Overview of Post-Closure Criticality Safety – RD&D Topics in Switzerland

The demonstration of post-closure criticality safety of spent nuclear fuel in a deep geological repository is a regulatory requirement in Switzerland and many other countries. One of the main challenges stems from the very long timescale (1 million years in Switzerland) that has to be considered. Nagra, the Swiss National Cooperative for the Disposal of Radioactive Waste, is presently elaborating the technical and scientific foundation of the criticality safety assessment in view of the upcoming general licence application for the Swiss Spent Fuel and HLW repository. In this context, Nagra supports and pursues a focussed RD&D programme in collaboration with several renowned research institutes. Nagra's safety concept relies on natural and technical barriers. For the initial thermal phase of the repository, a steel canister assures complete containment of the spent fuel. The canisters are foreseen to remain intact for approximately 10 000 years; however, the subcriticality of the system has to be ensured for a much longer period. In this context, an important part of the research activities pursued by Nagra address the nearfield evolution and the formulation of scenarios for the corresponding evolution of the canister and spent fuel system. The role that variations in the canister design and material composition have on the system's reactivity are also investigated. Other research topics focus on developing a reliable methodology for carrying out the criticality safety assessment. This symposium contribution gives an overview of the post-closure criticality RD&D activities pursued and envisioned by Nagra. The general context and Nagra's fundamental approach to elaborating the current phase of the criticality safety assessment are presented first. Following this, the current RD&D landscape and the most important technical considerations underpinning Nagra's technical basis for the post-closure criticality safety assessment in particular are discussed. Future planned research topics and points of interest are also presented as an outlook of this presentation.

CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff value of the proposed dry cask design for the RDNK spent fuel. The keff values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition

Criticality analysis of spent fuel assumes that the fuel material is unburned which means that it is in its most reactive condition. In fact, this is not the real situation for fuel as it is burned during reactor operation causing reduction in the reactivity. Considering the reduction in reactivity during spent fuel calculations is the Burn-up Credit concept (BUC). In addition, the control rods radial and axial positions have an effect on the reactivity which can be considered in the criticality safety analysis. This paper studies the effect of burnup and control rods (CRs) movement on reactivity and isotopes inventory. Calculations are carried out in two phases, first kinf is calculated for different burnup profiles with control rods are either fully withdrawn or fully inserted. In the second phase keff is calculated for different control rods insertion levels. For both phases, burnup calculations are performed for a UO2 assembly then multiplication factor calculations of burned UO2 assemblies in cold state are done. The burnup calculations are performed using MCNP6 code and ENDF/B-VII library for different burnup levels up to 45 GWd/tU. The results obtained can be taken in consideration in criticality safety analysis performed for the spent fuel to improve the economic efficiency for manufacture, storage and transportation of fissile materials.