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.

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

2016 ◽  
Vol 6 (2) ◽  
pp. 21-30
Author(s):  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Minh Tuan Nguyen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Benchmark Problem ◽  
Fuel Assemblies ◽  
Effective Multiplication Factor ◽  
Neutronic Characteristics ◽  
Effective Multiplication ◽  
Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

scholarly journals Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code

2021 ◽  
Vol 8 (1) ◽  
pp. 10-16
Author(s):  
Nguyen Thanh Vinh Ho ◽  
Vinh Vinh Le ◽  
Nhi Dien Nguyen ◽  
Kien Cuong Nguyen ◽  
Ton Nghiem Huynh ◽  
...  
Keyword(s):  
Power Distribution ◽  
Nuclear Physics ◽  
Research Reactor ◽  
Reactor Core ◽  
Calculated Data ◽  
Thermal Hydraulics ◽  
Fuel Type ◽  
Fuel Assemblies ◽  
Enriched Uranium ◽  
New Research

VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems to be suitable for VVR-KN fuel type. Before using PLTEMP code for thermal-hydraulics analysis of the new RR, a calculation for code validation was performed based on the data of the VVR-KN fuel lead test. First, MCNP5 code was used to calculate the power distribution of WWR-K reactor core with lead test fuel assemblies (LTAs) at the core center. Then, thermal-hydraulics parameters of the LTAs were obtained by using PLTEMP code together with calculated data of the power distribution and the lead test conditions. A comparison between the analytic results and the lead test data was made to confirm the suitability of PLTEMP code for thermal-hydraulics analysis of VVR-KN fuel under forced convection and downward flow conditions.

A Small PWR-Core Physical Calculation Based on PWR-Core Analysis Code CORAL

2021 ◽  
Author(s):  
Wen Yang ◽  
Lun Zhou ◽  
Junrong Qiu ◽  
Yun Tai
Keyword(s):  
Density Distribution ◽  
Power Distribution ◽  
Channel Model ◽  
Three Dimensional ◽  
Least Square ◽  
Object Oriented Programming ◽  
Core Analysis ◽  
Fuel Temperature ◽  
Fuel Assemblies ◽  
Calculation Results

Abstract Three dimensional PWR-core analysis code CORAL is developed by Wuhan Second Ship Design and Research Institute. This code provides basic functions including three-dimensional power distribution, fine power reconstruction, fuel temperature distribution, critical search, control rod worth, reactivity coefficients, burnup and nuclide density distribution, etc. CORAL employ nodal expansion method to solve neutron diffusion equation, and the least square method is used to achieve few group constants, and sub-channel model and one-dimensional heat transfer is used to calculate fuel temperature and coolant density distribution, and burnup distribution and nuclide nuclear density could be obtained by solving macro-depletion and micro-depletion equation. The CORAL code is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to analyze the computational accuracy of the CORAL code in small PWR-core and its capability to deal with heterogeneous, calculation analysis are carried out based on the material and geometry parameters of the SMART core. The core has 57 fuel assemblies, with 8, 20 or 24 gadolinium rods arranged in the fuel assemblies. In this paper, a quantitative comparison and analysis of the small PWR problem calculation results are carried out. Numerical results, including effective multiplication factor, assembly power distribution and pin power distribution, all agree well with the calculation results of OpenMC or Bamboo at both hot zero-power (HZP) and hot full-power (HFP) conditions.

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 VALIDATION OF PWR-FUEL CODE FOR STATIC PARAMETERS IN THE LWR CORE BENCHMARK

2018 ◽  
Vol 20 (3) ◽  
pp. 111 ◽  
Author(s):  
Iman Kuntoro ◽  
Surian Pinem ◽  
Tagor Malem Sembiring
Keyword(s):  
Reference Values ◽  
Finite Difference Methods ◽  
Reactor Core ◽  
Maximum Deviation ◽  
Multiplication Factor ◽  
Static Parameter ◽  
Calculation Results ◽  
Effective Multiplication Factor ◽  
Different Characteristics ◽  
Effective Multiplication

