Safety Analysis of the Control Rod Drop Event Applying 3-D Neutronics/Thermal-Hydraulics Coupling Code SPARKLE

2009 ◽  
Author(s):  
Yuta Maruyama ◽  
Satoshi Imura ◽  
Junto Ogawa ◽  
Shuhei Miyake
Keyword(s):  
Power Distribution ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Safety Margin ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Analysis Method ◽  
Dynamic Changes ◽  
Control Rod ◽  
Ejection Event

Mitsubishi Heavy Industries (MHI) has developed the SPARKLE code, which is a PWR plant system transient analysis code that includes a three-dimensional (3D) neutronics module coupled with a thermal-hydraulics module. MHI has performed a study of the applicability of the SPARKLE code to the events which are associated with dynamic changes in power distribution, such as the rod ejection event or the steam line break event. In this paper, MHI has applied the SPARKLE code to the control rod drop event (drop of multiple rods), which features such a power distribution change. In addition, the neutron flux detection is dependent on the location of the dropped rods in this event, which can be dynamically calculated in the SPARKLE code. By applying the SPARKLE code to the control rod drop event, it was confirmed that the safety margin for this event is sufficiently larger than the margin calculated using the current safety analysis method, even if the appropriate conservative assumptions are made.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo
Keyword(s):  
Electric Power ◽  
Supercritical Water ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Water Density ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Protection System ◽  
Supercritical Water Cooled Reactor

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

scholarly journals Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
A. Rais ◽  
D. Siefman ◽  
G. Girardin ◽  
M. Hursin ◽  
A. Pautz
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Transient Behavior ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Nuclear Data Library ◽  
And Control

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

scholarly journals THE THERMAL-HYDRAULICS ANALYSIS ON RADIAL AND AXIAL POWER FLUCTUATION FOR AP1000 REACTOR

2015 ◽  
Vol 17 (2) ◽  
pp. 79 ◽  
Author(s):  
Muh. Darwis Isnaini ◽  
Surip Widodo ◽  
Muhammad Subekti
Keyword(s):  
Heat Flux ◽  
Power Distribution ◽  
Safety Margin ◽  
Fissile Material ◽  
Thermal Hydraulics ◽  
Power Fluctuation ◽  
Operating Cycle ◽  
Maximum Heat ◽  
Control Rod ◽  
Ap1000 Reactor

