In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.

For the spent fuel modeling, the plastic model of the cladding used in FRAPCON uses the σ􀷥 = K􀟝̃􀯡 􁉂 􀰌􁈶 􀬵􀬴􀰷􀰯􁉃 􀯠 format. Strength coefficient (K), strain hardening exponent (n), strain rate sensitivity constant (m) are derived as the function of temperature. The coefficient m related to the strain rate shows dependence on the strain rate only in the α-β phase transition section, 1,172.5~1,255 K. But this is the analysis range of the FRAPTRAN code, which is an accident condition nuclear fuel behavior evaluation code. It does not apply to evaluate spent fuel. This coefficient in FRAPCON is used as a constant value (0.015) below 750 K (476.85°C), and at a temperature above 750 K, it is assumed that it is linearly proportional to the temperature without considering the strain rate dependence, also. In order to confirm the effect of strain rate, multiple test data performed under various conditions are required. However, since the strain rate dependence is not critical and test specimen limitation in the case of spent fuel, it is needed to replace with a new plastic model that does not include the strain rate term. In the new plastic model, the basic form of the Ramberg-Osgood equation (RO equation) is the same as ε􀷤 = 􀰙􀷥 􀮾 + 􀜭􀯥 􁉀􀰙􀷥 􀮾􁉁 􀯡􀳝. If the new variable α is defined as α = 􀜭􀯥􁈺􀟪􀯢/􀜧􁈻􀯡􀳝􀬿􀬵, this equation can be transformed into ε􀷤 = 􀰙􀷥 􀮾 + 􀟙 􀰙􀷥 􀮾 􁉀 􀰙􀷥 􀰙􀰬 􁉁 􀯡􀳝􀬿􀬵 . The procedure for expressing the stress-strain curve of the cladding with the RO equation is as follows. First, convert the engineering stress-strain into true stress-strain. Second, using a data analysis program such as EXCEL or ORIGIN, obtain the slope of the linear trend-line on the linear part and use it as the elastic modulus. Third, using the 0.2% offset method, find the yield point and the yield stress. Finally, using the solver function of EXCEL, find the optimal values of α and 􀝊􀯥 that minimize the sum of errors. The applicability of the suggested RO equation was evaluated using the results of the Zircaloy-4 plate room temperature tensile test performed by the KAERI and the Zircaloy cladding uniaxial tensile test results presented in the PNNL report. Through this, the RO equation was able to express the tensile test results within the uncertainty range of ±0.005. In this paper, the RO equation is suggested as a new plastic model with limited test data due to the test specimen limitation of spent fuel and its applicability is confirmed.

A tensile test is performed to obtain the mechanical property data of the spent fuel cladding. In general, the elastic modulus, elongation, yield stress, tensile stress, etc. are obtained by axial tensile test of cladding attaching an extensometer. However, due to the limitation in the number of specimens for spent nuclear fuel that can be made, the ring tensile test (RTT) whose required length of the specimen is short is mainly performed. In the case of RTT, an extensometer or strain gauge cannot be attached because the gauge part of the specimen is formed around the cladding and is short. In addition, since a load is applied in the radial direction of the cladding, a curved portion of the circular cladding is spread out and becomes straight, and then the cladding is tensioned. For this reason, it is difficult to obtain the stress-strain curve directly from the RTT results. Isight, which is used to identify the optimization design parameters, was used to build an optimization process that minimizes the difference between the RTT and the analysis to estimate the material property. For this, the elastic modulus, plastic strain, and the radius of the RTT jig were taken as fixed variables. As variables, isotropic hardening data and plastic stress were taken. The objective function was taken as the minimization of the area difference of the load-displacement curve obtained from the tests and analysis, of the difference in the magnitude of the maximum reaction force, and of the difference in the location where the maximum reaction force occurred. Optimization workflow was configured in the following order. First, using the calculator component, plastic stress design variables were created. Next, ABAQUS was placed to perform analysis using design variables, and the reaction force or displacement was calculated. After that, the reaction force was calculated considering the 1/4 symmetry condition using the script component. After that, the data matching component performed quantitative comparison of test and analysis data. Finally, by utilizing the exploration component, the plastic stress design variable that minimizes the difference in the objective function was obtained by automatically changing six optimization algorithms. In this paper, the constructed optimization process and the obtained plastic stress by applying it to the SUS316 RTT results are briefly described. The established optimization process can be utilized to obtain mechanical property from the results of the cladding RTT of spent nuclear fuel or new material.

The management before disposal of spent nuclear fuel is an essential process for safe management. It is important to determine the amount of nuclide inventory in order to ensure the integrity of spent nuclear fuel, as radiation generated from the nuclides is generated along with residual heat in the spent nuclear fuel. Based on the data on the characteristics of spent nuclear fuel generated in Korea, the correlation equation between burnup and enrichment was derived by referring to overseas cases (Sweden). Source term analysis was performed using the SCALE ORIGEN ARP code by securing the burnup history of nuclear fuel. Calculation was performed by inputting the combustion history of the fuel WH14×14 and WH17×17 as a reference for CE16×16 spent fuel. Through this study, the relationship was identified using the burnup, enrichment, and cooling time factors that influence the characteristics of spent nuclear fuel. In addition, the total source and spectrum data from neutrons and gamma sources were used to find out the characteristics of fuel.