The Re-Evaluation of the AVR Melt-Wire Experiment Using Modern Methods With Specific Focus on Bounding the Bypass Flow Effects

2008 ◽  
Author(s):  
Carel F. Viljoen ◽  
Sonat Sen ◽  
Frederik Reitsma ◽  
Onno Ubbink ◽  
Peter Pohl ◽  
...  
Keyword(s):  
Gas Flow ◽  
Coupled System ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Melting Temperatures ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Flow Effects ◽  
The Many

The AVR (Arbeitsgemeinschaft Versuchsreaktor) is a pebble bed type helium cooled graphite moderated high temperature reactor that operated in Germany for 21 years and was closed down in December 1988 [1]. The AVR melt-wire experiments [2], where graphite spheres with melt-wires of different melting temperatures were introduced into the core, indicate that measured pebble temperatures significantly exceeded temperatures calculated with the models used at the time [3]. These discrepancies are often attributed to the special design features of the AVR, in particular the control rod noses protruding into the core, and to inherent features of the pebble bed reactor. In order to reduce the uncertainty in design and safety calculations the PBMR Company is re-evaluating the AVR melt-wire experiments with updated models and tools. 3-D neutronics thermal-hydraulics analyses are performed utilizing a coupled VSOP99-STAR-CD calculation. In the coupled system VSOP99 [4] provides power profiles on a geometrical mesh to STAR-CD [5] while STAR-CD provides the fuel, moderator and solid structure temperatures to VSOP99. The different fuel histories and flow variations can be modelled with VSOP99 (although this is not yet included in the model) while the computational fluid dynamics (CFD) code, STAR-CD, adds higher-order thermal and gas flow modelling capabilities. This coupling therefore ensures that the correct thermal feedback to the neutronics is included. Of the many possible explanations for the higher-than-expected melt-wire temperatures, flow bypassing the pebble core was identified as potentially the largest contributor and was thus selected as the first topic to study. This paper reports the bounding effects of bypass flows on the gas temperatures in the top of the reactor. It also presents preliminary comparisons between measured temperatures above the core ceiling structure and calculated temperatures. Results to date confirm the importance of correctly modelling the bypass flows. Plans on future model improvements and other effects to be studied with the coupled VSOP99-STAR-CD tool are also included.

Optimization of a Radially Cooled Pebble Bed Reactor

2008 ◽  
Author(s):  
B. Boer ◽  
J. L. Kloosterman ◽  
D. Lathouwers ◽  
T. H. J. J. van der Hagen ◽  
H. van Dam
Keyword(s):  
Pressure Drop ◽  
Temperature Profile ◽  
Flow Direction ◽  
Outer Region ◽  
Fissile Material ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
The Core ◽  
Power Profile

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as 1 cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = −2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = −50 °C) can be achieved compared to present axially cooled designs.

Investigations on the Movement of Fuel Elements in the Core of a Pebble Bed Reactor

1969 ◽  
Vol 91 (2) ◽  
pp. 390-394
Author(s):  
D. Bedenig ◽  
C. B. v. d. Decken ◽  
W. Rausch
Keyword(s):  
Theoretical Model ◽  
Flow Behavior ◽  
Experimental Studies ◽  
Experimental Equipment ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
The Core ◽  
Method Of Measurement ◽  
Fuel Elements ◽  
Type Fuel

For several years gas-cooled high temperature reactors have been developed in Germany, the main feature of which are their pebble-type fuel elements. The pebble bed is in the state of a continuous circulation process which is the reason for a series of nuclear and technical advantages. To make use of these advantages, comprehensive experimental studies on the flow behavior of a pebble bed were carried out. First, experimental equipment and the most successful method of measurement are described. Then typical results of parameter studies are reported as well as a theoretical model to calculate the pebble bed flow behavior. At last typical functions describing the flow behavior in the core of the THTR 300 MWe Prototype Reactor are reported.

scholarly journals GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE-BED REACTOR

2019 ◽  
Vol 25 (1) ◽  
Author(s):  
Hery Adrial ◽  
Amir Hamzah ◽  
Entin Hartini
Keyword(s):  
Gamma Radiation ◽  
Dose Rate ◽  
Gamma Dose ◽  
Gamma Dose Rate ◽  
Minimum Thickness ◽  
Critical Height ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
The Core ◽  
Rate Analysis

GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE BED REACTOR. HTGR-10 MWth is a high-temperature gas-cooled reactor. The fuel and moderator are pebble shaped with a radius of 3 cm. One fuel pebble consists of thousands of UO2 kernels with a density of 10.4 gram/cc and the enrichment rate of 17%. The core of HTGR-10 MWth is the center of origin of neutrons and gamma radiation resulting from the interaction of neutrons with pebble fuel, moderator and biological shield. The various types of radiations generated from such nuclear reactions should be monitored to ensure the safety of radiation workers. This research was conducted using MCNP-6 Program package with the aim to calculate and analyze gamma radiation dose in biological shield of HTGR-10 MWth. In this study, the biological shield is divided into 10 equal segments. The first step of the research is to benchmark the created program against the critical height of HTR-10. The results of the benchmarking show an error rate of ± 1.1327%, while the critical core height of HTGR 10 MWth for the ratio of pebble fuel and pebble moderator (F:M) of 52: 48 occurs at a height of 134 cm. The rate of gamma dose at the core is 3.0052E + 05 mSv/hr. On the biological shield made of regular concrete with a density of 2.3 grams/cc, the rate of gamma dose decreases according to an equation y = 0.0042 e-0.03x. Referring to Perka Bapeten no 4 of 2013, the safe limits for workers and radiation protection officers will be achieved if the minimum thickness of biological shield is 115 cm with gamma dose rate of 0 mSv/hour.Keywords: Gamma dose rate, HTGR 10 MWth, biological shield, pebble

scholarly journals The effects of fuel type on control rod reactivity of pebble-bed reactor

Nukleonika ◽  
2019 ◽  
Vol 64 (4) ◽  
pp. 131-138
Author(s):  
◽  
Topan Setiadipura ◽  
Jim C. Kuijper ◽  
Keyword(s):  
Reactor Core ◽  
Fuel Type ◽  
Reactor Model ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Reactor Geometry ◽  
Calculation Results ◽  
Control Rods ◽  
Selection Of

Abstract As a crucial core physics parameter, the control rod reactivity has to be predicted for the control and safety of the reactor. This paper studies the control rod reactivity calculation of the pebble-bed reactor with three scenarios of UO2, (Th,U)O2, and PuO2 fuel type without any modifications in the configuration of the reactor core. The reactor geometry of HTR-10 was selected for the reactor model. The entire calculation of control rod reactivity was done using the MCNP6 code with ENDF/B-VII library. The calculation results show that the total reactivity worth of control rods in UO2-, (U,Th)O2-, and PuO2-fueled cores is 15.87, 15.25, and 14.33%Δk/k, respectively. These results prove that the effectiveness of total control rod in thorium and uranium cores is almost similar to but higher than that in plutonium cores. The highest reactivity worth of individual control rod in uranium, thorium and plutonium cores is 1.64, 1.44, and 1.53%Δk/k corresponding to CR8, CR1, and CR5, respectively. The other results demonstrate that the reactor can be safely shutdown with the control rods combination of CR3+CR5+CR8+CR10, CR2+CR3+CR7+CR8, and CR1+CR3+CR6+CR8 in UO2-, (U,Th)O2-, and PuO2-fueled cores, respectively. It can be concluded that, even though the calculation results are not so much different, however, the selection of control rods should be considered in the pebble-bed core design with different scenarios of fuel type.

Gas Flow Simulations in Randomly Distributed Pebbles

2010 ◽  
Vol 133 (5) ◽  
Author(s):  
Xiang Zhao ◽  
Trent Montgomery ◽  
Sijun Zhang
Keyword(s):  
Heat Transfer ◽  
Gas Flow ◽  
Flow Fields ◽  
Navier Stokes ◽  
Complex Flow ◽  
Pebble Bed Reactor ◽  
Flow Simulations ◽  
Geometry Modeling ◽  
Eddy Simulation ◽  
Computational Fluid Dynamics Cfd

In this paper, computational fluid dynamics (CFD) gas flow simulations are carried out for the pebble bed reactor. In CFD calculations, geometry modeling and physical modeling are crucial to CFD results. The effects of the treatments of the interpebble contacts on gas flow fields and heat transfer are examined. A sensitivity analysis for the gap size is conducted with two spherical pebbles, in which the interpebble region is modeled by means of two types of interpebble gap and two kinds of direct contact. Both large eddy simulation and Reynolds-averaged Navier–Stokes models are employed to investigate the turbulent effects. It is found that the flow fields and relevant heat transfer are significantly dependent on the modeling of the interpebble region. The calculations indicate the complex flow structures present within the voids between the fuel pebbles.

