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

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

2012 ◽  
Author(s):  
Lorenzo Damiani ◽  
Alessandro Pini Prato
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Coolant Temperature ◽  
Feed Water ◽  
Steam Generators ◽  
Fast Reactors ◽  
Primary Coolant ◽  
Reaction Heat ◽  
Generator System ◽  
Decay Heat

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is LEADER (Lead cooled European Advanced DEmonstrator Reactor) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, ALFRED (Advanced Lead Fast Reactor European Demonstrator). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The Authors have addressed the issue of bayonet-tube steam generators proposing the EBBSG (External Boiling Bayonet Steam Generator) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the Authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

2014 ◽  
Vol 137 (3) ◽  
Author(s):  
Lorenzo Damiani ◽  
Alessandro Pini Prato
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Coolant Temperature ◽  
Feed Water ◽  
Steam Generators ◽  
Fast Reactors ◽  
Primary Coolant ◽  
Reaction Heat ◽  
Decay Heat

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is lead cooled European advanced demonstrator reactor (LEADER) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, advanced lead fast reactor European demonstrator (ALFRED). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The authors have addressed the issue of bayonet-tube steam generators proposing the external boiling bayonet steam generator (EBBSG) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Experimental and Numerical Study of Water Distillation Performance of Small-Scale Direct Contact Membrane Distillation System

2017 ◽  
Author(s):  
Danielle Park ◽  
Elnaz Norouzi ◽  
Chanwoo Park
Keyword(s):  
Water Temperature ◽  
Direct Contact ◽  
Membrane Distillation ◽  
Small Scale ◽  
Feed Water ◽  
Water Production ◽  
Water Distillation ◽  
Direct Contact Membrane Distillation ◽  
Feed Temperature ◽  
Feed Water Temperature

A small-scale Direct Contact Membrane Distillation (DCMD) system was built to investigate its water distillation performance for varying inlet temperatures and flow rates of feed and permeate streams, and salinity. A counterflow configuration between the feed and permeate streams was used to achieve an efficient heat exchange. A two-dimensional Computational Fluid Dynamics (CFD) model was developed and validated using the experimental results. The numerical results were compared with the experiments and found to be in good agreement. From this study, the most desirable conditions for distilled water production were found to be a higher feed water temperature, lower permeate temperature, higher flow rate and less salinity. The feed water temperature had a greater impact on the water production than the permeate water temperature. The numerical simulation showed that the water mass flux was maximum at the inlet of the feed stream where the feed temperature was the highest and rapidly decreased as the feed temperature decreased.

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.

Preliminary Probabilistic Safety Assessment of the IRIS Plant

2002 ◽  
Author(s):  
Yuko O. Mizuno ◽  
Katsunori Ogura ◽  
Hisashi Ninokata ◽  
Lawrence E. Conway
Keyword(s):  
Safety Assessment ◽  
Fault Tree ◽  
Fault Tree Analysis ◽  
Heat Removal ◽  
Primary System ◽  
Event Tree ◽  
Secondary System ◽  
Probabilistic Safety Assessment ◽  
Loss Of Coolant Accident ◽  
Probabilistic Safety

A preliminary level-1 probabilistic safety assessment of the IRIS plant has been performed. The first focus is on five internal initiating events, such as primary system break (loss-of-coolant accident and steam generator tube rupture) and transients (secondary system line break and loss-of-off-site power). In this study, the event tree for each initiating event was developed and the fault tree analysis of the event tree headings was carried out. In particular, since one of the IRIS safety systems, the passive emergency heat removal system, is unique to the IRIS plant and its reliability is key to the core damage frequency evaluation, it received more extensive fault-tree development. Finally the dominant sequences that lead to severe accidents and the failures in the main and support systems are identified.

Loss of Feed Water Accident in an Integral Pressurized Water Reactor (IPWR)

2010 ◽  
Vol 171-172 ◽  
pp. 379-384
Author(s):  
Khan Salah Ud Din ◽  
Min Jun Peng ◽  
Muhammad Zubair
Keyword(s):  
Heat Removal ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Water Accident ◽  
Relap5 Code ◽  
Removal System ◽  
Residual Heat

In this paper research has been carried out on Loss of Feed Water Accident (LOFW) scenario of the Integral Pressurized Water Reactor ( IPWR) under two circumstances by the use of thermal hydraulic system code i.e Relap5/Mod3.4. In the first one, Passive Residual Heat Removal System (PRHRS) which is designed to absorb core residual heat in case of transient conditions is included which has the function of operating under the accident vulnerabilities. Concerning with the second case i.e without the use of PRHRS rather a tank of water which has the capacity of about 8% of the total feed water supply and is operated under accident scenario is considered. Taken into account these conditions,first the nodalization diagram of the two cases have been figured out then according to the LOFW accident time event scenario use the Relap5 code to simulate the accident. Finally the graphical explanation (separately) of the two cases with graphical approach as well as the conclusion is given at the end.

Performance Improvement of Tubular-Ice Making Machine by Reducing the Feed Water Temperature with Shell and Tube Heat Exchanger

2014 ◽  
Vol 564 ◽  
pp. 298-303 ◽  
Author(s):  
Sukkapop Nakornsri ◽  
Ratchaphon Suntivarakorn ◽  
Khanison Thanutwutthigorn
Keyword(s):  
Energy Consumption ◽  
Heat Exchanger ◽  
Water Temperature ◽  
Feed Water ◽  
Tube Heat Exchanger ◽  
Before And After ◽  
Energy Expense ◽  
Shell And Tube ◽  
Ice Production ◽  
Feed Water Temperature

The paper presents a study on performance of improvement carried out on a tubular ice making machine by reducing feed water temperature using a shell and tube heat exchanger. An ice making machine with a capacity of 20 ton per day was used to in this study. The shell and tube heat exchanger was designed to reduce the feed water temperature. It has a length of 1.5 meters and 3.5 m2 of heat transfer area. The heat exchanger were installed at the inlet of an evaporator. Then, the performance and energy consumption of the ice making machine were examined with the experimental condition of cooling temperature at 20°C and 30°C. The comparison of the performance and energy consumption before and after heat exchanger installation were also studied. From the experiment, the results showed that the heat exchanger can reduce the feed water temperature by an average of 7.5°C, and the energy consumption was decreased by 17.1 %. The cycle time for ice production was decreased to 7 minute/cycle, and the capacity of the ice making machine was increased by 17.5%. If the cost of running the heat exchanger is 200,000 baht, this solution can potentially save the energy expense of up to 269,570 baht/year with a payback period of 0.93 year.

Case study on the effects of feed water temperature on the performance of a reverse osmosis desalination system

2020 ◽  
Vol 205 ◽  
pp. 46-52
Author(s):  
Abdulrahim Kalendar ◽  
Aboelyazied Kulaib ◽  
Shafqat Hussain ◽  
Yousuf Alhendal
Keyword(s):  
Water Temperature ◽  
Reverse Osmosis ◽  
Feed Water ◽  
Feed Water Temperature ◽  
Reverse Osmosis Desalination
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