The development of passive design features for the Korean Next Generation Reactor

Cho, Sung Jae; Kim, Byong Sup; Kang, Myung Gi; Kim, Han Gon

doi:10.1016/s0029-5493(00)00257-0

Cited by 25 publications

(4 citation statements)

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“…The PCCS of a KNGR [6] ( Figure 3) is rather similar to the passive containment cooling system described for the SBWR ICs. It is provided with two heat exchangers, relevant lines and water tanks, and guarantees a way of external containment cooling through a natural circulation circuit.…”

Section: Review Of Passive Safety Systems For Decay Heat Removalmentioning

confidence: 99%

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Papini

Cammi

2010

Science and Technology of Nuclear Installations

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Decay Heat Removal (DHR) is a fundamental safety function which is often accomplished in the advanced LWRs relying on natural phenomena. A typical passive DHR system is the two-phase flow, natural circulation, closed loop system, where heat is removed by means of a steam generator or heat exchanger, a condenser, and a pool. Different condenser tube arrangements have been developed for applications to the next generation NPPs. The two most used configurations, namely, horizontal and vertical tube condensers, are thoroughly investigated in this paper. Several thermal-hydraulic features were explored, being the analysis mainly devoted to the description of the best-estimate correlations and models for heat transfer coefficient prediction. In spite of a more critical behaviour concerning thermal expansion issues, vertical tube condensers offer remarkably better thermal-hydraulic performances. An experimental validation of the vertical tube correlations is provided by PERSEO facility (SIET labs, Piacenza), showing a fairly good agreement.

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Section: Review Of Passive Safety Systems For Decay Heat Removalmentioning

confidence: 99%

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Papini

Cammi

2010

Science and Technology of Nuclear Installations

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“…The modeling and analyses results using the MAAP severe accident analysis code are provided in Sections 3 and 4, and a summary of the SBO coping capabilities in APR1400 are discussed in Section 5. [10,11] and these documents were modified [12,13] to reflect the APR1400 plant specific design characteristics [14]. When a SBO occurs, the reactor is tripped by either Reactor Protection System (RPS) trip signal or gravity drop of control rod due to the loss of electric power.…”

Section: Introductionmentioning

confidence: 99%

Extended Station Blackout Coping Capabilities of APR1400

Lee

Hong

Seo

et al. 2014

Science and Technology of Nuclear Installations

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The Fukushima Dai-ichi nuclear power plant accident shows that an extreme natural disaster can prevent the proper restoration of electric power for several days, so-called extended SBO. In Korea, the government and industry performed comprehensive special safety inspections on all domestic nuclear power plants against beyond design bases external events. One of the safety improvement action items related to the extended SBO is installation of external water injection provision and equipment to RCS and SG. In this paper, the extended SBO coping capability of APR1400 is examined using MAAP4 to assess the effectiveness of the external water injection strategy. Results show that an external injection into SG is applicable to mitigate an extended SBO scenario. However, an external injection into RCS is only effective when RCS depressurization capacity is sufficiently provided in case of high pressure scenarios. Based on the above results, the technical basis of external injection strategy will be reflected on development of revised severe accident management guideline.

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“…Some advanced pressurized water reactors such as the APR1400 [1] have adopted the new safety feature of a DVI (Direct Vessel Injection) system that supplies the ECC (Emergency Core Coolant) directly into the reactor vessel down-comer. The ECC is supplied from the safety injection system through DVI nozzles that are located in the upper part of the reactor vessel.…”

Section: Introductionmentioning

confidence: 99%

Major Thermal-Hydraulic Phenomena Found During Atlas Lbloca Reflood Tests for an Advanced Pressurized Water Reactor Apr1400

Park¹,

Choi²,

Cho³

et al. 2011

Nuclear Engineering and Technology

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The development of passive design features for the Korean Next Generation Reactor

Cited by 25 publications

References 0 publications

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Extended Station Blackout Coping Capabilities of APR1400

Major Thermal-Hydraulic Phenomena Found During Atlas Lbloca Reflood Tests for an Advanced Pressurized Water Reactor Apr1400

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