Core Safety Parameters Research Articles

Full Core Criticality Modeling of Gas-Cooled Fast Reactor using the SCALE6.0 and MCNP5 Code Packages

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals as the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working environment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 with the results of the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analyzed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be useful technique to monitor both fission source convergence (stationarity) and core eigenvalue convergence (keff) to fundamental eigenmode with MCNP5. All calculations were done with ENDF/B-VII.0 library. The obtained results showed high similarity with reference results.

Impact of core power density on economics of a small integral PWR

Increasing the core power density of nuclear reactors allows higher power generation from the same volume, reducing the cost of the plant. The lower power rating and higher degree of inherent safety for small modular reactors motivate exploring the highest achievable power density to improve their economics and overcome lack of economy of scale. Given the urgency to deploy carbon free energy sources in the near term, the current nuclear fuel supply must be adopted. Assuming standard 17x17 pressurized water reactor (PWR) fuel assemblies with enrichment below 5%, a parametric study on core power density of an integral PWR (iPWR) is performed. The reference core design is rated at 60 kW/L and composed of 77 reduced-height fuel assemblies. The reactivity control during normal operation is assumed to utilize soluble boron and fuel burnable poisons. The commercial STUDSVIK code package is used to design cores ranging from 20 kW/L to 170 kW/L. This range results in respective output powers between 50 and 420 MWe for an assumed 32% plant thermal efficiency. Core safety parameters, such as peaking factors, Minimum Departure from Nucleate Boiling Ratio and maximum fuel temperatures are evaluated for all designs. Simplified economic evaluation of lifetime-levelized unit costs for the fuel cycle, operation and maintenance and capital expenditures was performed as a function of power density. The relatively low core discharge burnup (32 MWD/kgU) leads to high lifetime-levelized fuel cycle unit cost for the reference core. However, increase in power density allows significant reduction in operation, maintenance and capital unit costs. Doubling power density resulted in 80% reduction in total levelized cost of electricity. The unique features of a forced circulation iPWR technology combined with compact steam generator technology could allow achieving high power densities (greater than 125 kW/L). Specific examples on relationship of cost and safety margins are also discussed.

Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

The objective of this paper is to analyze the ability of a VVER-1000 core and its control system to cope with a hypothetical main steam line break (MSLB) accident in case of multiple equipment failures. The study involves the use of advanced 3D core calculation models benchmarked and validated for reactivity accidents in preceding studies. A MSLB core boundary condition problem is solved on a coarse (nodal) mesh with the coupled COBAYA/CTF neutronic/thermal hydraulic codes. The core thermal-hydraulic boundary conditions are obtained from a preceding full-plant MSLB simulation. The assessment of the core safety parameters is supplemented by a fine-mesh (sub-channel) thermal-hydraulic analysis of the hottest assemblies with the CTF code using information from the 3D nodal COBAYA/CTF calculations. Thirteen variants of a pessimistic MSLB scenario are considered, each of them assuming a number of equipment failures aggravated by eight control rods stuck out of the core after scram at different locations in the overcooled sector. The results (within the limitations of the adopted modeling assumptions) show that the core safety parameters do not exceed the safety limits in the simulated aggravated reactivity accidents.

Predication of the SMR Critical Core Performance Under Zero Power

In Nuclear Power plants, Reactivity Induced Accidents can lead to sever accidents. Rod Ejection Accidents are part of Reactivity Induced Accidents that induced through driven by reactivity insertion due to many failures. Thus, safety analysis of core behaviour under many external rod reactivities in Nuclear power plants are mandatory by regulators or safety authorities. In this research, a new dynamic model is proposed for core safety analysis under Rod Ejection Accidents. Thermal Power and other core parameters predictions are the most important goals for any reactor operation policy, during all periods and specifically at zero power to avoid severe accidents. The proposed model involves of a point kinetics explanation of neutronics combined with thermal hydraulic dynamics in the reactor core to predict its variation of parameters during transients using MATLAB environment. The proposed model is validated through comparing with the transient dynamic responses obtained through previous research for a chosen design of NuScale small modular reactor. In addition, the proposed model is verified through determining the dynamic reactor responses of Rod Ejection Accidents at hot zero power with many perturbations of different control rod ejection. The Performed safety analysis results of validation and the verification demonstrate that, the proposed model represents the reactor core behavior during the rod ejection transients with good prediction of thermal power of core peaks. Moreover, it allowed large explorations of core safety parameters and predicting the performance of its rector core during Rod Ejection Accidents under critical Hot zero power.

