A New Concept of the 4S Reactor and Thermal Hydraulic Characteristics

2004 ◽  
Author(s):  
Yoshihisa Nishi ◽  
Nobuyuki Ueda ◽  
Izumi Kinoshita ◽  
Tomonari Koga ◽  
Satoshi Nishimura ◽  
...  
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Electric Power Industry ◽  
Reactor Vessel ◽  
Air Cooling ◽  
Central Research ◽  
Neutron Absorption ◽  
Similar System ◽  
Decay Heat

CRIEPI (Central Research Institute of Electric Power Industry) has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application to dispersed energy supply and multipurpose use, with Toshiba Corporation [1,2,3,4]. Electrical output of the 4S reactor is from 10MW to 50MW, and burn-up reactivity loss is regulated by neutron reflectors. The reflector that surrounds the core is gradually lifted up to control the reactivity according to core burn-up. 30year core lifetime without refueling can be achieved with the 10MW 4S (4S-10M) reactor. All temperature feedback reactivity coefficients, including coolant void reactivity, of the 4S-10M are negative during the 30year lifetime. A neutron absorption rod is set at the center of the reactor core with the ultimate shutdown rod. The neutron absorption rod used during the former 14 years is moved to the upper part of the reactor core, and the operation is continued through the latter 16 years. The pressure loss of the reactor core is lower than 2kg/cm2 to enable effective utilization of the natural circulation force, and the average burn-up rate is 76GWD/t. To suppress the influence of the scale disadvantage, loop-type reactor design is one of the candidates for the 4S-10M. The size of the reactor vessel can be miniaturized by adopting the loop type design (4S-10ML). In the 4S-10ML design, integrated equipment which includes primary and secondary electromagnetic pumps (EMPs), an intermediate heat exchanger (IHX) and a steam generator (SG) is adopted and collocated by the reactor vessel. The decay heat removal systems of 4S-10ML consist of the reactor vessel air cooling system (RVACS) and SGACS (a similar system to the RVACS, with air cooling of the outside of the integrated equipment vessel). They are completely passive systems. To decrease the construction cost of the reactor building, a step mat structure and the horizontal aseismic structure are adopted. 4S-10ML has unique features in the cooling systems such as integrated equipment and two separate passive decay heat removal systems which operate at the same time. To evaluate the design feasibility, the transition analyses were executed by the CERES code developed by CRIEPI [5]. In this paper, the design concept of 4S-10ML, and the results of the plant transition analyses are described.

Performance Evaluation of DRACS System for FHTR and Time Assessment of Operation Procedure

2013 ◽  
Author(s):  
Junya Nakata ◽  
Mikihiro Wakui ◽  
Michitsugu Mori ◽  
Hiroto Sakashita ◽  
Charles Forsberg
Keyword(s):  
Performance Evaluation ◽  
High Temperature ◽  
Cooling System ◽  
Heat Removal ◽  
Severe Accident ◽  
Air Cooling ◽  
Fluoride Salt ◽  
Nuclear Power Reactor ◽  
Decay Heat ◽  
Operation Procedure

The Fluoride-salt-cooled High-temperature Reactor (FHR) is a new concept of nuclear power reactor being investigated mainly in U.S. and China. The coolant is a liquid salt with a melting point of about 460°C and a boiling point of over 1400°C. As the baseline decay heat removal system, a passive Direct Reactor Air Cooling System (DRACS) is utilized. Though DRACS system has been developed in Sodium Fast reactors (SFR), there are some differences between both. For example, the system in FHR must decrease heat removal when temperatures are low to avoid freezing of the salt and blocking the flow of liquid. Therefore, considering its characteristics, a numerical investigation of DRACS system is needed to evaluate whether FHR has proper ability to remove decay heat and to be robust for a long-time cooling operation after even a severe accident. Furthermore, in addition to its performance evaluation, it is required to make up the operation plan of FHR considering features of this system. It is highly important, with the view of avoiding severe accident, to determine by when the system should be started up. In both countries mentioned above, Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently in progress to build. Reviewing its design and system is a crucial step needed to develop the FHR technology. In this research, a performance of DRACS system under some thermal-hydraulic basic events was evaluated by numerical simulation. This paper also suggested the adequate operation procedure suitable for FHTR to avoid a severe accident.

Study on the Turbulent Mixed Convection Phenomena Inside the Air-Cooled RCCS Riser

2018 ◽  
Author(s):  
Dong-Ho Shin ◽  
Sin-Yeob Kim ◽  
Chan Soo Kim ◽  
Goon-Cherl Park ◽  
Hyoung Kyu Cho
Keyword(s):  
Heat Transfer ◽  
Mixed Convection ◽  
Cooling System ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Transfer Coefficients ◽  
Concrete Wall ◽  
Experimental Conditions ◽  
Decay Heat ◽  
Reactor Types

