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As a crucial core physics parameter, the control rod reactivity has to be predicted for the control and safety of the reactor. This paper studies the control rod reactivity calculation of the pebble-bed reactor with three scenarios of UO2, (Th,U)O2, and PuO2 fuel type without any modifications in the configuration of the reactor core. The reactor geometry of HTR-10 was selected for the reactor model. The entire calculation of control rod reactivity was done using the MCNP6 code with ENDF/B-VII library. The calculation results show that the total reactivity worth of control rods in UO2-, (U,Th)O2-, and PuO2-fueled cores is 15.87, 15.25, and 14.33%Δk/k, respectively. These results prove that the effectiveness of total control rod in thorium and uranium cores is almost similar to but higher than that in plutonium cores. The highest reactivity worth of individual control rod in uranium, thorium and plutonium cores is 1.64, 1.44, and 1.53%Δk/k corresponding to CR8, CR1, and CR5, respectively. The other results demonstrate that the reactor can be safely shutdown with the control rods combination of CR3+CR5+CR8+CR10, CR2+CR3+CR7+CR8, and CR1+CR3+CR6+CR8 in UO2-, (U,Th)O2-, and PuO2-fueled cores, respectively. It can be concluded that, even though the calculation results are not so much different, however, the selection of control rods should be considered in the pebble-bed core design with different scenarios of fuel type.

A current effort focused on mechanisms of maximizing plutonium consumption is one of the solutions to control large stockpile of plutonium produced from nuclear power plants and military programs. The pebble-bed reactor with plutonium fuel is a unique case in the high-temperature reactors intended to reduce the concentration of plutonium that has been accumulated in the world. However, the plutonium core with tendency to produce a lower effective delayed neutron fraction (βeff), which is a crucial parameter in reactor safety, should be examined. The purpose of this study is to investigate the characteristics of effective delayed neutron fraction (βeff) for pebble-bed reactor with plutonium fuel. The uranium and thorium contents with a 1:1 and 1:2 ratio of heavy metal mass per pebble were combined into the plutonium core to study their effect on the characteristics of βeff. A series of calculations with various mass of plutonium per pebble were performed by using the MCNP6 code and ENDF/B-VII library. The βeff of pebble-bed reactor with plutonium fuel and its combination with uranium and thorium fuels were analyzed in term of core safety aspect because it could affect the transient behavior into a fatal accident situation.

The 10 MW Indonesia’s Reaktor Daya Eksperimental (RDE) is a Pebble Bed High Temperature Reactor-type and to be constructed in Serpong Nuclear Zone, Puspiptek, Indonesia. The reactor applies helium gas coolant, graphite moderator, and low uranium enriched fuel (17%). A water-ingress accident may take place in the RDE core and such the accident is caused by a pipe break in the steam generator system of the reactor and hence the accident influences the reactivity of the core. To begin with, an initial core of the RDE should be modelled and the model contains the identification of material zones in the reactor core. To estimate the reactivity change due to the water-ingress accident in the initial RDE core, a void core condition from any coolant should be firstly deliberated. The core reactivity change is then estimated based on effective multiplication factor (k eff) as a function of water density which comes into void cells among pebble containing helium coolant in the RDE core and that of the variation of 235U enrichment. The data library of thermal scattering S(α,β) as a function of temperature has been applied. The data base of thermal scattering for graphite was also utilized to comprehend the validity calculation using MCNPX. The ENDF/B-VII continuous energy nuclear data library was also smeared for all calculations at temperature of 900 0K. For the calculation using MCNPX, two options of KCODE and KSRC were taken into account to estimate k eff of the RDE core. To achieve the accuracy of keff in the RDE reactor core, the amount of 20,000 neutrons per cycle for 150 non-active cycles and 650 active cycles was taken into account. The results showed that water-ingress accident in the RDE core is a dynamical behaviour which may result in reactivity core change of the RDE core. Finally, from all caclulated results taking into account the varietes of water density and 235U fuel enrichment, the RDE reactor core is totally safe in the event of water-ingress accident ensuing during the RDE reactor operation.

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP–ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP–ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.