[논문]설계수명 이후 해체를 위한 금속 겸용용기의 방사화 특성 평가

김태만; 구지영; 도호석; 조천형; 고재훈

doi:10.7733/jnfcwt.2016.14.4.343

설계수명 이후 해체를 위한 금속 겸용용기의 방사화 특성 평가
Activation Analysis of Dual-purpose Metal Cask After the End of Design Lifetime for Decommission 원문보기

Journal of nuclear fuel cycle and waste technology = 방사성폐기물학회지, v.14 no.4, 2016년, pp.343 - 356

김태만 (한국원자력환경공단) , 구지영 (한국원자력환경공단) , 도호석 (한국원자력환경공단) , 조천형 (한국원자력환경공단) , 고재훈 ((주)코네스코퍼레이션)

초록
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한국원자력환경공단에서는 국내 경수로 원전에서 발생한 사용후핵연료를 건식으로 저장하기 위하여 안전성을 최우선으로 국내/외 기술기준을 준수하여 금속겸용용기를 개발하였다. 이러한 금속용기는 50년 동안 주요 안전성요소(구조, 열제거, 격납, 임계방지, 방사선차폐 등)에 대한 건전성을 유지하고, 운영기간 중 유지보수 과정에 폐기물의 발생을 최소화 하고 이를 안전하게 관리할 수 있도록 설계하였다. 본 논문은 설계수명이 종료된 금속용기 본체 및 내/외부 구조물에 대한 방사화 평가를 통해 정량적인 방사능 재고량에 대한 정보를 제공한다. 본 논문에서는 금속용기 본체 및 구성품의 방사화 방사능 재고량은 MCNP5 ORIGEN-2 평가체계를 이용하여 계산하였으며, 각 구성품의 화학조성, 중성자속 분포, 반응률 및 저장기간 동안 중성자조사 기간을 반영하여 평가하였다. 평가결과, 설계수명 이후 10년 경과시 모든 금속재질에서 $^{60}Co$의 방사능이 기타 핵종들에 비하여 가장 큰 방사능을 띄는 것으로 나타났으며, 중성자차폐체인 수지에서는 수명직후 $^{28}Al$ 및 $^{24}Na$등의 고에너지 감마선을 방출하는 핵종은 반감기가 짧아 0.5년 이후에는 무시할 수 있는 수준으로 나타났다. 또한, 사용후핵연료 제거후 캐니스터 및 금속용기 본체에 대한 표면 선량률 평가결과, 상당히 낮은 값을 나타내어, 해체 시 작업자가 받는 피폭선량은 무시할 수 있는 수준으로 평가되었다. 본 평가방법은 사용후핵연료 금속겸용용기 해체 시 계획의 수립 및 해체작업 종사자의 피폭선량 예측, 방사성폐기물의 관리/재활용 등의 기본자료로 활용할 수 있을 것으로 사료된다.

Abstract ▼ AI-Helper

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

주제어

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제안 방법

Based on the evaluations of this study it is believed that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when the decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure of workers engaged in decommissioning operations, the management of radioactive wastes, etc. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime of 50 years were evaluated to be at very low levels, and the radiation exposure doses to which radiation workers are subjected during the decommissioning process appeared to be at negligible levels.
The neutron cross-section was then obtained by calculating the reaction rate of a parent nuclide activated by neutron flux and irradiation in the main body and each component of a metal cask, using input data such as chemical composition based on the material quality of the components of the metal cask, etc. Based on this method, the neutron cross-section of an activated parent nuclide was corrected in the light-water fuel library of the ORIGEN-2 computation code, and the radioactivity of the radionuclide was calculated again based on neutrons accumulated over the design lifetime [9]. The MCNP code includes libraries of (n,γ), (n,p), (n,d), (n,α), (n,³He), etc.
Finally, to establish the decontamination and decommissioning procedures, the surface exposure dose rates of the main body of the cask and the canister, from which all of the spent nuclear fuel had been removed, were evaluated at the point of design lifetime expiration. However, because the inside of the canister, which comes in direct contact with the fuel, can be decontaminated, the contamination conditions were not taken into consideration for the evaluation.
In this study, an evaluation was conducted on the radioactive inventory of the components of a spent nuclear fuel metal cask over 10 years after its design lifetime using MCNP and ORIGEN-2. The results of the calculation of neutron flux in the main body and components of a metal cask, including the basket that is closest to the fuel, and the neutron shield that is farthest from the fuel, were between 2.
This study applies the MCNP5 computation code, which enables a realistic radiation transport analysis of three-dimensional geometrical structures [7]. It was used for calculating the neutron flux and reaction rate of the main body and components of a metal cask, as well as for detailed modeling. The total neutron flux emitted by the 21 bundles of design basis fuel at the initial stage of loading was evaluated using the ORIGEN-S module of the SCALE 6.
The radioactive inventory of the main body and components of the metal cask was calculated by applying the MCNP5·ORIGEN-2 evaluation system and by considering each component’s chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period.
An activation evaluation of the main body and internal/external components of metal casks whose design lifetime has ended provides quantitative data on their radioactive inventory. This study can be utilized as basic data necessary for the decommissioning of the metal cask (i.e., estimation of exposure doses to workers during decommissioning operations, determination of a decontamination technology, assessment of residual radioactivity at facilities/sites, etc.) [3].

