Small Pressurized Water Reactor Research Articles

Concurrent fault diagnosis of small pressurized water reactors based on long-short term memory networks

The control systems of small pressurized water reactors (SPWR) with complex structures, compact layouts, and variable operating environment may be involved in various types of signal and concurrent faults. Concurrent faults can be taken as two or more single faults occurring simultaneously, which usually cause much larger damage to the system and are more difficult to be diagnosed than single faults. However, their fault diagnosis methods are rarely studied because of the numerous fault types and tremendous diagnostic difficulty. This paper explores the concurrent fault diagnosis method for sensors and actuators in SPWR control systems. An intelligent current fault diagnosis model is developed using long short-term memory network with the training and test datasets generated based on a fault simulation platform of the target SPWR. The test results show that both single and concurrent faults of the SPWR can be diagnosed rapidly in an average of 1.06 s after their occurrence with the classification and diagnosis accuracies reaching up to 96.61% and 97.27%, respectively. Moreover, by injecting different noise signals on the faulty dataset for training and validation, it is shown that the proposed LSTM network has strong noise immunity. This demonstrate the excellent diagnosis accuracy and efficiency of the model under both single and concurrent fault conditions. This study provides valuable guidance for the accuracy diagnosis of complex concurrent faults of nuclear power plants.

Comparison study of two control strategies for the small pressurized water reactor

The control strategy has a significant impact on the control performance of the reactor system. This paper investigates the feedwater flow and steam flow control strategies for small pressurized water reactors. First, the mathematical model of the SPWR Nuclear Steam Supply System is introduced. Second, a dual-constant control strategy was proposed, and the feedwater flow rate and steam pressure control systems under the two control modes were designed. Moreover, a steam bypass control system was designed to avoid excessive steam pressure under the feedwater flow control mode, and the reactor power and pressurize control systems were introduced. Simulation results indicate that the two control modes both meet the SPWR control requirements. However, the feedwater flow control mode has better reactor coolant temperature and steam pressure control performances during the load rejection transient, owing to the timely discharge of heat from the reactor coolant system through the steam bypass system.

A multi-objective co-optimization method of controller parameters for the overall system of small pressurized water reactor

Small pressurized water reactors (SPWRs) normally have complicated and varying working conditions, and become an important direction for nuclear energy development. The overall system of SPWR, including the reactor coolant average temperature control system, the feedwater control system of the once-through steam generator (OTSG), the steam turbine speed control system and the steam dump control system, is used to guarantee the safe and stable operation. The overall system of SPWR is designed with multiple PI controllers. To accomplish intelligent self-tuning of PI controller parameters, reduce the dependence on human engineering experience, and improve the control performance of the overall system, a multi-objective co-optimization method(MMOCO) with PI controller parameters for overall system of SPWR is proposed. Firstly, based on the coordinated control theory and PI control algorithm, the overall system of SPWR is established. Then, a MMOCO for the overall system of SPWR is designed with the Non-Dominated Sorting Genetic Algorithm-II (NSGA-II). Finally, to compare the performance of the controller parameters derived from the MMOCO and the controller parameters obtained via the engineering tuning method, step load decrease transients are selected. The results show that the ability to control the overall system of SPWR is effectively improved after applying the MMOCO.

Scale effects on core design, fuel costs, and spent fuel volume of pressurized water reactors

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.

A study on the application of artificial neural network to predict k-eff and peaking factor of a small modular PWR

Machine learning (ML) using artificial neural network (ANN) methods is being applied to predict required parameters for nuclear reactors based on learning from big data sets. The ML models usually give faster calculation speed while the accuracy is good agreement with physical simulators. In this work, a multi-layer perceptron network was built and trained to predict k-eff and peaking factor of a small modular pressurized water reactor (PWR). The results are compared with those obtained by using a reactor physics code system, i.e. SRAC2006. The comparison shows good agreement accuracy and higher performance of the ML models.

