scholarly journals Fuel depletion calculation in MTR-LEU NUR reactor

2008 ◽  
Vol 23 (1) ◽  
pp. 11-18
Author(s):  
Foudil Zeggar ◽  
Amrane Halilou ◽  
Brahim Meftah ◽  
Mohamed Mokeddem ◽  
Fathi Letaim ◽  
...  
Keyword(s):  
Data Base ◽  
Fission Products ◽  
Reactor Fuel ◽  
Core Level ◽  
Plutonium Isotopes ◽  
The Core ◽  
Fuel Burn ◽  
Fuel Assemblies ◽  
Calculation Results ◽  
Good Agreement

In this article, we present the results of a few energy groups calculations for the NUR reactor fuel depletion analysis up to 45 000 MWd/tU taken as the maximum fuel burn up. The WIMSD-4 cell code has been employed as a calculation tool. In this study, we are interested in actinides such as the uranium and plutonium isotopes, as well as fission products Xe-135, Sm-149, Sm-151, Eu-155, and Gd-157. Calculation results regarding the five energy groups are in a good agreement with those obtained with only two energy groups which can, therefore, be used in all subsequent calculations. Calculation results presented in this article can be used as a microscopic data base for estimating the amount of radioactive sources randomly dispersed in the environment. They can also be used to monitor the fuel assemblies inventory at the core level.

scholarly journals Validation of Doppler Temperature Coefficients and Component Power Distribution for the Advanced Neutronics Component Program KYLIN V2.0

2021 ◽  
Vol 9 ◽  
Author(s):  
Lei Jichong ◽  
Xie Jinsen ◽  
Chen Zhenping ◽  
Yu Tao ◽  
Yang Chao ◽  
...  
Keyword(s):  
Engineering Design ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Engineering ◽  
Temperature Coefficients ◽  
The Core ◽  
Fuel Assemblies ◽  
Design Requirements ◽  
Design Software ◽  
Good Agreement

This work is interested in verifying and analyzing the advanced neutronics assembly program KYLIN V2.0. Assembly calculations are an integral part of the two-step calculation for core design, and their accuracy directly affects the results of the core physics calculations. In this paper, we use the Doppler coefficient numerical benchmark problem and CPR1000 AFA-3G fuel assemblies to verify and analyze the advanced neutronics assembly program KYLIN V2.0 developed by the Nuclear Power Institute of China. The analysis results show that the Doppler coefficients calculated by KYLIN V2.0 are in good agreement with the results of other well-known nuclear engineering design software in the world; the power distributions of AFA-3G fuel assemblies are in good agreement with the results of the RMC calculations, it’s error distribution is in accordance with the normal distribution. It shows that KYLIN V2.0 has high calculation accuracy and meets the engineering design requirements.

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.

scholarly journals Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

2013 ◽  
Vol 2013 ◽  
pp. 1-7 ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Power Distribution ◽  
Outer Core ◽  
Monte Carlo Code ◽  
The Core ◽  
Fuel Burn ◽  
Fuel Assemblies ◽  
Core Characteristics ◽  
Power Flattening ◽  
Reactivity Insertion ◽  
Operational Time

An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

scholarly journals Determination of 18-month fuel cycle parameters for the purpose of fuel costs minimization at the basis of use constructions of fuel assemblies in VVER-1200 reactors

2019 ◽  
Vol 5 (1) ◽  
pp. 9-15
Author(s):  
Taha M. Hashlamoun ◽  
Sergey B. Vygovsky ◽  
Sergey T. Leskin ◽  
A. Safa Duman
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fuel Cycle ◽  
Life Time ◽  
Reactor Vessel ◽  
Reactor Fuel ◽  
The Core ◽  
Fuel Assemblies ◽  
Fuel Costs

