Operational Safety Analysis of HANARO Research Reactor using STAMP/STPA

Prior to core reloading, planned power upgrading, or as a part of required analyses of past events, accurate safety evaluations should be carried out. Generally speaking, the content of a safety report has to be modified whenever a new type or design of fuel is to be used in a reactor core. As the existing plants have well established licensing procedures, including well founded analysis methods, the application of new analysis methods has to be thoroughly evaluated, with specific emphasis on their capability of producing results beneficial to reactor operation. The detailed study presented here was carried out so as to insure that the allowed operational safety limits of the NUR research reactor are not exceeded under any circumstances.

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Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

Annals of Nuclear Energy ◽

10.1016/j.anucene.2015.02.027 ◽

2015 ◽

Vol 80 ◽

pp. 409-415 ◽

Cited By ~ 6

Author(s):

F. Mohamed ◽

A. Hassan ◽

R. Yahaya ◽

I. Rahman ◽

M. Maskin ◽

...

Keyword(s):

Research Reactor ◽

Safety Analysis ◽

Reliability Study ◽

Probabilistic Safety

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Upgraded safety analysis document including operations policies, operational safety limits and policy changes. Revision 2

10.2172/226065 ◽

1996 ◽

Author(s):

K. Batchelor

Keyword(s):

Safety Analysis ◽

Policy Changes ◽

Operational Safety

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Steady-State Safety Analysis of Ghana Research Reactor-1 With Low-Enriched-Uranium Core

Journal of Thermal Science and Engineering Applications ◽

10.1115/1.4046598 ◽

2020 ◽

Vol 12 (5) ◽

Author(s):

Edward Shitsi ◽

Prince Amoah ◽

Emmanuel Ampomah-Amoako ◽

Henry Cecil Odoi

Keyword(s):

Steady State ◽

Research Reactor ◽

Flow Instability ◽

Argonne National Laboratory ◽

Safety Analysis ◽

Nucleate Boiling ◽

Pressurized Water ◽

Safety Margins ◽

Water Reactors ◽

Enriched Uranium

Abstract Research reactors all over the world are expected to operate within certain safety margins just like pressurized water reactors and boiling water reactors. These safety margins mainly include onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and flow instability ratio (FIR) in addition to the maximum clad or fuel temperature and saturation temperature or boing point of the coolant inside the core of the reactor. This study carried out steady-state safety analysis of the Ghana Research Reactor-1 (GHARR-1) with low enriched uranium (LEU) core. Monte Carlo N-particle (MCNP) code was used to obtain radial and axial power peaking factors used as inputs in the preparation of the input file of plate temperature code of Argonne National Laboratory (PLTEMP/ANL code), which was then used to obtain the mentioned safety parameters of GHARR-1 with LEU core in this study. The data obtained on the ONBR were used to obtain the initiation of nucleate boiling boundary data with respect to the active length of the reactor core for various reactor powers. The obtained results for LEU core were also compared with that of the high enriched uranium (HEU) core. The results obtained show that the 34 kW GHARR-1 with LEU core is safe to operate just as the previous 30 kW HEU core was safe to operate.

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Analysis of Rod Ejection Accident in a Research Reactor by the Coupled Technique

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

10.1115/icone17-75211 ◽

2009 ◽

Author(s):

Tewfik Hamidouche ◽

El Khider Si-Ahmed ◽

Anis Bousbia-Salah ◽

Jack Legrand

Keyword(s):

Nuclear Power ◽

Research Reactor ◽

Safety Margin ◽

Safety Analysis ◽

Monte Carlo Code ◽

Standard Material ◽

Computer Tools ◽

Operation Conditions ◽

Coupled Technique ◽

Test Reactor

This paper investigates the possibility to extend standard computer tools and methods, commonly used in the safety technology of nuclear power reactors, to research reactor safety analysis. A 3-D Neutron Kinetics Thermal-Hydraulic code (3D-NKTH), based on coupling PARCS and RELAP5/3.3 codes, was developed for a standard Material Test Reactor (MTR). The assessment of the model has been performed by comparison of steady state calculations against conventional diffusion codes and Monte Carlo code results. The model is applied for the analysis of a rod ejection accident. The comparison of the 3D-NKTH code, with conventional conservative research reactor tools showed that 3D-NKTH provided a more realistic course of the accident and did not require to define hot channel parameters. This approach could also open new frontiers in the safety analysis of research reactor such as setting realistic safety margin and adequate limits and operation conditions for optimal utilization of research reactors.

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