Small Pressurized Water Reactor Research Articles

Robust power control design for a small pressurized water reactor using an H infinity mixed sensitivity method

The objective of this study is to design a robust power control system for a small pressurized water reactor (PWR) to achieve stable power operations under conditions of external disturbances and internal model uncertainties. For this purpose, the multiple-input multiple-output transfer function models of the reactor core at five power levels are derived from point reactor kinetics equations and the Mann's thermodynamic model. Using the transfer function models, five local reactor power controllers are designed using an H infinity (H∞) mixed sensitivity method to minimize the core power disturbance under various uncertainties at the five power levels, respectively. Then a multimodel approach with triangular membership functions is employed to integrate the five local controllers into a multimodel robust control system that is applicable for the entire power range. The performance of the robust power system is assessed against 10% of full power (FP) step load increase transients with coolant inlet temperature disturbances at different power levels and large-scope, rapid ramp load change transient. The simulation results show that the robust control system could maintain satisfactory control performance and good robustness of the reactor under external disturbances and internal model uncertainties, demonstrating the effective of the robust power control design.

Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

Conceptual design of a small-pressurized water reactor using the AP1000 fuel assembly design

This paper presents the conceptual design of a 300 MWt small modular reactor (SMR)using fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system and the JENDL-4.0 data library. The analysis showed that Doppler, moderator temperature, void, and power reactivity coefficients are all negative over the core lifetime. Semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel andclad surface temperature. The minimum departure from nucleate boiling ratio (MDNBR) is also calculated. The results indicate that a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

Mathematical modeling of a pressurizer in a pressurized water reactor for control design

• A nonequilibrium nonlinear three-region pressurizer model was proposed and linearized. • Transfer function models of pressurizer during insurge and outsurge transients were developed for controller design . • Pressure and water level controllers of a small PWR pressurizer were designed based on the transfer function models. • Simulation studies show effectiveness of the pressurizer models and robustness of the designed controllers. A pressurizer is a key equipment to ensure the safe operation of a pressurized water reactor by maintaining the reactor coolant system pressure within allowed tolerances. In this paper, a nonequilibrium three-region pressurizer model for the pressurized water reactor was developed first base on the basic conservation laws of mass and energy for the steam and water in the pressurizer. Then the nonlinear model was linearized to introduce the transfer function models of the pressurizer during insurge and outsurge transients for the controller design of a small pressurized water reactor pressurizer. Based on the developed transfer function models, the closed-loop pressurizer pressure and water level control systems including conventional proportional-integral-derivative controllers were designed employing the control strategy that uses a spray valve and two electric heaters for pressure control and regulates the charging flow rate with the letdown flow rate keeping constant for water level control. Finally, a simulation platform for the small pressurized water reactor pressurizer was developed in MATLAB/Simulink with implementation of the proposed nonlinear and nonequilibrium three-region pressurizer model, a widely-used nonlinear and nonequilibrium three-region pressurizer model and the designed pressure and water level controllers. Two typical load change transients, including the 100% to 90% of full power step load decrease and load rejection from 100% to 25% of full power, were simulated based on the platform. Dynamic responses of the pressure and water level obtained using the two nonlinear pressurizer models were analyzed, and the controller performances were assessed. The analysis and assessment results have shown satisfactory control performance and good robustness of the designed controllers regardless of the pressurizer simulation model adopted, demonstrating the feasibility, effectiveness and accuracy of the developed nonlinear and transfer function models of the pressurizer for dynamic simulation and controller design.

Feasibility of passive autonomous frequency control operation in a Soluble-Boron-Free small PWR

Nuclear power plants should be capable of contributing safely and flexibly to the electrical grid stability. This research deliberates the feasibility of a new innovative concept of a passive and autonomous frequency control operation (PAFO) in soluble-boron-free (SBF) pressurized water reactors (PWRs). The new concept eliminates any required active control in the reactor core for attaining primary and secondary frequency control operational modes in PWRs. In particular, control rods and soluble-boron are not utilized in a PAFO. Meanwhile, the reactor power maneuvering is achieved entirely by the passive reactor feedback to the coolant temperature change. This approach is motivated by the strongly negative coolant temperature coefficient available in an SBF coolant system. In order to ascertain judiciously the practical viability of the proposed concept, numerical simulations of passive primary and secondary frequency operations are performed using a lumped PWR system model solved by an in-house FORTRAN-95 computer code. The various simulation parameters are obtained from 3D Monte Carlo simulations of an SBF 450 MWt PWR by using Serpent-2 code with ENDF/B-7.1 data library. The numerical simulations are carried out at the beginning, middle and end of the fuel cycle to emphasize the effect of fuel burnup on the proposed PAFO. Moreover, sensitivity analyses on xenon worth and nominal fuel temperature are performed to deliberate their impacts in tailoring the reactor performance. Because the reactor power matches well with the power demand with a small variation in coolant temperature during the frequency control, we conclude that there is promising potential for the proposed PAFO concept.