This article presents the results of research, that were focused on determining the optimal parameters of the extension of (reactor life-time) reactor fuel cycle in order to reduce the total operating costs of nuclear power plants during the transition from 12-month reactor fuel cycle to 18-month fuel cycle. The relevance of the research is related to the fact that, in recent years, there is a transition at all operating nuclear power plants VVER-1000 (1200) from 12-month reactor fuel cycle to extended 18-month fuel cycle. At the same time, represent the interests to solve the problem of conservation the extension of reactor life-time while reducing the number of loaded fuel assemblies with fresh fuel assemblies, which would reduce the total operating, and fuel costs. Search for solutions of this problem is associated with mandatory implementation of all requirements for the safe operation of the reactor and the reduction of the maximum fast neutron fluence on the reactor vessel in comparison with its value at the operating nuclear power plants. In the present work, with using the program PROSTOR software complex researched the neutron-physical characteristics of the core at the nominal parameters of the VVER-1200 reactor through the implementation of various fuel cycle strategies. The article developed various schemes of fuel-reloading for an 18-month fuel cycle with a different number of fuel assemblies. The article carries out a comparative analysis of the main parameters in the core for fuel-reloading schemes options of an 18- and 12-month fuel cycle with each other. Determine the minimum amount of fuel assemblies and provide the necessary duration of the reactor life-time for 18-month fuel cycle with using the extension of reactor life-time by reducing the power at the end of the reactor cycle to 70% of the nominal power. In the article, the arrangements of fuel assemblies were developed to provide limitations of local power by volume of the core, which reduce the fluence of fast neutrons on the reactor vessel in comparison with the projected value of the fluence. This article shows that the 18-month fuel cycle for the VVER-1200 reactor is more economical than the 12-month fuel cycle. These studies were carried out for the VVER-1200 reactor at the power of 100% of the nominal.

scholarly journals VERIFIKASI PROGRAM PWR-FUEL DALAM MANAJEMEN BAHAN BAKAR PWR

2015 ◽  
Vol 16 (1) ◽  
pp. 53 ◽  
Author(s):  
Surian Pinem ◽  
Tagor M. Sembiring ◽  
Mr Tukiran
Keyword(s):  
Power Distribution ◽  
Transport Theory ◽  
Neutron Transport ◽  
Fission Products ◽  
Diffusion Method ◽  
Fuel Management ◽  
The Core ◽  
Neutron Transport Theory ◽  
Fuel Assemblies ◽  
Nodal Diffusion

ABSTRAKVERIFIKASI PROGRAM PWR-FUEL DALAM MANAJEMEN BAHAN BAKAR PWR. Majemen bahan bakar dalam teras PWR tidak mudah karena jumlah perangkat bahan bakar dalam teras sebanyak 192 perangkat sehingga banyak kemungkinan penempatan bahan bakar dalam teras. Konfigurasi perangkat bahan bakar dalam teras harus tepat dan akurat sehingga reaktor beroperasi aman dan ekonomis. Untuk itu perlu dilakukan verifikasi program PWR-FUEL yang akan digunakan dalam manajemen bahan bakar PWR. Program PWR-FUEL didasarkan pada teori transport neutron dan diselesaikan dengan pendekatan metode difusi nodal banyak dimensi banyak kelompok dan metode difusi beda hingga (FDM). Tujuannya untuk memeriksa apakah program berfungsi dengan baik terutama untuk desain dan mana-jemen bahan bakar teras PWR. Verifikasi dilakukan dengan model pencarian teras setimbang pada tiga kondisi yaitu bebas boron, konsentrasi boron 1000 ppm dan konsentrasi boron kritis. Hasil perhitungan distribusi fraksi bakar rata-rata perangkat bahan bakar dan distribusi daya pada BOC dan EOC menunjukkan tren yang konsisten dimana perangkat bahan bakar dengan dengan daya yang tinggi pada BOC akan menghasilkan fraksi bakar yang tinggi pada EOC. Pada teras tanpa boron diperoleh faktor multiplikasi yang tinggi karena tidak adanya boron dalam teras dan efek produk fisi pada teras sekitar 3,8 %. Efek reaktivitas larutan boron 1000 ppm pada BOC dan EOC masing-masing 6,44 % dan 1,703 %. Distribusi fluks neutron dan kerapatan daya menggunakan metode NODAL dan FDM mempunyai hasil yang sama. Hasil verifikasi menunjukkan bahwa program PWR-FUEL berfungsi dengan baik terutama untuk desain dan pengolahaan bahan bakar dalam teras PWR. ABSTRACTTHE VERIFICATION OF PWR-FUEL CODE FOR PWR IN-CORE FUEL MANAGEMENT. In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron consentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is  6.44% and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR.

