ROSA/AP600 Testing: Facility Modifications and Initial Test Results.

Kukita, Yutaka; Yonomoto, Taisuke; Asaka, Hideaki; Nakamura, Hideo; Kumamaru, Hiroshige; Anoda, Yoshinari; Boucher, T.J.; Ortiz, M.G.; Shaw, R.A.; Schultz, Richard R.

doi:10.3327/jnst.33.259

Cited by 8 publications

(4 citation statements)

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“…This setpoint will actuate the opening of ADS stage 1, followed by stages 2 and 3 sequentially. Those ADS actuations result in a decrease of pressure at the pressurizer steam space and cause the coolant in the upper plenum to flow into the pressurizer [19]. This event is shown as an increase of the pressurizer's two-phase water level [20] until the actuation of the ADS stage 4 to drain the pressurizer water once more again.…”

Section: Transient Simulationmentioning

confidence: 99%

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Ekariansyah

Widodo

2016

Atom Indo.

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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.

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Section: Transient Simulationmentioning

confidence: 99%

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Ekariansyah

Widodo

2016

Atom Indo.

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“…As a conservative assumption to the core cooling, the ACC system injected coolant into a cold leg in the loop with pressurizer only. The low-pressure injection (LPI) system of the ECCS injected coolant into vessel downcomer through direct vessel injection lines [13] in both loops. The two LSTF tests were analyzed by RELAP5/MOD3.2.1.2 code [14] to clarify the remaining subjects.…”

Section: Introductionmentioning

confidence: 99%

ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

Takeda

Ohnuki

Kanamori

et al. 2016

Science and Technology of Nuclear Installations

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Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

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“…The second integral facility APEX emphasized the depressurization, IRWST injection, and long-term cooling during small break LOCA [9]. Besides, US NRC conducted independent integral test programs at ROSA in Tokai-mura, Japan, focusing on the high pressure and depressurization phase, the initiation of IRWST injection [10], and the passive residual heat removal [11].…”

Section: Introductionmentioning

confidence: 99%

Research and Evaluation for Passive Safety System in Low Pressure Reactor

Chuanxin¹,

Wenbin²,

Bingde³

et al. 2015

Science and Technology of Nuclear Installations

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Low pressure reactor is a small size advanced reactor with power of 180 MWt, which is under development at Nuclear Power Institute of China. In order to assess the ability and feasibility of passive safety system, several tests have been implemented on the passive safety system (PSS) test facility. During the LOCA and SBO accident, the adequate core cooling is provided by the performance of passive safety system. In addition the best-estimate thermal hydraulic code, CATHARE V2.1, has been assessed against cold leg LOCA test. The calculation results show that CATHARE is in a satisfactory agreement with the test for the steady state and transient test.

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ROSA/AP600 Testing: Facility Modifications and Initial Test Results.

Cited by 8 publications

References 0 publications

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

Research and Evaluation for Passive Safety System in Low Pressure Reactor

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