Band 86, Heft 2

An experimental study has been conducted to examine the effect of different concentrations of α-Al 2 O 3 nanofluids on boiling heat transfer for downward facing heating. The experimental results indicated that the surface heat transfer was enhanced with the rise of the nanofluids concentration, due to the bubble generation rate and disturbance increased. For downward facing heating, bubbles were not able to escape since the bouncy force influenced, and the vapor film appeared earlier as the increase of bubble generation rate. However, the heat transfer coefficient remains at a relatively high value in the early stage of film boiling, which suppresses the deterioration of heat transfer, due to the influences of nanofluids. As the various concentrations of nanofluids increased from 0 g/L to 0.012 g/L, it was found that the enhancement of the CHF (critical heat flux) up to 20.5%. After the nanofluids boiling, the surface roughness decreased and the wettability became worse. From the experimental phenomena, under the influence of these two factors, the bubble activity was enhanced.

In this report the expected rate of buildup of 244 Cm in a CANDU neutron flux is evaluated and used to explain cases of high 244 Cm in alpha-active samples from Bruce and Pickering Units. It is demonstrated that 244 Cm is enriched on the surface of irradiated pressure tubes where it is associated with Zr/Nb activation products. It is further shown, using 94 Nb as a fluence monitor, that the 244 Cm/( 239 Pu + 240 Pu) ratio for Bruce and Pickering irradiated pressure tube deposits exhibits a power law = 0.0042 (Fluence) 3.1982 . For non-pressure tube samples, such as feeder pipes, steam generator deposits and PHTS cruds, it is observed that Zr/Nb activation products are also associated with elevated 244 Cm activities. Thus, based on the data presented in this report, the inference is that all CANDU Units may be expected to exhibit significant levels of 244 Cm activity on PHTS surfaces, both in and out-of-core, with 244 Cm/( 239 Pu + 240 Pu) ratios significantly greater than one.

An integral analysis methodology for BWR ATWS has been developed. The method covers main scopes about ATWS events, including radiological consequence, primary system integrity, fuel integrity, containment integrity, and long-term shutdown and cooling capability. The primary techniques about this methodology were described herein. The methodology has been applied on Kuosheng nuclear power station to show the applicability. Under this framework, some suggestions were proposed for further development of this methodology. This methodology can give a way to evaluate safety of BWR plants confidently. Further, it can be a tool for developing emergency procedures about severe accidents, or exploring strategies and checking their effectiveness.

This study established an RCS-Containment coupled model that integrates the reactor coolant system (RCS) and the containment system by using the TRACE code. The coupled model was used in both short-term and long-term loss of coolant accident (LOCA) analyses. Besides, the RELAP5/CONTAN model that only contains the containment system was also developed for comparison. For short-term analysis, three kinds of LOCA scenarios were investigated: the recirculation line break (RCLB), the main steam line break (MSLB), and the feedwater line break (FWLB). For long-term analysis, the dry-well and suppression pool temperature responses of the RCLB were studied. The analysis results of RELAP5/CONTAN and TRACE models are benchmarked with those of FSAR and RELAP5/GOTHIC models, and it appears that the results of the above four models are consistent in general trends.

Anticipated Transients without Scram (ATWS) are evaluated to demonstrate that during ATWS the fuel’s integrity prevents fission product release, and that the safety features the BWR-6 plant is equipped with mitigate the ATWS event. For analyzing this scenario, a RETRAN-3D system model and hot channel model of Kuosheng NPP have been developed. From all possible events those are analyzed where the main steam isolation valves close (MSIVC) when the reactor is operating at rated power conditions. For cases with low core flows this results in higher peak cladding temperatures and thicker oxidation thicknesses. All results confirm the ATWS acceptance criteria. For sensitivity cases, increasing the boron concentration of standby liquid control system (SLCS) or injecting boron of SLCS earlier were investigated. The results show, that these measures cannot decrease the PCT value immediately, but can shorten the period of peak cladding temperature over 850 °C which results in smaller oxidation thickness.

In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95 th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.

The safe and secure designs of any nuclear power plant together with its cost-effective operation without accidents are leading the future of nuclear energy. As a result, the Reliability, Availability, Maintainability, and Safety analysis of NPP systems is the main concern for the nuclear industry. But the ability to assure that the safety-related system, structure, and components could meet the safety functions in different events to prevent the reactor core damage requires new reliability analysis methods and techniques. The Fault Tree Analysis (FTA) is one of the most widely used logic and probabilistic techniques in system reliability assessment nowadays. The Dynamic fault tree technique extends the conventional static fault tree (SFT) by considering the time requirements to model and evaluate the nuclear power plant safety systems. Thus this paper focuses on developing a new Dynamic Fault Tree for the Auxiliary Feed-water System (AFWS) in a pressurized water reactor. The proposed dynamic model achieves a more realistic and accurate representation of the AFWS safety analysis by illustrating the complex failure mechanisms including interrelated dependencies and Common Cause Failure (CCF). A Simulation tool is used to simulate the proposed dynamic fault tree model of the AFWS for the quantitative analysis. The more realistic results are useful to establish reliability cantered maintenance program in which the maintenance requirements are determined based on the achievement of system reliability goals in the most cost-effective manner.

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 k inf is calculated for different burnup profiles with control rods are either fully withdrawn or fully inserted. In the second phase k eff is calculated for different control rods insertion levels. For both phases, burnup calculations are performed for a UO 2 assembly then multiplication factor calculations of burned UO 2 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.

The change in the temperature of the nuclear reactor components (fuel, moderator, coolant, and structural materials) is considered to be a significant source of reactivity variation. This change must be taken in account during criticality calculations for safety analysis of the reactor. Hence, the exact representation of temperature in the calculations is very important. In this paper, two PWR assemblies are simulated, solid 16 ⨯ 16 and annular 12 ⨯ 12 fuel assemblies. The infinite multiplication factor and its temperature dependent parameters are calculated for both fuel assemblies. Adjusted temperature dependent libraries are created using makxsf code to exactly represent the different temperature values used in the calculations. It is shown that the results obtained using adjusted cross section libraries are more reliable. The two fuel assembly types follow the same behavior despite the differences in their geometrical configuration. The introduction of annular fuel has a very small effect on the investigated neutronic parameters because the moderator to fuel ratio is preserved.