The excess reactivity management in small pressurized water reactor utilizing fully ceramic microencapsulated fuel

After Fukushima Daiichi nuclear disaster, different techniques to overcome the weaknesses of pressurized water reactor (PWR) were developed, such as a change of the cladding and fuel materials. The Fully Ceramic Microencapsulated (FCM) fuel is one of those potential materials to replace traditional UO2 pellets fuel to make PWR safer.In this paper, a neutronics study has been performed to apply burnable poison (BP) for an excess reactivity management in a 200 MWt small PWR adopted soluble boron free (SBF) concept. In this core design, fuel enrichment is set as 19.9 w/o to get long enough cycle length due to the low packing faction of FCM fuels. As a result, the excess reactivity control of the core would be huge at the beginning of core life. In order to suppress excess reactivity efficiently, burnable poison such as Pu-238, Integral Fuel Burnable Absorber (IFBA) and Wet Annular Burnable Absorber (WABA) were evaluated for a cycle in terms of the major core performance parameters such as the suppressing excess reactivity, shutdown margin, and core cycle length. MCNP4C and ORIGEN were used as a tool to perform the results. It has been demonstrated that the burnup reactivity swing over a long cycle period can be reduced from 50,000 pcm to about 20,000 pcm due to utilizing different type burnable poison. It is shown that combinations of burnable poison with control rods make it possible to reduce excess reactivity and get enough shutdown margin during the core life for soluble boron free operation.

Development of a computer code for thermal–hydraulic design and analysis of helically coiled tube once-through steam generator

Abstract The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal–hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal–hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

An innovative core design for a soluble-boron-free small pressurized water reactor

This research investigates the neutronic feasibility of a high-performance soluble-boron-free (SBF) small modular reactor (SMR) core based on a new burnable absorber concept called the “Burnable Absorber-Integrated Guide Thimble” (BigT). Three unique BigT designs were loaded into the core; each BigT design was judiciously ascertained from the core radial power profile to tailor the required reactivity depletion patterns for an SBF operation. The approach is demonstrated to work well as the SBF SMR design exceeds the targeted cycle length while successfully controlling its burnup reactivity swing between 634 and 800 pcm throughout most of its operation. This paper also describes the use of hafnium-doped stainless steel as mechanical shim (MS) rods to attain the core criticality. Because the worth of the MS rods is relatively small, insertion and withdrawal of the rods during operation hardly alter the core radial power distributions. The resulting axial power profile, meanwhile, displays a more refined bottom-skewed pattern during the early portion of the irradiation cycle due to partial top-half insertion of the MS rods. This investigation further deliberates on a modified checker board control rod pattern to assure safe cold shutdown of the core. All calculations in this multiphysics assessment of the 3D SBF SMR core were completed by using a 2-step Monte Carlo deterministic hybrid procedure based on the Monte Carlo Serpent and COREDAX diffusion codes with the ENDF/B-7.1 nuclear data library.

Controller design and optimization of reactor power control system for ASPWR

The objective of this paper is to design a reactor power control system (RPCS) for the advanced small pressurized water reactor (ASPWR) that adopts constant primary loop average coolant temperature and steam generator secondary side steam pressure programs. In this paper, a two-input and two-output linear nuclear steam supply system (NSSS) model is developed first based on the linearization of the nonlinear point reactor core model with one-group delayed neutron and simplified nonlinear once-through steam generator model. Then, both the conventional trial-and-error design method and the improved version of multi-objective, non-dominated sorting genetic algorithm (NSGA-II) are applied for the RPCS controller design. It has been demonstrated by the numerical simulation results that the RPCS control parameters optimized by the NSGA-II provide better control performance than those tuned by conventional design method. Therefore, the NSGA-II based control parameter optimization method is recommended to tune the RPCS controllers.

