• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1999년도 추계학술발표회요약집
• /
• pp.373.2-373
• /
• 1999

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
• /
• pp.90-91
• /
• 2018

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2016년도 추계학술논문요약집
• /
• pp.111-112
• /
• 2016

• Journal of Radiation Protection and Research
• /
• 제9권2호
• /
• pp.90-96
• /
• 1984

장주기핵연료(長週期核燃料) 노심기법(爐心技法)에 의한 사용후핵연료(使用後核燃料)가 기존(旣存) 사용후핵연료저장시설(使用後核燃料貯藏施設)의 설계변경(設計變更)없이 동(同) 시설(施設)에 수용(受容) 가능(可能)한지를 결정(決定)하기 위하여 저장시설(貯藏施設)에서의 예상(豫想) 방사선피폭선량률(放射線被曝線量率) DLC-23/CASK (22n, 18g) 단면적(斷面積)자료(資料)와 ANISN-W 전산(電算)코드로 계산(計算)하여 설계기준치(設計基準値)와 비교(比較) 검토(檢討)하였다. 사용후핵연료내(使用後核燃料內)의 방사능량(放射能量) 및 감마선스펙트럼은 핵연료교체(核燃料交替)모델에 따라 ORIGEN 전산(電算)코드로 계산(計算)하였다. 방사선량률(放射線量率)의 계산(計算)에 있어서 저장조(貯藏槽)의 기하학적(幾何學的) 모델은 무한평판모형(無限平板模型)이며 저장(貯藏)된 사용후핵연료(使用後核燃料)의 구성물질(構成物質)과 방사선원(放射線源)은 핵연료집합체내(核燃料集合體內)에 균일(均一)하게 분포(分布)되었다고 가정(假定)하였다. 사용후핵연료저장조(使用後核燃料貯藏槽)에 저장(貯藏)된 핵연료집합체(核燃料集合體) 및 저장용수중(貯藏用水中) 방사성핵종(放射性核種)에 의한 방사선량률(放射線量率)의 계산(計算) 결과(結果)는 정상(正常) 및 사고수면시(事故水面時) 계산(計算)된 방사선량률(放射線量率)이 설계기준치(設計基準値)를 만족(滿足)시켜주는 것으로 나타났다.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.627-638
• /
• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• 한국전자통신학회논문지
• /
• 제14권3호
• /
• pp.505-512
• /
• 2019

본 논문에서는 원자력발전소의 전력이 상실되는 사고가 발생하더라도 사용후핵연료 저장조의 수위를 감시할 수 있는 방법을 제시하고, 이의 성능 및 한계점을 분석하였다. 제안한 방식은 광분배기를 통과한 광신호가 저장조의 냉각수 및 공기에서 반사되는 반사광의 세기를 측정하는 방식이다. $1{\times}8$ 광분배기를 사용하는 경우에는 저장조의 수위를 측정하는데 문제가 없을 것으로 분석되나, $1{\times}16$ 광분배기를 사용하는 경우에는 Rayleigh 역산란으로 인해 16개의 출력포트 중 15개만이 수위측정 용도로 사용이 가능하고, $1{\times}32$ 광분배기를 사용하는 경우에는 32개의 출력포트 중 25개만이 수위측정 용도로 사용이 가능한 것으로 분석되었다.

• 전기학회논문지
• /
• 제67권2호
• /
• pp.337-342
• /
• 2018

Spent fuel is moved from the reactor into the spent fuel pool when nuclear power plant permanently shutdown. The sole function of a permanently defueled facility is to store spent fuel in a quiescent state. The function of electric system and loads are reduced. It is necessary to establish an operating strategy of electric system in the permanent shutdown nuclear plant. This paper reviews required loads and design criteria considering transition to permanent shutdown. An operating strategy of onsite electric system is proposed considering decommissioning strategy and stage of defueled condition.

• Nuclear Engineering and Technology
• /
• 제49권2호
• /
• pp.442-449
• /
• 2017

Risk management of the Fukushima Daiichi Nuclear Power Station decommissioning is a great challenge. In the present study, a risk management framework has been developed for the decommissioning work. It is applied to fuel assembly retrieval from Unit 3 spent fuel pool. Whole retrieval work is divided into three phases: preparation, retrieval, and transportation and storage. First of all, the end point has been established and the success path has been developed. Then, possible threats, which are internal/external and technical/societal/management, are identified and selected. "What can go wrong?" is a question about the failure scenario. The likelihoods and consequences for each scenario are roughly estimated. The whole decommissioning project will continue for several decades, i.e., long-term perspective is important. What should be emphasized is that we do not always have enough knowledge and experience of this kind. It is expected that the decommissioning can make steady and good progress in support of the proposed risk management framework. Thus, risk assessment and management are required, and the process needs to be updated in accordance with the most recent information and knowledge on the decommissioning works.

• 방사성폐기물학회지
• /
• 제11권1호
• /
• pp.31-39
• /
• 2013

핵연료는 원자로 운전 중 예기치 못한 상황에서 연료 결함을 초래할 수 있다. 핵연료 결함은 연료봉의 수소화나 이물질에 의한 금속 마모, 그리고 펠렛과 피복관의 상호작용에 의해 피복관이 손상된다. 이렇게 손상된 핵연료의 결함원인을 규명하는 것은 원자력발전의 안전운전에 중요하다고 사료된다. 핵연료가 손상되면 원자로 냉각재가 오염되어 원자로 출력을 낮추거나, 발전소를 정지할 수도 있다. 모든 사용후연료는 건식저장고로 이동 보관되어야 하나, 결함연료는 이동할 수 없으므로 이 연구의 목적은 중수로형 원자로에서 연료가 인출된 후 사용후연료 저장조에서 보관된 연료에 대하여 결함 여부를 판단할 수 있는 기술을 개발하고자 하였다. 이 연구를 통하여 핵종 누설 검출 기술을 이용한 사용후연료 검사기술을 개발하였으며, 이 기술을 월성발전소에 적용함으로써, 검사기술 및 검사시스템에 대한 성능을 입증하였다.

• 한국압력기기공학회 논문집
• /
• 제18권2호
• /
• pp.43-49
• /
• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.