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.

iB1350: Part 2 — Level1 PRA Considering Optimization of Safety Systems for the iB1350

2018 ◽  
Author(s):  
Tanaka Go ◽  
Sato Takashi ◽  
Komori Yuji ◽  
Matsumoto Keiji
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Lessons Learned ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Active Safety ◽  
Decay Heat ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

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.

Quality Management during Manufacturing of High Tempertaure Thin Walled Austenitic Stainless Steel Sodium Tanks of Prototype Fast Breeder Reactor

2013 ◽  
Vol 794 ◽  
pp. 507-513
Author(s):  
R.G. Rangasamy ◽  
Prabhat Kumar
Keyword(s):  
Stainless Steel ◽  
High Temperature ◽  
Austenitic Stainless Steel ◽  
Quality Management ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Heat Removal ◽  
Fast Breeder Reactor ◽  
Decay Heat ◽  
Thin Walled

Austenitic stainless steels are the major material of construction for the fast breeder reactors in view of their adequate high temperature mechanical properties, compatibility with liquid sodium coolant, good weldability, availability of design data and above all the fairly vast and satisfactory experience in the use of these steels for high temperature service. All the Nuclear Steam Supply System (NSSS) components of FBR are thin walled structure and require manufacture to very close tolerances under nuclear clean conditions. As a result of high temperature operation and thin wall construction, the acceptance criteria are stringent as compared to ASME Section III. The material of construction is Austenitic stainless steel 316 LN and 304 LN with controlled Chemistry and calls for additional tests and requirements as compared to ASTM standards. Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type, 500 MWe reactor which is at advanced stage of construction at Kalpakkam, Tamilnadu, India. In PFBR, the normal heat transport is mainly through two secondary loops and in their absence; the decay heat removal is through four passive and independent safety grade decay heat removal loops (SGDHR). The secondary sodium circuit and the SGHDR circuit consist of sodium tanks for various applications such as storage, transfer, pressure mitigation and to take care of volumetric expansion. The sodium tanks are thin walled cylindrical vertical vessels with predominantly torispherical dished heads at the top and bottom. These tanks are provided with pull-out nozzles which were successfully made by cold forming. Surface thermocouples and heaters, wire type leak detectors are provided on these tanks. These tanks are insulated with bonded mineral wool and with aluminum cladding. All the butt welds in pressure parts were subjected to 100% Radiographic examination. These tanks were subjected to hydrotest, pneumatic test and helium leak test under vacuum. The principal material of construction being stainless steel for the sodium tanks shall be handled with care following best engineering practices coupled with stringent QA requirements to avoid stress corrosion cracking in the highly brackish environment. Intergranular stress corrosion cracking and hot cracking are additional factors to be addressed for the welding of stainless steel components. Pickling and passivation, Testing with chemistry controlled demineralised water are salient steps in manufacturing. Corrosion protection and preservation during fabrication, erection and post erection is a mandatory stipulation in the QA programme. Enhanced reliability of welded components can be achieved mainly through quality control and quality assurance procedures in addition to design and metallurgy. The diverse and redundant inspections in terms of both operator and technique are required for components where zero failure is desired & claimed. This paper highlights the step by step quality management methodologies adopted during the manufacturing of high temperature thin walled austenitic stainless steel sodium tanks of PFBR.

Severe Accident Analysis for BWR With Containment Outer Pool

2012 ◽  
Author(s):  
Koki Yoshimura ◽  
Kohei Hisamochi
Keyword(s):  
Cooling System ◽  
Water Injection ◽  
Heat Removal ◽  
Severe Accident ◽  
Gas Temperature ◽  
Passive System ◽  
Active Components ◽  
Active System ◽  
Station Blackout ◽  
Materials Used

Newly designed plants, e.g., next-generation light water reactor or ESBWR, employ a passive containment cooling system and have an enhanced safety with RHRs (Residual Heat Removal system) including active components. Passive containment cooling systems have the advantage of a simple mechanism, while materials used for the systems are too large to employ these systems to existing plants. Combination of passive system and active system is considered to decrease amount of material for existing plants. In this study, alternatives of applying containment outer pool as a passive system have been developed for existing BWRs, and effects of outer pool on BDBA (Beyond Design Basis Accident) have been evaluated. For the evaluation of containment outer pool, it is assumed that there would be no on-site power at the loss of off-site power event, so called “SBO (Station BlackOut)”. Then, the core of this plant would be uncovered, heated up, and damaged. Finally, the reactor pressure vessel would be breached. Containment gas temperature reached the containment failure temperature criteria without water injection. With water injection, containment pressure reached the failure pressure criteria. With this situation, using outer pool is one of the candidates to mitigate the accident. Several case studies for the outer pool have been carried out considering several parts of containment surface area, which are PCV (Pressure Containment vessel) head, W/W (Wet Well), and PCV shell. As a result of these studies, the characteristics of each containment outer pool strategies have become clear. Cooling PCV head can protect it from over-temperature, although its effect is limited and W/W venting can not be delayed. Cooling suppression pool has an effect of pressure suppressing effect when RPV is intact. Cooling PCV shell has both effect of decreasing gas temperature and suppressing pressure.

Sign in / Sign up

Export Citation Format

Share Document