Wavy-attention network for real-time cyber-attack detection in a small modular pressurized water reactor digital control system

Global interest in advanced reactors has been reignited by recent investments in small modular reactors and micro-reactor design. The use of digital devices is essential for meeting the size and modularity requirements of small modular reactor controls. By fully digitizing the small modular reactor control systems, critical information can be obtained to optimize control, reduce costs, and extend the reactor’s lifetime. However, the potential for cyber-attacks on digital devices leaves digital control systems vulnerable. To address this risk, this study presents a novel wavy-attention network for sensor attack detection in nuclear plants. The wavy-attention network comprises stacks of batch-normalized, dilated, one-dimensional convolution neural networks, and sequential self-attention modules, superior to conventional single-layer networks on sequence classification tasks. To evaluate the proposed wavy-attention network architecture, the International Atomic Energy Agency’s Asherah Nuclear Simulator and a false data injection toolbox found in the literature, both implemented in MATLAB/SIMULINK, are utilized. This approach leverages changes in process measurements to identify and classify cyber-attacks on priority signals using the proposed wavy-attention network. Three false data injection attacks are simulated on the simulator’s pressure, temperature, and level sensors to obtain representative process measurements. The wavy-attention network is trained and validated with normal and compromised process variables obtained from the simulator. The performance of the wavy-attention network to discriminate between the reactor states using the test set shows 99% accuracy, as opposed to other baseline models such as vanilla convolution neural networks, long short-term memory networks, and bi-directional long short-term memory networks with 90%, 77%, and 91% accuracy, respectively. An ablation study is also conducted to test the contribution of each component of the proposed architecture. The theoretical framework of the proposed wavy-attention network and its implementation for nuclear reactor digital sensor attack detection are discussed in this paper.

Analysis of transient characteristics and design improvement of the passive residual heat removal system of NHR-200-II

NHR-200-II is a small integrated pressurized water reactor with 200 MW core thermal power. The core heat is transferred to two independent intermediate circuits via fourteen in-vessel primary heat exchangers (PHE), and the heat in the intermediate circuits is transferred to feedwater by two steam generators (SG) in the two intermediate circuits respectively. A passive residual heat removal (PRHR) branch is connected to each intermediate circuit to remove core decay heat under postulated accidents. During normal operation, PRHR branches are isolated by valves while SG branches in intermediate circuits are open. The valves in PRHR branches will be opened and the isolation valves of SG branches will be closed during decay heat removal scenarios. The decay heat removal capacity of NHR-200-II PRHRS could be seriously deteriorated once the isolation valves for SG branches fail to close, which was confirmed in a scaled integral test loop previously. Current understanding of PRHRS’s thermal-hydraulic characteristics with possible isolation failure in SG branches is limited. In this paper, the NHR-200-II PRHRS is modeled with RELAP5 considering the case of success and fail to isolate SG branches. A series of numerical simulations are carried out to study the impact of various parameters, such as the initial temperature, the size of the intermediate circuits’ header, and the initial flow direction in the intermediate circuits. Oscillatory flow is found when SG branches fail to be isolated under certain parameters combinations. An improved PRHRS design is purposed to eliminate possible flow oscillations, and the purposed improved design are tested by numerical simulations.

Predicting Model Biases of the APOLLO3 Neutronics Code Using Machine Learning: Toward a Multiscale and Multifidelity Approach

As part of the verification, validation, and uncertainty quantification process applied to neutronics deterministic codes, there is a requirement to expand the validation domain, especially to accommodate new third-generation reactors. The objective of the present work is to estimate the numerical biases arising from the several approximations used in deterministic codes and across different points in the phase space. Typically, this is accomplished by comparing the deterministic code to be validated with a Monte Carlo or stochastic reference code (without significant approximations). Since these reference calculations are computationally expensive, this paper proposes an alternative approach for predicting model biases of the APOLLO3® deterministic code for third-generation pressurized water reactors using machine learning algorithms. Three types of metamodels are employed (polynomial regression, kriging, and neural networks). Two scales are investigated, from a single assembly to a cluster of 3 × 3 assemblies [small two-dimensional (2-D) core], with model biases evaluated for APOLLO3 schemes with various levels of accuracy (lattice and core solvers, with high- to low-fidelity approaches). For the small 2-D core, numerical biases are observed for reactivity and power peak, representing both global and local quantities of interest. Throughout the study, the best results are achieved using kriging or neural networks, even if polynomial regression provides satisfactory predictions in some cases. The possibility of predicting biases for different quantities is also introduced. In conclusion, this paper discusses the prospects of extending the applicability of these metamodels to small and large third-generation pressurized water reactor cores, the idea being to potentially use these metamodels to support safety demonstrations for the new reactors in the long term.