ABSTRACT THE THERMAL-HYDRAULICS ANALYSIS ON RADIAL AND AXIAL POWER FLUCTUATION FOR AP1000 REACTOR. The reduction of fissile material during reactor operation affects reactivity reduction. Therefore, in order to keep the reactor operating at fixed power, it must be compensated by slowly withdrawing the control-rod up. However, it will change the shape of the horizontal/axial power distribution and safety margin as well. The research carries out the calculations of the core thermal-hydraulics to determine the effect of the fluctuations of the power distribution on the thermal-hydraulic AP1000’s parameters and study their impacts on the safety margin. The calculation is done using the COBRA-EN code and the result shows that the maximum heat flux at the Beginning of Cycle (BOC) is 1624.02 kW/m2. This heat flux will then decrease by 22.75% at the Middle of Cycle (MOC) and by 0.29% at the End of Cycle (EOC). The peak fuel centerline temperature at the BOC, MOC and EOC, are 1608.15°C, 1232.15°C, and 1301.75°C, respectively. These temperature differences are caused by the heat flux effects on sub-cooled boiling regions in the cladding surface. Moreover, the value of MDNBRs at the MOC and EOC are 3.23 and 3.00, which are higher than the MDNBR at the BOC of 2.49. It could be concluded that the operating cycle of the AP1000 reactor should be operated in the MOC and the EOC, which will be more safely than be operated in the BOC. Keywords: Core thermal-hydraulics, AP1000, fluctuation of power distribution, COBRA-EN.  ABSTRAK ANALISIS TERMOHIDRAULIKA PADA FLUKTUASI DAYA AXIAL DAN RADIAL UNTUK REAKTOR AP1000. Berkurangnya material fisil selama operasi reaktor, mengakibatkan reaktivitas berkurang. Oleh karena itu, agar reaktor tetap beroperasi pada daya yang tetap, maka harus dikompensasi dengan menarik batang kendali ke atas sedikit demi sedikit. Akan tetapi, hal ini akan berakibat pada berubahnya bentuk distribusi daya ke arah horisontal/aksial dan berdampak ke perubahan marjin keselamatan. Penelitian ini melakukan perhitungan termohidrolika teras untuk mengetahui pengaruh fluktuasi distribusi daya pada parameter termohidrolika AP1000 dan mempelajari dampaknya terhadap marjin keselamatan. Hasil perhitungan dilakukan dengan menggunakan kode COBRA-EN dan hasilnya menunjukkan bahwa fluks kalor maksimum pada awal siklus (BOC) sebesar 1624,02 kW/m2 berkurang 22,75% di tengah siklus (MOC) dan berkurang lagi 0,29% di akhir siklus (EOC). Temperatur puncak tengah bahan-bakar di awal, tengah dan akhir siklus adalah sebesar 1608,15°C; 1232,15°C; dan 1301,75°C akibat dari fluks kalor pada daerah kelongsong yang mengalami pendidihan tak jenuh. Sedangkan nilai MDNBR pada tengah dan akhir siklus adalah 3,23 dan 3,00; meningkat dibanding MDNBR pada awal siklus 2,49. Dari hasil tersebut dapat disimpulkan bahwa pada kondisi tengah dan akhir siklus operasi reaktor AP1000 memiliki marjin keselamatan yang lebih baik dibanding kondisi awal siklus.  Kata kunci : Termohidrolika teras, AP1000, fluktuasi distribusi daya, COBRA-EN

Neutronics and Thermal Hydraulics Coupled 3-D Calculation of Pressure-Tube Deformation

2020 ◽  
Author(s):  
Eunhyun Ryu ◽  
Hangyu Joo ◽  
Seungyul Yoo ◽  
Jongyub Jung
Keyword(s):  
Dimensional Space ◽  
Three Dimensional ◽  
Safety Margin ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Main Concern ◽  
Thermal Hydraulics ◽  
Pressure Tubes ◽  
Fuel Pin ◽  
Heavy Water Reactor

Abstract Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs). When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on. In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.

An Extended Application Research of RELAP5 to Typical Non-LOCA of Passive NPP

2017 ◽  
Author(s):  
Wang Haitao ◽  
Liu Zhan
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Transient Analysis ◽  
Safety Margin ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Application Research ◽  
Extended Application ◽  
Accident Consequence ◽  
Variation Tendency

As to passive nuclear power plant (NPP) transient/accident analysis, conservative codes rather than best estimated code such as RELAP5 are mainly used to do the non-LOCA analysis work. While considering the severer accident consequence, RELAP5 is selected in some cases as a tool for LOCA analysis so that safety margin can be gained and meanwhile, the safety NPP condition is guaranteed. In order to extend the RELAP5 application rang to passive NPP, a simulation capability research of RELAP5 to typical non-LOCA consisting of condition II, III and IV transient analysis of CAP1000 was conducted, and the results were compared with that analyzed by safety analysis code. Meanwhile, as to simulation skill, an innovative RELAP5 model was developed. In despite of different conservative degree of the key parameters, the response of the key equipment together with the variation tendency of the key parameters were consistent with that predicted by safety analysis code. Besides, the results met the acceptance criteria. It was showed that RELAP5 has the capability to simulate the CAP1000 non-LOCA.

Analysis of Bridge Design Based on Three-Dimensional Space

2012 ◽  
Vol 529 ◽  
pp. 210-214
Author(s):  
Weiwei Suo ◽  
Fu Ting Wang
Keyword(s):  
Seismic Wave ◽  
Seismic Behavior ◽  
Transient Analysis ◽  
Dimensional Space ◽  
Three Dimensional ◽  
Analysis Method ◽  
Fem Model ◽  
Arch Bridges ◽  
Main Bridge ◽  
Three Dimensional Space

Considering the effect of non-linearity of geometry and material to the arch bridge, this paper creates the FEM model of the whole bridge, uses the transient analysis method to analyze the seismic behavior of the main bridge of GuCheng Great Bridge independently under the action of Tianjin seismic wave. To give some suggestion on the design of rigid-tied rigid-arch rib CFST arch bridges by studying the seismic behavior of each component on the different positions with the seismic wave of different orientations.

