• Nuclear Engineering and Technology
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• 제15권4호
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• pp.248-255
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• 1983

하나의 PWR 핵연료 집합체를 수송할 수 있는 사용후 핵연료 수송용기에 대한 열해석을 수행하였다. 정상 및 화재사고 조건하에서 수송용기에 대한 온도분포는 10CFR Part 71에서 제시한 조건에 맞도록 계산하였다. 붕괴열은 연소도가 45,000 MWD/MTU이고 사용후 핵연료 저장실에서 300일 냉각기간을 가질 KNU 5&6 핵연료 집합체를 고려하였다. 계산결과 화재사고시 dry cavity조건하에서 핵연료 피복관의 최대온도가 455$^{\circ}C$로 계산되었으며, 이 간은 10CFR Part 50.46에 규정된 최대 피복관 제한치 보다 훨씬 낮게 나타났다. 이것은 수송용기의 운반중에 화재사고 조건하에서도 핵연료 피복관의 파손이 일어나지 않는 것으로 설명된다. 그리고 중요 차폐체인 납의 용융도 일어나지 않았다.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.54-64
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• 1988

FLHT-2 실험 결과를 사용하여 원자력발전소의 중대사고발생시 노심의 거동을 해석하기 위한 전산코드인 SCDAP코드를 평가하였다. 계산결과에 의하면 코드는 실험시 측정된 노심의 온도경향, 총수소발생량 및 순간최대수소발생율, 그리고 연료봉내압과 피복재파열시간을 잘 예측하는 것으로 평가되었다. 그러나 이상유체높이와 복사열전달 및 zircaloy의 급격한 산화 시작 온도에 대한 모델은 수정되어야 할 것으로 평가되었다. 또한 핵 연료봉에서의 gap을 고려하여주는 것은 노심손상현상의 정확한 예측에 커다란 도움을 줄 수 있다는 것이 밝혀졌다.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1525-1531
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• 2019

The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.

• 수산해양기술연구
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• 제28권1호
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• pp.71-78
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• 1992

Zr-4 used for a cladding and an end plug of reactor component has creep deformation under operation at high temperature. Creep is regarded as the time dependent deformation of a material under constant applied stress. Although the major source of the deformation of zirconium component in water-cooled reactors is irradiation creep, the thermal creep may give a rise to significant deformation in reactor component especially at relatively high temperatures and at various constant stresses, and therefore it must be predicted accurately. Stress relaxation is the time dependent change of stress at constant strain and it is a process related intimately to creep. In this paper, the creep behavior and stress relaxation of Zr-4 is examined at the temperature of 50$0^{\circ}C$ that is 40% of the absolute melting temperature of Zr-4 under the stress below yield stress and under the various constant strains. The results obtained are summarized as follows: 1) With an increase of stress, the steady state creep rate increases and the creep rupture time decreases. 2) The steady state creep rate $\varepsilon$(%/s) for the stress $\sigma$sub(c) (kgf/mm super(2)) of Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 are in accord with Norton's model equation($\varepsilon$=K$\sigma$ sub(c) super (n)). The constants of materials computed are as follows: K=3.9881$\times$10 super(-5), n=1.9608 3) The rupture time T sub(r) (hr) decreases linearly with the increase of stress on the log-log scaled graph. The empirical equations computed for Zr-4 are in accord with Bailey's model equation (T sub(r)=K sub(1)$\sigma$sub(c) super(m)). The constants of materials computed are as follows: K sub(1)=1.2875$\times$10 super(16), m=-3.467 4) It seems clear that the strain could be quantitatively dependent on the high temperature creep properties such as creep stress, rupture time, steady state creep rate and total creep rate. It is found that these relationships are linear on the log-log graph. 5) In stress relaxation test, as the critical constant strain that can be allowed to the specimen is larger, stress relaxation becomes more rapid, and as the constant strain is smaller, the stress relaxation becomes slower.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• 제17권3호
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• pp.211-215
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• 1985

온도 및 연신율 변촤가 Zircaloy-4의 요드 응력부식 거동에 미치는 영향을 30$0^{\circ}C$에서 일정 하중법과 300, 350, 40$0^{\circ}C$에서 일정 연신율법으로 ($10^{-5}$sec~$10^{-6}$ sec) 3.34mg $I_2$/㎤의 요드분위기에서 연구하였다. 요드 응력부식균열에 대한 저항성은 온도가 상승하거나 변형속도가 감소하면 감소했고 파손 시간과 응력과의 관계는1/tｆ∝exp (0.3$\sigma$/$\sigma$uTs-31.5)로 표시할 수 있었다. 30$0^{\circ}C$에서 요드 응력 부식 균열에 대한 저항성을 불활성 분위기에서의 파손에너지에 대한 요드분위기에서의 파손 에너지의 비율로 표시할 때 변형속도가 7.6$\times$$10^{-6}$ sec 부근에서 저항성이 가장 낮았고 온도가 35$0^{\circ}C$, 40$0^{\circ}C$ 로 증가함에 따라 보다 높은 변형속도에서 최저 저항성을 나타내는 경향을 보였다. 요드 응력부식 균열의 파단면에서 준-벽계 파면(quasi-cleavage fracture)을 관찰했다. 전술한 결과에 의하면 Zircaloy-4의 요드 응력부식균열의 기구에 있어서 보호 피막파손단계 (film rupture step)가 중요한 과정으로 추정된다.