The safety analysis of a small pressurized water reactor utilizing fully ceramic microencapsulated fuel

For post-Fukushima nuclear power plants, accident-tolerant fuel (ATF) is strongly desired in order to prevent radioactive release in the event of a severe accident. The Fully Ceramic Microencapsulated (FCM) fuel is one of these materials. The neutronics and thermal hydraulics analysis of small pressurized water reactor (PWR) utilizing FCM fuels needs to be carried out to demonstrate this fuel performance. In this study, first, a three dimensional full core model of a small PWR design is developed to assess the cycle lengths and power distribution. In order to achieve the acceptable cycle lengths, some design optimizations are taken, including higher TRISO particle packing fraction, higher enrichment and using higher density fuels. The results demonstrate that the acceptable cycle lengths can be achieved after this design optimizations. The results reveal that the cycle lengths can be improved to about 550days after using the 19.9w/o enrichment fuels with UO2 fuel kernel, and the cycle lengths can be maintain about 800days with UN fuel kernel due to the higher density than UO2 fuel kernel. Then, suppressing the excess reactivity, the core power distributions and reactivity coefficients are performed to prepare for thermal hydraulics analysis.The steady state and transient state analysis for LOCA is carried out using the best estimate system code RELAP5. For the conventional UO2 fuel, fission product releases simultaneously when the cladding fails. But for the FCM fuel, the radioactive material can be contained in the fuel pellet even after the cladding failure since enveloping the TRISO particles within a dense SiC matrix provides multiple barriers to fission product release. Consequently, although the Zircalloy cladding fails, there is still 3500s available for operator actions before the start of FCM fuel melting occurs even through the SB-LOCA occurring without ECCS, and there is still 1443s available even through the LB-LOCA occurring without ECCS. The results of the accident analysis without ECCS would be useful in understanding cladding and fuel failure progression and could provide guidelines in establishing detailed accident management procedures to design the proper ECCS in the future.

Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as keff, axial and radial power, fission and capture rates were compared and found to be in good agreement.

Neutronic-thermal hydraulic coupling analysis of the fuel channel of a new generation of the small modular pressurized water reactor including hexagonal and square fuel assemblies using MCNP and CFX

The main purpose of this paper is neutronic-thermal hydraulics coupling analysis of the fuel channel of an advanced Small Modular Reactor (SMR), which is nominated as a near term option of the generation IV reactors. The CAREM25 is chosen as the reference SMR. By considering hexagonal and square lattices of the fuel assemblies in the SMR reactor core, neutronic-thermal hydraulic coupling of hot and average fuel channels were conducted via utilizing of the MCNPX and CFX codes. Central fuel assemblies are determined as the hottest fuel assembly in the proposed SMR cores with hexagonal and square fuel assemblies with power peaking factor of 1.778 and 1.674, respectively. The hottest fuel rod power peaking factor is calculated as 1.846 and 1.954 for hexagonal and square fuel channels, respectively. The calculated linear power densities of the hottest and the average fuel rod of hexagonal channels are 20.18(kWm)and10.77(kWm), respectively, as well as of square channels are24.75(kWm) and10.36(kWm), respectively. Results show that the maximum axial power along the fuel rods occurred below the mid-plane of the rod. The ratio of the hot to average rod axial power peaking factor as a safety parameter used to calculate the maximum heat flux in the hottest channel, is calculated close to 2 and 2.4 for the core with hexagonal and square fuel assemblies, respectively. The calculated power peaking factors and linear power densities of fuel rods are in suitable range and in good agreement with PWR safety issues.The maximum temperature of fuel centerline in both types of the proposed SMRs is much lower than that of large PWRs (e.g., 2156 K for hot rod in VVER-1000). Between hexagonal and square fuel assemblies, the hexagonal FAs are preferred because of the two reasons; first, the axial fuel centerline temperature in the hottest hexagonal channel is lower than that of hottest square channel, so it is a good feature with respect of safety margins. Second, the outlet temperatures of the channels in the SMR core with hexagonal FAs are higher than the outlet temperature of the channels in the SMR core with square FAs. Thus the thermal efficiency of the reactor core with hexagonal FAs is more than that of the core with square FAs.

Conventional Controller Design for the Reactor Power Control System of the Advanced Small Pressurized Water Reactor

The objective of this paper is to design a reactor power control system for the advanced small pressurized water reactor that adopts a constant average coolant temperature and a secondary-side steam pressure program. Based on the nonlinear core model with the one-group delayed neutron and simplified nonlinear once-through steam generator model, a two-input and two-output linear nuclear steam supply system (NSSS) model is obtained. Three types of control systems are then proposed and designed on a Bode diagram using analytical methods and second-order approximation. The comparison of the control performance and robustness of the three control systems shows that the double feedback control (DFC) with both power feedback and temperature feedback provides the best performance for reactor power and average coolant temperature with parameter uncertainty due to control rod differential worth variation. The simulations based on the high-order nonlinear NSSS model also show good performance of the DFC system.

New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B4C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370ppm, low power peaking factors, and large shutdown margins of control element assemblies.

