Earticle

Background: As Egypt ventures into nuclear energy, ensuring the safety and security of its nuclear facilities is paramount. The ongoing construction of the El Dabaa Nuclear Power Plant (NPP) directs the necessity for Egypt to meticulously evaluate the potential outcomes of any conceivable accidents. A focused analysis on design-based accidents such as steam generator tube ruptures (SGTR) is imperative for assessing their impact on the local populace and the environment. Through comprehensive assessments, Egypt aims to enhance the safety protocols of its NPP. This preemptive strategy facilitates the identification and minimization of risk, along with the formulation of efficient emergency response frameworks. Such diligent preparations are essential for cultivating public confidence while ensuring compliance with international safety norms and securing the long-term sustainability of Egypt’s nuclear energy program. Materials and Methods: The hypothetical scenario of a SGTR at the El Dabaa NPP was modeled using the Radiological Assessment System for Consequence Analysis (RASCAL) code. This model examined the potential consequences of such an accident by incorporating meteorological data from 2013 to 2023 across all seasons to compute the source term, total effective dose equivalent (TEDE), and thyroid dose. Then, these metrics were evaluated against established safety thresholds to determine compliance. The outcomes of this analysis provide critical insights and aid in formulating strategies for effective response to an accident, especially if the calculated doses exceed the permissible limits. Results and Discussion: The analysis of the worst-case scenario for the SGTR accident at the El Dabaa NPP involves a U-tube breakage above the water level, exacerbated by a concurrent station blackout. This condition potentially leads to a considerable dispersal of radioactive materials, especially within a 0.4 km radius, with a TEDE reaching 4.80×103 mSv during autumn. However, the severity of this worst-case scenario fluctuates with seasonal weather conditions; notably in spring, the highest TEDE was 15 mSv at a 40 km distance. The source term distribution indicates that noble gases account for 31.6%, iodine group for 32.0%, and other sources constitute 36.4% of the total radioactive release. These findings confirmed that TEDE and thyroid dose exceeded the permissible thresholds, thereby highlighting the importance for protective measures to mitigate the potential risks of such accidents. Conclusion: A comprehensive assessment of SGTR accidents at the El Dabaa NPP emphasizes the critical necessity for stringent safety protocols and protective interventions in worst-case scenarios. By preemptively addressing these risks, Egypt can fortify its nuclear safety framework, emergency response capabilities, and adherence to global safety standards. This proactive stance not only assures the long-term sustainability of Egypt’s nuclear energy program but also solidifies public confidence.

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. Thespent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARPthat is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if thestructure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT arecompared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on thespent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gammadose rate varies irrespective of type of the material that the crack is filled with.

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

발산형 바닥 경사 생성항(DFB, Divergence Form for Bed slope source term)을 엄밀하게 유도하였으며, DFB 중에서 격자의 변에서 평균 수심을 이용하는 mDFB의 오차를 명백하게 입증하였다. 또한, DFB 기법은 바닥 경사 생성항에 대해 정확한 방법임을 밝혔다. 완전히 잠기기 않은 격자에 대한 기존의 체적-수심 관계의 오류를 수정하였으며, C-특성의 충족을 위해 완전히 잠기지 않은 변에 대한 처리가 필요함을 검토하였다. 이 연구를 통해 근사 Riemann 해법으로 천수방정식을 해석할 때 보다 정확한 수단을 제공할 수 있을 것으로 기대한다.

DFB (Divergence Form for Bed slope source term) was rigorously derived and the error of mDFB using mean water depth at the cell face in DFB was clearly demonstrated. In addition, DFB technique turned out to be an exact method to the bed slope source term. The existing volume/free-surface relationship to the PSC (Partially Submerged Cell) has been corrected. It was discussed that treatment for the partially submerged edge is required to satisfy the C-property in PSC. It is expected that this study will provides a more accurate means in analyzing the shallow water equations with the approximate Riemann solver.