The PWR-FUEL code is a multi dimensional, multi group diffusion code with nodal and finite difference methods. The code will be used to calculate the fuel management of PWR reactor core. The result depends on the accuracy of the codes in producing the core effective multiplication factor and power density distribution. The objective of this research is to validate the PWR-FUEL code for those cases. The validation are carried out by benchmarking cores of IAEA-2D, KOERBERG-2D and BIBLIS-2D. The all three cases have different characteristics, thus it will result in a good accuracy benchmarking. The calculation results of effective multiplication factor have a maximum difference of 0.014 %, which is greater than the reference values. For the power peaking factor, the maximum deviation is 1.75 % as compared to the reference values. Those results show that the accuracy of PWR-FUEL in calculating the static parameter of PWR reactor benchmarks are very satisfactory.Keywords: Validation, PWR-FUEL code, static parameter. VALIDASI PROGRAM PWR-FUEL UNTUK PARAMETER STATIK PADA TERAS BENCHMARK LWR. Program PWR-FUEL adalah program difusi multi-dimensi, multi-kelompok dengan metode nodal dan metode beda hingga. Program ini akan digunakan untuk menghitung manajemen bahan bakar teras reaktor PWR. Akurasi manajemen bahan bakar teras PWR tergantung pada akurasi program dalam memprediksi faktor multiplikasi efektif teras dan distribusi rapat daya. Untuk itu dilakukan validasi program PWR-FUEL sebagai tujuan dalam penelitian ini.  Validasi PWR-FUEL dilakukan menggunakan teras benchmark IAEA-2D, KOERBERG-2D dan BIBLIS-2D. Ketiga kasus ini mempunyai karaktristik yang berbeda sehingga akan memberikan hasil benchmark yang akurat. Hasil perhitungan faktor multiplikasi efektif terdapat perbedaan maksimum adalah 0,014 % lebih besar dari referensi. Sedangkan untuk perhitungan faktor puncak daya, terdapat perbedaan maksimum 1,75 % dibanding harga referensi. Hasil perhitungan menunjukkan bahwa akurasi paket program PWR-FUEL dalam menghitung parameter statik benchmark reaktor PWR menunjukkan hasil yang sangat memuaskan.Kata kunci: Validasi, program PWR-FUEL, parameter statik

Validation of HELIOS Neutron Cross-Section Library for RBMK Reactors Against the Data From the Critical Facility Experiments

2002 ◽  
Author(s):  
Audrius Jasiulevicius ◽  
Bal Raj Sehgal
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Cross Sections ◽  
Measurement Errors ◽  
Reactor Core ◽  
Rbmk Reactor ◽  
Cross Section Data ◽  
Section Data ◽  
Fuel Assemblies ◽  
Calculation Results

The RBMK reactors are channel type, water-cooled and graphite moderated reactors. The first RBMK type electricity production reactor was put on-line in 1973. Currently there are 13 operating reactors of this type. Two of the RBMK-1500 reactors are at the Ignalina NPP in Lithuania. Experimental Critical Facility for RBMK reactors, located at Kurchiatov Institute, Moscow was designed to carry out critical reactivity experiments on assemblies, which imitate parts of the RBMK reactor core. The facility is composed of Control and Protection Rods (CPR’s), fuel assemblies with different enrichment in U-235 and other elements, typical for RBMK reactor core loadings, e.g. additional absorber assemblies, CPR imitators, etc. A simulation of a set of the experiments, performed at the Experimental Critical Facility, was carried out at the Royal Institute of Technology (RIT), Nuclear Power Safety Division, using CORETRAN 3-D neutron dynamics code. The neutron cross sections for assemblies were calculated using HELIOS code. The aim of this work was to evaluate capabilities of the HELIOS code to provide correct cross section data for the RBMK reactor. The calculation results were compared to the similar CORETRAN calculations, when employing WIMS-D4 code generated cross section data. For some of the experiments, where calculation results with CASMO-4 code generated cross sections are available, the comparison is also performed against CASMO-4 results. Eleven different experiments were simulated. Experiments differ in size of the facility core (number of assemblies loaded): from simple core loadings, composed only of a few fuel assemblies, to complicated configurations, which represent a part of the RBMK reactor core. Diverse types of measurements were carried out during these experiments: reactivity, neutron flux distributions (both axial and radial), rod reactivity worth and the voiding effects. Results of the reactivity measurements and relative neutron flux distributions were given in the Experiment report [1] as parameters, to be obtained using static calculations, i.e. the reported results were already processed numerically using the facility equipment, e.g. the reactimeter. The reported measurement errors consist only of instrumentation errors, i.e. measurement method errors and the influence from the space–time effects were not included in the error evaluation.

