Analysis of Protected Accidental Transients in the EFIT Reactor With the RELAP5 Thermal-Hydraulic Code

2008 ◽  
Author(s):  
Giacomino Bandini ◽  
Maddalena Casamirra ◽  
Francesco Castiglia ◽  
Mariarosa Giardina ◽  
Paride Meloni ◽  
...  
Keyword(s):  
Proton Beam ◽  
Thermal Power ◽  
Heat Removal ◽  
Reactor Power ◽  
Lead Target ◽  
Primary System ◽  
Safety Behavior ◽  
Forced Circulation ◽  
Detailed Model ◽  
Decay Heat

The European Facility for Industrial Transmutation (EFIT) is aimed at demonstrating the feasibility of transmutation process through the Accelerator Driven System (ADS) route on an industrial scale. The conceptual design of this reactor of about 400 MW thermal power is under development in the frame of the European EUROTRANS Integrated Project of the EURATOM Sixth Framework Program (FP6). EFIT is a pool-type reactor cooled by forced circulation of lead in the primary system where the heat is removed by steam generators installed inside the reactor vessel. The reactor power is sustained by a spallation neutron source supplied by a proton beam impinging on a lead target at the core centre. 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 in case of loss of secondary circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety requirements, and confirm the inherent safety behavior of the reactor, such as decay heat removal in accidental conditions relying on natural circulation in the primary system. The accident analyses for the EFIT reactor include both protected and unprotected transients, on whether the reactor automatic trip, consisting in proton beam switch off, is actuated or not by the protection system. In this paper, the main results of the analyses of some protected transients with RELAP5 are presented. The analyzed transients concern the Protected Loss of Heat Sink (PLOHS), in which the DHR system plays a key role in bringing the reactor in safe conditions, and the Protected Loss of Flow (PLOF) transients with partial or total loss of forced circulation in the primary system.

Validation Experiment for the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) With PASCAL Facility

2012 ◽  
Author(s):  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yu-Sun Park ◽  
Bok-Deuk Kim ◽  
Kyoung-Ho Kang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heat Exchanger ◽  
System Analysis ◽  
Thermal Power ◽  
Heat Removal ◽  
Steady State Condition ◽  
Single Tube ◽  
Decay Heat ◽  
Validation Experiment

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.

Decay Heat Benchmark for Uranium-Free Fuels With Minor Actinides

2010 ◽  
Author(s):  
Andrei Rineiski ◽  
Ge´rald Rimpault ◽  
Georgios Glinatsis ◽  
Nadia Messaoudi ◽  
Sandro Pelloni ◽  
...  
Keyword(s):  
Heat Removal ◽  
Reactor Power ◽  
Heat Energy ◽  
Minor Actinides ◽  
Decay Heat ◽  
Unstable Nuclei ◽  
Safety Studies ◽  
Heat Removal System ◽  
Type Fuel ◽  
Removal System

The decay heat (energy due to decay of unstable nuclei) is a small fraction of reactor power at nominal conditions, but after reactor shut-down it is the most important heat source. For taking this source into account in design and safety studies, recommendations are available for fuels of operating reactors, such as UOX and MOX. Fuels for EFIT (European Facility for Industrial Transmutation), unlike UOX and MOX, should contain a significant amount of Minor Actinides (MAs) that would influence decay heat. CEA, CIEMAT, ENEA, FZK (now KIT), PSI and SCK•CEN established a benchmark case and computed decay heat curves for MA-bearing fuels and a MOX-type fuel. The decay heat in the fuels with MAs is appreciably higher than in MOX, except for low burnup cases after short cooling times. This should be taken into account in the design of the decay heat removal system for EFIT. The obtained differences between the decay heat in MA-bearing and MOX fuels are supposed to be representative for the benchmark (or similar) conditions. More effort is needed to evaluate the uncertainties of the computed results.

scholarly journals Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

2002 ◽  
Author(s):  
S. Michael Modro ◽  
James Fisher ◽  
Kevan Weaver ◽  
Pierre Babka ◽  
Jose Reyes ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Thermal Power ◽  
Heat Removal ◽  
Steam Pressure ◽  
Light Water Reactor ◽  
Primary System ◽  
Light Water ◽  
Water Reactor ◽  
System Pressure ◽  
Environmental Laboratory