High Temperature Gas cooled Reactor (HTGR) that is one of GEN-IV reactor types enhances its safety adopting a passive cooling system, Reactor Cavity Cooling System (RCCS). When the active cooling system for the reactor core in a HTGR doesn’t work, the decay heat from the reactor core is transferred to the reactor vessel and the concrete wall of the reactor cavity, which are cooled by the RCCS. The RCCS consists of the vertical rectangular ducts, called risers, surrounding reactor vessel at a certain distance and chimneys that are connected to risers. Risers receiving the decay heat from the reactor vessel, the air inside the riser is heated and flows up to the chimney which accelerates exhalation of the air to the external atmosphere. The RCCS performance depends on the heat transfer rate inside the riser ducts, but the turbulent mixed convection that may occur in the riser ducts can complicate the verification. In this study, an experimental facility had been constructed to investigate heat transfer phenomena inside a riser duct and experiments with various heat flux and flow rate conditions were carried out. The experimental results showed that the turbulent mixed convection occurred for certain experimental conditions in the riser duct, which leads to heat transfer deterioration. Therefore, a correlation was suggested to predict how much the heat transfer could decrease compared to the forced convection. Large eddy simulation was carried out for CFD analysis. The heat transfer coefficients from LES showed consistent results with the correlation. The unique velocity profile and secondary flow was observed when the heat transfer was deteriorated. This study will contribute to the evaluation of the RCCS performance.

Passive Safety Features in Sodium Cooled Super-Safe, Small and Simple Reactor

2002 ◽  
Author(s):  
N. Ueda ◽  
I. Kinoshita ◽  
Y. Nishi ◽  
A. Minato ◽  
H. Matsumiya ◽  
...  
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Passive Safety ◽  
Primary Coolant ◽  
Decay Heat ◽  
The Core ◽  
Design Requirements ◽  
Safety Features ◽  
Dynamics Analyses

This paper describes the passive safety features utilized in the updated sodium cooled Super-Safe, Small and Simple fast reactor, which is the improved 4S reactor. This reactor can operate up to ten years without refueling and neutron reflector regulates the reactivity. One of the design requirements is to secure the core against all anticipated transients without reactor scram. Therefore, the reactor concept is to design to enhance the passive safety features. All temperature reactivity feedback coefficients including whole core sodium void worth are negative. Also, introducing of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. Safety analyses are carried out to simulate various transient sequences, which are loss of flow events, transient overpower events and loss of heat sink events, in order to evaluate the passive safety capabilities. A calculation tool for plant dynamics analyses for fast reactors has been modified to model the 4S including the unique plant system, which are reflector control system, circulation pumps and RVACS. The analytical results predict that the designed passive features improve the safety in which temperature variation in transients are satisfied with the safety criteria for the fuel element and the structure of the primary coolant boundary.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Incorporation of the New Irradiation Embrittlement Correlation Method Into the Japanese Code of Surveillance Tests for Reactor Vessel Materials

2008 ◽  
Author(s):  
Seiji Asada ◽  
Takashi Hirano ◽  
Naoki Soneda ◽  
Norimichi Yamashita ◽  
Akira Yonehara ◽  
...  
Keyword(s):  
Nuclear Reactors ◽  
Correlation Method ◽  
Electric Power Industry ◽  
Central Research Institute ◽  
Reactor Vessel ◽  
Central Research ◽  
Irradiation Embrittlement ◽  
New Correlation ◽  
Sensitivity Studies ◽  
Reactor Pressure

A new irradiation embrittlement correlation method for reactor pressure vessel (RPV) materials was developed by the Central Research Institute of Electric Power Industry (CRIEPI). In order to incorporate this new correlation method into the Japan Electric Association Code, JEAC 4201, “Method of Surveillance Tests for Structural Materials of Nuclear Reactors”, sensitivity studies for PWR and BWR reactor vessel materials were performed and the features of the new correlation method was investigated. Based on those studies, the prescription of the new correlation method in JEAC 4201 was formulated. The results of the studies and the overview of the revised JEAC 4201 are presented in this paper.

Test and Evaluation of a Filtering System for Retention of Fibres in a Coolant Flow After Loss-of-Coolant Accident

2012 ◽  
Author(s):  
T. Gocht ◽  
W. Kästner ◽  
A. Kratzsch ◽  
M. Strasser
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Time Interval ◽  
Insulation Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In case of an accident the safe heat removal from the reactor core with the installed emergency core cooling system (ECCS) is one of the main features in reactor safety. During a loss-of-coolant accident (LOCA) the release of insulation material fragments in the reactor containment can lead to malfunctions of ECCS. Therefore, the retention of particles by strainers or filtering systems in the ECCS is one of the major tasks. The aim of the presented experimental investigations was the evaluation of a filtering system for the retention of fiber-shaped particles in a fluid flow. The filtering system consists of a filter case with a special lamellar filter unit. The tests were carried out at a test facility with filtering units of different mesh sizes. Insulation material (mineral rock wool) was fragmented to fiber-shaped particles. To simulate the distribution of particle concentration at real plants with large volumes the material was divided into single portions and introduced into the loop with a defined time interval. Material was transported to the filter by the fluid and agglomerated there. The assessment of functionality of the filtering system was made by differential pressure between inlet and outlet of the filtering system and by mass of penetrated particles. It can be concluded that for the tested filtering system no penetration of insulation particles occurred.

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