Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems

The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents.

High Flux Reactor Review and Reactivity Control Analysis

For over 60 years, research reactors have provided the world with a versatile tool to test materials and promote irradiation research, as well as to produce radioisotopes for medical treatments. The High Flux Reactor (HFR), as a water moderated and cooled, beryllium-reflected reactor has awarded more attention in recent years. There is a wide range of designs and applications for HFRs that based on their own situation to meet research requirements. For the purpose of reducing the volume and mass of the reactor, as well as ensuring the safety operation, it is necessary to determine the most effective reactivity control scheme, and analyze the corresponding reactivity insertion accidents. This paper is going to investigate typical high flux reactors within the same type with HFETR, summarize general description and characteristics, the uses of the high flux reactor, and reactivity control mechanisms. In addition, the associated reactivity insertion accidents were presented and analyzed. The analysis result will provide some references to further design and construction of high flux reactor.

ANALYSIS OF REACTIVITY INSERTION AS A FUNCTION OF THE RSG-GAS FUEL BURN-UP

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.