Comparison of the neutronic properties of the (Th-233U)O2, (Th-233U)C, and (Th-233U)N fuels in small long-life PWR cores with 300, 400, and 500 MWth of power

Abstract The neutronic characteristics of (Th-233U)O2, (Th-233U)C, and (Th-233U)N have been compared in small long-life pressurized water reactors (PWRs). Neutronic calculations were carried out at 300 MWth, 400 MWth, and 500 MWth with two cladding types: zircaloy-4 and ZIRLO (Zr low oxygen). They were performed using the Standard Reactor Analysis Code (SRAC) and JENDL-4.0 nuclide data, dividing the reactor core into three fuel zones with varying 233U enrichment levels, ranging from 3% to 9% and fluctuating by 1%, employing the PIJ module at the fuel cell level and the CITATION module at the reactor core level. In addition, 231Pa was added as burnable poison (BP). The (Th-233U)N fuel demonstrated superior criticality compared to the other fuel types, as it consistently achieves critical conditions throughout the reactor’s operating cycle with excess reactivity <1.00% dk/k for several fuel configurations at the 300 MWth and 400 MWth power levels. Moreover, the (Th-233U)N and (Th-233U)C fuels exhibited similar and flatter power density distribution patterns compared to the (Th-233U)O2 fuel. The power peaking factor (PPF) value was relatively higher for (Th-233U)O2 fuel than the other two fuels. The (Th-233U)N fuel exhibited the most negative Doppler coefficient, followed by (Th-233U)C and (Th-233U)O2 fuels. Analysis of burnup levels revealed that the (Th-233U)O2 fuel achieved significantly higher burnup than the other two fuels.

Analysis of coolant flow distribution characteristics at core inlet of small pressurized water reactor under rolling condition

The flow distribution characteristics at core inlet directly determines the thermal and hydraulic characteristics in the core, and it is closely related to the safety of nuclear reactor. The primary coolant is subjected to additional inertial force in ocean motion, which influence the flow distribution characteristics at core inlet. In the present paper, numerical simulation method is adopted to analyze the coolant flow distribution characteristics at core inlet of the Reactor Pressure Vessel (RPV) of a Small Reactor under rolling condition. Then the results of the flow distribution characteristics were compared with those under steady state. The results show the variation range of flow distribution factors in each channel under rolling condition is within 5%. And the uneven distribution of coolant at the core inlet increases with the increase of rolling amplitude, and decrease with the rolling period in short period. Standard deviation of flow distribution factor at the core inlet changes little under rolling condition, which indicates the flow distribution at the core inlet keeps good uniformity.

Numerical simulation and optimization of In-Vessel primary heat exchangers of NHR200-II

NHR200-II is a multipurpose small integral pressurized water reactor. The primary heat exchangers of NHR200-II are in-vessel single-phase heat exchangers. The primary heat exchangers adopted a tube-in-tube bundle design to improve compactness; however, a significant weakness of the tube-in-tube bundle is the nonuniform thermal expansion between the inner and outer tubes owing to the inhomogeneous tube-temperature distribution. To minimize the thermal stress, the inner and outer tube diameters were optimized using numerical simulations. The nonuniform mass flow rates and temperature fields in each channel and tube were calculated using the optimal tube sizes, and the axially averaged temperatures of the inner and outer tubes were also calculated. The present prediction of nonuniform temperature fields can be used as the input to estimate the magnitude of thermal stress in the assessment of the structural integrity of the primary heat exchangers of NHR200-II.

Reactivity control optimization of boron-free small modular pressurized water reactor with helical-cruciform metallic fuel