scholarly journals Development of a criterion for assessment of fuel washout during operation of WWER power units

2020 ◽  
Vol 6 (4) ◽  
pp. 307-312
Author(s):  
Igor A. Evdokimov ◽  
Andrey G. Khromov ◽  
Petr M. Kalinichev ◽  
Vladimir V. Likhanskii ◽  
Aleksey A. Kovalishin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Reactor Operation ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Reliable Detection ◽  
Coolant Activity

Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.

scholarly journals THE IGNALINA NUCLEAR POWER PLANT AND THE ENVIRONMENT

2006 ◽  
Vol 1 ◽  
pp. 18
Author(s):  
G. Adlys ◽  
D. Adlienė
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Calculated Data ◽  
Fission Products ◽  
Computer Code ◽  
Reactor Fuel ◽  
Fuel Burn ◽  
Data Files

Lithuania has to solve a lot of problems when closing Ignalina nuclear power plant (INPP). One of them is predicting, monitoring and managing of radiation situation and its impact to environment during this period. Calculated data files for nuclides in nuclear fuel could be used for this aim.This work presents our calculations of concentrations and activities of actinides and fission products in RBMK-1500 reactor fuel in dependence on fuel burn up and cooling time.Calculations were made using French computer code Apollo1 [1].

scholarly journals FUEL BURN-UP DISTRIBUTION AND TRANSURANIC NUCLIDE CONTENTS PRODUCED AT THE FIRST CYCLE OPERATION OF AP1000

2016 ◽  
Vol 18 (2) ◽  
pp. 101 ◽  
Author(s):  
Jati Susilo ◽  
Jupiter Sitorus Pane
Keyword(s):  
Fuel Assembly ◽  
Input Data ◽  
Reactor Core ◽  
Fission Products ◽  
Life Time ◽  
Fuel Loading ◽  
The Core ◽  
Fuel Burn ◽  
Burnable Absorber ◽  
High Level

