Volume 86, Issue 3

Preventing the leakage of radioactive materials is important to nuclear safety. During a station blackout accident in pressurized water reactors, the hot leg creep rupture caused by hot leg countercurrent flow occurs before the reactor pressure vessel failure that caused by lower head rupture. The secondary fission products barrier is lost after hot leg creep rupture. An analysis for this phenomenon was done using the Modular Accident Analysis Program version 4.0.4 code. A station blackout accident for CPR1000 is simulated and the occurrence and influence of hot leg creep rupture phenomenon are analyzed in detail. After that, a sensitivity analysis of the opening of different pressurizer pilot-operated relief valves at five minutes after entering severe accident management guideline (before the hot leg creep rupture occurs) is studied. The results show that reactor pressure vessel failure time can be extended by at least 4 h if at least one pilot-operated relief valve is opened and direct containment heating phenomenon can be eliminated if at least two pilot-operated relief valves are opened.

The fuel burn-up rate has been raised in recent years to improve the efficiency of nuclear LWRs (light water reactors). Therefore, surveying and estimating changes in fuel properties and structural materials during radiation exposure is of paramount importance. In the present study, the researchers focused on analyzing the role of LWR fuel rod initial gap pressure (initial gas pressure when a fuel rod is fabricated) on the rod’s thermal and mechanical performance. FRAPCON-4.0 steady-state fuel performance code was used to simulate the effect of initial gap pressure on the behavior of a specific BWR-type fuel rod that was irradiated under the HALDEN research program. This fuel rod is similar to commercial BWR fuel rods in all respects, except that the research reactors have a height limit. The important fuel design criteria, such as the centerline temperature, effective stresses, total released fission gas to the fuel rod’s void volumes, and the cladding strains, were included in the analysis. According to the present study, a potential initial gap pressure range could be suggested to increase fuel rods’ lifetime by improving the safety criteria margins, especially fuel centerline temperature and the released amount of gaseous fission products. As we know, lower fuel temperature leads to having a reactor with a higher power density and, consequently, a maximum fuel burn-up rate, which can affect the economy and safety of nuclear power plants.

Subchannel analysis codes are widely used for the thermal-hydraulic design of nuclear reactor rod bundle. The effectiveness of subchannel analysis codes depends on turbulent mixing between these subchannels. Turbulent mixing has no direct contribution to the axial mass flow rate through subchannel but it will cause exchange of momentum and energy between the neighboring subchannels. Thus, it is important to evaluate the turbulent mixing coefficient for reactor rod bundle as it is a significant factor in the lateral energy and momentum equation for subchannel analysis codes like COBRA IIIC, COBRA-IV and MATRA LMR-FB. With the rapid developments in computational fluid dynamics and computer performance, three-dimensional analyses of turbulent flows occurring in the nuclear rod bundle have become more prominent. Several numerical analyses have already been attempted to investigate the flow behavior in rod bundles of different reactors. Much of these are dedicated to find out the structure of turbulence in rod bundle but a few analyses has been done to evaluate the magnitude of the turbulent mixing coefficient. In view of this, CFD analyses were carried out to determine the turbulent mixing coefficient in the simulated sub-channels of the reactor rod bundle. Previous studies on the structure of turbulence reveals that it is highly anisotropic. Hence, the Reynolds Stress Model (RSM), finer mesh and near wall distance ( y + ≤ 2) is required to capture turbulent mixing phenomena. The validation of results is done by comparing with subchannel mixing experiments.

A methodology for analyzing accidents with radioactive material release with EPZDose code was established. This code assesses doses and it is designed and developed by NTHU (National Tsing Hua University). To confirm the capacity of EPZDose, three postulated accident scenarios Taiwanese NPPs Chinshan (BWR/4) and Maanshan (PWR) are analyzed. All these scenarios are SBO (station blackout) transients because it is assumed that they result in a release of radioactive material. In this study, the source term data for EPZDose are taken from MELCOR or RASCAL calculations. In addition, calculated results of RASCAL code are compared with the results of EPZDose for these scenarios. The comparison show that the EPZDose predictions are consistent with the data of RASCAL. This indicates that the EPZDose has a respectable accuracy in the analysis of radioactive material release accidents.

In order to reduce the work burden during the training and drilling of the severe accident and severe accident management guidelines, and improve the implementation efficiency of the guidelines, a method of digitizing the SAMG program files is proposed. A set of digital SAMG system supported by information technology and combined with software and hardware is developed to transform the manual processes of paper file browsing, data searching, logical judgment and auxiliary calculation into automatic and digital processes, which can be used for SAMG training and drilling, and also for verifying the SAMG execution process and the effectiveness of mitigation measures.

In this study, the Doppler reactivity coefficient has been investigated for UO 2 , MOX, and (Th/U)O 2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor k eff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO 2 ). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O 2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

The objective of this research is to understand the condition of the structural material of the 30 MW RSG-GAS research reactor as input for the aging management program. Furthermore, this should enable a prediction of the remaining life of the components. In the current experiment, corrosion surveillance was carried out at Interim Storage for Spent Fuel (ISSF), that has similar water quality as in reactor pool by using a corrosion probe which is made of aluminum alloy and stainless steel. The probe set is designed to understand the effect of water quality in the ISSF pond. The corrosion processes observed were pitting, crevice and galvanic corrosion. Two sets of corrosion probes were immersed into the ISSF pool in 2007, hanging by steel wire, 1-meter height from the bottom surface. One probe set consists of horizontal and vertical positions. The soaking time was 7 years. The observations made were water chemical content, corrosion rate and visual analysis, macro and micro. For macro visual observations an optical microscope was used, for micro-observations SEM-EDX. From the results of macro-observations, information on the presence of galvanic corrosion, crevice and pitting was obtained. SEM-EDX provides information on the influence of chloride ions on corrosion products. This experience will be very useful in dealing with the aging process of Indonesia’s nuclear power plants in the future.

The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.