이론/모형

It was used for calculating the neutron flux and reaction rate of the main body and components of a metal cask, as well as for detailed modeling. The total neutron flux emitted by the 21 bundles of design basis fuel at the initial stage of loading was evaluated using the ORIGEN-S module of the SCALE 6.1 computation code [8]. From among fuels generated from domestic light-water reactors, the selection of the design basis fuel was made based on the following conditions: the degree of burnup of 45,000MWD/MTU or less; ²³⁵U enrichment degree of 4.
Estimating the radioactive inventory of the main body and components of a metal cask resulting from neutron irradiation can be conducted using computation codes. This study applies the MCNP5 computation code, which enables a realistic radiation transport analysis of three-dimensional geometrical structures [7]. It was used for calculating the neutron flux and reaction rate of the main body and components of a metal cask, as well as for detailed modeling.

성능/효과

From the evaluation results, it was found that 5 years after end of the cask’s design lifetime, 60Co had greater radioactivity than other nuclides among the metal materials.
The radioactive inventory was then calculated for each time point of decommissioning. The calculation results of reaction rates by major nuclides were greatly affected by the characteristics of the relevant nuclide, and the relative errors of the calculation were mostly below 2%.
Refer to Tables 5 ~ Table 12 for the radioactivity of major nuclides over time, from immediately after the design lifetime to various time points. The results of the evaluation of radioactive inventory of the main body and each component of a metal cask indicated that ⁵⁴Mn, ⁵⁵Fe, ⁶⁰Co, ⁵⁹Ni, and ⁶³Ni were the major radionuclides generated by elements of the stainless steel and carbon steel, which were the two major materials forming the metal cask. In particular, the radioactivity of ⁶⁰Co was found to be relatively larger in all the metal cask components, due to the electron capture and decay of Ni and Fe, the major metal elements of the steels, after neutron irradiation.

후속연구

In addition, it will be possible to use this evaluation method to predict the radiation exposure doses of radiation workers during the decommissioning of other component parts of the metal cask, and the obtained results later can be used as basic data for decommissioning storage casks. The confinement system of a storage cask is based on a dry process and limits any contamination risks to the inside of the canister.

참고문헌 (10)

S.T. Yoon. Technology development for implementation of spent nuclear fuel transportation & storage system, 2nd Final Report. Korea Radioactive Waste Agency Report, KORAD-TR-2014-01 (2014).
S.B. Hong, B.G. Seo, D.G. Cho, G.H. Jeong, and J.K. Moon, "A study on the inventory estimation for the Activated bioshield concrete of KRR-2", Journal of Radiation Protection, 37(4), 202-207 (2012).

원문보기 상세보기
Nuclear Safety and Security Commission (NSSC), "Technical standards for the structure and equipment of interim storage facility for spent nuclear fuel", NSSC 2015-19 (2015).
American Society of Mechanical Engineering (ASME), An international code 2010 ASME boiler & pressure vessel code section.II Part A. ferrous material specifications, ASME Press, New York (2010).
United States Nuclear Regulatory Commission (U.S.NRC), Long lived activation products in reactor materials, NUREG/CR-3474, US (1984).
G.Y. Cha, S.Y. Kim, J.M. Lee, and Y.S. Kim, "The effects of impurity composition and concentration in reactor structure materials on neutron activation inventory in pressurized water reactor", JNFCWT, 14(2), 91-100 (2016).
D.B. Pelowitz, "MCNP - A General Mote Carlo N Particle Transport Code, Version 5", LA-CP-11-00438, Version 2.7.0, Oak Ridge National Laboratory, Oak Ridge (2011).
I.C. Gauld, "ORIGEN-S: Depletion Module to Calculate Neutron Activation, Actinide Transmutation, Fission Product Generation, and Radiation Source Terms", ORNL/TM-2005/39, Version 6.1, Sect.F7, Oak Ridge National Laboratory, Oak Ridge (2011).
A.G. Croff, "A User's Manual for the ORIGEN-2 Computer Code", ORNL/TM-7175, Oak Ridge National Laboratory, Oak Ridge (1980).
Nuclear Safety and Security Commission (NSSC), "Enforcement of Decree of the Nuclear Safety Act", NSSC 26760 (2015).

저자의 다른 논문 :

표제어: PCR

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용어 설명 출처 목록 (6)

용어 설명: PCR은 세균 특이성이 있는 primer를 이용하여 적은 수의 세균이 있을지라도 쉽게 검출할 수 있는 유용한 방법이며, 이를 이용하여 구강 내 치면세균막이나 타액에서 직접 세균을 검출할 수 있게 되었다[8].

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