• Nuclear Engineering and Technology
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• 제41권6호
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• pp.765-774
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• 2009

For over 30 years, investigations of the thermohydraulic behavior of pressurized-water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany. The PKL facility models the entire primary side and significant parts of the secondary side of a of pressurized water reactor at a height scale of 1:1. Volumes, power ratings and mass flows are scaled with a ratio of 1:145. The experimental facility consists of four primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermohydraulic phenomena. The PKL tests began in the mid 1970s with the support of the German Research Ministry. Since the mid 1980s, the project has also been significantly supported by the German PWR operators. Since 2001, 25 partner organizations from 15 countries have taken part in the PKL investigations with the support and mediation of the OECD/ NEA (Nuclear Energy Agency). After an overview of PKL history and a short description of the facility, this paper focuses on the investigations carried out since the beginning of the international cooperation, and shows, by means of some examples, what insights can be derived from the tests.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.411-425
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• 2017

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose-risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• Nuclear Engineering and Technology
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• 제32권1호
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.108-127
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• 2000

Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facility, SNUF(Seoul National University Facility). This facility simulated YGN 3&4 with volume ratio of 1/1140 based on Ishii's three level scaling. Experiment showed that SI(Safety Injection) water refilled cold leg first and core later. SI water was vaporized in the core, which resulted in the repressurization of reactor. This increase of pressure drove the water in cold leg to flow up half height of U tubes. However, since the water was drained back soon, the release through the SG side broken section by evaporation was negligibly small. This study also provided experimental assessment of RELAP5 results by KAERI for the release through the SG side broken section.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• 방사성폐기물학회지
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• 제6권3호
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• pp.233-244
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• 2008

본 연구에서는 경수로 사용후핵연료로부터 핵연료 물질(예: 차세대형 원자로의 연료)로 재사용할 수 있는 우라늄과 초우라늄원소군(TRU)을 분리, 회수하기 위한 파이로 처리공정(pyroprocess) 시설의 개념설계연구를 수행하였다. 이 시설의 목적은 공학적 실증시험을 통하여 상용 규모의 확대(scale-up) 자료를 확보하는 것과 운전 경험을 쌓을 수 있도록 하자는 것이고 그 용량은 비교적 작은 공학적 규모인 20 kg HM/batch 로 설정하였다. 처리 대상 핵연료로는 경수로의 전형적인 핵연료 형태인 3.5 % 농축우라늄, 35,000 MWd/tU 그리고 5년 냉각시킨 경수로 사용후핵연료를 선택하였다. 본 개념설계연구에서 고려한 주요 항목은 차폐셀을 포함한 파이로 처리공정 시설의 배치, 공정 운전에 대비한 시설 안전 관리, 방사선 안전, 차폐셀 내 불활성 분위기 관리, 연료 물질의 계량 관리, TRU 제품의 핵임계 관리 등이다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.1447-1448
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• 2004

As the plants are classified as a bulk facility by the Agency's safeguards criteria, the Material Balance Evaluation is a good tool to timely detect diversion that will be accomplished through the creation of defects as small as bias defects. Through all evaluations made by the Agency, it Is strongly recommended to report SRD based on both weight and enrichment, maintain the reliable MUF declaration and improve the gamma spectrometry measurement procedure. These recommendations have been now applied and are going on.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1555-1562
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• 2023

Natural circulation phenomena have been nowadays largely revisited aiming to investigate the performances of passive safety systems in carrying-out heat removal under accidental conditions. For this purpose, assessment studies using CFD (Computational Fluid Dynamics) and also 3D thermal-hydraulic system codes are considered at different levels of the design and safety demonstration issues. However, these tools have not being extensively validated for specific natural circulation flow regimes involving flow mixing, temperature stratification, flow recirculation and instabilities. In the present study, an experimental test case based on a small-scale pool test rig experiment performed by Korea Atomic Energy Research Institute, is considered for code-to-code and code-to-experimental data comparison. The test simulation is carried out using the FLUENT and the 3D thermal-hydraulic system CATHARE-2 codes. The objective is to evaluate and compare their prediction capabilities with respect to the test conditions of the experiment. It was observed that, notwithstanding their numerical and modelling differences, similar agreement results are obtained. Nevertheless, additional investigations efforts are still needed for a better representation of the considered phenomena.