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This paper which is a thesis for the title teknologie licentiat is a summary text of several works performed by the author regarding spectroscopic measurements on spent nuclear fuel. Methods for determining the decay heat of spent nuclear fuel by means of gammaray spectroscopy and for verifying the integrity of nuclear fuel by means of tomography is presented. A summary of work performed regarding gammaray detector technology for studies of fission gas release is presented.

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The molten core-concrete interaction (MCCI) is treated as one of the important phenomena that may lead to the late containment failure by basemat penetration in a hypothetical severe accident of light water reactors (LWRs). The earlier research has showed that heat transfer limitation exists for the coolability of ex-vessel corium by atop water flooding due to crust formation on the melt/water interface that will isolate melt from water. However, several cooling mechanisms were identified in a series of intense investigations. A code (CORQUENCH) was developed and updated to incorporate the newly identified cooling mechanisms for the better predictions of cavity erosion and corium cooling behaviors. A description about such cooling mechanisms (i.e., bulking cooling, water ingression, eruption and crust breach) and the concrete ablation models implemented in the code is presented in this thesis. The technical work in the thesis includes two parts: first, the verification and validation of the code were performed against the CCI tests from the OECD/MCCI projects; and then a reactor-scale simulation was carried out for MCCI and ex-vessel corium coolability of a reference PWR with LCS concrete. The calculations of CCI tests have a plausible agreement with the experimental data. The calculation predicts an optimistic result for the reactor case, and a fast quenching achieved at about 145 minutes. In addition, a sensitivity study was also conducted on several important parameters, i.e., concrete type, corium composition, water flooding time, atmosphere pressure, concrete ablation temperature, initial temperature, decay power, cavity geometry, concrete decomposition model and melt upper heat transfer model. An attempt to explain the physics of the different predicted phenomena is presented as well. Finally, comparative calculations were performed by the other codes (ASTEC and FinCCI) for the same reactor-scale configuration. Discrepancies are found in the results. Some suggestions are proposed to improve the CORQUENCH code.

Heavy liquid metal (HLM - lead or lead bismuth eutectic) is considered as a candidate coolant for next-generation fast reactor and accelerate-driven systems (ADS), due to its favorable chemical, thermo-physical and neutronic properties in comparison with sodium which has been used as coolant in fast breeder reactors (FBRs). To perform design-base-accident analysis for the HLM-cooled reactors, the well-known transient thermal-hydraulic analysis codes (e.g., RELAP5 and TRACE) are being applied to the reactors with the new coolant. Since these codes were originally developed for light water reactors (LWRs), validations are necessary to ensure the codes to count all influences of the thermo-physical properties of HLM. In this thesis, the TRACE code is employed to simulate the transients performed on the TALL test facility which is a medium-size loop for thermal-hydraulic study of lead bismuth eutectic (LBE). The objectives of the present work are two-fold: i) to interpret the transients performed on the test facility; ii) to qualify the capabilities of the TRACE code for HLM-cooled system by using the experimental data and then perform separate-effect study beyond the experimental conditions. The transients related to safety issues such as loss of heat sink, loss of primary pump, loss of both primary/secondary pumps, overpower and accelerator trip are chosen in the simulations. In addition, the operational transients, the startup and the shut-down of the system are also simulated, respectively. Two configurations of the facility are considered: Configuration-A with a core tank, and Configuration-B with a fuel rod simulator. The separate-effect study is conducted to investigate the effects of coolant inventory, LBE flow resistance and mass flowrate in the secondary loop on natural circulation in the primary loop. Generally speaking, for all the cases analyzed in the present study, the calculation results have a good agreement with the experimental data for the primary side (LBE) parameters (e.g., variations in temperature and mass flowrate). Specifically, the simulation for the transient loss of heat sink indicates the same tendency as in the experiment in term of temperature: it is rising at the inlet and outlet of the core simulator, as well as at those of the intermediate heat exchanger. The temperature keeps going up till the resumption of the heat sink as a protective measure, and then it decreases sharply at the very beginning and gradually returns to steady-state conditions. For the transient loss of primary pump, the temperature level is elevated and a significant natural circulation in the primary loop obtained. The simulation well reproduces the establishing process of natural circulation and final flowrate, but it overestimates the peak temperature. Such simulation outcome II applies to the transient ―loss of both primary and secondary pumps‖, except for a further elevated temperature. The calculation results of the transients startup and overpower are perfectly matching the experimental data, especially when approaching the final steady state, while the transients of shutdown and heater trip are underestimated in their final temperatures but the trends are well captured in both the transients. In general, the transient time to a steady state or the maximum temperature level of Configuration-B is much shorter than that of Configuration-A, mainly because of the larger inventory of LBE in Configuration-A. The separate-effect study also shows that the LBE inventory in the primary loop plays a mitigative role in the transients. For the loss of primary pump transient, decrease in primary flow resistance and increase in secondary flowrate all contributes to passive safety. Keywords: Heavy metal liquid, transient thermal-hydraulics, safety analysis, TRACE

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Accelerator-driven systems (ADS) are being investigated and designed for transmutation of the long-lived nuclear waste. Application of ADS allows to safely transmute large fractions of minor actinides (MA) per reactor core, while the fraction in critical reactors is limited to a few percent due to the safety constraints. Additional imposed costs of ADS introduction into the nuclear fuel cycle can be decreased by improving their effciency, particularly the external source effciency. Design of the European Facility for Industrial Transmutation (EFIT) with transuranium (TRU) oxide fuel has been recently developed in the frame of  the EUROTRANS project. In this thesis it is shown that the neutron and proton source effciency of EFIT can be significantly improved by application of advanced TRU nitride fuel. Thanks to the good neutron economy of the nitride fuel, the EFIT core size can be reduced, which permits reducing the size of the spallation target. This provides a twice higher proton source effciency and therefore lowers the demand for the proton accelerator current. Additionally, the nitride version of EFIT features two times lower coolant void worth improving the core safety. The pulsed neutron source (PNS) methods for ADS reactivity control have been studied experimentally at the YALINA facility in Minsk (Belarus) and shown good agreement with numerical simulation. The PNS methods will be most probably used for calibration of online reactivity monitoring system in future ADS. / QC 20110412

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