Development of a phenomena identification ranking table for simulating a station blackout transient of a pressurized water reactor with a thermal-hydraulic integral effect test facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

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Simulation of JAERI Downcomer Effective Water Head Experiments With WCOBRA/TRAC-TF2

Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation ◽

10.1115/icone26-82610 ◽

2018 ◽

Author(s):

Jeffrey R. Kobelak ◽

Jun Liao ◽

Katsuhiro Ohkawa

Keyword(s):

Hydraulic System ◽

Large Scale ◽

Reactor Vessel ◽

Test Facility ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

System Pressure ◽

Water Head ◽

Effective Water

During the reflood phase of a postulated large break loss-of-coolant accident (LBLOCA), the liquid head in the reactor vessel downcomer provides the driving force to reflood the core. Since the reflood rate is a function of the downcomer inventory, the calculation of the downcomer liquid inventory is critical in simulating the reflood phase of a postulated LBLOCA accident in a pressurized water reactor. Since the reactor coolant system pressure decreases rapidly after the onset of a LBLOCA transient, the walls surrounding the downcomer become superheated for the duration of the transient. The Japan Atomic Energy Research Institute (JAERI) downcomer effective water head test facility was designed to study boiling and steam-water interaction in the reactor vessel downcomer under prototypical reflood conditions. A number of tests were conducted at this facility with varying degrees of wall superheating (among other things) that cover the expected degree of superheating in a pressurized water reactor. The wall superheating achieved at the JAERI facility is greater than that of other large-scale facilities that are typically simulated to validate thermal-hydraulic system codes. WCOBRA/TRAC-TF2 is the thermal-hydraulic system code utilized in the FULL SPECTRUM™ LOCA (FSLOCA™) evaluation model (EM). The ability of the WCOBRA/TRAC-TF2 code to predict phenomena occurring in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA has been previously validated. However, only limited wall superheating was present in the existing validation basis. As such, two experiments conducted at the JAERI downcomer effective water head test facility are simulated to provide additional information on the capability of WCOBRA/TRAC-TF2 to predict the liquid inventory in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA. The code captured all the trends observed in the experimental data for both Run 115 and Run 121. The various collapsed liquid levels tended to be well-predicted or under-predicted by the code after the initial simulated accumulator injection period.

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Integral Effect Test for Performance Evaluation of the PAFS (Passive Auxiliary Feedwater System) During a SBO (Station Blackout) Transient

Volume 2B: Thermal Hydraulics ◽

10.1115/icone22-30563 ◽

2014 ◽

Author(s):

Byoung-Uhn Bae ◽

Seok Kim ◽

Yu-Sun Park ◽

Yun-Je Cho ◽

Kyoung-Ho Kang

Keyword(s):

Steam Generator ◽

Flow Rate ◽

Water Levels ◽

Experimental Result ◽

Test Facility ◽

Decay Heat ◽

The Core ◽

Integral Effect ◽

Station Blackout ◽

Effect Test

Station blackout (SBO) accident is considered as one of the most significant design extension conditions (DECs), which has been extensively focused after the Fukushima Dai-chi accident. When the SBO accident occurs in the APR+ (Advance Power Reactor Plus), the PAFS (Passive Auxiliary Feedwater System), which is an advanced safety feature adopted in the APR+, should play a significant role to cool down the core decay heat without any operation of active safety systems. This study focuses on validation of the cooling and operational performance for the PAFS during the SBO transient with utilizing an integral effect test facility, ATLAS-PAFS. In order to simulate the SBO transient of the APR+ as realistically as possible, a pertinent scaling approach was taken into account. The initial steady-state conditions and the sequence of event in the SBO scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. In the transient simulation, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate at the return-water line were measured and investigated. Following the reactor trip at the initiation of the transient, the coolant inventory of the secondary system of the steam generator was reduced by the repeated opening and closing of the MSSV. When the collapsed water level reached 25% of wide range, the PAFS was actuated to cool down the primary system by the condensation heat transfer at the PCHX (Passive Condensation Heat Exchanger). The pressure and the temperature of the reactor coolant system continuously decreased during the heat removal by the PAFS operation. It points out that the PAFS can supply auxiliary feedwater to the steam generator and remove the core decay heat without any active system. From the present experimental result, it could be concluded that the APR+ has the capability of coping with the hypothetical SBO scenario with adopting the PAFS and proper set-points of its operation. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS, RELAP5 as well as the SPACE code and to identify any code deficiency for a SBO simulation with an operation of the PAFS.

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Perhitungan Perubahan Reaktivitas Pada Reaktor Serba Guna G.A. Siwabessy Akibat Pengoperasian Power Ramp Test Facility

Prosiding SNFA (Seminar Nasional Fisika dan Aplikasinya) ◽

10.20961/prosidingsnfa.v4i0.35919 ◽

2019 ◽

Vol 4 ◽

pp. 121

Author(s):

A Suparmi ◽

Tuti Dwi Setyaningsih ◽

Suharyana Suharyana ◽

Fuad Anwar ◽

Riyatun Riyatun

Keyword(s):

Reactor Core ◽

Test Facility ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Ramp Test ◽

Fuel Pin ◽

Safe Limit ◽

Reactor Operating ◽

Test Fuel

Abstract: Power Ramp Test Facility (PRTF) is one of the irradiation facility contained in the Multipurpose Reactor GA Siwabessy. This facility is used to test the reactor fuel element pin-type Pressurized Water Reactor. As a result of the entry of foreign bodies cause changes reactor conditions, one of which is expressed with the amount of reactivity to assess the safety of the reactor due to the operation PRTF. PRTF operation simulation and calculation is done using software neutronics MCNP6. Test UO2 fuel enriched assumed at 5% with constant power reactor operating at 15 MW and test fuel pin placed on PRTF within 0, 20, 40, 60, 80, 100, 120, and 140 mm from the centre of the reactor core. Change of reactivity values required in order to secure the reactor, maximal value is 0,5%. The calculation were obtained at each position is (; ; ; ;; ; ; ). Change of reactivity values smaller than the safe limit. Therefore, the study of reactivity changes PRTF operation to test fuel pin is secure.Abstrak: Power Ramp Test Facility (PRTF) merupakan salah satu fasilitas iradiasi yang terdapat pada Reaktor Serba Guna G.A. Siwabessy. Fasilitas ini digunakan untuk menguji pin elemen bahan bakar reaktor tipe Pressurized Water Reactor. Akibat dari masuknya benda asing menyebabkan perubahan kondisi reaktor, salah satunya dinyatakan dengan besaran reaktivitas untuk mengkaji keselamatan reaktor akibat pengoperasian PRTF. Simulasi pengoperasian PRTF dan perhitungan netronik dilakukan menggunakan perangkat lunak MCNP6. Bahan bakar uji UO2 diasumsikan diperkaya sebesar 5% dengan daya operasi reaktor konstan sebesar 15 MW. Pin bahan bakar uji diletakkan pada PRTF berjarak 0, 20, 40, 60, 80, 100, 120, dan 140 mm dari arah pusat teras reaktor. Nilai perubahan reaktivitas yang dipersyaratkan agar reaktor aman adalah , sedangkan nilai perubahan reaktivitas dari penelitian pada masing-masing posisi dari pusat reactor adalah (; ; ; ;; ; ; ) . Nilai perubahan reaktivitas akibat masuknya pin bahan bakar di PRTF mempunyai nilai perubahan reaktivitas 1/10 kali lebih kecil daripada batas aman. Oleh karena itu, ditinjau dari kajian nilai perubahan reaktivitas maka pengoperasian PRTF untuk uji pin bahan bakar adalah aman.

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