Small modular pressurized water reactors have the advantages of small size, modularity, inherent and passive safety, and flexible power generation for a wider range of users and applications. Helical-cruciform metallic fuel has the potential to be applied to small modular pressurized water reactors capable of its axial self-spacing plane, high power density, increasing heat transfer area, and extending the cycle length. A novel boron-free small modular pressurized water reactor NETH-HCF175M design using helical-cruciform metallic fuel is proposed in this paper. The reactivity of the core is completely controlled by burnable absorbers and control rods. Two different geometric structures of separate pins and integral pins are researched. Results indicate that Gd2O3 is a suitable neutron absorbing candidate, while Er shows the best reactivity control effect in the integral pin structure. The separate pin structure with Gd2O3 is selected as the burnable absorber design and the core depletion result shows that the core can achieve a cycle length of about 1360 effective full power days. Control rods are divided into 60 compensating rods and 60 shutdown rods for excess reactivity controlling and shutting the core. The genetic algorithm is applied to the optimization work of layout and concentration design for burnable absorbers and layout of compensating rods. Based on the optimized burnable absorbers and compensating rod designs, excess reactivity can be controlled and the core can achieve criticality at beginning of cycle. The core can be transitioned from hot full power to cold zero power with shutdown rods at BOC, which obtains a shutdown margin more than 3000 pcm. Finally, the physical characteristics of core are analyzed, including depletion simulation, pin power distribution, and reactivity equilibrium. This paper provides an overview and information practical for the reactivity control design of novel SMPWR with high-enrichment and geometric complex helical-cruciform metallic fuel.

A coordinated control methodology for small pressurized water reactor with steam dump control system

Small pressurized water reactors (SPWRs) have complex and variable working environments with high external perturbations that challenge the operational balance and control of the primary and secondary circuits. To improve the control performance of SPWRs, a two-level hierarchy coordinated control system for SPWRs is designed in this paper. First, the equipment model of the primary and secondary circuit systems is developed. Then, based on ideal steady-state programming (ISP) with a constant coolant average temperature and a constant steam pressure, the reactor coolant average temperature, feedwater of once-through steam generator (OTSG), steam dump and steam turbine speed control systems are established by implementing PID controller. Finally, the SPWR coordinated control system is designed according to the logical relationships among the above sub-control systems, and the system is simulated and analyzed under load rejection conditions. The results show that the SPWR coordinated control system using ISP has good operational performance, and the overshoot of the coolant average temperature and steam pressure is minor under load rejection conditions, ensuring the safe operation of SPWRs.

Fuel lifetime extension for the KLT-40S small modular reactor by means of thorium-uranium fuel cycle

This paper demonstrates the results of the Monte Carlo simulation conducted with MCU-PTR software for a small modular pressurized water reactor KLT-40S. Four dispersed fuel compositions, (U238 + U235)O2, (U238 + Pu239)O2, (Th232 + U235)O2, and (Th232 + U233)O2, were examined for fuel lifetime and fuel burnup calculations.The transition to the (Th232 + U233)O2 fuel composition extends fuel lifetime by 50% and fuel burnup by 44% at design value of the fuel rod diameter. If the fuel rod diameter for (Th232 + U233)O2 fuel composition is increased to 7.2 mm, fuel lifetime is extended by 105%, and fuel burnup is extended by 38% compared to the design values.

Uncertainty quantification of once-through steam generator for nuclear steam supply system using latin hypercube sampling method

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

Control of small pressurized water reactors under single-loop operation based on MFAPC theory

Small pressurized water reactors can supply power to ships, and single-loop operation is adopted to improve maneuverability under some abnormal conditions. Large key parameter fluctuations and the potential risk of human errors occur in conventional single-loop operation. Therefore, an automatic rapid power reduction method is proposed to improve the control performance of single-loop operation. An automatic rapid power-reduction method based on conventional control systems was tested. Poor power control performance and large fluctuations in steam generator water level were observed. Reactor power and steam generator water level control systems improved by the model-free adaptive predictive control (MFAPC) theory were developed, and the parameters of the MFAPC controller were optimized by the differential evolution method. The simulation results under single-loop operation show that the improved control systems reduce the overshoot of the reactor power and effectively suppress the water level fluctuation of the steam generator.

LQG/LTR Controller Design for Power Control of Small Pressurized Water Reactors Under Four Feedback-Controlled Strategies

To improve the economy and safety of small pressurized water reactors (SPWRs) with flexible operating characteristics, the reactor power control system should process excellent robustness to provide satisfactory control performances at different operating conditions. This paper proposes four control strategies for reactor power control of SPWRs based on the linear quadratic Gaussian with loop transfer recovery (LQG/LTR) robust control method, including the single-loop reactor power feedback control (RPFC), single-loop average temperature feedback control, dual-loop feedback control, and modified dual-loop feedback control (MDFC) strategies. The corresponding LQG/LTR controllers in the reactor power control system of a SPWR were designed to assess the performance of the four control strategies. The simulation results show that the LQG/LTR controller with the MDFC strategy can provide good control performances for both reactor power and average coolant temperature among the four control strategies while the controller-based single-loop feedback control shows poor control of the reactor power or average coolant temperature. Meanwhile, compared with the existing conventional reactor power control system, the designed robust control system employing the MDFC strategy can provide better control performance for the reactor power and average coolant temperature in full-power operation of 100% to 90% rated power and low-power operation of 25% to 35% rated power with the differential control rod worth taken as 4 pcm/step and 24 pcm/step, indicating its effectiveness and superiority.

