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"Generation II Nuclear Power Plants (NPPs) have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR) type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI) line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis."

"Any events presumed to risk the safety of a nuclear reactor should be analyzed. In a research reactor, the applicability of best estimate thermal-hydraulic codes has been assessed for safety analysis purposes. In this paper, the applicability of the RELAP/SCDAPSIM/MOD3.4 thermal-hydraulic code to one Indonesian research reactor, which is named TRIGA-2000, is performed. The aim is to validate the model and use the model to analyze the thermal-hydraulic characteristics of TRIGA-2000 for main transient events considered in the Safety Analysis Report. The validation was done by comparing the calculation results with experimental data mainly in steady state conditions. The comparison of calculation results with the measurement data showed good agreement with little discrepancies. Based on these results, simulations for thermal-hydraulic analyses were performed for loss of coolant transients. The calculation results also properly depicted the physic of the thermal-hydraulic phenomena following the loss of coolant transients. These results showed the adequacy of the model. It could be shown that the engineered safety features of TRIGA-2000 play an important role in keeping the reactor safe from the risk of postulated loss of a coolant accident."

"Teknologi pembuangan panas peluruhan secara pasif pada reaktor daya nuklir masih bergantung pada penggunaan penukar kalor yang memindahkan energi kalor ke dalam tangki atau kolam air dengan volume tertentu sebagai heat sink pamungkas dengan memanfaatkan konveksi alam atau gravitasi. Teknologi ini memiliki keterbatasan kapasitas dalam jangka waktu tertentu sehingga tetap mengandalkan sistem pendinginan secara aktif. Salah satu teknologi pendukung pengambil kalor peluruhan yang dapat digunakan adalah teknologi two-phased closed termosyphon (TPCT) yang telah digunakan sebagai objek penelitian dalam bentuk pemodelan maupun kegiatan eksperimen. Di sisi lain, sistem reaktor daya memiliki ukuran besar yang lebih praktis untuk dimodelkan dan disimulasikan. Salah satu program pemodelan adalah RELAP5/Mod3.4/SCDAP yang didesain untuk mensimulasikan proses pembangkitan dan pemindahan kalor pada reaktor daya nuklir berpendingin air saat kondisi operasi normal dan abnormal. Tujuan penelitian ini adalah melakukan simulasi kinerja termal TPCT yang telah dimodifikasi dengan menambahkan tangki air pendingin di sekeliling evaporator sebagai sumber kalor. Pemodelan RELAP5 yang diperoleh perlu divalidasi dengan hasil pemodelan lain seperti FLUENT dan hasil eksperimen yang telah dilakukan. Hasil validasi dengan memberikan beban kalor berbeda pada bagian evaporator menunjukkan tidak adanya fenomena dryout pada RELAP5 yang berbeda dengan hasil simulasi FLUENT. Hal ini disebabkan oleh perbedaan konsep pemodelan antara FLUENT dan RELAP5, namun demikian nilai tahanan termal TPCT memiliki kesamaan karakteristik di antara kedua metoda. Validasi simulasi dengan hasil eksperimen pada filling ratio (FR) 30 %, 45 %, 60 % menunjukkan karakterisasi perubahan temperatur dinding TPCT yang sama pada variabel FR 60 %. Pada FR lebih rendah, terdapat variasi perbedaan hasil simulasi dengan eksperimen terutama pada pencapaian temperatur pendidihan air tangki dan pola pengambilan kalor kondenser. Simulasi tambahan dengan FR 70 % dan 100 % menunjukkan kinerja termal TPCT yang lebih optimum pada FR 70 % dan menurun pada FR 100 %. Pemodelan TPCT dengan 2 volume radial menghasilkan simulasi yang lebih baik, namun demikian dibutuhkan pemahaman yang lebih tepat mengenai peranan parameter gas non-kondensibel pada nodalisasi model dengan kondisi vakum. Berdasarkan hasil validasi, simulasi TPCT dengan RELAP5 lebih tepat digunakan untuk tujuan prediksi kinerja termal TPCT berdasarkan FR yang optimum.

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The current technology in removing residual heat in passive way in nuclear power reactor still depends on the use of heat exchanger to transfer heat energy into a water tank or pool with certain volume as the ultimate heat sink utilizing natural circulation or gravitation. This technology still requires active cooling supply system in certain time period due to the limited capability. On of the proposed support technology to remove residual heat is two-phased closed thermosyphon (TPCT), which has been used as research object in form of modelling or experimental activity. In other side, power reactor system has big dimension, which is more practical to be modelled to simulate its performance and safety level. On of the modelling code is RELAP5/Mod3.4/SCDAP, designed to simulate the heat generation and removal process inside the water-cooled nuclear power reactor in normal and abnormal operation. Therefore, the research purposes are to model and simulate the TPCT, which has been modified by adding the external water tank around the evaporator. The generated model of TPCT using RELAP5 has to be validated using other modelling tool such as FLUENT and experimental results. The generated model of TPCT using RELAP5 has to be validated using other modelling tool such as FLUENT and experimental results. One of validation results by giving different heat load into the evaporator shows no indication of dry-out in the RELAP5 simulation as obtained otherwise in the FLUENT simulation. This is because the different modelling concept between FLUENT and RELAP5. However, the calculated thermal resistances had similar characteristics from both calculation tools. Simulation validation with experimental results using filling ratio (FR) of 30 %, 45 %, and 60 % shows a much more similar characteristics of the evaporator, adiabatic, and condenser wall temperature with higher FR of 60 %. In lower FR, there are several output variations in achieving the water tank boiling temperature and condenser heat transfer. Additional simulations with FR 70 % and 100 % indicated more optimum thermal performances of TPCT, especially in FR 70 %, which decreased in FR 100 %. TPCT modelling with 2 radial volumes results in a better simulation, however it requires a better understanding regarding the role of non-condensable gas in the model in the vacuum condition. Based on the validation results, TPCT simulation using RELAP5 is better conducted based on the optimum condition in term of filling ratio to predict the thermal performances of TPCT."