사용후핵연료를 파이로 건식처리하면 사용후핵연료 자체 내에 존재하는 세슘, 스트론튬, 초우라늄 계열등이 중간저장 되어 영구처분 방사선원항에서 제외되므로 사용후핵연료집합체를 구성하는 구조재, 즉 금 속폐기물의 방사선원항이 중요해지게 된다. 따라서 본 연구에서는 17×17 KOFA 사용후핵연료 10 톤이 파 이로 건식처리 되었을 경우를 가정하여 각 구조재 부품별로 방사선원항 특성을 분석하였다. 우선 구조재 부품별로 질량 및 부피를 상세히 계산하였다. 핵연료 상단 및 하단 고정체에서의 중성자스펙트럼이 노심과 다르므로 각 구조재 부품별로 핵반응단면적라이브러리를 KENO-VI/ORIGEN-S 모듈로 직접 생산하였으 며, 이를 적용하여 ORIGEN-S 코드로 방사화 방사선원항을 평가하였다. 평가결과 원자로 방출후 10 년 시 점에서의 방사능세기, 붕괴열, 위해지수 값은 각각 1.40×1015 Bequerels, 236 Watts, 4.34×109 m3-water 로 나타났으며, 이는 사용후핵연료 자체 값의 0.7 %, 1.1 %, 0.1 %에 해당하는 값이다. 방사능세기, 붕괴 열, 위해지수 모든 측면에서는 금속폐기물 전체물량의 1 %만을 차지하는 인코넬 718 그리드판이 가장 중 요한 것으로 평가되었으며, 특히 이를 따로 분리하여 관리하면 금속폐기물 전체 방사능세기를 20∼45 % 정도, 위해지수를 30∼45 % 정도 감소시킬 수 있는 것으로 나타났다. 전체적으로 볼 때, 금속폐기물의 방 사능세기 및 위해지수는 처분시스템 설계 시 중요한 인자로 고려되어야 하나, 붕괴열은 그 열량이 작아 중 요하지 않은 것으로 나타났다.

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.40 1015 Bequerels, 236 Watts, 4.34 109 m3-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20 45 % and 30 45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

현재 전 세계적으로 설계단계에서 부식 생성물과 방사성 핵종의 양을 예측하는 프로그램에 대해서는 개발되거나 개발중인프로그램이 다양하다. 그러나 원자력 발전소 해체 시 발생하는 방사화 부식생성물의 양을 평가하는 코드에 대한 개발은 이루어지지 않고 있어 정확한 산정에 어려움이 있다. 원자로 용기, 원자로 구성품 및 인접 구조물에서의 특성 원소의 중성자 조사로 인한 방사화재고량을 평가하기 위해서는 원자로의 고정된 구조물을 대표하는 모든 영역에서의 평균 중성자속과 구조물의 물질조성 및 원자로 운전이력 등을 이용하여 평가해야 한다. 본 논문에서는 설계단계에서 사용되는 1차 계통의 부식생성물과 방사성 핵종의 양을 예측하는 CORA, PACTOLE, CRUDSIM, CREAT 및 ACE 코드를 분석하였다. 향후 연구에서는 제염해체 폐기물 발생량 평가에 대한 사용가능성과 개선점을 찾아 부식생성물량 산정에 정확성을 높이고자 한다.

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosionproduct at the stage of designing NPP. However, since there are not many developments in evaluating quantity ofactivation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculatingaccurately. In order to evaluate activation products inventory for the research of effect of neutron activation in thereactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history ofnuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure ofnuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantityof radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracyin calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvementfor evaluation of the decontamination.