The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure 1.52 MPa and the net electrical output is 35–50 MWe.

scholarly journals PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS

2021 ◽  
Vol 23 (2) ◽  
pp. 47
Author(s):  
Andi Sofrany Ekariansyah ◽  
Surip Widodo ◽  
Susyadi Susyadi ◽  
Hendro Tjahjono
Keyword(s):  
Heat Exchanger ◽  
Heat Removal ◽  
Air Cooling ◽  
Water Tank ◽  
Forced Circulation ◽  
Design Assessment ◽  
Decay Heat ◽  
Time Period ◽  
Station Blackout ◽  
Safety Features

The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.  

The Safety Analysis of the 600 MWe Sodium-Cooled Fast Reactor With MARS-LMR

2012 ◽  
Author(s):  
Kwi Lim Lee ◽  
Kwi Seok Ha ◽  
Hae Yong Jeong ◽  
Won Pyo Chang
Keyword(s):  
Heat Exchanger ◽  
Heat Transport ◽  
Transport System ◽  
Fast Reactor ◽  
Thermal Power ◽  
Heat Removal ◽  
Decay Heat ◽  
Metal Fuel ◽  
Heat Removal System ◽  
Independent Loops

Korea Atomic Energy Research Institute (KAERI) has been developing a conceptual design of the demonstration fast reactor (DFR), which is the pool type sodium cooled fast reactor with the thermal power of 1548.2 MW and the core loaded with metal fuel. The DFR is composed of a Primary Heat Transport System (PHTS), an Intermediate Heat Transport System (IHTS), a Steam Generating System (SGs) and a decay heat removal system (DHRS). The DHRS is composed of 2 units of Passive Decay-heat Removal Circuits (PDRC) and 2 units of Active Decay-heat Removal Circuits (ADRC). The PDRC consists of two independent loops with sodium-to-sodium Decay Heat eXchanger (DHX) and natural-draft sodium-to-Air Heat eXchanger (AHX). The ADRC consists of two independent loops with sodium-to-sodium DHX and Forced-Draft sodium-to-air Heat eXchanger (FDHX) located in the upper region of the reactor building. The PDRC is very different from that of KALIMER-600 on the points of the submerged location and the heat transfer mechanism. For the identification of safety characteristics, 5 DBE’s (Design Bases Events) are analyzed using the MARS-LMR code. The representative DBE’s are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Reactor Vessel Leak and Pipe Break. As a result, it is identified that the DFR were appropriately performed as designed and the temperatures of the fuel and the structure were evaluated to satisfy the criteria.

Reliability comparison of natural convection and forced circulation configurations of a FBR thermo-siphon decay heat removal system

2020 ◽  
Vol 121 ◽  
pp. 103242
Author(s):  
M. Ramakrishnan ◽  
A. John Arul ◽  
U. Parthasarathy ◽  
V. Bhuvana ◽  
V. Magesh Mari Raj ◽  
...  
Keyword(s):  
Natural Convection ◽  
Heat Removal ◽  
Forced Circulation ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

scholarly journals SIMULATION OF FEED WATER TEMPERATURE DECREASE ACCIDENT IN NUSCALE REACTOR

2018 ◽  
Vol 20 (3) ◽  
pp. 133 ◽  
Author(s):  
Susyadi Susyadi
Keyword(s):  
Water Temperature ◽  
Heat Removal ◽  
Coolant Temperature ◽  
Feed Water ◽  
Primary System ◽  
Temperature Decrease ◽  
Secondary System ◽  
Decay Heat ◽  
Feed Water Temperature ◽  
Relap5 Code