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.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Kerntechnik ◽  
2021 ◽  
Vol 86 (4) ◽  
pp. 302-311
Author(s):  
M. E. Korkmaz ◽  
N. K. Arslan
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Study ◽  
Minor Actinides ◽  
Monte Carlo Code ◽  
Effective Multiplication Factor ◽  
Conceptual Core ◽  
Effective Multiplication

Abstract Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

scholarly journals Distribution of gaps between fuel assemblies in the VVER 1000 core

2018 ◽  
pp. 18-28
Author(s):  
A.M. Abdullayev ◽  
A.I. Zhukov ◽  
S.V. Maryokhin ◽  
S.D. Riabchykov
Keyword(s):  
Distribution Function ◽  
Size Distribution ◽  
Reactor Core ◽  
Energy Functional ◽  
Size Distribution Function ◽  
Design Parameters ◽  
Fuel Loading ◽  
Fuel Rods ◽  
The Core ◽  
Fuel Assemblies

The safety of fuel loading of VVER reactors is justified by calculations of the neutronic characteristics of the forthcoming campaign. These calculations are based on the design parameters of fuel assemblies (FA) — fuel enrichment, materials, design features, etc. However, during operation, some parameters change in an uncontrolled manner. In particular, FA can deform — bend or twist, this leads to the appearance of increased gaps between the fuel assemblies. These regions filled with a moderator lead to an increase in comparison with the calculations for the generation of thermal neutrons and, as a result, to a surge in the power of the fuel rods surrounding these regions. Safety requirements limit the power of fuel rods. Therefore, design capacities are increased by means of the so-called engineering margin factor to account for random outbursts. The deviation of the size of the water gaps between the fuel assemblies from the design ones should be known to calculate this coefficient, for example, the size distribution function of the gaps. This information is most often obtained by modeling the mechanical state of fuel assemblies in the reactor core. Other approaches are based on experimental data. Measurements in the core during the campaign are not possible. Therefore, the geometric parameters of the fuel assembly after the discharging from the core are measured. The presented paper uses the data of such measurements obtained after the 24th fuel campaign of ZNPP unit 4. It is assumed that the fuel assemblies tend to retain the form they have in the “free” state, and mechanical interaction with neighboring fuel assemblies leads to a certain equilibrium state that can be easily analyzed. In contrast to similar calculations, the elastic energy functional of interacting fuel assemblies is proposed whose minimum gives the required size distribution function of the gaps. 24 and 25 campaigns were modelled; the role of inter-sector FA shuffling was studied. The distribution of the gaps between the fuel assemblies in VVER-1000 core is calculated based on the measured deformations of the fuel assemblies discharged from the core and the elastic characteristics of the fuel assemblies. It was demonstrated that 95 % of gaps in the cores both with FA-A and FA-WR do not exceed 7.6 mm. The results can be used to calculate the engineering margin factor in determining the peaking factors of energy release.

Spent Fuel Canister Criticality Safety Calculation in Groundwater Immersion Accident

2014 ◽  
Author(s):  
Siyuan Wu ◽  
Bo Yang ◽  
Hexi Wu ◽  
Qianglin Wei ◽  
Yibao Liu
Keyword(s):  
Safety Issue ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Safety Standard ◽  
Fuel Assemblies ◽  
Calculation Results ◽  
Computer Codes ◽  
Effective Multiplication Factor ◽  
Effective Multiplication ◽  
Criticality Safety

This paper studies the KBS-3 spent fuel canister criticality safety issue in the event of a groundwater penetration accident. First, the study calculates the composition of the PWR spent fuel assemblies by using MCBURN software, and the coupling of the MCNP and ORIGEN 2.1 computer codes. Then, with the help of Isotope Generation and Depletion Code, it calculates the composition of various types of actinides and their daughters for 100,000 years. Finally, the above calculation results are used in MCNP5 to calculate the effective multiplication factor keff of the canister for different degrees of groundwater penetration. The study finds that the maximum keff of the canister is 0.79609 in groundwater penetration accident, satisfying criticality safety standard.

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