Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

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.

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PRELIMINARY POWER TRANSIENT ANALYSIS OF THE SUPER FR WITH AXIALLY HETEROGENEOUS CORE

EPJ Web of Conferences ◽

10.1051/epjconf/202124707002 ◽

2021 ◽

Vol 247 ◽

pp. 07002

Author(s):

Tsutomu Okui ◽

Akifumi Yamaji

Keyword(s):

Temperature Gradient ◽

Transient Analysis ◽

Large Temperature ◽

Control Rod ◽

Mox Fuel ◽

Fuel Layer ◽

Axial Temperature ◽

Short Period ◽

Full Power ◽

Supercritical Water Cooled Reactor

The Super FR is one of the SuperCritical Water cooled Reactor (SCWR) concepts with once-through direct cycle plant system. Recently, new design concept of axially heterogeneous core has been proposed, which consists of multiple layers of MOX and blanket fuels. To clarify the safety performance during power transient, safety analyses have been conducted for uncontrolled control rod (CR) withdrawal and CR ejection at full power. RELAP/SCDAPSIM code was used for the safety analysis. The results show that the peak cladding surface temperature (PCST) is high in the upper MOX fuel layer. It is also shown that axial temperature gradient of cladding greatly increases in a short period. Suppressing such large temperature gradient may be a design issue for the axially heterogeneous core from the viewpoint of ensuring fuel integrity.

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

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications ◽

10.1115/icone17-75729 ◽

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.

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Transient Analysis of UO2 and MOX-Fueled Cores Based on Microscopic Reactor Physics

Volume 1: Plant Operations, Maintenance, Installations and Life Cycle; Component Reliability and Materials Issues; Advanced Applications of Nuclear Technology; Codes, Standards, Licensing and Regulatory Issues ◽

10.1115/icone16-48344 ◽

2008 ◽

Author(s):

Toshikazu Takeda ◽

Hiroaki Tagawa ◽

Tadafumi Sano

Keyword(s):

Cross Sections ◽

Present Method ◽

Peak Power ◽

Transient Analysis ◽

Effective Cross Section ◽

Light Water Reactor ◽

Reactor Physics ◽

Control Rod ◽

Fuel Rod ◽

Averaged Model

A transient analysis has been performed for UO2 and MOX-fueled light water reactor cores based on Microscopic Reactor Physics, which treats the detailed distribution of temperature and effective cross section within a rod. Conventionally the volume-averaged temperature and the Rowlands’ effective temperature are used to calculate fuel rod-averaged cross sections, and applied to the transient analysis. The present method is considered as a reference and the result is compared with the conventional method for a mini fuel core containing eight fuel rods and a control rod. From numerical results, it is found that the Rowlands’ model underestimates the peak power and the volume averaged model produces rather good peak power results. After 1.0 sec, the Rowlands’ model yields the similar power as the reference, while the volume averaged model yields less power than the reference one.

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CYNOD: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis

Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 1 ◽

10.1115/htr2008-58255 ◽

2008 ◽

Author(s):

Hikaru Hiruta ◽

Abderrafi M. Ougouag ◽

Hans D. Gougar ◽

Javier Ortensi ◽

David W. Nigg ◽

...

Keyword(s):

Transient Analysis ◽

Fuel Kernel ◽

Pebble Bed ◽

Feedback Effects ◽

Control Rod ◽

One Dimensional ◽

Neutron Kinetics ◽

Pebble Bed Reactors ◽

Transient Problems ◽

Kernel Model

In this paper, a new neutron kinetics solver for cylindrical R-Z geometry, CYNOD, is presented for the simulation of coupled transient problems for pebble bed reactors. The code utilizes the Direct Coarse Mesh Finite Difference method, in which a set of one-dimensional equations in each transverse direction is solved by means of the analytic Green’s function method. A method that deals with control rod cusping problems is also presented. A heterogeneous fuel kernel model is implemented in order to accurately take into account Doppler feedback effects. Numerical results that demonstrate the accuracy of the code are also presented.

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