Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Antariksawan, Anhar Riza; Wahyono, Puradwi Ismu; Taxwim,

doi:10.1088/1742-6596/962/1/012048

Cited by 4 publications

(5 citation statements)

References 13 publications

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“…Salah satu program perhitungan yang digunakan untuk analisis adalah RELAP5. Penelitian seperti ini sekaligus dimaksudkan untuk menguji validitas RELAP5 untuk aplikasi pada reaktor riset, termasuk pada reaktor jenis TRIGA (Antariksawan dkk., 2006;Antariksawan, Widodo, & Tjahyono, 2017;Antariksawan, Wahyono, & Taxwim, 2017;Antariksawan, Umar, dkk., 2017).…”

Section: Loca DI Reaktor Risetunclassified

“…Penggunaan RELAP5 untuk analisis termohidraulika reaktor Kartini juga telah dilakukan, baik untuk kondisi operasi normal maupun jika diasumsikan terjadi kebocoran, baik dari pipa pendingin primer maupun tabung berkas neutron. Model yang disusun telah memberikan hasil yang secara fenomena dapat diterima (Antariksawan, Wahyono, & Taxwim, 2017) Gambar 24. Evolusi Ketinggian Air di Tangki Reaktor pada LOCA Pipa Pecah…”

Section: Loca DI Reaktor Risetunclassified

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Keselamatan Reaktor Nuklir: Kecelakaan Dasar Desain dan Kecelakaan Parah

Antariksawan¹,

Juarsa²

2023

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Pembangkit Listrik Tenaga Nuklir (PLTN) telah memberikan kontribusi yang siginifikan pada pembangkitan energi dunia. Meskipun demikian, pertanyaan tentang keselamatan PLTN masih menjadi perdebatan. Tiga kecelakaan terburuk pada PLTN komersial yang terjadi dalam sejarah dunia, yaitu kecelakaan PLTN Three Mile Island unit 2 di Amerika Serikat, PLTN Chernobyl unit 4 di Ukraina, dan PLTN Fukushima Dai-ichi unit 1–4 di Jepang, menjadi salah satu subjek debat tentang keselamatan reaktor nuklir. Di sisi lain, ketiga kecelakaan tersebut juga menjadi dasar dari perbaikan keselamatan PLTN. Kecelakaan-kecelakaan tersebut telah mendorong dilakukannya banyak penelitian dalam rangka melakukan perbaikan dan mengevaluasi standar keselamatan PLTN. Buku ini menjelaskan apa yang terjadi pada ketiga kecelakaan tersebut dan upaya perbaikan melalui riset yang dilakukan serta hasil-hasilnya, terutama yang dilakukan oleh penulis. Pembahasan berfokus pada kecelakaan dasar desain dan kecelakaan parah seperti yang terjadi pada ketiga reaktor nuklir di atas. Lingkup pembahasan buku ini adalah kecelakaan pada reaktor berbasis air ringan, khususnya pada reaktor air bertekanan (pressurized water reactor, PWR). Dari aspek keilmuan, buku ini membahas kecelakaan reaktor nuklir dari sisi keilmuan termohidraulika. Buku ini sangat tepat dibaca bagi mahasiswa, dosen, peneliti, praktisi, serta masyarakat pemerhati nuklir. Buku ini diharapkan dapat memberikan pengetahuan tentang keselamatan reaktor nuklir dengan contoh kecelakaan PLTN dan hasil-hasil riset terkait kecelakaan PLTN.

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Section: Loca DI Reaktor Risetunclassified

Keselamatan Reaktor Nuklir: Kecelakaan Dasar Desain dan Kecelakaan Parah

Antariksawan¹,

Juarsa²

2023

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“…This code was de vel oped by the IDAHO Na tional En gi neer ing Lab o ra (INEL) in the USA. Fur ther more, the In no va tive Sys tem Soft ware (ISS) devel oped this code as RELAP/SCDAPSIM/Mod3.4 [15]. In this code, for this study, the re ac tor core was mod eled into a hot and av er age chan nel, with a time-de pend ent vol ume as a model of wa ter above the re ac tor core, and wa ter be low the re ac tor core is rep re sented as the heat sink as shown in fig.…”

Section: Steady-state Cal Cu La Tionmentioning

confidence: 99%

“…Fur ther more, LOFA could also be an a lyzed us ing EUREKA2/RR by add ing a by pass flow model [14]. On the other hand, the RELAP5 code [15] is also widely used to cal cu late the tran sient be hav ior of the light wa ter re ac tor cool ing system, in clud ing the ther mal-hy drau lic char ac ter is tics of the re search re ac tor [16].…”

Section: Introductionmentioning

confidence: 99%

Transient calculation on TRIGA 2000 of plate-type fuel element using RELAP5, EUREKA2/RR, and MTR-DYN codes

2021

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The TRIGA 2000 reactor Bandung is proposed to convert from rod-type fuel to plate-type fuel because there are no manufacturers in the world that still produce rod-type fuel. Calculating the safety parameters and their reliability related to reactor operation has become an important thing to do. The objective of this study was to perform the reactor transient calculation of TRIGA 2000 that uses plate-type fuel. The reactor core modification is based on the capability of the existing primary coolant system, without changing its flow rates. Thermal-hydraulics calculations related to reactivity insertion accident and loss of flow accident were analyzed using EUREKA2/RR, MTR-DYN, and RELAP5 codes. The chosen loss of flow accident scenario is a decrease in flow caused by a sudden stop of the pump power supply, while reactivity insertion accident is conducted by the withdrawal of control rods with maximum reactivity insertion of 32.33 ? 10?5 s?1. The calculated parameters are reactivity, reactor power, and temperature of the coolant, cladding, and fuel in the hottest channel position. In general, from a safety point of view, those computer codes were capable of performing the transient calculations with appropriate results. Based on the calculation results, during the transient condition of both reactivity insertion accident and loss of flow accident scenario, the reactor operation safety parameters do not exceed the allowable safety limit.

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“…RELAP5 (Reactor Excursion and Leak Analysis Program) can be used for reactor safety analysis and design, operators simulator training, and as an educational tool by universities [6][7][8][9]. BATAN has also been using this tool for several reactor safety analyses, both for research reactors and power reactors [10][11][12][13] . In the method, the NuScale reactor system and component is modeled using generic models available in the code, which then executed for transient simulation.…”

Section: Introductionmentioning

confidence: 99%

Simulation of Feed Water Temperature Decrease Accident in Nuscale Reactor

Susyadi

2018

Tri Dasa Mega

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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

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Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Cited by 4 publications

References 13 publications

Keselamatan Reaktor Nuklir: Kecelakaan Dasar Desain dan Kecelakaan Parah

Keselamatan Reaktor Nuklir: Kecelakaan Dasar Desain dan Kecelakaan Parah

Transient calculation on TRIGA 2000 of plate-type fuel element using RELAP5, EUREKA2/RR, and MTR-DYN codes

Simulation of Feed Water Temperature Decrease Accident in Nuscale Reactor

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