• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 1997년도 추계학술대회 논문집
• /
• pp.617-620
• /
• 1997

In this paper, the thermal stress analysis of spent nuclear fuel disposal canister in a deep repository at 500m underground is done for the underground pressure variation. Since the nuclear fuel disposal usually emits much heat and radiation, its careful treatment is required. And so a long term safe repository at a deep bedrock is used. Under this situation, the canister experiences some mechanical external loads such as hydrostatic pressure of underground water, swelling pressure of bentonite buffer, and the thermal load due to the heat generation of spent nuclear fuel in the basket etc.. Hence, the canister should be designed to designed to withstand these loads. In this paper, the thermal stress analysis is done using the finite element analysis code, NISA.

• Nuclear Engineering and Technology
• /
• 제56권5호
• /
• pp.1603-1618
• /
• 2024

Emergency operation plans are indispensable elements for effective process safety management especially when unanticipated events occur under extreme situations. In the paper, a synthesis framework is proposed for the integration success path planning and performance analysis. Within the synthesis framework, success path planning is implemented through flow-directed signal tracing, renaming and reconstruction from a complete collection of Minimal Path Sets (MPSs) that are obtained using graph traversal search on GO-FLOW model diagram. The performance of success paths is then evaluated and prioritized according to the task complexity and probability calculation of MPSs for optimum action plans identification. Finally, an Auxiliary Feed Water System of Pressurized Water Reactor (PWR-AFWS) is taken as an example system to demonstrate the flow-directed MPSs search method for success path planning and performance analysis. It is concluded that the synthesis framework is capable of providing procedural guidance for emergency response and safety management with optimal success path planning under extreme situations.

• 분석과학
• /
• 제16권2호
• /
• pp.110-116
• /
• 2003

사용 후 핵연료의 화학특성 연구를 위하여 요오드의 분리와 정량에 관한 연구를 수행하였다. 사용 후 핵연료를 용해시키는 과정에서 핵연료 중에 CsI로 존재하는 요오드가 $I_2$로 산화되어 휘발되지 않도록 질산과 염산의 혼합산 (80:20 mol%)을 이용하여 비휘발성 ${IO_3}^-$­로 안정화시켰다. 2.5 M $HNO_3$ 매질에서 $NH_2OH{\cdot}HCl$을 이용하여 $I_2$로 환원시킨 후 사염화탄소로 추출하여 우라늄과 핵분열생성물로부터 분리, 회수하였다. 0.1 M $NaHSO_3$을 사용하여 요오드를 역추출하였으며 수용액층으로 회수된 요오드를 이온 크로마토그래피로 정량하였다. 방사성 물질 분석에 적합한 이온 크로마토그래피/차폐 시스템을 구성하였으며 42,000~44,000 MWd/MtU 의 연소도를 갖는 사용후핵연료를 대상으로 요오드를 분석한 결과 Origin 2 연소도 전산코드에 의한 계산결과인 $324.5{\sim}343.6{\mu}g/g$와는 -8.3~-0.5%의 편차를 나타내었다.

• Nuclear Engineering and Technology
• /
• 제54권9호
• /
• pp.3567-3579
• /
• 2022

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

• 한국전산구조공학회논문집
• /
• 제17권4호
• /
• pp.413-421
• /
• 2004

본 논문에서는 깊은 지하 500m에 처분된 가압경수로(PWR) 고준위폐기물 처분용기에 지하수압과 벤토나이트 팽윤압이 가해지는 동안 처분용기에 발생하는 크립변형을 예측하기 위하여 처분용기에 대한 구조해석을 수행하였다. 보통 이러한 크립변형은 처분용기에 추가적인 외력이 작용하지 않더라도 처분용기에 작용하는 압력과 내부의 높은 열에 의하여 발생될 수 있다. 처분용 기내부의 열분포의 복잡성 덴 시간의존성으로 인하여 일단 외부 지하수압 및 팽윤압만 고려하여 크립해석을 수행하였다. 이를 위하여 적당한 크립함수를 사용하였으며, 해석은 1억$(10^8)$초 동안 수행하였다. 해석결과 1억초 동안 발생하는 크립 변형률은 매우 작으며 주희 처분용기의 위아래 덮개에 발생함을 알 수 있었다. 그러나 처분용기의 구조강도에 중요한 내부 주철삽입물에는 훨씬 더 작은 미소한 변형률만 발생하여 처분용기에 발생하는 크립변형은 처분용기의 구조적인 안전성에는 큰 영향을 미치지 않음을 알 수 있었다. 해석 초기에 처분용기 내에 급격히 응력이 증가하여 최고치에 도달한 후 잠깐동안 이 응력 값을 유지하다가 그 이 후에는 급격히 응력 값이 감소하는 응력이완현상을 보이고 있기 때문에 발생 응력 측면에서도 전혀 처분용기의 구조적인 안전성에 문제가 없음이 확인되었다.

