Spent nuclear fuel continues to be generated domestically and abroad, and various studies are actively being conducted for interim dry storage and disposal of spent nuclear fuel. The characteristics vary depending on the type of spent nuclear fuel and the initial specifications, and based on these characteristics, it is essential to estimate the burnup and enrichment of spent nuclear fuel as a nondestructive assay. In particular, it is important to estimate the characteristics of spent nuclear fuel with non-destructive tests because destructive tests cannot be performed on all encapsulated spent nuclear fuel in case of intrusion traces in safeguards. Data is made by measuring spent nuclear fuel directly to evaluate burnup of spent nuclear fuel, but computer simulation research is also important to understand its characteristics because past burnup history is not accurately written, and destructive testing is difficult. In Sweden, the dependency of the burnup history in source strength and mass of light-water reactor-type spent nuclear fuel was evaluated, and this part was also applied to MAGNOX in consideration of the possibility of being used to verify DPRK’s denuclearization. SCALE 6.2 TRITON modeling was performed based on public information on DPRK’s 5 MWe Yongbyon reactor, and the source strength of Nb-95, Zr-95, Ru-106, Cs-134, Cs-137, Ce-141, Ce- 144, Eu-154 nuclides were evaluated. Since the burnup of MAGNOX is lower than that of lightwater reactors, major nuclides in decay heat were not considered. The cooling period was evaluated based on 0, 5, 10, and 20 years. In case the discharge timing was different, the total period of discharge and reloading was the same, and the end-cycle burnup was the same, calculations showed that the source strength emitted from major nuclides was evaluated within 2-3% except for Ru-106 and Ce-144 nuclides. Even the burnup step of nuclear fuel is the same, and the reloaded length after discharge is different, i.e., the cooling period between is different at 5, 10, and 20, the source strength of Nb-95, Zr-95, Ce-144, and Cs-137 was evaluated as an error of 1%. Except for Ru-106 and Ce-144, nuclides are highly dependent on burnup. Compared to the case of light-water reactors, the possibility of a decrease in error needs to be considered later because the specific power is low. As a result, radionuclides in released fuel depend on the effects of burnup, discharged and reloaded period, and a cooling period after release, and research is needed to correct the cooling period within the future burnup history. In addition, in this study, it is necessary to select a scenario -based burnup because the standard burnup due to the statistical treatment of discharged fuels was not considered as conducted in previous studies.

Spent nuclear fuel (SNF) characterization is important in terms of nuclear safety and safeguards. Regardless of whether SNF is waste or energy resource, the International Atomic Energy Agency (IAEA) Specific Safety Guide-15 states that the storage requirements of SNF comply with IAEA General Safety Requirement Part 5 (GSR Part 5) for predisposal management of radioactive waste. GSR Part 5 requires a classifying and characterizing of radioactive waste at various steps of predisposal management. Accordingly, SNF fuel should be stored/handled as accurately characterized in the storage stage before permanent disposal. Appropriate characterization methods must exist to meet the above requirements. The characterization of SNF is basically performed through destructive analysis/non-destructive analysis in addition to the calculation based on the reactor operation history. Burnup, Initial enrichment, and Cooling time (BIC) are the primary identification targets for SNF fuel characterization, and the analysis mainly uses the correlation identified between the BIC set and the other SNF characteristics (e.g., Burnup - neutron emission rate) for characterizing. So further identification of the correlation among SNF characteristics will be the basis for proposing a new analysis method. Therefore, we aimed to simulate a SNF assembly with varying burnup, initial enrichment, and cooling time, then correlate other SNF properties with BIC sets, and identify correlations available for SNF characterization. In this study, the ‘CE 16×16’ type assembly was simulated using the SCALEORIGAMI code by changing the BIC set, and decay heat, radiation emission characteristics, and nuclide inventory of the assembly were calculated. After that, it was analyzed how these characteristics change according to the change in the BIC set. This study is expected to be the basic data for proposing new method for characterizing the SNF assembly of PWR.

After spent fuel is stored in a dry storage container, it becomes difficult to obtain information on the fuel’s characteristics. As a result, it is necessary to identify the characteristics of spent nuclear fuel in advance and secure the information necessary to establish delivery acceptance requirements for interim storage and disposal in the future. Therefore, it is necessary to evaluate the characteristics of spent fuel before loading dry storage casks. In order to prepare for the dry storage of spent fuel, information on the basic characteristics of the fuel is required. As part of this information, it is also necessary to establish calculation criteria for spent fuel burnup. Spent fuel burnup can be classified into three categories. The first is burnup evaluated using design codes (design burnup), the second is burnup measured by furnace instruments during power plant operation (actual burnup), and the third is burnup measured through measurement equipment (measured burnup). This paper describes a comparative evaluation of design burnup, actual burnup, and measured burnup for specific fuels (40 bundles).

한국방사성폐기물관리공단 주관 하에 개념 설계된, 연소도이득효과 적용 대용량 수송용기에 대해 방사 선 차폐 안전성을 평가하였으며 여러 방사선원들이 수송용기 주변 선량률 분포에 미치는 영향을 분석하였다. 가능한 모든 방사선원(중성자선원, 감마선원, 방사화선원)들을 고려하였으며 보수적인 가상의 핵 연료(너비: WH 17 RFA, 축방향: CE Type)를 선정, 실제 상황과 동일한 조건이 되도록 계산모델을 구축 하였다. 모든 조건(정상 및 가상사고 조건)에서 표면선량률과 외부선량률이 법적기준치를 만족하고 있었 으며 축방향 높이에 따라 각 선원들의 기여도가 변하고 있었지만 정상조건에서의 최대 표면선량률과 외 부선량률은 방사화선원에 의한 영향이 가장 높은 것으로 확인되었다. 가상사고 조건에서는, 중성자선원 의 선량률 기여도가 대략 90%에 달하고 있었으나 수송용기 끝단에서는 방사화선원에 의한 선량률이 급 격하게 상승함에 따라 BUC 적용 수송용기의 방사선 차폐해석시 충분히 보수적으로 해석되도록 방사화선 원을 정밀하게 분석하여 설정하여야 할 것으로 판단되었다.