Neutronic and thermal-hydraulic aspects of loading pattern optimization during the first cycle of VVER-1000 reactor using Polar Bear Optimization method

Loading Pattern Optimization (LPO) is an extremely important issue for the safe and economic operation at nuclear power plants because the core configuration affects many of the neutronic and thermal-hydraulic parameters (NTPs) of the reactor. Most researches were done in this area limited to employing different heuristic optimizing algorithm and the neutronic aspects at the beginning of the cycle (BOC). In this research, a comprehensive code package has been developed using coupling the PARCS code and the COBRA-EN code with the new Polar Bear Optimization (PBO) algorithm, which tries to find an optimal loading pattern with maximized cycle length, maximized departure from the nucleate boiling ratio (DNBR), and flatter power distribution. Other core safety parameters such as maximum boric acid concentration at the BOC, maximum power peaking factor (PPF), the maximum temperature of the fuel center and the maximum clad surface temperature are considered as penalties for accepting or rejecting optimal configuration alongside the fitness function. This code Package which named PBLPO (Polar Bear Loading Pattern Optimization) code, was implemented for the core of VVER-1000 reactor in Bushehr Nuclear Power Plant (BNPP). The optimized pattern that found with PBLPO improved safety parameters comparing to loading pattern that the proposed in the Final Safety Analysis Report (FSAR) and shows that the polar bear algorithm has a good performance in LPO.

Calculation of core safety parameters and uncertainty analyses during unprotected transients for the ALLEGRO and a sodium-cooled fast reactor

The calculation of the essential core safety parameters in fast spectrum reactors are loaded with large uncertainties due to the particularly large uncertainty of nuclear data, modelling uncertainties and possible mistakes by users. These core safety parameters such as coefficients of reactivity, power peaking factors and point-kinetics parameters play an important role during unprotected transients. The contributions to the uncertainties of the target parameters (e.g. peak cladding temperature) from the core safety parameters can be determined by uncertainty and sensitivity analyses.Three-dimensional models of the ALLEGRO (demonstrator gas-cooled fast reactor) and a sodium-cooled fast reactor (SFR) concept were created in the Serpent Monte Carlo code, and several coefficients of reactivity were calculated. Furthermore, thermal hydraulic analyses with point kinetics were made for the selected transients by the ATHLET code system. An unprotected loss of flow and an unprotected overpower transient were studied at first using the best estimate approach to understand the behaviour of the reactors in these scenarios. In addition, uncertainty analyses were performed applying the GRS method, considering the uncertainties of the coefficients of reactivity, which were taken from benchmark results. Finally, the importance of the uncertain parameters was determined from sensitivity measurement.The results show that the thermal expansion of the core structural elements has a significant effect on the reactivity of these fast spectrum reactors. Moreover, it was found that during unprotected overpower transients the most important uncertain parameters are the Doppler coefficient and the fuel thermal expansion coefficient of reactivity. In addition, the large uncertainty of the positive coolant and cladding temperature coefficients of reactivity have a significant influence on the uncertainty of the calculated peak cladding temperature, especially in the case of unprotected loss of flow transients. The results imply that the uncertainty of some core safety parameters should be reduced. Lastly, the results adumbrate that additional safety shutdown systems are to be applied to mitigate the consequences of unprotected transients.

Nuclear Reactors Safety Core Parameters Prediction using Artificial Neural Networks

The present work investigates an appropriate algorithm based on Multilayer Perceptron Neural Network (MPNN), Apriori association rules and Particle Swarm Optimization (PSO) models for predicting two significant core safety parameters; the multiplication factor Keff and the power peaking factor Pmax of the benchmark 10 MW IAEA LEU research reactor. It provides a comprehensive analytic method for establishing an Artificial Neural Network (ANN) with self-organizing architecture by finding an optimal number of hidden layers and their neurons, a less number of effective features of data set and the most appropriate topology for internal connections. The performance of the proposed algorithm is evaluated using the 2-Dimensional neutronic diffusion code MUDICO-2D to obtain the data required for the training of the neural networks. Experimental results demonstrate the effectiveness and the notability of the proposed algorithm comparing with Trainlm-LM, quasi-Newton (Trainbfg-BFGS), and Resilient Propagation (trainrp-RPROP) algorithms.

Neutronic and fuel cycle comparison of uranium and thorium as matrix for minor actinides bearing-blankets

Minor actinides transmutation is one of the three main axes defined by the 2006 French law for nuclear waste management, along with long-term storage and use of a deep geological repository. In the heterogeneous approach, minor actinides are loaded in specially designed targets assemblies which are located in the periphery of the core, in order to limit the impacts on core operations. In this paper, we compare the use of uranium and thorium dioxide as support matrix in which minor actinides are diluted in the target assemblies. Both UO2 and ThO2 exhibit sufficiently good irradiation behavior to withstand the long residence time associated with heterogeneous transmutation. Five different reprocessing strategies are compared in which some or all the elements in the blankets are reused after reprocessing. The impacts on core safety parameters and fuel cycle parameters are also evaluated for each case and it is found that using thorium as support matrix with reuse of uranium 233 leads to transmutation performances similar to the one obtained with the reuse of plutonium from uranium blankets with slightly lower global impacts on reactor operation and fuel cycle.