Study on thermal hydraulic behavior of the NuScale reactor during secondary system malfunction that causes a feed water temperature decrease has been conducted using RELAP5 code. This study is necessary to investigate the performance of safety system and design in dealing with an accident. The method used involves simulation of reactor transient through numerical modeling and calculation in RELAP5 code covering primary and secondary system, including the decay heat removal system (DHRS). The investigation focuses on the flow and heat transfer characteristics that occurs during the transient. The  calculation result shows that at the beginning, core power increases up to trip set point of 200 MW which is driven by positive feedback reactivity of coolant overcooling and automatic control rod bank adjustment. Meanwhile, the core exit coolant temperature increases up to 600 K. and primary system circulation flow rate speeds up to 556 kg/s. After that, the reactor trips and power drops sharply, followed by opening of DHRS valves and closing of steam line and feed water isolation valves. The simulation shows that, the DHRS are capable to transfer decay heat to the reactor pool and as a result the primary system temperature and pressure decreases. The reactor could stay in safe shutdown state afterward.Keywords: NuScale, RELAP5, feed water, decay heat, simulation SIMULASI KECELAKAAN PENURUNAN TEMPERATUR AIR UMPAN DI REACTOR NUSCALE. Studi tentang perilaku termalhidraulik reaktor NuScale saat terjadi kerusakan sistem sekunder yang menyebabkan penurunan suhu air umpan telah dilakukan dengan menggunakan kode RELAP5. Penelitian ini penting untuk menyelidiki kinerja disain dan sistem keselamatan reaktor dalam menghadapi kecelakaan. Metoda yang digunakan melibatkan simulasi transien reaktor melalui pemodelan dan kalkulasi numerik dengan RELAP5 yang meliputi sistem primer dan sekunder serta sistem pembuangan panas peluruhan (DHRS). Investigasi berfokus pada aliran dan karakteristik perpindahan panas yang terjadi selama transien. Hasil perhitungan menunjukkan bahwa pada awalnya, terjadi peningkatan daya teras hingga mencapai titik seting pemadaman (trip) 200 MW, sebagai akibat dari umpan balik reaktivitas positif dari pendinginan fluida sistem primar dan respon otomatis penaikan batang kendali. Sementara itu, suhu keluaran teras meningkat menjadi 600 K serta laju aliran sirkulasi sistem primer meningkat menjadi 556 kg/s. Setelah itu, reaktor padam dimana daya menurun tajam dan diikuti pembukaan katup DHRS dan penutupan katup pada jalur uap dan air umpan. Simulasi ini menunjukkan bahwa, DHRS mampu membuang panas ke kolam reaktor, dimana suhu serta tekanan sistem primer menurun. Reaktor tetap dalam keadaan shutdown aman sesudahnya.Kata kunci: NuScale, RELAP5, air umpan, panas peluruhan, simulasi

Environmentally-Friendly HTGR: MHR-50/100—Concept and Characteristics

2011 ◽  
Author(s):  
Isao Minatsuki ◽  
Tomomi Otani ◽  
Katsusuke Shimizu ◽  
Tetsuo Saguchi ◽  
Sunao Oyama ◽  
...  
Keyword(s):  
High Temperature ◽  
Conceptual Design ◽  
Financial Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Power ◽  
Primary System ◽  
Plant Design ◽  
Plant Dynamics ◽  
High Temperature Gas

A business plan and a new concept of the Mitsubishi small-sized High temperature gas-cooled modular Reactors (MHR-50/100) had been developed as reported in a paper at the HTR-2010 conference in Prague. The present paper reports the results of ensuing conceptual design study including updated market researches, improved safety features of the plant, and the plant dynamics analysis. Market researches on Japan, the USA, Southeast Asia and the Middle East have been updated applying the latest energy outlook data. The result shows that the potential market share for the type of HTGR (high temperature gas reactor) reactors is expected to be 10–20% in new construction of heat source plants in those market areas. A financial analysis made based on the results of the updated market research and the plant cost evaluations indicates that the feasibility of an HTGR business potentially exists. Concerning about the conceptual design, as main themes of the study, a plant design, safety design and plant dynamics have been carried out. The MHR-50/100 high safety characteristics have been confirmed based on the results of the following studies as reported in the present paper: (1) An investigation of a safety scenario during occurrence of a Total Black Out event; (2) An analysis of the reactor decay heat removal via a natural circulation. Lastly, the control methods for the reactor and associated steam cycle system for the MHR-50 have been studied. The results show that the reactor power changes can be effectively achieved by controlling the primary system helium flow rate. The ASURA code developed by MHI is used for simulation of such typical plant transients as 10% step load reduction and full load rejection. The results confirm the easy operability and controllability of the plant.

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

2008 ◽  
Vol 2008 ◽  
pp. 1-8 ◽  
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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