Most of the radioactive wastes generated during the nuclear fuel processing activities conducted by KEPCO Nuclear Fuel Co., Ltd. are classified as the categories of intermediate and low-level radioactive waste. These radioactive waste materials are intended for permanent disposal at a designated disposal site, adhering strictly to the waste acceptance criteria. To facilitate the safe transportation of radioactive waste to the disposal site, it is necessary to ensure that the waste drums maintain a level of criticality that complies with the waste acceptance criteria. This necessitates the maintenance of subcritical conditions, under immersion or optimal neutron moderation conditions. This paper presents a criticality safety assessment of concrete radioactive waste under the most conservative conditions of immersion and moderation conditions for waste drums. Specifically, In order to send radioactive waste, which is the subject of criticality analysis, to a disposal facility, pre-processing operations must be performed to ensure compliance with waste accepatance criteria. To meet the physical characteristics required by the accepance criteria, particles below 0.2 mm should not be included. Thus, a 0.3 mm sieve is used to separate particles lager than 0.3 mm, and only those particles are placed in drums. The drums should be filled to achieve a filling ratio of at least 85%. A criticality analysis was conducted using the KENO-VI of SCALE. The Criticality Safety Analysis Results of varying the filling ratio of concrete drums from 85% to 100% presented in an effective multiplication factor of 0.22484. Additionally, the effective multiplication factor presented to be 0.25384 under the optimal moderation conditions. This demonstrates full compliance with the USL and criticality technology standards set as 0.95.

Since the Fukushima nuclear accident in 2011, the development of accident tolerant fuel (ATF) has been actively pursued as an alternative to improve the safety of nuclear power plants. In addition, nuclear power plants containing ATF have recently been included as green energy in the 2022 EU taxonomy bill, receiving a lot of attention. Many countries are considering increasing 235U enrichment from 5 to 10 235U % for higher burnup and long cycle operation with ATF improving safety. To utilize ATF, the applicability of fuel storage systems such as new fuel storage vault, Region 1, and Region 2 must be determined. The purpose of this paper is to confirm the applicability of applying ATF, which is being developed in Korea, to the nuclear fuel storage system of Korean nuclear power plants. The nuclear power plant model used in the analysis is APR-1400, a representative Korean nuclear power plant model, and ATF model used in the analysis is Mo microcell UO2 pellet with CrAl coating, which is being developed in Korea. MCNP 6.2 has been used for multiplication factor calculations, and the TRITON/NEWT and ORIGEN-S modules of the SCALE code have been used for depletion calculations. From the analysis results, solutions and additional analysis would be necessary to satisfy criticality regulatory requirements to utilize ATF with increased enrichment.

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the keff value, 0.95. The STAD canister had the lower keff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in keff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

Currently, as the saturation capacity of wet storage pool for spent nuclear fuel (SNF) of PWR in Korea has reached approximately 75%, Dry Storage Facilities (DSF) are necessary for sustainable operation of nuclear power plants. It is necessary to develop acceptance requirements for the delivery of SNF from reactor storage site to Centralized DSF. To do this end, the mechanical integrity of SNF is directly related to its repacking, retrieving, and transporting/handling performances. And also, this integrity is a key factor associated with the criticality safety that is connected to the damaged status of SNF. According to the NUREG/CR-6835, the NRC expects that the potential for nuclear fuel failures will increase because of the increase of the fuel discharge burnup and the degradation of fuel and clad material properties. Due to such damages and/or degradation, the fuel rods in the fuel assembly may be extracted and empty for following treatments (transportation, storage, handling etc). This condition can have a detrimental effect on the criticality safety of SNF. Thus, this study investigated whether extracted and empty of damaged SNF rod affects criticality safety. In this analysis, it is assumed that up to four fuel rods are missed. As a result of the analysis, As the number of fuel rods miss up to a certain number, the value of multiplication factor value of the fuel assembly increases. In addition, since the fuel rods located at the outermost layer contained relatively less fissile material than the fuel rods located center of the lattice, and neutrons were lost by the absorption material, the effective multiplication factor value gradually decreased. Nevertheless, the criticality safety was assessed to be maintained.

The criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of PWR SNF disposal canister: the KBS-3 type canister and the small standardized transportation, aging and disposal (STAD) canister. The criticality analyses were carried out for four cases as follows: (1) the calculation of isotopic compositions within a SNF using a depletion assessment code and (2) the calculation of the effective multiplication factor (keff) value using a criticality assessment code. Firstly, the KBS-3 type canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU was modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.74407, 0.74102, and 0.73783, respectively. Secondly, the STAD canister was modelled. The SNFs contained in the STAD canister were assumed to be the enrichment of 4.0wt% and the burnup of 45,000 MWD/MTU. The keff values for the cooling times of 40, 50, and 60 years were estimated to be 0.71448, 0.70982, and 0.70743, respectively. Thirdly, the KBS-3 canister with four SNFs of which the enrichment was 4.5wt% and the burnup was 55,000 MWD/MTU was modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.73366, 0.72880, and 0.72634, respectively. Finally, the calculations were carried out for the STAD canister containing four SNFs of the enrichment of 4.5wt% and the burnup of 55,000 MWD/MTU. The keff values for the cooling times of 40, 50, and 60 years were 0.70323, 0.69946, and 0.69719, respectively. Therefore, all of four cases met the performance target with respect to the keff values, 0.95. The STAD canister showed lower keff values than the KBS-3 canister. This appears to be the neutron absorber plate installed in the STAD canister although the distance among the four SNFs in the STAD canister was shorter than the KBS-3 canister.

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS+). The results confirmed that the calculated effective multiplication factors (keff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy‧hr-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS+ system without any shielding issues. The container thickness applied to the current KRS+ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS+ design is suitable in terms of nuclear criticality and radiation shielding.

Robots for a wide range of purposes have been developed along with the rapid industrialization. On the basis of higher convenience, the robots have been creating new industrial environment. The robots are generally classified into service robots and industrial robots. Robots in various shapes have been developed on the basis of the autonomous mobile robots. The autonomous mobile robots have the possibility to crash against any object in their moving range. This paper suggests a collision avoidance method to prevent collision of robots. The collision avoidance method analyzes the road context data and makes a robot move to a safe area. The collision avoidance method proposed in this paper converts the road context data into the information value. The collision avoidance method analyzes the present risk on the basis of the converted information value. The collision avoidance method makes a robot move to a safe area when crash is estimated by the information analysis.

미임계 시설은 정상 또는 사고상태에서 핵임계안전성이 확보되어야 한다. 이를 위해선 계산된 임계도가 바이어스와 불확실도로 결정된 미임계상한치(USL)를 초과하지 않는다는 것을 검증하는 절차가 반드시 필요하다. 하지만 핵임계안전성 검증 방법론은 여러 가지가 존재하며, 방법론이 달라지면 USL도 달라지므로 가장 적절한 한가지의 방법론으로 평가하는 것이 중요하다. 본 연구에서는 핵임계안전성 검증 방법론이 기술된 두 개의 문서를 비교 분석하여 한 가지 방법론으로 정립하였고, SCALE6.1 코드를 이용한 용기 설계에서의 미임계상한치 결정에 적용하였다.