A Study of Nonlinear Vibrations in a Frictionally-Damped Turbine Bladed Disc With Comprehensive Modelling of Aerodynamic Effects

2012 ◽  
Author(s):  
E. P. Petrov ◽  
Z.-I. Zachariadis ◽  
A. Beretta ◽  
R. Elliott
Keyword(s):  
Gas Flow ◽  
Aerodynamic Characteristics ◽  
Forced Response ◽  
New Method ◽  
Mode Shapes ◽  
Response Analysis ◽  
Resonance Frequencies ◽  
Computational Fluid Dynamics Cfd ◽  
Flow Effects

A new effective method for comprehensive modelling of gas flow effects on vibration of nonlinear vibration of bladed discs has been developed for a case when effect of the gas flow on the mode shapes is significant. The method separates completely the structural dynamics calculations from the significantly more computationally expensive computational fluid dynamics (CFD) calculations while provides the high accuracy of modelling for aerodynamic effects. A comprehensive analysis of the forced response using the new method has been performed for a realistic turbine bladed disc with root-disc joints, tip and underplatform dampers. The full chain of aerodynamic and structural calculations are performed: (i) determination of boundary conditions for CFD; (ii) CFD analysis; (iii) calculation of the aerodynamic characteristics required by the new method; (iv) nonlinear forced response analysis using the MAIM. The efficiency of the friction damping devices has been studied and compared for several resonance frequencies and engine orders. Advantages of the method for aerodynamic effect modelling have been demonstrated.

scholarly journals Impact of Different Nuclear Data Library on Control Rod Reactivity Worth Calculation of Small Pebble Bed Reactor

2021 ◽  
Vol 2048 (1) ◽  
pp. 012029
Author(s):  
Suwoto ◽  
H Adrial ◽  
T Setiadipura ◽  
Zuhair ◽  
S Bakhri
Keyword(s):  
Nuclear Data ◽  
Pebble Bed ◽  
Integral Control ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Nuclear Data Library ◽  
Data Libraries ◽  
And Control ◽  
Nuclear Data Libraries ◽  
Data Library

Abstract One of the main critical issues on a nuclear reactor is safety and control system. The control rod worth plays an important role in the safety and control of nuclear reactors. The control rods worth calculation is used to specify the safety margin of the reactor. The main objective of this work is to investigate impact of different nuclear data libraries on calculating the control rod reactivity worth on small pebble bed reactor. Calculation of the control rod reactivity worth in small high temperature gas cooled reactor has been conducted using the Monte Carlo N-Particle 6 (MCNP6) code coupled with a different nuclear data library. Famous evaluated nuclear data libraries such as JENDL-40u, ENDF/B-VII.1 and JEFF-3.2 continuous cross section-energy data libraries were used. The overall calculation results of integral control rod worth show that the ENDF/B-VII.1, JENDL-40u and JEFF-3.2 files give values of - 17.814%☐k/k, -18.0204 %☐k/k and -18.0267%☐k/k, respectively. Calculations using ENDF/B-VII.1 give a slightly lower value than the others, while the JENDL-4.0u file gives results that are close to JEFF-3.2 file. The different nuclear data libraries have a relatively small impact on the control rod worth of small pebble bed reactor. Accurate prediction by simulation of control rod worth is very important for the safety operation of all reactor types, especially for new reactor designs.

A Proposed Model to Describe the Relationship Between Online Burnup Assay and Economy and Safety of Pebble Bed Reactor

2016 ◽  
Author(s):  
Yushi Tang ◽  
Liguo Zhang ◽  
Quiju Guo ◽  
Jianzhu Cao ◽  
Jiejuan Tong
Keyword(s):  
Unique Feature ◽  
Reactor Core ◽  
Fuel Cost ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
The Core ◽  
Proposed Model ◽  
Theoretical Hypothesis ◽  
Radioactive Release ◽  
The Relationship

Online burnup measurement is a unique feature for pebble bed gas-cooled reactor and the fuel balls undergo a multi-circulation on the basis of the online burnup assay. It is ascertained that the accuracy of the online burnup assay is related with the economy and safety of pebble bed reactor. In the economical perspective, the burnup assay accuracy allow some part of pebbles that are below the burnup limit in the orifice to be discharged out of the core. In the safety view, the burnup assay allow some part of pebbles in the reactor core to exceed the burnup limit. In this paper, a mathematical model is proposed to establish the relationship. The model is implemented based on some reasonable theoretical hypothesis, and the influence of assay accuracy on the reactor safety and fuel cost issues are discussed based on the simulated results given by different assay accuracy. It is ascertained that improvements on burnup assay accuracy could save the fuel cost and improve the PBR economical efficiency as well as reduce the probability of radioactive release due to over-irradiation and enhance the safe reliability of PBR. Further research on the burnup distribution of pebbles in and out of the core and the burnup assay model are expected to provide some implications on proposing reasonable requirements for accuracy of online burnup assay.