"Untuk meningkatkan keselamatan termal pada saat terjadi kecelakaan akibat station blackout, vertical straight wickless-heat pipe pipa kalor lurus tanpa sumbu kapiler yang diletakkan secara vertikal diusulkan sebagai sistem pendingin pasif baru untuk pembuangan panas sisa hasil peluruhan di kolam penyimpanan bahan bakar bekas nuklir. Pipa kalor akan membuang panas peluruhan dari kolam penyimpanan bahan bakar bekas nuklir dan dapat menjaga sistem tetap aman. Tujuan penelitian ini adalah untuk menginvestigasi karakteristik, fenomena perpindahan kalor, dan unjuk kerja termal pipa kalor yang digunakan mencari pengaruh kecepatan pendinginan dengan besarnya kalor yang harus dibuang, menganalisis keserupaan dimensi dari pipa kalor yang digunakan, dan mengetahui teknologi pipa kalor yang dapat digunakan sebagai sistem keselamatan pasif di instalasi nuklir pada kondisi kecelakaan akibat station blackout. Investigasi secara eksperimen dilakukan dengan mempertimbangkan pengaruh tekanan awal pipa kalor, evaporator filling ratio, beban kalor evaporator, dan laju aliran pendingin di water jacket. Air pendingin disirkulasikan dalam water jacket sebagai penyerap kalor di bagian condenser. Simulasi dengan program perhitungan termohidraulika RELAP5/MOD3.2 dilakukan untuk mendukung dan membandingkan dengan hasil eksperimen yang didapatkan. Hasil eksperimen menunjukkan bahwa unjuk kerja termal terbaik pipa kalor didapatkan pada tahanan termal 0,016 C/W. Unjuk kerja termal terbaik didapatkan pada saat pipa kalor diberikan filling ratio 80 , tekanan awal terendah, laju aliran pendingin tertinggi, dan beban kalor evaporator tertinggi. Dari nilai tahanan termal tersebut didapatkan bahwa pipa kalor ini memiliki kemampuan memindahkan kalor 199 kali lebih besar jika dibandingkan dengan batang pejal tembaga dengan geometri yang sama. Model pipa kalor dalam simulasi dengan RELAP5/MOD3.2 dapat digunakan untuk mendukung investigasi secara eksperimen dalam memprediksi fenomena yang berlangung di bagian dalam pipa kalor. Analisis dimensi dan keserupaan pipa kalor yang didapatkan bisa digunakan untuk merancang pipa kalor lain dengan geometri yang berbeda namun tetap menghasilkan unjuk kerja termal yang sama. Kesimpulan investigasi yang dilakukan menunjukkan bahwa pipa kalor ini memiliki unjuk kerja termal yang tinggi dan dapat digunakan sebagai sistem pendingin pasif di kolam penyimpanan bahan bakar bekas nuklir pada saat terjadinya kecelakaan akibat station blackout.

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To enhance the thermal safety when station blackout accident occurs, a vertical straight wickless heat pipe is proposed as a new passive residual heat removal system in nuclear spent fuel storage pool. The heat pipe will remove the decay heat from nuclear spent fuel pool and keep the system safe. The objective of this research is to investigate the characteristics, heat transfer phenomena, and thermal performance of heat pipe, to analyse the effect of coolant flowrate against heat to be removed, analysing the dimensional similarity of heat pipe, and to know the heat pipe technology that could be used as passive safety system in nuclear installation during to station blackout accident. The experimental investigation was conducted to investigate the heat transfer phenomena and heat pipe thermal performance with considering the influence of heat pipe initial pressure, evaporator filling ratio, evaporator heat load, and coolant volumetric flow rate of water jacket. Cooling water was circulated in water jacket as condenser cooling system. A numerical simulation with nuclear reactor thermal hydraulic code RELAP5 MOD3.2 was performed to support and to compare with the experimental results. The experimental results showed that the best thermal performance was obtained at thermal resistance of 0.016 C W, with filling ratio of 80 , the lower initial pressure, higher coolant volumetric flow rate, and higher heat load of evaporator. From thermal resistance analysis, it is found that the heat pipe has the ability to remove heat 199 times greater than copper rod with the same geometry. The RELAP5 MOD3.2 simulation model can be used to support experimental investigation and to predict the phenomena inside the heat pipe. The dimensional analysis and similitude of the heat pipe can be applied to design the other heat pipe with different geometries with produces the same thermal performance. The conclusion of investigation showed that vertical straight wickless heat pipe has higher thermal performance and can be used as passive residual heat removal system of nuclear spent fuel pool when station blackout occurs."