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[en] Highlights: • A flux reconstruction method is presented that uses a 3D transport theory form factor. • 3D form factor is a 2D xy-plane component times an approximate 1D z-axis component. • Method is used to simulate travelling flux detector scan (TFD scan) readings. - Abstract: Even with current computing capabilities, detailed full core three-dimensional (3-D) transport calculations are still not practical. However, if we are satisfied with knowing only the average values of spatial flux distributions, the 3-D diffusion solution will constitute the final solution. On the other hand, in reactor design and safety analysis, direct information about the local flux distribution for the heterogeneous assemblies is required to assess the design and determine the safety margins. For this reason, after having solved the full-reactor-core problem, we have to look into the possibilities of recovering in a second step the information on local properties of single heterogeneous assemblies. In particular, the detector readings at detector locations are derived using these global homogenized parameters by applying appropriate numerical methods such as advanced interpolations. In this paper, we propose a method based on flux reconstruction to calculate the simulated detector readings in three-dimensions with high fidelity. Data from detector readings are very important in ensuring optimal reactor operations as well as in detecting any deviations from normal operations. Thus, calculating the detector readings with high fidelity will allow improvements to operating and safety margins. To validate this method, comparisons between detector reading simulation results and measurements from an operating CANDU reactor will be conducted and results will be presented.
[en] Highlights: • An efficient MC method for the sensitivity calculation of reactivity coefficients is developed. • The sensitivity of reactivity coefficient is calculated by MC second-order perturbation techniques. • Its effectiveness is examined in a two-group homogeneous problem and Godiva. • S/U analyses are performed for MDC of a LWR pin cell and FTC of a CANDU 6 lattice model. - Abstract: The uncertainty quantification of the reactivity coefficients such as the fuel temperature coefficient (FTC) and the moderator density coefficient (MDC) is crucial for the nuclear reactor safety margin evaluation. This paper proposes a continuous-energy MC second-order perturbation (MC2P) method as a new way to estimate efficiently the sensitivity of reactivity coefficients to nuclear cross section data. The proposed MC2P method takes into account the second-order effects of the fission operator and the fission source distribution. The effectiveness of the MC2P method implemented in a Seoul National University MC code, McCARD, is demonstrated in a Godiva 235U density coefficient problem via comparison of its results with direct subtraction MC calculation. It is shown that the new method can predict the cross section sensitivities of the reactivity coefficient more accurately even with much smaller number of MC history simulations than the direct subtraction MC method. It is also shown that the proposed method is applicable for quantifying the uncertainties of the MDC of a LWR pin cell problem and the FTC of a CANDU 6 lattice cell problem due to the uncertainties of the nuclear cross section input data represented by nuclear cross section covariance data.
[en] Highlights: • Wilks’ method of uncertainty quantification was confirmed with a realistic problem. • Thermal-hydraulics modeling of a BWR spray cooling licensing experiment was used. • Critical points of the method (input distribution, sidedness, order) were assessed. - Abstract: Wilks’ formula has been frequently used to quantify the minimum amount of computational work required to meaningfully assess a model’s uncertainty, due to its nonparametric statistical nature that does not require knowledge of the distribution of the qualifying parameters of interest. Additionally, this method allows for any number of input uncertain parameters in the simulation model. This is favorable due to considerable computational expense of typical nuclear safety simulations, providing a quantifiable number of code executions that can statistically verify a desired level of safety. However, there are various existing definitions and uses of Wilks’ theorem in such scenarios, which the present study aims to investigate and quantify for a real thermal-hydraulics experiment used for reactor safety licensing. In this work, the U.S. NRC TRACE thermal-hydraulics code was chosen to simulate the separate-effect spray cooling tests performed by ASEA-ATOM for licensing of BWR SVEA-64 fuel. The computational model was evaluated by performing forward uncertainty quantification (UQ) using Dakota as the analysis tool and code driver on 31 identified sensitive parameters. Using this validated model, the TRACE model was sampled 1000 times with four different input parameter probability distributions to produce four model data sets to assess the applicability of Wilks’ theorem within a realistic nuclear safety analysis scenario. The obtained results compared various Wilks-defined ‘sample sizes’ according to one-sided confidence intervals for the 1st, 2nd and 3rd-order statistics, and with the two-sided confidence interval for the 1st-order statistics. The comparison demonstrated that Wilks’ method satisfies the reactor safety modeling requirements at the 95%/95% tolerance/confidence level as determined by the U.S. NRC.