2009년말 기준으로 11,811 다발의 경수로 사용후핵연료가 방출되었으며, 지금까지 각 사용후핵연료에 대해 방사선원항을 가중하여 설계에 반영하기는 사실상 불가능하여, 원자력 관련시설 설계시 보수성을 갖는 기준 사용후핵연료를 선정하고 이를 바탕으로 시스템 설계를 수행하여 왔다. 방사선원항에 대한 단순모델을 적용하면 각 사용후핵연료에 대한 방사선원항을 가중함으로써 이와 같은 보수성을 배제할 수 있으므로 본 연구에서 웨스팅하우스형 원전에 사용된 사용후핵연료를 대상으로 방사선원항, 즉, 붕괴열, 방사능세기, 섭취위해도 등을 예측하기 위한 회귀모형을 개발하였다. 개발된 회귀식을 통해 예측된 방사선원항값은 ORIGEN-ARP 코드로 계산된 값과 약 5% 이내에서 잘 일치함을 확인하였으며, 이의 유용성을 검토한 결과 각각의 사용후핵연료에 대한 방사선원항을 가중하여 설계에 반영하면 보수성을 줄일 수 있음을 확인하였다. 따라서 본 연구에서 개발된 회귀식은 사용후핵연료의 저장 및 처분과 관련한 원자력시설 설계시 개념설계 단계에서 유용하게 사용될 수 있을 것으로 판단된다.

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude of conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.

원자력발전소 해체를 준비하기 위해서는 해체대상 발전소에 대한 선원항 평가가 선행되어야 한다. 해체전략 수립단계에서선원항 평가 결과를 토대로 해체 폐기물을 분류하고 비용평가를 수행한다. 본 연구에서는 월성 1호기의 예비 선원항 계산을수행할 수 있도록 MCNP/ORIGEN-2 모델의 타당성 평가를 수행하였다. 연소도가 다른 핵연료 다발의 악티나이드 계열과핵분열 생성물의 핵종 수밀도는 싱글 채널 모델을 이용하여 MCNPX 코드로 연소 계산하여 구하였다. 선원항의 정확도에 영향을 미치는 두가지 요인에 대해 조사하였다. 첫번째 요인으로 선원항 계산에 영향을 미치는 중성자 스펙트럼을 MCNP로 계산하여 해당 핵종의 1군 미시 핵단면적에 반영하였다. 중성자 스펙트럼이 반영된 라이브러리로 계산한 선원항과 ORIGEN-2코드 package에 내장된 library (CANDUNAU.LIB)로 구한 선원항을 비교하였다. 두번째 요인으로 선원항에 대한 출력이력의 영향을 조사하였다. 해체 폐기물의 저준위 폐기물 처분 가능성을 살펴보기 위해, 2010년도 교체된 압력관, 칼란드리아관과 기존 칼란드리아 동체에 대하여 중성자 스펙트럼을 반영한 library를 적용하여 MCNP/ORIGEN-2로 선원항 평가 계산을 수행하였다.

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planningphase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based onthe results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminarysource term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles withdifferent burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factorsaffecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activationcalculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNPsimulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library werecompared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operationhistory effect on activation calculation was also investigated. The source terms on both pressure tubes and calandriatubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum correctedlibrary if the decommissioning wastes can be classified as a low level waste.

The release source term of radioactivity becomes a critical foundation for emergency response and accident consequence assessment after a nuclear accident Rapidly and accurately inverting the source term remains an urgent scientific challenge. Today source term inversion based on meteorological data and gamma dose rate measurements is a common method. But gamma dose rate actually includes all nuclides information, and the composition of radioactive nuclides is generally uncertain. This paper introduces a novel nuclear accident source term inversion model, which is Improve Snow Ablation Optimizer-Sensitivity Analysis Pruning Extreme Learning Machine (ISAO-SAELM) model. The model inverts the release rates of 11 radioactive nuclides (I-131, Xe-133, Cs-137, Kr-88, Sr-91, Te-132, Mo-99, Ba-140, La-140, Ce-144, Sb-129). It does not require the use of the physical field of the reactor to obtain prior information and establish a dispersion model. And the robustness is validated through noise analysis test. The mean absolute errors of the release rates of 11 nuclides are 15.52 %, 15.28 %, 15.70 %, 14.99 %, 14.85 %, 15.61 %, 15.96 %, 15.42 %, 15.84 %, 15.13 %, 17.72 %, which show the significant superiority of ISAO-SAELM. ISAO-SAELM model not only achieves notable advancements in accuracy but also receives validation in terms of practicality and feasibility.