## 3 Safety Evaluation of the DSM
### Criticality Analysis  
The design of fuel canisters in the DSM ensures that the fuel will remain in a configuration that has been predetermined to be subcritical during loading, storage, and retrieval. This criticality design should allow for any consequences likely to result from redistribution or the intrusion of a moderator as a consequence of an internal or external event. The technical standards for criticality evaluation are in accordance with the US 10 CFR72.124 \"Criteria for Nuclear Criticality Safety\", the American Nuclear Society regulatory guidelines ANSI/ANS-57.9 [13], and the US NRC standard review guidelines NUREG-1536. The technical standards for critical evaluation applied to the conceptual SDSS design are as follows:  
* \(K_{\mathit{eff}}\) (effective neutron multiplication factor) at the 95% probability and 95% confidence level should not exceed 0.95 under the normal, off-normal, and accident conditions.
* The criticality safety of the DSM design should be ensured on the basis of favorable geometry, with a fixed neutron absorber. Where solid neutron absorbers are used, the canister design should provide for a positive means to verify their continued efficacy during the storage. The neutron absorbers' continued efficacy may be confirmed by verification before use, showing that the degradation of these materials cannot occur over the lifetime of the facility.
* The criticality safety of the canister design should not rely on credit for more than 75% of the neutron poison material in the fixed neutron absorber. Comprehensive tests are needed in order to verify the presence and uniformity of the neutron absorber.  
In short, critical evaluation should be designed to maintain subcriticality in consideration of normal, off-normal, and accident conditions according to the technical standards presented above, but in this analysis it was designed in consideration of only the full flood condition, as the most conservative assumption was considered. The canister model used for critical evaluation was a 24 FA container, which consists of a basket cell, a neutron absorber, and a neutron-absorber cover. For critical evaluation, the SCALE6.1.3 program developed by ORNL was used, and the library of the ENDF 238 group was applied. Validation of the SCALE6.1.3 program was carried out in accordance with the NUREG-6698 methodology based on the 236 Benchmark Experiments presented by the OECD/NEA. These uncertainties may be statistically combined, and the calculation for the maximum \(K_{\mathit{eff}}\) may therefore be described as follows:  
\[\mathit{Maximum}\ K_{\mathit{eff}}=K_{\mathit{cal}}+K_{\mathit{bias}}+K_{ \mathit{uncertainty}}\]  
where: \(K_{\mathit{cal}}\) is the calculated SCALE6.1.3 reactivity in three dimensions;  
\(K_{\mathit{bias}}\) includes the bias from the 236 benchmark calculations;  
\(K_{\mathit{uncertainty}}\) includes the effects of all uncertainties combined statistically.  
\(K_{\mathit{uncertainty}}\) includes uncertainties of the manufacturing tolerances of the fuel assembly --such as UO2 density and cladding thickness--and the concrete module, such as canister thickness and the B10 density of the neutron absorber. As a result of the validation of the SCALE codes, bias and uncertainty were assessed as 0.00749 and 0.01079, respectively. Figure 3 shows the radial and axial models used in the evaluation, while Table 2 shows the results of the critical assessment, indicating that the maximum \(K_{\mathit{eff}}\) for normal conditions at the 95% probability and 95% confidence level is 0.94153, which is below the design basis of 0.95.