• 전기화학회지
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• 제4권3호
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• pp.87-93
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• 2001

외부 Ag/AgCl 기준 전극은 가압형 및 비등형 경수로 환경에 널리 사용되었다. 전극의 채움 용액 (Siting solution)으로 통상 KCl을 사용하는데, 다공성 지르코니아로 만들어지는 플러그를 통한 Cl 이온의 누설이 전극의 전위차 변동을 유발하는 문제가 있다. 누설로 인한 전위차 변동의 문제를 해결하기 위해 채움 용액으로 순수를 사용하였다 순수를 사용하는 경우 상온에서의 AgCl용해도에 의해 Cl이온의 농도가 결정된다. 붕산과 수산화리튬 혼합용액으로 $288^{\circ}C$에서 전극의 안정성 실험을 실시하였다. 약 일주일간 전위차 변화는 10mV 이내였으며, $288^{\circ}C$와 $240^{\circ}C$에서의 온도 사이클링 시험 전후의 전위차 변화는 15mV 이내였다. 이온의 limiting equivalent conductances와 Agar의 수역학적 이론을 토대로 하여 전극의 TLJP을 계산하였다. 전극 채움 용액 내의 Cl이온 농도를 상온에서 측정한 값으로 보정하여 이론값을 계산할 경우, 실험값과 비교적 잘 일치하는 것을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Nuclear Engineering and Technology
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• 제16권1호
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• pp.36-46
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• 1984

가압경수로의 정상운전상태는 물론 예상 과도상태에서도 노심내에서 DNB가 발생하지 않아야 된다는 설계근거를 만족시키는 새로운 설계방법 즉, 통계처리에 의한 열설계 방법이 개발되어 이에 대하여 검토하였다. 이같은 설계방법을 사용하여 설계변수에 대한 불확실도를 통계적으로 처리함으로써 노심설계에 따른 설계여유도를 정량적으로 계산할 수 있어 원자로심의 안전성을 충분히 유지하면서도 DNB비례산에 따른 불필요한 보수성을 배제할 수 있다. 본 기술검토보고서는 미국의 Westing-house와 B & W원자로 제작회사가 개발한 통계적 열설계방법을 소개하고 본 설계방법의 특성을 설명하며 이어서 불확실도의 통계처리 과정, DNB설계 제한치 설정방법, 그리고 본 방법의 응용 결과를 비교하여 보여준다. 본 검토를 통하여 두 회사의 설계방법은 근본적으로 유사하나 통계처리를 위한 설계변수의 선택과 이들 불확실도의 처리방법이 다소 상이하다는 것을 알았으며 또한 본 방법의 사용으로 노심설계에 있어서 설계여유도가 현저히 증가한다는 것을 알았다.

• Nuclear Engineering and Technology
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• 제17권4호
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• pp.247-255
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• 1985

본 논문은 가변구조모델추종제어(VSMFC)계 설계의 새로운 방법을 고찰한 것이다. 설계 개념은 가변구조계(VSS)와 슬라이드모드 이론을 사용하여 비선형 시변다변수계가 파라미터 변동이 있을지라도 모델추종을 정확히 하게끔 제어측이 가변구조를 갖게 하는 것이다. 본 논문의 방법을 실제 물리계에 적용할 때 컴퓨터 계산시간의 감소와 파라미터변동에 무관한 동적응답을 기대할 수 있다. 이론의 유효성을 밝히기 위해 VSMPC를 1000MWe의 불등경수형 원자로(BWE)에 적용하였다. 즉 원자로의 출력요구가 정격출력의 85∼90% 범위에서 변할 때 부하추종출력제어가 원활히 이루어지는가를 컴퓨터 시뮬레이션하였다. 12개의 비선형미분방정식으로 동특성이 주어지는 원자로에서 6차계 선형모델을 85% 정격치에서 구하고 여러범위에 걸쳐서 부하변동이 있을 때 파라미터변동을 극복하면서도 출력제어를 원활히 하는가를 연구하였다.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.426-433
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• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• Corrosion Science and Technology
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• 제19권1호
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• pp.1-7
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• 2020

Chalk River Unidentified Deposits (CRUDs) were collected from the Korean pressurized water reactor (PWR) plant (A, B, and C) where the axial offset anomaly (AOA) occurred. AOA, also known as a CRUD-induced power shift, is one of the key issues in maintaining stable PWR plant operations. CRUDs were sampled from spent nuclear fuel rods and analyzed using an electron probe micro-analyzer (EPMA). This paper describes the characteristics of boron-deposits from the CRUDs sampled from twice-burnt assemblies from the Korean PWR. The primary coolant of a PWR contains boron and lithium. It is known that boron deposition occurs in a thick CRUD layer under substantial sub-cooled nucleate boiling (SNB). The results of this study are summarized as follows. Boron was not found at the locations where the existence was confirmed in simulated CRUDs, in other words, the cladding and CRUD boundaries. Nevertheless, we clearly observed the presence of boron and confirmed that boron existed as a lump in crystalline form. In addition, the study confirmed that CRUD existed in a crystal form with a unique size of about 10 ㎛.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2094-2099
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• 2004

An experimental study on heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with the increase of the system pressure For a fixed inlet mass flux and subcooling, the CHF falls sharply at about 3.8 MPa and shows a trend toward converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall because the CHF occurred at remarkably low power levels. In the pressure reduction transient experiments, as soon as the pressure passed through the critical pressure, the wall temperatures rise rapidly up to a very high value due to the occurrence of the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, then tends to decrease gradually.

• Nuclear Engineering and Technology
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• 제36권5호
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• pp.403-414
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• 2004

An experimental study of heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) tests, and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with increase of the system pressure for fixed inlet mass flux and subcooling. The CHF falls sharply at about 3.8 MPa and shows a trend towards converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall, because the CHF occurs at remarkably low power levels. In the pressure reduction transients, as soon as the pressure passes below the critical pressure from the supercritical pressure, the wall temperatures rise rapidly up to very high values due to the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, and then tends to decrease gradually.