[en] Highlights: •Inner reflector improves reactor safety feature. •Annular cylinder reactor is safer than solid cylinder one. •Uniform power density profile throughout a core enhances the passive safety margin. •Longer life cycle is assured for annular reactor compared to solid cylinder one. -- Abstract: We studied the capability of an annular, prismatic HTGR to remove decay heat passively. The purpose of the study was to obtain the design parameters relationship of the annular, prismatic HTGR with passive decay heat removal depending on power density profile and to compare them with those for solid cylinder one. The results showed that the safety feature of the annular reactor is improved a lot compared with that of cylinder one. The safety margin could be increased further by flattening the power density profile. Then fundamental neutronic analysis was performed for the annular reactor whose design parameters are obtained from the condition.
[en] Highlights: • The law of dynamic distribution of the heat and flow load are studied using three experiments. • The strain response of the pressure vessel to temperature and pressure load is obtained. • The frequency range of strong impact on structural response is obtained. - Abstract: An emergency core cooling system for a pressurized water reactor adopts direct safety injection with reactor pressure vessel. In this design, a special flow guide device is introduced in order to minimize the heat effects on the reactor internals. But this design makes the pressure vessel bear stronger heat and current impact. To investigate the dynamic response of the pressure vessel during safety injection, the law of dynamic distribution of the heat and flow load are studied using three experiments: In the visualization experiment, the relationship between injection condition and distribution pattern in the downcomer is obtained. In the heat mixing experiment, measuring the temperature and pressure near the inner wall of the pressure vessel, enables us to find out the law that governs dynamic distribution of the heat and pressure load as well as the main distribution area of these loads, and analyze how the temperature oscillation generated. In the structural response experiment, the strain response of the pressure vessel to temperature and pressure is obtained. Moreover, the frequency range of its response to hot oscillation under safety injection is also obtained by analysis. This study provides support to recognize the action law of heat, pressure and structural response in the reactor during safety injection.
[en] Highlights: • The severe accident issues of LBE-cooled reactors are summarized. • Key processes and phenomena during CDAs of LBE-cooled reactors are discussed. • The systematic framework of severe accident safety analysis of LBE-cooled reactors is given. - Abstract: The safety analysis work is always a key issue for the nuclear power plant licensing. After Fukushima Accident of Japan, the severe accident safety analysis has been more concerned, which would be also very important and essential for the Research and Development (R&D) work of LBE-cooled reactors. So far, a great number of studies on severe accident of LBE-cooled reactors have been carried out. In this paper, a summary of severe accident of LBE-cooled reactors is conducted, which contains almost all the relevant issues on the hypothetical Core Disruptive Accidents (CDAs). Detailed contents of these issues are discussed. The systematic framework of severe accident safety analysis of LBE-cooled reactors is preliminarily given, which aims at providing a useful reference for the further safety analysis research work of this kind of advanced reactor.
[en] The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process and to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.
[en] Highlights: • This work analyzed the models that governing the rewetting phenomenon in PWR reactor. • An analysis of temperature evolution and of reweting front was made for two experiments. • The results obtained with REWET code were compared with the experimental results. • The result simulated for the rewetting time came closer to the experimental results. - Abstract: The safety of nuclear power plants is determined by their protection against the possible outcomes of postulated accidents. One of the most important accidents is the loss of coolant in the core (Loss-of-Coolant Accident – LOCA). A process of fundamental importance in the event of a LOCA in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Centre (CDTN) has been developing programmes since the 1970s to make Brazil independent in the field of reactor safety analysis. To that end, a Rewetting Test Facility (ITR in Portuguese) was designed, assembled and commissioned. This facility aims at investigating the phenomena involved in the thermal hydraulic reflood phase of a LOCA in a PWR nuclear reactor. The objective of this work is the analysis of physical and mathematical models that govern the rewetting phenomenon. Thus, a simulation code was developed for the Rewetting Test Facility (ITR), which represents a PWR core cooling channels. The thermohydraulic code was named Rewet. The results obtained with Rewet code were compared with the experimental results of the ITR and with the results of the Hydroflut code, the program used until then. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the rewetting front for two typical tests using the two-code calculation and experimental results. In all cases, there was a better adjustment of the Rewet results in relation to those of the Hydroflut. The result simulated by the Rewet code for the rewetting time also came closer to the experimental results more than that obtained with the Hydroflut code.
