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 

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>

Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

1981 ◽  
Vol 103 (4) ◽  
pp. 627-636 ◽  
Author(s):  
B. M. Ma
Keyword(s):  
Reactor Core ◽  
Fatigue Cracking ◽  
Computer Code ◽  
Fuel Burnup ◽  
Fuel Pellet ◽  
Gas Release ◽  
Unsteady State ◽  
Fission Gas ◽  
Fuel Elements ◽  
Fission Gas Release

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.

scholarly journals Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling

2019 ◽  
Vol 5 (4) ◽  
pp. 297-303
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Temperature Field ◽  
Mass Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Cfd Modeling ◽  
Problem Solution ◽  
Conductivity Equation ◽  
The Core ◽  
Significant Difference

Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

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.  

scholarly journals Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Soroush Heidari Sangestani ◽  
Mohammad Rahgoshay ◽  
Naser Vosoughi ◽  
Mitra Athari Allaf
Keyword(s):  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Accurate Method ◽  
Transient Simulation ◽  
Transient Pressure ◽  
Pressurized Water ◽  
Thermal Feedback ◽  
Acoustic Phenomenon

This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR). Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.

scholarly journals ANALISIS KONVEKSI ALAM TERAS REAKTOR TRIGA BERBAHAN BAKAR TIPE PELAT MENGGUNAKAN COOLOD-N2

2015 ◽  
Vol 17 (2) ◽  
pp. 67 ◽  
Author(s):  
Sudjatmi K A ◽  
Endiah Puji Hastuti ◽  
Surip Widodo ◽  
Reinaldy Nazar
Keyword(s):  
Natural Convection ◽  
Fuel Element ◽  
Flow Instability ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Coolant Temperature ◽  
Triga Reactor ◽  
Fuel Elements ◽  
Type Fuel ◽  
Plate Type

ABSTRAK Analisis Konveksi Alam Teras Reaktor Triga Berbahan Bakar Tipe Pelat MENGGUNAKAN COOLOD-N2. Rencana penghentian produksi elemen bakar jenis TRIGA oleh produsen elemen bakar reaktor TRIGA, sudah seharusnya diantisipasi oleh badan pengoperasi reaktor TRIGA untuk menggantikan elemen bakar tipe silinder tersebut dengan tipe pelat yang tersedia di pasaran. Pada penelitian ini dilakukan perhitungan untuk model teras reaktor dengan spesifikasi utama menggunakan bahan bakar U3Si2Al dengan pengayaan uranium  sebesar 19,75% dan tingkat muat 2,96 gU/cm3. Analisis dilakukan menggunakan program COOLOD-N2 yang tervalidasi pada konfigurasi teras TRIGA konversi berbahan bakar tipe pelat, yang tersusun atas 16 elemen bakar, 4 elemen kendali dan 1 fasilitas iradiasi yang terletak tepat di tengah teras. Hasil analisis menunjukkan bahwa dengan temperatur pendingin masuk ke teras sebesar 37oC, dan rasio faktor puncak daya radial ≤ 1,92 maka daya maksimum yang dapat dioperasikan pada moda operasi konveksi bebas adalah 600 kW. Karakteristik termohidrolika yang diperoleh antara lain adalah temperatur pendingin di sisi outlet, kelongsong dan meat masing-masing sebesar 82,39oC, 108,88oC, dan 109,02oC, pada ΔTONB (Temperature Onset of Nucleate Boiling) =7,18oC dan nilai OFIR (Onset of flow instability ratio) =1,03 Hasil yang diperoleh dari perhitungan ini diharapkan dapat dijadikan acuan untuk menentukan tingkat daya reaktor TRIGA berbahan bakar pelat. Kata kunci: TRIGA Konversi, COOLOD-N2, karakteristik termohidrolika, konveksi alam, elemen bakar tipe pelat.  ABSTRACT ANALYSIS OF NATURAL CONVECTION IN TRIGA REACTOR CORE PLATE TYPES FUELED USING COOLOD-N2. Any pretensions to stop the production of TRIGA fuel elements by TRIGA reactor fuel elements manufacturer should be anticipated by the operating agency of TRIGA reactor to replace the cylinder type fuel element with plate type fuel element that available on the market. In this study, the calculation of U3Si2Al fuel with uranium enrichment of 19.75 % and a load level of 2.96 gU/cm3 was performed. Analyses were performed using the validated COOLOD - N2 program. TRIGA conversion core configurations of fuel plate type are composed of 16 fuel elements, 4 control elements and 1 irradiation facilities which are located in the middle of core. The calculation results showed that if the cooling temperature was 37°C, and the ratio of radial power peaking factor ≤ 1.92, then the maximum power that can be operated on free convection mode of operation was 600 kW. The thermalhydraulic characteristic obtained such as coolant temperature at the outlet side, cladding and meat were 82.39°C, 108.88°C and 109.02°C respectively, while the ΔTONB (Temperature Onset of Nucleate Boiling) was 7.18°C and OFIR (Onset of flow instability ratio) value was 1.03. The results are expected to be used as a reference for determining the power level of the TRIGA reactor core plate types fueled. Keywords: TRIGA Convertion, COOLOD-N2, Thermalhydraulics characteristic, natural convection, plate type fuel element.

