The coupled neutronics and thermal hydraulics code system PANBOX for PWR safety analysis / Das gekoppelte neutronisch-thermohydraulische Programmsystem PANBOX zur Sicherheitsanalyse von Druckwasserreaktoren.

Kerntechnik ◽  
1992 ◽  
Vol 57 (1) ◽  
pp. 49-54
Author(s):  
R. Böer ◽  
R. Böhm ◽  
H. Finnemann ◽  
R. Müller
Keyword(s):  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Code System

Transient Validation of Areva’s New Arcadia™ Code System

2014 ◽  
Author(s):  
Sebastian Kuch ◽  
Mario Leberig ◽  
Richard Brock ◽  
Florian Reiterer ◽  
Michael Riedmann ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Leading Edge ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Code System ◽  
The Core ◽  
Wide Range ◽  
Fully Coupled

AREVA has developed a new leading edge code suite to meet the challenges arising from increasing expectations in nuclear power plant availability and fuel performance while satisfying stricter safety requirements. ARCADIA™ [1] is an advanced 3D coupled neutronics/thermal-hydraulics/thermal-mechanics code system for Light Water Reactor (LWR) fuel assembly and core design calculations as well as safety analysis, using a new software architecture allowing for nodal and pin-by-pin calculation capability. ARCADIA™ was licensed by the US Nuclear Regulatory Commission (NRC) for applications for PWR UO2 cores in 2013. It is on the way to be licensed in other countries for AREVA customers. ARCADIA™ contains the steady-state and transient core-simulator ARTEMIS™ [2] for core design and coupled transient safety analysis. ARTEMIS™ can be used in a coupled mode with S-RELAP5 and CATHARE 2 to allow fully coupled transient analysis, combining the sophisticated 3D core model of ARTEMIS™ with the proven system thermal-hydraulics of S-RELAP5 and CATHARE 2 including a detailed simulation of the Instrumentation and Control (I&C). This allows simulating complex transients affecting the core as well as the primary and secondary side including I&C signals and responses. For the validation of ARTEMIS™ a comprehensive set of validation cases was selected, including international benchmarks and measurements covering various classes of transients. These cases include a ‘Load Rejection to station service’ event at a German 1300 MWe plant, where a wide range of system and core parameters was measured that allow the validation of the fully coupled code system. Another validation case is a nodal recalculation of the core behavior during the pump shaft break transient that occurred in the Gösgen nuclear power plant in 1985 [3]. The paper will provide representative example results for the abovementioned validation cases.

scholarly journals MELT-III: a neutronics, thermal-hydraulics computer program for fast reactor safety analysis. Volume II

1974 ◽  
Author(s):  
A.E. Waltar ◽  
W.L. Partain ◽  
D.C. Kolesar ◽  
L.D. O'Dell ◽  
A. Jr. Padilla ◽  
...  
Keyword(s):  
Computer Program ◽  
Fast Reactor ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Reactor Safety

Development and Application of the Neutronics/Thermal-Hydraulics Coupling Code for Safety Analysis of EBR-II Loss of Heat Sink Tests Without Scram

2016 ◽  
Author(s):  
Daogang Lu ◽  
Chao Guo ◽  
Danting Sui
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Fuel Cycle ◽  
Transient Behavior ◽  
Safety Analysis ◽  
Cold Pool ◽  
Thermal Hydraulics ◽  
Closed Fuel Cycle ◽  
Zone Method ◽  
Gen Iv

In the GEN IV technology evaluations, the LMFBR (Liquid Metal Fast Breeder Reactor) system which includes SFR (Sodium-cooled Fast Reactor) and LFR (Lead-cooled Fast Reactor) was top-ranked in sustainability due to its closed fuel cycle and it is top-ranked in proliferation resistance and physical protection because it employs a long-life core. It is necessary to conduct the coupled neutronics and thermal-hydraulics simulation when the feedback effects are significant in the safety analysis of Anticipated Transients Without Scram (ATWS) in LMFBR. Thus, a neutronics-thermalhydraulics coupling code for safety analysis of LMFBR was developed and used to analyze whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under Loss of Heat Sink Without Scram (LOHSWS) tests in this paper. The two mixing zone method for cold pool coupled with SAC-CFR was used and the predicted results agree well with measurements which are taken from EBR-II LOHSWS test data.

The Principle and Preliminary Validation of a New Subchannel Code

2013 ◽  
Author(s):  
Ning Bai ◽  
Yuanbing Zhu ◽  
Zhihao Ren ◽  
Haibo He ◽  
Haoliang Lu ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Constitutive Relations ◽  
Reactor Design ◽  
Safety Analysis ◽  
Industrial Applications ◽  
Transfer Model ◽  
Design Codes ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Pressurized Water

Following China’s road map of nuclear technology development, the development of self-reliant nuclear design codes becomes one of the most significant steps in the plan. Among the nuclear design codes, thermal-hydraulic analysis code is indispensable because it is the foundation of reactor safety analysis and reactor design. Recently, China Guangdong Nuclear Power Group has launched a series of R&D projects of reactor design code development. The sub-channel analysis code-LINDEN becomes one of the key subprojects. Since the sub-channel code is developed for thermal-hydraulic design and safety analysis of pressurized water reactors (PWRs), the basic requirements for the LINDEN code are reliability and stability. Therefore, the mathematical model and numerical method developed in the code are based on the matured approaches that have been used in various industrial applications. These models and methods includes: four-equation drift framework model of two-phase flow; the classical heat transfer model and fuel rod model (Poisson equation) as well as the constitutive relations; explicit formulation and stepping algorithms for equation solutions. The solver module of the code is programmed using object-oriented C/C++ language with the structural design.. With all these features, the code was developed to be stable, scalable and compatible. The code’s applicability has been further improved after model improvement and design optimization according to characteristics of China’s proprietary type of reactor. COBRA-IV and LINDEN have been used to conduct the thermal-hydraulics analysis for the Daya bay unit 1 and 2 nuclear plants at the steady-state conditions. The results demonstrate that the two codes agree well with each other. The preliminary tests show that the LINDEN code should be suitable for thermal-hydraulics analysis of large PWRs.

Comprehensive Safety Analysis Code System for Nuclear Fusion Reactors IV: Preliminary Estimation of Dust Production Due to Plasma Disruptions

1996 ◽  
Vol 30 (1) ◽  
pp. 95-103 ◽  
Author(s):  
Takuro Honda ◽  
Takashi Okazaki ◽  
Yasushi Seki ◽  
Isao Aoki ◽  
Tomoaki Kunugi
Keyword(s):  
Nuclear Fusion ◽  
Safety Analysis ◽  
Fusion Reactors ◽  
Code System ◽  
Dust Production ◽  
Preliminary Estimation

scholarly journals Editorial: Safety Analysis of Nuclear Reactor Thermal-Hydraulics

2021 ◽  
Vol 9 ◽  
Author(s):  
Liangming Pan ◽  
Jun Wang ◽  
Yanping Huang ◽  
Ki-Yong Choi
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Thermal Hydraulics

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.

Sign in / Sign up

Export Citation Format

Share Document