Control system design and validation platform development for small pressurized water reactors (SPWR) by coupling an engineering simulator and MATLAB/Simulink

Significant progress has been made in the development of the small pressurized water reactors (SPWR). Unique characteristics of the SPWR deliver challenges to its control system design. In order to facilitate the control system design process and enhance its efficiency, it is important and necessary to establish a control system design and validation platform. Using shared memory technology, an engineering simulator coupled with MATLAB/Simulink is employed to achieve this objective. Shared memory is an efficient method to exchange data within programs. Dynamic data exchange and simulation time synchronization methods are particularly treated. To verify the platform, an SPWR with its control system is modeled using the platform and the simulator. Thermal-hydraulic modeling of the SPWR is carried out using Relap5, and its nodalization is introduced. The objectives of the control strategy are to maintain the average coolant temperature linearly varying with the reactor power and steam pressure constant. A preliminary SPWR control system is designed with proportional-integral-derivative (PID) controllers, and is implemented in MATLAB/Simulink associated with the engineering simulator. Subsequently, in order to evaluate the performance of the established simulation platform, transients of abrupt load changes and wide range load changes are simulated and simulation results are verified against those obtained from the engineering simulator alone. It is demonstrated that simulation results of both platforms are consistent with each other, which proves that the coupling of engineering simulator and MATLAB/Simulink is successful. Therefore, the established simulation platform can be committed to control system design and validation. Furthermore, it is revealed that the designed control strategy is able to regulate the process parameters such as coolant average temperature and steam pressure to the desired values.

Neutronic study of a new generation of the small modular pressurized water reactor using Monte-Carlo simulation

Small Modular Reactors (SMRs) are the innovative design of nuclear reactors making remarkable interest during recent years. Since there is not enough available operating experience on SMRs, it might be possible to initiate extensive investigations on these types of reactors for the purpose of improving the current performance level of these systems, significantly. The main purpose of this present study is neutronic study of a typical small modular pressurized water reactor via Monte-Carlo method using the MCNPX code. The CAREM25 is chosen as the reference SMR. The reactor core geometry is simulated and neutronic parameters are visualized and analyzed via high qualified 3-D figures. They are figure out neutronic nature of the chosen case study. They capture the simulation of the reactor core geometry, and skim the neutron flux and power distribution, radial and axial power peaking factors and the influence of the control rods on the thermal flux. Central fuel assembly is determined as the hottest fuel assembly with power peaking factor of 1.778. The hottest fuel rod power peaking factor is calculated as 1.846 in the hottest assembly. The maximum calculated axial power peaking factor of the hot rod is 2.85. Results show that the maximum axial power along the fuel rods occurred below the mid-plane of the rod. The ratio of the hot to average rod axial power peaking factor as a safety parameter used to calculate the maximum heat flux in the hottest channel, is calculated close to 2 in almost 70% of the core height.The core reactivity at cold and hot shut down without safety injections of boron acid is calculated as 0.09712(Δkk) and −0.00103(Δkk), respectively.Results show suitable neutronic behavior and responses during virtual tests and analyses.

Cutback sensitivity test for boron-free small modular PWR

A soluble boron-free small modular pressurized water reactor (SMPWR) uses burnable absorbers (BA) instead of soluble boron to reduce excess reactivity. As a consequence, the fuel cycle length can be shortened by the residual penalty of BA. This paper performs cutback sensitivity tests to extend the cycle length. The influence of the height of the cutback, of the 235U enrichment rate, and of the BA material on the power peaking factor (Fq), the axial offset (AO) and the fuel cycle length is analyzed with the reactor core design system, CASMO-4E/SIMULATE-3 code system.

Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

The Integral, Inherently Safe Light Water Reactor (I2S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I2S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

Boron-free small modular pressurized water reactor design with new burnable absorber

Summary This paper presents a preliminary design of a Small Modular Pressurized Water Reactor (SMPWR) aimed at boron-free operation which can reduce the size of a nuclear power plant (NPP) and the amount of liquid radioactive waste, and also reduce the corrosion issues caused by boric acid in the coolant. The design parameter limits, such as reactivity swing, axial offset (AO), 3D pin peaking factor (Fq), and the required shutdown margin, have been established for the boron-free SMPWR. Furthermore, a new ring type of burnable absorber (R-BA) with Zr–167Er is adopted as a burnable absorber (BA) and HfB2 is adopted as a control rod material to satisfy those design limits without liquid boron. Optimal fuel assemblies (FAs) and loading patterns have been searched through a sensitivity study, and the excess reactivity control capacity and design requirements have been demonstrated to be satisfactory. Copyright © 2016 John Wiley & Sons, Ltd.

Technology Selection for Offshore Underwater Small Modular Reactors

Abstract This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead–bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.