scholarly journals Modification of Neutron Kinetic Code for Plate Type Fuel Nuclear Reactor

2013 ◽  
Vol 2013 ◽  
pp. 1-6
Author(s):  
Salah Ud-Din Khan ◽  
Shahab Ud-Din Khan ◽  
Yang Zhifei
Keyword(s):  
Power Distribution ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Small Modular Reactor ◽  
Main Emphasis ◽  
Fuel Assemblies ◽  
Fortran Language ◽  
Fuel Elements ◽  
Type Fuel ◽  
Plate Type

The research is conducted on the modification of neutron kinetic code for the plate type fuel nuclear reactor. REMARK is a neutron kinetic code that works only for the cylindrical type fuel nuclear reactor. In this research, our main emphasis is on the modification of this code in order to be applicable for the plate type fuel nuclear reactor. For this purpose, detailed mathematical studies have been performed and are subjected to write the program in Fortran language. Since REMARK code is written in Fortran language, so we have developed the program in Fortran and then inserted it into the source library of the code. The main emphasis is on the modification of subroutine in the source library of the code for hexagonal fuel assemblies with plate type fuel elements in it. The number of steps involved in the modification of the code has been included in the paper. The verification studies were performed by considering the small modular reactor with hexagonal assemblies and plate type fuel in it to find out the power distribution of the reactor core. The purpose of the research is to make the code work for the hexagonal fuel assemblies with plate type fuel element.

Preliminary Evaluation of the Fixed Bed Nuclear Reactor Concept Using IAEA-INPRO Methodology

2004 ◽  
Author(s):  
Farhang Sefidvash
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Reactor Core ◽  
Primary Circuit ◽  
Preliminary Evaluation ◽  
Spent Fuel ◽  
Energy Agency ◽  
Reactor Technology ◽  
Fuel Elements

The Atomic Energy Agency (IAEA) through its INPRO Project has developed a methodology to evaluate the innovative nuclear reactors. The main objectives of INPRO are to help to ensure that nuclear energy is available in the 21st century in a sustainable manner; and bring together both technology holders and technology users to achieve desired innovations in nuclear reactors and nuclear fuel cycles which are to be acceptable to the public because they are economic, safe, proliferation resistant, sustainable, and having reduced environmental impact. Here is a preliminary application of this methodology (IAEA-TECDOC-1362) to evaluate the Fixed Bed Nuclear Reactor Concept (FBNR). Some of the characteristics of the proposed reactor are: The FBNR is based on pressurized light water reactor technology. It is a small, modular, and integrated primary circuit reactor. The fuel elements of FBNR are 8 mm diameter spherical uranium dioxide pellets cladded by zircaloy or made of compacted TRISO type fuel particles. The reactor core is suspended by the flow of water coolant. The stop in flow causes the fuel elements leave the reactor core by the force of gravity and fall into a passively cooled fuel chamber or even leave the reactor completely and become deposited in the spent fuel pool. It is an inherently safe and passively cooled reactor concept. FBNR in its advanced versions can use supercritical steam or helium gas as coolant, and utilize MOX or thorium fuel.

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 Calculation of critical core configurations of a research reactor using MTR, IRT-4M, VVR-KN fuel assemblies

2021 ◽  
Vol 8 (2) ◽  
pp. 10-18
Author(s):  
Quoc Duong Tran ◽  
Nhi Dien Nguyen ◽  
Ton Nghiem Huynh ◽  
Kien Cuong Nguyen ◽  
Minh Tuan Nguyen
Keyword(s):  
Power Distribution ◽  
Flux Distribution ◽  
Research Reactor ◽  
Reactor Core ◽  
Fuel Loading ◽  
Neutron Spectra ◽  
Fuel Assemblies ◽  
Calculation Results ◽  
Effective Multiplication ◽  
Irradiation Facilities

This paper presents calculation results to determine critical core configurations and aminimum number of fuel assemblies (FAs) or uranium mass of a research reactor loaded with three types of FAs such as MTR, IRT-4M and VVR-KN. The MCNP5 code and ENDF/B7.1 library were applied to estimate characteristics parameters of the fuel types and the whole core. Infinitive multiplication factor kinf, neutron flux distribution and neutron spectra of the fuels were calculated. The reactor core configurations with three fuel types were modeled in 3-dimensions, and then the effective multiplication factors keff, relative radial power distribution of each configuration were also evaluated. From calculation results, twelve fuel loading schemes were chosen based on lowest uranium mass or smallest number of FAs loaded into the core. In addition, two full core configurations using VVR-KN and MTR FAs and consisting of beryllium reflectors, vertical irradiation facilities, horizontal neutron beam ports, etc. have been proposed for further consideration in thermal hydraulic calculations and safety analysis.

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