scholarly journals Study on Control Rod Reactivity of Small Pebble Bed Reactor with Wallpaper Fuel Design

2021 ◽  
Vol 1772 (1) ◽  
pp. 012021
Author(s):  
Zuhair ◽  
Suwoto ◽  
S. Permana ◽  
T. Setiadipura
Keyword(s):  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Fuel Design

scholarly journals PRELIMINARY ANALYSIS OF CORE TEMPERATURE DISTRIBUTION OF EXPERIMENTAL POWER REACTOR USING RELAP5

2018 ◽  
Vol 20 (3) ◽  
pp. 159
Author(s):  
Andi Sofrany Ekariansyah ◽  
Surip Widodo ◽  
Hendro Tjahjono ◽  
Susyadi Susyadi ◽  
Puradwi Ismu Wahyono ◽  
...  
Keyword(s):  
High Temperature ◽  
Temperature Distribution ◽  
Core Temperature ◽  
Power Reactor ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
The Core ◽  
Detail Design ◽  
Core Section ◽  
High Temperature Gas

High Temperature Gas Cooled Reactor (HTGR) is a high temperature reactor type having nuclear fuels formed by small particles containing uranium in the core. One of HTGR designs is Pebble Bed Reactor (PBR), which  utilizes helium gas flowing between pebble fuels in the core. The PBR is also the similar reactor being developed by Indonesia National Nuclear Energy Agency (BATAN) under the name of the Reaktor Daya Eksperimental (RDE) or Experimental Power Reactor (EPR) started in 2015. One important step of the EPR program is the completion of the detail design document of EPR, which should be submitted to the regulatory body at the end of 2018. The purpose of this research is to present preliminary results in the core temperature distribution in the EPR using the RELAP5/SCDAP/Mod3.4 to be complemented in the detail design document. Methodology of the calculation is by modelling the core section of the EPR design according to the determined procedures. The EPR core section consisting of the pebble bed, outlet channels, and hot gas plenum have been modelled to be simulated with 10 MWt. It shows that the core temperature distribution under assumed model of 4 core zones is below the limiting pebble temperature of 1,620 °C with the highest pebble temperature of 1,477.0 °C. The results are still preliminary and requires further researches by considering other factors such as more representative radial and axial power distribution, decrease of core mass flow, and heat loss to the reactor pressure vessel.Keywords: Pebble bed, core temperature, EPR, RELAP5 ANALISIS AWAL DISTRIBUSI TEMPERATUR TERAS REAKTOR DAYA EKSPERIMENTAL MENGGUNAKAN RELAP5. High Temperature Gas Cooled Reactor (HTGR) adalah reaktor tipe temperatur tinggi yang memiliki bahan bakar nukir dalam bentuk bola-bola kecil yang mengandung uranium. Salah satu desain HTGR adalah reaktor pebble bed (Pebble bed reactor/PBR) yang memanfaatkan gas helium sebagai pendingin yang mengalir di celah-celah bahan bakar bola di dalam teras. PBR juga merupakan tipe reaktor yang sedang dikembangkan oleh BATAN dengan nama reaktor daya eksperimental (RDE) yang dimulai pada 2015. Salah satu tahapan penting dalam program RDE adalah penyelesaian dokumen desain rinci yang harus dikirimkan ke badan pengawas pada akhir 2018. Tujuan penelitian adalah untuk menyajikan hasil-hasil awal pada distribusi temperatur di teras RDE menggunakan RELAP5/SCDAP/Mod3.4  sehingga dapat melengkapi isi dokumen desain rinci. Metode perhitungan adalah dengan memodelkan bagian teras RDE sesuai hasil penelitian sebelumnya.  Bagian teras RDE yang dimodelkan terdiri dari pebble bed, kanal luaran, dan plenum gas bawah yang disimulasikan pada daya 10 MWt. Hasil simulasi menunjukkan bahwa distribusi temperatur teras dengan asumsi pembagian 4 zona teras mendapatkan temperatur tertinggi sebesar 1477 °C yang masih di bawah batasan temperatur di bola bahan bakar yaitu 1620 °C. Hasil yang diperoleh masih estimasi awal dan membutuhkan penelitian lebih lanjut dengan mempertimbangkan faktor-faktor lainnya seperti distribusi daya aksial dan radian yang lebih representatif, pengurangan aliran teras, dan kehilangan panas teras yang diserap oleh bejana reaktor.Kata kunci: Pebble bed, temperatur teras, RDE, RELAP5

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