• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4042-4051
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• 2021

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.429-436
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• 1995

This paper describes a methodology for determination of representative CANDU feeder geometry and the pressure drops between inlet/outlet header and fuel channel in the primary loop. A code, MEDOC, was developed based on this methodology and helps perform a calculation of equivalent feeder geometry for a selected channel group on the basis of feeder geometry data (fluid volume, mass flow rate, loss factor) and given property data pressure, quality, density) at inlet/outlet header. The equivalent feeder geometry calculated based on this methodology will be useful fur the transient thermohydraulic analysis of the primary heat transport system for the CANDU heavy water-cooled pressure tube reactor.

• Nuclear Engineering and Technology
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• 제27권2호
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• pp.226-233
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• 1995

CANDU 원자로에서 심각하게 대두되는 압력관 파손을 방지하기 위해 압력관의 두께를 증가시키는 방안이 연구되었다. 본 연구에서는 압력관 두께변화가 Zr-2.5Nb 압력관의 응력, 수소농도 및 수소지연파괴에 미치는 영향에 대해 연구를 수행하였다. 압력관 두께가 현재의 4.2 mm에서 5.2 mm로 증가할 경우에 압력관이 받는 응력과 발전소 가동중에 누적되는 중수소 흡수량은 19% 줄어드는 것으로 나타났으며, 압력관에 균열이 발생할 경우 발전소 냉각동안에 일어나는 균열 성장은 상당히 감소한다. 수소지연파괴는 압력관이 받는 응력과 누적되는 수소량에 비해 지배되는데 이와 같은 결과로부터 두꺼운 압력관은 수소지연파괴 관점에서 상당한 이점이 있는 것으로 평가되었다. 그러나 압력관 두께 증가는 수소지연파괴의 성장속도를 가속할수도 있으므로 앞으로 연구할 사항이다.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.901-910
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• 2021

KHNP recently has obtained the approval for the commercialization of the modified 37-element (or 37 M) fuel bundle and now is loading the 37 M fuel bundles in CANDU-6 reactors in KOREA. One of the main issues for approval was the burnup limit. Due to CANDU design and neutronic characteristics, there was no specific burnup restriction of a fuel bundle. The absence of a burnup limit does not mean that a fuel bundle can stay in the CANDU reactor without a time limit. However, the regulator requested traditional design values as well as the burnup limit reflecting the computer code uncertainty. The method for the PWR burnup limit was not applied to the CANDU fuel bundle. Since there was no approved methodology to build the burnup limit with uncertainties, KHNP introduced a Monte-Carlo method coupled with a 95/95 approach to determine the conservative burnup limit from the viewpoint of the centerline temperature, internal pressure, strain measurement deviation. Moreover, to consider the uncertainties of various computing models, a converted power uncertainty was introduced. This paper presents the methodology and puts forward the limits on burnup, evaluated for each of the existing and modified fuel bundles, in consideration of the pressure tube aging condition.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 추계학술대회논문집
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• pp.345-349
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• 2003

The dynamic behavior of CANDU nuclear fuel channel was analyzed by the use of 3-dimensional finite element method, under the various fault conditions such as a fault in the end fitting support and the removal/migration of the garter spring in the fuel channel, in order to predict the dynamic behavior for a degraded symptoms of CANDU nuclear fuel channel. Moreover, the frequency response analysis for possible fault conditions was also peformed considering the effects of the pressure tube vibration and flow-induced vibration by the coolant flow. From the analysis of the frequency responses, defects in the garter spring have influenced the changes of 2nd and 3rd modes and all the important modes are varied for the failure in the journal bearing in the end fitting body.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.252.2-252.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.315.2-315
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.319-319
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.395.2-395
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.207-207
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• 1999