Study on the influence of source term release on accident consequences in severe accidents for small pressurized water reactor

In this paper, a small PWR with no pressure vessel external cooling system (ex-vessels) is used as the research object. It is found that for the small low burnup marine reactor without ex-vessels, the more core releases at an early stage, the later core damage occurs. Considering in the case of large break loss of coolant and loss of power accident, and the ORIGEN program is coupled with the MELCOR program is used as a severe accident process simulation tool. The significant parameters have been took into account, include core melting time, core damage rate, and lower head failure time, etc. It could be found through calculation that, for a certain reactor, in case of the reactor core meltdown happened, different released amount of source term will results in a difference of core heat release in the pressure vessel. For the small reactors without ex-vessels, the amount difference of source term release can reach 20%, and the lower head failure time can be delayed by 70%.

Computational model for thermohydraulic analysis of an integral pressurized water reactor with mixed oxide fuel (Th, Pu)O2

The use of advanced generation III+ and IV nuclear reactors, and their applications, has become important, seen as a means capable of contributing to the global transition to more sustainable, affordable and reliable energy systems. This technology, which could be integrated into future carbon-free electric power generation systems with high proportions of different renewable energy sources, includes Small Modular Reactors (SMR). There are about 100 different proposed projects of Generation III+ and IV, of which about 50 are SMR concepts, in various stages of development and of different types of technologies. Other important issues for achieving the long-term sustainability of nuclear energy are the proper use of its fuel sources and the improvement of nuclear waste management. Therefore, fuels based on a mixture of oxides have been used successfully in several countries. In addition, the incorporation of thorium-based fuel is a current challenge for the new designs of advanced reactors. The present paper focuses on the analysis of a small modular integral pressurized water reactor (iPWR) with Thorium-Uranium Oxide (Th-U MOX) mixtures. A thermohydraulic model is developed using the Ansys CFX program, which allows the calculation of the temperature distribution in the section where the highest power is produced within the SMR IPWR core (critical section). The temperature distributions in the fuel, clad and coolant were calculated with the objective of verifying that they were within the safety limits.

Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutronic analysis has been carried out. PWR in module is one type of reactor that can be utilized because of its small size so that it can be placed on demand. Neutronic calculations were performed using SRAC version 2006, the data library using JENDL 4.0. The first calculation was fuel pin (PIJ) calculation with hexagonal fuel pin cell type. And the second calculation was reactor core (CITATION) calculation using homogeneous and heterogeneous core configurations. ThN and UN fuels use heterogeneous configurations with 3 fuel variations. The reactor geometry was used in two fuels are the same, with diameter and height active core was 300 cm and 100 cm. In this research, Np-237 was added as a minor actinide in the UN fuel to reduce the amount of Np-237 in the world and also reduce the k-eff value. For ThN fuel, Pa-231 also added in the fuel to reduce the k-eff value. The optimum configuration of UN fuel reached when used heterogeneous core configuration case four with percentage of U-235 in F1=5.5 %, F2=7 % and F3=8.5 % also with the addition of Np-237 0.2 % and fuel fraction 56 %. It has a maximum excess reactivity value 12.56 % %∆k/k. And then, the optimum configuration of ThN fuel reached when used heterogeneous core configuration case three with percentage of U-233 in F1=2 %, F2=4 % and F3=6 % with the addition of Pa-231 0.5 % and fuel fraction 53 %. It has a maximum excess reactivity value 7.67 % %∆k/k. The comparison of optimum design of UN and ThN fuel shows that the ThN fuel has the k-eff value closer to critical than UN fuel. Therefore, in this study, ThN fuel is more suitable for use in PWR reactors because it has a small excess value and can operate for 10 years without refueling