Design Studies of a Low Sodium Void Reactivity Core Able to Accommodate Degraded TRU Fuel

As part of the Fast Reactor Cycle Technology Development (FaCT) Project, JSFR (Japan Sodium-Cooled Fast Reactor) core design efforts have been made to cope with the transuranic (TRU) fuel compositions expected during the light water reactor (LWR)–to–fast breeder reactor transition period, during which various kinds of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and their recycling method. The sodium void reactivity, which is one of the major core safety parameters, is considerably influenced by TRU fuel compositions. The criteria assigned to the JSFR core include a void reactivity effect limited to ∼6 $; therefore, designing a core with reduced sodium void reactivity will offer a greater margin for the core to host TRU fuel.To this end, a new core concept called BUMPY is proposed. This homogeneous core exhibits a low sodium void reactivity, due to partial-length fuels with an upper sodium plenum interspersed within the core, among other standard fuel assemblies. This core configuration enhances the upward and lateral neutron leakage from the core fuel region toward the sodium plenum when voiding to reduce void reactivity. The BUMPY core is applied to the 750-MW(electric) JSFR core design. The core can meet the design target by adjusting the loading fraction of the partial-length fuels and the height of the step in fuel lengths. The calculated void reactivity of the selected BUMPY core is 2.5 $ (25% loading fraction, 30-cm step height), which is considerably reduced from the 5.3 $ value of the reference core. This allows the BUMPY core to accommodate 5% to 9% more minor actinides in the core compared to the reference core.

Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix

An uncertainty reduction method for core safety parameters, for which measurement values are not obtained, is proposed. We empirically recognize that there exist some correlations among the prediction errors of core safety parameters, e.g., a correlation between the control rod worth and the assembly relative power at corresponding position. Correlations of errors among core safety parameters are theoretically estimated using the covariance of cross sections and sensitivity coefficients of core parameters. The estimated correlations of errors among core safety parameters are verified through the direct Monte Carlo sampling method. Once the correlation of errors among core safety parameters is known, we can estimate the uncertainty of a safety parameter for which measurement value is not obtained.

A conceptual design study of proliferation-resistant PWR fuel using reprocessed uranium

A conceptual design study was carried out to enhance proliferation-resistant nature of current light water reactor fuels. Main features of the proliferation-resistant fuel design are adoption of alloy instead of oxide and utilization of enriched reprocessed uranium (∼10 wt% 235U). Major dimensions of the fuel assembly were not changed because of thermal-hydraulic considerations and back-fittability to current PWRs. Its smaller 238U inventory reduces generation of plutonium and 236U in the reprocessed uranium promotes generation of 238Pu that has large decay heat. The assembly calculation results of the fuel indicated that the fuel has good proliferation-resistant nature in the viewpoint of decreased plutonium generation, worse plutonium composition and increased decay heat. Neutronic analyses of an equilibrium core loaded with the proliferation-resistant fuels were carried out and calculation results indicate that variations of major core safety parameters are not very large. Therefore, basic feasibility of the proliferation-resistant fuel design using reprocessed uranium was confirmed in the course of this study.

Application of artificial neural network for safety core parameters prediction in LWRRS

This paper reports on the use of Artificial Neural Networks (ANNs) in predicting core safety parameters, such as multiplication factor K eff and fuel powers peaks P max, in Light Water Research Reactors (LWRRs) using a personal computer. The purpose of this emulation, when it is associated with optimal incore fuel management, is to allow rapid and extensive exploration of potential good configurations of fuel in LWRRs cores. The key idea of the developed program lies in the use of an adaptive learning rate procedure in a typical back-propagation error model. The performance of the algorithm was improved to the proper setting of the learning rate. Which is adjusted according to the minimal error predicted. Using the 2-Dimensional neutronic diffusion code MUDICO-2D for the referenced target values, the neural network development process is reported for a slight modified version of the IAEA 10Mw benchmark LEU core.

Fast Reactor Core Concepts for Minor Actinide Transmutation Using Hydride Fuel Targets

Fast reactor core concepts are studied which reduce long-term radiotoxicity of nuclear waste by using minor actinides (MAs) in the form of zirconium-hydride fuel targets. A systematic parameter survey is carried out to investigate the fundamental characteristics of MA transmutation and the core safety parameters such as sodium void reactivity in a 1,000 MWe-class fast reactor core. Two core concepts are proposed, using 36 target assemblies, by adjusting the composition of hydride fuels. One is the MA burner core to transmute a large amount of MAs in a short time combined with Pu multi-recycling in fast reactors, whereby the MAs produced in about 13 LWRs can be transmuted every year with a 58% MA transmutation rate in discharged targets. The other is the MA once-through core to incinerate a small amount of MAs by fission, whereby the MAs produced in about 2 LWRs can be incinerated every year with a 64% MA incineration rate (a 93% transmutation rate) in discharged targets. This study shows these concepts have great potential to achieve good transmutation characteristics of MAs while providing the improved safety characteristics of a fast reactor core.