, LLC. This paper will describe the ATR and discuss the research opportunities for university (faculty and students) and industry researchers to use this unique facility for nuclear fuels and materials experiments in support of advanced reactor development and life extension issues for currently operating nuclear reactors. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration (Fig. 1) of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and chemistry can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ∼1.0 x1015 n/cm2-sec with a maximum fast flux of ∼5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

scholarly journals A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

2016 ◽  
Vol 2 ◽  
pp. 3 ◽  
Author(s):  
Yury Alekseevich Bezrukov ◽  
Vladimir Ivanovitc Schekoldin ◽  
Sergey Ivanovich Zaitsev ◽  
Andrey Nikolaevich Churkin ◽  
Evgeny Aleksandrovich Lisenkov
Keyword(s):  
Reactor Core ◽  
Vver Reactor ◽  
Core Flooding ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 229-235
Author(s):  
Y. Alzahrani ◽  
K. Mehboob ◽  
F. A. Abolaban ◽  
H. Younis
Keyword(s):  
Reactor Core ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
The Monte Carlo Method ◽  
Burnable Absorber ◽  
Effective Multiplication Factor ◽  
Reactivity Coefficient ◽  
Increasing Temperature ◽  
Effective Multiplication

Abstract In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

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