• Nuclear Engineering and Technology
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• 제42권4호
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• pp.434-441
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• 2010

The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.303.2-303.2
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• 2003

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1109-1116
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• 2019

It is essential to predict the radioactivity distribution around the reactor cooling system (RCS) during obligatory cyclic operation of AP-1000. A home-developed program CPA-AP1000 is upgraded to predict the response of activated corrosion products (ACPs) in the RCS. The program is written in MATLAB and it uses state of the art MCNP as a subroutine for flux calculations. A pair of cyclic power profiles were superimposed after initial full power operation. The effect of cyclic operation is noticed to be more prominent for in-core surfaces, followed by the primary coolant and out-of-core structures. The results have shown that specific activity trends of $^{56}Mn$ and $^{24}Na$ promptly follow the power variations, whereas, $^{59}Fe$, $^{58}Co$, $^{99}Mo$ and $^{60}Co$ exhibit a sluggish power-following response. The investigations pointed out that promptly power-following response of ACPs in the coolant is vital as an instant radioactivity source during leakage incidents. However, the ACPs with delayed power-following response in the out-of-core components are perceived to cause a long-term activity. The present results are found in good agreement with those for a reference PWR. The results are useful for source term monitoring and optimization of work procedures for an innovative reactor design.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1974-1982
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• 2020

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 학술논문집
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• pp.397-398
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• 2004

It is important to estimate the radiation activity of the nuclear fuel which is a source term of the loss of coolant accident. The purpose of this study is to identify the most important parameters of the source term calculation based on three fuel types: typical natural uranium CANDU fuel, slightly enriched uranium and DUPIC fuel. The characteristics of the radiation source term were analyzed through sensitivity calculations of the linear power, fuel turnup, and the power shape.(omitted)

• Journal of Radiation Protection and Research
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• 제39권1호
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• pp.14-20
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• 2014

원전 내 방사선작업종사자 피폭량의 대부분은 계획예방정비기간 중 냉각재계통에 존재하는 $^{58}Co$, $^{60}Co$등과 같은 CRUD에 의하여 일어난다. 따라서 원전 내 방사선작업종사자의 피폭 최적관리를 위해서는 냉각재계통의 선원항을 사전에 파악할 필요가 있다. 이 연구는 원전 내 선원항을 알아보기 위해 국내 최초로 계획예방정비 기간 중 증기발생기 부근에서 CZT 반도체 검출기를 이용한 배관 직접 측정법을 사용하였다. 또한 신규원전과 노후원전에서 선원항의 차이를 알아보기 위해 두 원전에서 측정한 결과를 비교 하였고 노후원전에 대하여는 정지화학처리에 따른 선원항의 변화를 측정하였다. 노후원전에서 정지화학처리에 따른 선원항 변화는 발견되지 않았으며, 신규원전 및 노후원전의 주요 선원항은 $^{58}Co$와 $^{60}Co$ 였고, $^{59}Fe$는 신규원전에서만 $^{137}Cs$와 $^{95}Zr$는 노후원전에서만 보였다. $^{58}Co/^{60}Co$의 비율은 노후원전보다 신규원전에서 크게 나타났으며 운전연한이 증가 할수록 반감기가 긴 $^{60}Co$의 비방사능이 커지기 때문이다.

• Journal of Radiation Protection and Research
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• 제20권1호
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• pp.53-64
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• 1995

월성 원자력발전소 2,3,4 호기의 보조건물 주요 지역에서 냉각재 다량상h7사고 (large LOCA) 후의 방사선장을 평가하였다. 핵분열 생성물의 총량은 ORIGEN2 코드를 사용하여 계산하였고 선원항은 2중고장 시나리오, 즉 LOCA 사고후 비상노심냉각 (ECC) 계통의 고장이 결부된 사고시의 방사능 방출에 근거하였다. 원자로건물, 보조건물 및 ECC 계통의 구조모형을 QAD-CG 모델에 포함하여 계산하였다. 사고시점부터 90일 경과시까지 시간대 별로 선량율과 누적선량을 계산하였다. 결과적으로, 연속출입이 요구되는 중요지역에서의 방사선장은 충분히 낮은 것으로 평가되었다. 그러나, 일부구역에서는 제한적인 출입을 허용할 정도로 상대적으로 높은 방사선장을 나타내었다.