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

• 한국전산구조공학회논문집
• /
• 제25권1호
• /
• pp.37-49
• /
• 2012

가압경수로(PWR)에서 배출되는 고준위폐기물을 지하 500m의 화강암 암반의 처분장에 장기간(약 10,000년 동안) 처분하기 위하여 여러 구조적 안전성 평가수행을 통하여 처분용기모델이 개발되었다. 기존에 설계 개발된 가압경수로용 처분용기 모델은 구조적으로 처분용기 내부에 정사각형 단면의 네 개의 고준위폐기물다발이 처분용기 단면의 중심에 대칭되게 나란히 배열된 형태를 취하고 있다. 그러나 이와 같은 배열형태가 최적의 구조인지는 아직 결정할 수 없다. 특히 경량화하는 데에는 여전히 문제가 있다. 이러한 문제를 해결하는 방법은 처분용기 단면 중심에 대하여 대칭으로 배열된 네 개의 고준위폐기물다발의 단면형상을 변경시키는 것이다. 단면형상을 변경시키는 방법에는 정사각형 형상을 유지시키면서 단면을 회전시키는 방법과 정사각형 형상을 다른 단면형상으로 변경시키는 두 가지 방법이 있다. 기 수행된 연구를 통하여 정사각형 단면형상을 유지시키면서 단면을 회전시키면 회전각도가 $30{\sim}35^{\circ}$인 배열구조의 처분용기가 나란한 정사각형 배열구조보다 구조적으로 더 안정적이어서 경량화할 수 있음을 알 수 있었다. 그러나 이 회전한 배열구조의 처분용기가 최적인지는 역시 아직 결정할 수 없다. 왜냐하면 정사각형이 아닌 다른 단면형상의 구조물에 대해서는 아직 구조적으로 더 안정한지가 확인되지 않았기 때문이다. 따라서 처분용기 단면 중심면에 대하여 대칭성을 유지하면서 고준위폐기물다발의 단면형상이 정사각형이 아닌 다른 단면형상의 처분용기구조에 대한 구조해석이 필요하다. 본 연구에서는 네 개의 고준위폐기물다발이 처분용기 중심 면에 대하여 대칭적으로 배열되면서 단면형상이 여러 가지로 변화된 가압경수로용 처분용기에 대하여 구조해석을 수행하였다. 구조해석을 수행한 결과 기존의 설계 개발된 처분용기 단면의 중심에 대칭되게 나란히 고준위폐기물다발이 배열된 정사각형 단면의 처분용기보다 다발의 단면형상이 원형인 처분용기가 구조적으로 좀 더 안정성이 있음이 밝혀졌다.

• 분석과학
• /
• 제17권5호
• /
• pp.438-442
• /
• 2004

To determine the contents of $^{99}Tc$ in the spent PWR (pressurized water reactor) nuclear fuels by ICP-MS (inductively coupled plasma-mass spectrometry), a technetium separation method using an extraction chromatographic resin (TEVA Spec resin) has been established. $^{99}Tc$ was separated from a spent PWR nuclear fuel solution by this separation procedure and its concentration was determined by ICP-MS. The result agrees well with the value calculated by the program ORIGEN 2 and also the value measured by AG MP-1 resin/ICP-MS method described in our previous paper. It can be concluded that the present separation procedure is superior to the AG MP-1 resin procedure with respect to the time required for technetium separation as well as the efficiency of decontamination from other radioactive nuclides.

• 대한용접접합학회:학술대회논문집
• /
• 대한용접접합학회 2005년도 추계학술발표대회 개요집
• /
• pp.70-72
• /
• 2005

Nuclear fuel assemblies for pressurized water reactors(PWR) are loaded in the reactor core throughout the residence time of three to five years. A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, laser welding qualities of the Zircaloy spacer grid assembly welded by two welding companies, such as weld strength, weld penetration depth, and weld bead size, are examined and compared.

• 대한금속재료학회지
• /
• 제50권8호
• /
• pp.583-589
• /
• 2012

This work is concerned with a statistical approach to the life prediction on irradiation-assisted stress corrosion cracking (IASCC) of stainless steel (SS) for core internals of a pressurized water reactor (PWR). The previous results of the time-to-failure of IASCC measured on neutron-irradiated stainless steel components were statistically analyzed in terms of stress and irradiation. The accelerating life testing model of IASCC of cold worked Type 316 SS was established based on an inverse power model with two stress-variables, the applied stress and irradiation dose. Considering the variation of the yield strength and applied stress with the irradiation dose in the model, the remaining life of the baffle former bolt was statistically predicted during operation under complex environments of stress and irradiation.