ABSTRACT FUEL BURN-UP DISTRIBUTION AND TRANSURANIC NUCLIDE CONTENTS PRODUCED AT THE FIRST CYCLE OPERATION OF AP1000. AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB2, Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and  produce energy, fission products and new neutron. Because of the U-238 neutron absoption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic crossection, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU.  Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. Keywords: Fuel Burn-Up, Transuranic, AP1000, EOC, SRAC2006   ABSTRAK DISTRIBUSI BURN-UP DAN KANDUNGAN NUKLIDA TRANSURANIUM YANG DIHASILKAN BAHAN BAKAR PADA SIKLUS OPERASI PERTAMA TERAS AP1000. Reaktor AP1000 didesain dengan daya nominal 1154 MWe (3415 MWth), mampu beroperasi selama umur reaktor sekitar 60 tahun dan memiliki panjang tiap siklus sekitar 18 bulan. Pada siklus operasi pertama, teras AP1000 menggunakan tiga jenis pengkayaan bahan bakar UO2 yaitu 2,35 w/o, 3,40 w/o dan 4,450 w/o. Penyerap neutron ZrB2, Pyrex dan larutan Boron digunakan sebagai kompensasi reaktivitas lebih pada awal siklus. Di dalam teras reaktor, bahan bakar U-235 mengalami pembakaran melalui reaksi fisi yang akan menghasilkan energi, produk fisi dan neutron baru. Karena adanya reaksi serapan neutron oleh U-238 maka reaktor juga menghasilkan limbah radioaktif tingkat tinggi berupa nuklida transuranium yang mempunyai waktu paruh sangat panjang seperti Np, Pu, Am, dan Cm. Dalam penelitian ini dilakukan analisis hasil perhitungan distribusi burn-up bahan bakar dan kandungan nuklida transuranium yang dihasilkan oleh teras AP1000 saat akhir siklus operasi pertama. Perhitungan model geometri 2 dimensi teras AP1000 bentuk ¼ bagian dilakukan dengan paket program SRAC2006 modul COREBN/HIST. Sedangkan input data berupa tabel tampang lintang makroskopik diperoleh dari perhitungan dengan modul PIJ. Hasil prhitungan menunjukkan bahwa burn-up perangkat bahan bakar (Fuel Assembly, FA) tertinggi  adalah sebesar 27,04 GWD/MTU dan ini masih jauh lebih rendah dari batas maksimum burn-up yang diijinkan yaitu 62 GWd/MTU. Posisi pemuatan perangkat bahan bakar di bagian tengah teras akan menghasilkan burn-up dan nuklida transuranium yang lebih besar dibandingkan dengan ditepi teras. Penggunaan bahan bakar Integrated Fuel Burnable Absorber hanya sedikit berpengaruh terhadap penurunan burn-up dan nuklida transuranium yang dihasilkan. Kata kunci: Fuel burn-up, kandungan nuklida transuranium, AP1000, siklus operasi pertama, SRAC2006 

Dynamic Modeling of a BWR Control Rod Insertion System for Seismic Analysis and its Experimental Validation

2013 ◽  
Author(s):  
Yuichi Koide ◽  
Yoshihiro Goto ◽  
Yuki Sato ◽  
Hirokuni Ishigaki ◽  
Tsuyoshi Takahashi ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Dynamic Modeling ◽  
Time History ◽  
Control Unit ◽  
Hydraulic Control ◽  
Control Rod ◽  
Fuel Assemblies ◽  
Calculation Results ◽  
History Of ◽  
Good Agreement

The assessment of the seismic scrammability, which means the control rod insertability during a seismic event, is one of the most important design tasks for ensuring the seismic safety of nuclear power plants in Japan. This paper discusses the dynamic modeling of the control rod insertion behavior of a boiling water reactor (BWR) during an earthquake. A dynamic model of a control rod insertion system for BWR was developed based on multi-body dynamics. The coupled vibration behavior of the fuel assemblies in the fluid was modeled as an inertial coupling system. The effect of the interaction force between the control rod and the fuel assemblies was considered in a three-dimensional contact analysis. The hydraulic control unit and the control rod drive, which provide the control rod with drive force, were modeled in the concentrated parameter system. The model parameters, such as the friction coefficient between the control rod and the fuel assembly and the discharge coefficient of the scram piping, were obtained by conducting experiments. The validity of the model was confirmed by comparing the analytical results with the experimental ones. First, the validity of the fuel assembly model was verified through a comparison with the vibration testing in an underwater condition. It was confirmed that the calculation results for the frequency response of the fuel assembly were in good agreement with the experimental ones. Second, the validity of the modeling method of the drive system consisting of the hydraulic control unit and the control rod drive was verified through a comparison with the scram testing under non-vibration condition. The calculation results for the time history of the control rod insertion, the accumulator pressure, and the flow through the scram piping were in good agreement with the experimental ones. Finally, the validity of the modeling method of the whole system consisting of the fuel assemblies, the control rod, and the drive system was verified through a comparison with the scram testing under vibration condition. The calculation results for the time history of the control rod insertion stroke and the time delay of the insertion motion during an earthquake were in good agreement with the experimental ones. The results of these comparisons show that the developed analysis model can simulate the control rod insertion behavior during an earthquake.

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