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 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.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

A Comparison of Fast Thorium Breeder Reactor Designs With Oxide and Metallic Fuels

2008 ◽  
Author(s):  
V. Jagannathan ◽  
Usha Pal ◽  
R. Karthikeyan ◽  
Devesh Raj
Keyword(s):  
Unit Volume ◽  
Core Region ◽  
Metallic Fuel ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Depletion Rate ◽  
Metallic Fuels ◽  
Core Characteristics

Loading of seedless thoria rods in internal blanket regions and using them later as part of seeded fuel assemblies is the central theme of the thorium breeder reactor (ATBR) concept [1]. The fast reactors presently consider seedless blanket region surrounding the seeded core region. This results in slower fissile production rate in comparison to fissile depletion rate per unit volume. The overall breeding is achieved mainly by employing blanket core with more than double the volume of seeded core. The blanket fuel is discharged with fissile content of ∼30g/kg, which is much less than the asymptotic maximum possible fissile content of 100g/kg. This is due to smaller coolant flow provided for in the blanket regions. In a newly proposed fast thorium breeder reactor (FTBR) [2], the blanket region is brought in and distributed through out the core. By this the fissile depletion and production rates per unit volume become comparable. The core considered simultaneous breeding from both fertile thoria and depleted uranium and hence the concept can be called as fast twin breeder reactor as well. Sodium is used as coolant. The blanket fuel rods achieve nearly 80% of the seed fuel rod burnup and also contain nearly the maximum possible fissile content at the time of discharge. In this paper a comparison of FTBR core characteristics with oxide and metallic fuel are compared.

scholarly journals THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR

2016 ◽  
Vol 4 ◽  
pp. 22
Author(s):  
Filip Fejt
Keyword(s):  
Atmospheric Pressure ◽  
Room Temperature ◽  
Nuclear System ◽  
Hydraulic Analysis ◽  
Research Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Coolant Circulation ◽  
Thermal Hydraulic Analysis ◽  
Reactivity Insertion

The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

scholarly journals Conceptual Study on Recriticality Prevention Core Having Duplex Pellets with Neutron Absorber in Outer Core in a Fast Reactor

2019 ◽  
Vol 2019 ◽  
pp. 1-6
Author(s):  
Toshio Wakabayashi ◽  
Makoto Takahashi ◽  
Naoyuki Takaki ◽  
Yoshiaki Tachi ◽  
Mari Yano
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Thermal Characteristics ◽  
Outer Core ◽  
Severe Accident ◽  
Monte Carlo Code ◽  
Core Concept ◽  
Neutron Absorber ◽  
The Core ◽  
Nuclear Data Library

In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

scholarly journals A study on the core loading pattern of the VVER-1200/V491

2021 ◽  
Vol 7 (1) ◽  
pp. 21-27
Author(s):  
Vinh Thanh Tran ◽  
Viet Phu Tran ◽  
Thi Dung Nguyen
Keyword(s):  
Power Distribution ◽  
Nuclear Power ◽  
Safety Standards ◽  
Loading Pattern ◽  
The Core ◽  
Fuel Burn ◽  
Loading Patterns ◽  
Safety Parameters ◽  
Safety Features ◽  
Effective Multiplication

The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant.However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER- 1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns of the VVER-1200/V491 were suggested. The first loading pattern was applied from the VVER-1000/V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution, the effective multiplication factor (k-eff), and fuel burn-up were then calculated by SRAC code. To verify several safety parameters of loading patterns of the VVER-1200/V491, the neutron delayed fraction (DNF), fuel andmoderator temperature feedbacks (FTC and MTC) were investigated and compared with the safety standards in the VVER-1200/V491 FS-SAR or the VVER-1000/V392 ISAR.

scholarly journals Calculation of excore detector weighting functions for a sodium-cooled TRU burner mockup using MCNP5

2015 ◽  
Vol 5 (2) ◽  
pp. 15-25
Author(s):  
Viet Ha Pham Nhu ◽  
Min Jae Lee ◽  
Sunghwan Yun ◽  
Sang Ji Kim
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Detector Response ◽  
Weighting Functions ◽  
Control Rod ◽  
The Core ◽  
Power Regulation ◽  
Fuel Assemblies ◽  
Control Rods ◽  
Horizontal Layers

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A.

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 SENSITIVITY ANALYSIS OF STAINLESS STEEL REFLECTOR FOR VR-1 TRAINING REACTOR

2021 ◽  
Vol 247 ◽  
pp. 08003
Author(s):  
Jan Frybort ◽  
Lubomir Sklenka ◽  
Filip Fejt ◽  
Pavel Suk ◽  
Lenka Frybortova
Keyword(s):  
Stainless Steel ◽  
Power Distribution ◽  
Reactor Core ◽  
Sensitivity Study ◽  
Fast Neutrons ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Radial Reflector

Pressurized water reactors are typically surrounded in the radial direction by neutron reflectors made from stainless steel and water. These reflectors decrease neutron leakage and provide protection of pressure vessel from fast neutrons damaging its integrity. Such a radial reflector influences multiplication factor of the core and distribution of neutron flux and fission power inside the core. All these effects can be analyzed by full-core simulations using macroscopic constants. Methodology for generation of the macroscopic constants for non-fuel regions will be tested for new stainless steel reflectors at the VR-1 reactor. Rods from SS 304l material will be used for construction of radial reflectors for the VR-1 reactor. They will be design to generate sufficient measurable response in selected core characteristics. The study is focused on core power distribution and reactivity worth of absorbing rods in a VR-1 reactor core. The core typically consists of about 20 IRT-4M fuel assemblies and seven absorbing rods UR-70. Replacing water surrounding the core by several reflector assemblies containing stainless steel will influence leakage and distribution of neutrons inside the core. The current analysis deals with local effects and employs the sensitivity study to discover the nature of reflectors’ impact on the reactor core. These effects were studied even for several past VR-1 reactor core configurations. All calculations were carried out in Serpent2 Monte-Carlo code with various evaluated libraries: ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 data.

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