Study on Calculation Difference of the SCWR Core Steady-State Analysis Code

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Lianjie Wang ◽  
Lei Yao ◽  
Ping Yang ◽  
Di Lu ◽  
Wenbo Zhao
Keyword(s):  
Steady State ◽  
Cross Section ◽  
Power Distribution ◽  
Group Structure ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Steady State Analysis ◽  
Energy Group ◽  
The Cross ◽  
State Analysis

Abstract The three-dimensional code system supercritical water-cooled reactor (SCWR) coupled neutronics/thermal-hydraulics analysis (SNTA) code is developed for SCWR core steady-state analysis by coupling neutronics/thermal-hydraulics (N/T). This paper studies the calculation difference between the SNTA code and the standard reactor analysis code (SRAC). By using the impacts exclusive method, it is confirmed that the calculation difference between these two codes is caused by different feedback of the cross section. The cross section data and the energy group structure of the SRAC code differ from the SNTA code, and the density coefficient of reactivity calculated by the SRAC code is higher, which means the feedback of the density and power distribution is bigger and the axial power distribution varies rapidly. The SNTA code with finer energy group structure is suitable for the performance analysis of SCWR core which has strong N/T coupling characteristics.

Coupled Three Dimensional Neutronics/Thermal-Hydraulics Code STTA for SCWR Core Transient Analysis

2014 ◽  
Author(s):  
Lianjie Wang ◽  
Wenbo Zhao ◽  
Ping Yang ◽  
Bingde Chen ◽  
Dong Yao
Keyword(s):  
Transient Analysis ◽  
Function Method ◽  
Three Dimensional ◽  
Supercritical Pressure ◽  
Channel Code ◽  
Neutron Diffusion ◽  
Thermal Hydraulics ◽  
The Core ◽  
Coupling Approach ◽  
Neutron Diffusion Equation

A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN_K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN_K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation.

Validation of TRACG05 for Application to BWR AOO

2012 ◽  
Author(s):  
Asuka Matsui ◽  
Masashi Tamitani ◽  
Yoshiro Kudo ◽  
Sho Takano ◽  
Tatsuya Iwamoto ◽  
...  
Keyword(s):  
Experimental Data ◽  
Three Dimensional ◽  
Reactor Core ◽  
Expansion Method ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Neutron Kinetics ◽  
Safety Issues ◽  
Best Estimate ◽  
Assessment Matrix

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.

The Numerical Simulation of Three-Dimensional Dynamic-Mesh Flow Field of a Hydraulic Buffer

2012 ◽  
Vol 588-589 ◽  
pp. 1264-1268 ◽  
Author(s):  
Song Jiang Peng ◽  
Cheng Zhou ◽  
Cun Gui Yu
Keyword(s):  
Flow Field ◽  
Three Dimensional ◽  
Internal Flow ◽  
Special Equipment ◽  
Dynamic Changes ◽  
Control Rod ◽  
Working Chamber ◽  
Computational Fluid Dynamics Cfd ◽  
Dynamic Meshing ◽  
Physical Quantities

In order to study the dynamic changes of the internal flow field physical quantities in the hydraulic buffer of a special equipment, based on the Computational Fluid Dynamics (CFD), a three-dimensional incompressible and viscous model under unsteady condition is created. The model keeps the control rod of varying diameter and non-working chamber cover, simulates the turbulent in the flow field of the hydraulic buffer of the special equipment with dynamic meshing technology. From the results, the distributions of velocity in flow field and pressure in chamber are got. It shows that there are negative pressure areas in the non-working chamber and that will lead to cavitation. The results give us a great reference to improve the structure of hydraulic buffer.

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