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.

AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

2006 ◽  
Author(s):  
Richard F. Wright ◽  
James R. Schwall ◽  
Creed Taylor ◽  
Naeem U. Karim ◽  
Jivan G. Thakkar ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cooling System ◽  
Heat Removal ◽  
Passive Safety ◽  
Tube Heat Exchanger ◽  
Removal Capacity ◽  
Safety Margins ◽  
Final Design ◽  
Safety Features ◽  
Residual Heat

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.

Conceptual Design and Feasibility Study for FAil-Safe Simple Economical SMR (FASES) System

2014 ◽  
Author(s):  
Ho Sik Kim ◽  
Hee Cheon No
Keyword(s):  
Conceptual Design ◽  
Feasibility Study ◽  
Cooling System ◽  
Heat Removal ◽  
Operating Conditions ◽  
Design Concepts ◽  
Decay Heat ◽  
Design Requirements ◽  
Core Cooling ◽  
Fail Safe

Because of high marketability of SMR, although many countries are trying to develop SMR, the SMR market is not formed yet. For early dominance of SMR market, the SMR system should be fail-safe, simple and economical. In order to develop FAil-safe Simple Economical SMR (FASES) system we applied the design characteristics of HTGRs into water-cooled reactors. In this study, we performed conceptual design and feasibility study for the FASES system. The feasibility study is focused on a thermal-hydraulic aspect in normal operating conditions and several accident conditions. The key design characteristic of the FASES system obtained from design concepts is to have sufficient decay heat removal capability even in the accidents involving extended station blackout (SBO), failure of passive decay heat removal system (PDHRS), and failure of emergency core cooling system (ECCS). Based on the design concepts, we could define several thermal-hydraulic design requirements. Then, we performed thermal-hydraulic analysis for feasibility study and proposed the specific design of the FASES system satisfying several design requirements.

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.

Study of Decay Heat Removal Approach for Advanced Passive PWR During Station Blackout

2014 ◽  
Author(s):  
Jie Zou ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Safety System ◽  
Injection System ◽  
Passive Safety ◽  
System Failure ◽  
Decay Heat ◽  
Passive Safety Systems ◽  
Station Blackout ◽  
Safety Systems

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.

scholarly journals REACTOR CAVITY COOLING SYSTEM WITH PASSIVE SAFETY FEATURES ON RDE: THERMAL ANALYSIS DURING ACCIDENT

2019 ◽  
Vol 21 (2) ◽  
pp. 87
Author(s):  
Rahayu Kusumastuti ◽  
Sriyono Sriyono ◽  
Mulya Juarsa ◽  
Hendro Tjahjono ◽  
I. D. Irianto ◽  
...  
Keyword(s):  
Thermal Analysis ◽  
Natural Convection ◽  
Cooling System ◽  
Power Level ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Normal Operation ◽  
Passive Safety ◽  
Diesel Generator ◽  
Cavity Cooling

Reaktor Daya Eksperimental (RDE) is an experimental power reactor based on HTGR technology that implements inherent safety system. Its safety systems are in compliance with “defense in depth” philosophy. RDE is also equipped with reactor cavity cooling system (RCCS) used to remove the heat transferred from the reactor vessel to the containment structure. The RCCS is designed to fulfil this role by maintain the reactor vessel under the maximum allowable temperature during normal operation and protecting the containment structure in the event of failure of all passive cooling systems. The performance and reliability of the RCCS, therefore, are considered as critical factors in determining maximum design power level related to heat removal. RCCS for RDE will use a novel shape to efficiently remove the heat released from the RPV through thermal radiation and natural convection. This paper discusses the calculation of RCCS thermal analysis during accident. The RPV temperature must be maintained below 65ºC. The accident is assumed that there is no electricity from diesel generator supplied to the blower. The methodology used is based on the calculation of mathematical model of the RCCS in the passive mode. The heat is released through cavity by natural convection, in which the RCCS is capable to withdraw the heat at the rate of 50.54 kW per hour.Keywords: Passive safety, RCCS, RDE, Thermal analysis

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