Emergency Decay Heat Removal by Reactor Vessel Auxiliary Cooling System from an Accelerator-Driven System

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.

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Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

Science and Technology of Nuclear Installations ◽

10.1155/2008/681847 ◽

2008 ◽

Vol 2008 ◽

pp. 1-8 ◽

Cited By ~ 3

Author(s):

Giacomino Bandini ◽

Paride Meloni ◽

Massimiliano Polidori ◽

Maddalena Casamirra ◽

Francesco Castiglia ◽

...

Keyword(s):

Natural Circulation ◽

Thermal Power ◽

Heat Removal ◽

Normal Operation ◽

Steam Generators ◽

Flow Paths ◽

Decay Heat ◽

Driven System ◽

Accelerator Driven System ◽

Relap5 Code

The development of a conceptual design of an industrial-scale transmutation facility (EFIT) of several 100 MW thermal power based on accelerator-driven system (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

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

10th International Conference on Nuclear Engineering, Volume 2 ◽

10.1115/icone10-22354 ◽

2002 ◽

Cited By ~ 1

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.

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Development of reactor vessel thermal hydraulic evaluation method under natural circulation decay heat removal conditions in sodium-cooled fast reactor

The Proceedings of Mechanical Engineering Congress Japan ◽

10.1299/jsmemecj.2020.s08111 ◽

2020 ◽

Vol 2020 (0) ◽

pp. S08111

Author(s):

Erina HAMASE ◽

Yasutomo IMAI ◽

Norihiro DODA ◽

Masaaki TANAKA

Keyword(s):

Fast Reactor ◽

Evaluation Method ◽

Natural Circulation ◽

Heat Removal ◽

Reactor Vessel ◽

Decay Heat

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

Volume 6: Beyond Design Basis Events; Student Paper Competition ◽

10.1115/icone21-16571 ◽

2013 ◽

Cited By ~ 3

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.

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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

Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation ◽

10.1115/icone2020-16796 ◽

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.

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Partial Redesign of an Accelerator Driven System Target for Optimizing the Heat Removal and Minimizing the Pressure Drops

Energies ◽

10.3390/en11082090 ◽

2018 ◽

Vol 11 (8) ◽

pp. 2090 ◽

Cited By ~ 2

Author(s):

Guglielmo Lomonaco ◽

Giacomo Alessandroni ◽

Walter Borreani

Keyword(s):

Cooling System ◽

Fluid Dynamic ◽

Heat Removal ◽

Dynamic Features ◽

Pressure Drops ◽

Driven System ◽

Reference Design ◽

Accelerator Driven Systems ◽

Definition Of ◽

Target Design

Accelerator Driven Systems (ADS) seem to be a good solution for safe nuclear waste transmutation. One of the most important challenges for this kind of machine is the target design, particularly for what concerning the target cooling system. In order to optimize this component a CFD-based approach has been chosen. After the definition of a reference design (Be target cooled by He), some parameters have been varied in order to optimize the thermal-fluid-dynamic features. The final optimized target design has an increased security margin for what regarding Be melting and reduces the maximum coolant velocity (and consequently even more the pressure drops).

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