[en] Highlights: •The multiple redundant controller, SPLC is configured as the combination of DMR and TMR architecture. •We construct the Markov model of SPLC using the concept of the system unavailability rate. •To satisfy the availability requirement of safety grade controller, the fault coverage factor (FCF) should be ≥0.8 and the MTTR of each module should be ≤100 h when FCF is 0.9. •The availability of SPLC is better than that of PLC having iTMR architecture however it is poorer than iTMR considering the off-line test and inspection on the assumption that MTTR of each module is ≤200 h. -- Abstract: We analyze the availability of the Safety Programmable Logic Controller (SPLC) having multiple redundant architectures. In the SPLC, input/output and processor module are configured as triple modular redundancy (TMR), and backplane bus, power and communication modules are configured as dual modular redundancy (DMR). The voting logics for redundant architectures are based on the forwarding error detection. It means that the receivers perform the voting logics based on the status information of transmitters. To analyze the availability of SPLC, we construct the Markov model and simplify the model adopting the system unavailability rate. The results show that the fault coverage factor should be ≥0.8 and Mean Time To Repair (MTTR) should be ≤100 h in order to satisfy the requirement that the availability of the safety grade PLC should be ≥0.995. Also we evaluate the availability of SPLC comparing to other PLCs such as simplex, processor DMR (pDMR) and independent TMR (iTMR) PLCs used in the existing nuclear safety systems. The availability of SPLC is higher than those of the simplex, pDMR but is lower than that of iTMR for one month which is the periodic off-line test and inspection. That’s why the number of redundant modules used in PLC is more dominant to increasing the availability than the number of fault masking methods such as voting logics used in PLC on the assumption that operation time is in the early stage. But the availability of iTMR, which has many redundant modules but has only a voting logic fast decrease and eventually is the lowest after 8000 h. Also if the MTTR of each module in PLC is required to be increased to 200 h, the availability of SPLC would be better than iTMR.
[en] Highlights: • Effect of gas phase on debris bed formation behavior investigated experimentally using gas-injection. • Increasing gas velocity leads to enhanced pool convection and weakened role of particle inertia. • Effect of gas phase on regime boundary verified. • Influence of gas flow on bed geometric properties confirmed. - Abstract: Studies on debris bed formation behavior are important for the improved evaluation of Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). To clarify the flow-regime characteristics underlying this behavior, in recent years a series of simulated experiments was performed at the Sun Yat-sen University by discharging various solid particles into Two-Dimensional (2D) water pools. Based on the experimental observation, it is found that, due to the different interaction mechanisms between solid particles and water pool, four kinds of regimes, termed respectively as the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertia dominant regime, are identifiable. In this work, aimed at providing some evidence for understanding the effect of coolant boiling on the regime transition, a number of new experiments are performed by percolating nitrogen gas uniformly through the water pool during the particle sedimentation. It is recognized that, possibly caused by the enhanced pool convection as well as the weakened role of particle inertia, increasing the gas velocity are confirmable to have an evident impact on the regime transition. On the other hand, even for the cases without regime transition, the gas flow injected is also verifiable to have a great influence on the particle-bed properties (e.g. specific geometric angles), regardless what regime it is. Knowledge and evidence from our work might be utilized for future development of a general model directly applicable for reactor safety analyses as well as the verifications of SFR severe accident analysis codes in China.