scholarly journals Analysis Of Doppler Reactivity Coefficient On The Typical Pwr-1000 Reactor With Mox Fuel

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Fuel Temperature ◽  
Safety Level ◽  
The Core ◽  
Mox Fuel ◽  
Calculation Results ◽  
Reactivity Coefficient ◽  
Fuel Elements

<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>

scholarly journals ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

2016 ◽  
Vol 18 (3) ◽  
pp. 127 ◽  
Author(s):  
Setiyanto Setiyanto ◽  
Tukiran Surbakti
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Empty Space ◽  
Computer Code ◽  
Safety Analysis ◽  
Design Calculation ◽  
Triga Reactor ◽  
Calculation Results ◽  
Fuel Elements ◽  
Irradiation Facilities

ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g), but very low value for Lazy Susan position (lest then 0,11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung, telah dilakukan kajian penggunaan bahan bakar jenis pelat seperti yang digunakan oleh RSG-GAS. Berbagai langkah analisis telah disiapkan, termasuk perhitungan desain teras, dan sistem keselamatannya. Penggunaan elemen bakar tipe pelat menghasilkkan reaktor dapat dioperasikan hanya dengan 20 elemen bakar. Dibandingkan teras aslinya, nampak bahwa teras baru menjadi lebih kecil dan kompak, rapat dayanya naik, tetapi menyisakan beberapa ruang kosong yang dimungkinkan untuk menempatkan fasilitas iradiasi di teras. Dengan adanya fasilitas iradiasi di dalam teras, maka pembangkitan panas gamma di teras menjadi faktor baru yang harus diperhatikan. Untuk alasan ini, telah dilakukan perhitungan pembangkitan panas gamma teras reaktor Triga 2000 Bandung mengunakan program Gamset. Perhitungan didasarkan pada persamaan atenuasi liner, sumber garis dan arah perambatan tiga dimensi. Selain panas gamma di teras, akan dihitung juga panas gamma di reflektor (Lazy Susan), dan di CIP untuk berbagai jenis bahan. Diperoleh hasil bahwa panas gamma di CIP cukup signifikan (0,87 w/g), tetapi di posisi Lazy Susan relatif kecil, rata-rata hanya 0,11 w/g. Dari hasil tersebut dapat disimpulkan bahwa penggunaan CIP untuk iradiasi perlu mempertimbangkan panas gamma dalam perhitungan LAK nya. Kata kunci: panas gamma, reaktor nuklir, reaktor penelitian, keselamatan reaktor 

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.

scholarly journals EFFECT OF PERMEABILITY OF PEBLE BED ON HEAT TRANSFER IN THE CORE OF NUCLEAR REACTOR WITH HELIUM COOLANT

2017 ◽  
Vol 39 (4) ◽  
pp. 55-60
Author(s):  
A. A. Avramenko ◽  
N. P. Dmitrenko ◽  
М. M. Kovetskaya ◽  
Yu. Yu. Kovetskaya
Keyword(s):  
Heat Transfer ◽  
Mass Transfer ◽  
Temperature Distribution ◽  
Heat And Mass Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Geometric Parameters ◽  
The Core ◽  
Fuel Elements ◽  
Helium Coolant

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.  

Fuel Assembly Bowing and Core Restraint Design in Fast Reactors

2014 ◽  
Author(s):  
J. J. Grudzinski ◽  
C. Grandy
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Transient Behavior ◽  
Restraint System ◽  
The Core ◽  
Fuel Elements ◽  
Over Time

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.

scholarly journals Gas-cooled nuclear reactor core shaping using heat exchange intensifiers

2019 ◽  
Vol 5 (1) ◽  
pp. 75-80
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Heat Exchange ◽  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Temperature Fields ◽  
Ansys Fluent ◽  
Safety Requirements ◽  
Cooling Channels ◽  
The Core ◽  
Nuclear Reactor Core

The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

Nodal Expansion and its Application in AP1000 Nuclear Reactor Core Monitoring System

2013 ◽  
Vol 448-453 ◽  
pp. 1907-1911
Author(s):  
Wei Zhi Jia ◽  
Rui Wang ◽  
Yun Zhou
Keyword(s):  
Monitoring System ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
The Core ◽  
Nodal Model ◽  
Nuclear Reactor Core ◽  
Nodal Expansion Method ◽  
Full Core

As the core monitoring system of AP1000, BEACON always uses a full-core nodal model for core monitoring based on the ANC-NEM nodal model. The theory behind the nodal expansion method is discussed, and the application of the method in BEACON is described. Finally, an ANC-NEM calculation simulation is proposed.

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

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
Keyword(s):  
Power Density ◽  
Hexagonal Lattice ◽  
Reactor Core ◽  
The Core ◽  
Intermediate Neutron ◽  
Density Factor ◽  
Fission Power ◽  
Concentric Circles ◽  
Thermionic Conversion ◽  
Fuel Elements

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.

RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis

2002 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
3D Model ◽  
Reactor Core ◽  
Core Analysis ◽  
Reactor Operation ◽  
The Core ◽  
Calculation Results ◽  
Plant Data ◽  
Comparison Of The Results ◽  
Real Plant ◽  
Good Agreement

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data.

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