Integral Pressurized Water Reactor Research Articles

Simulation and conceptual analyses of a stable natural core cooling system in an integrated small modular PWR

Some of advanced Small Modular Reactors (SMRs) have much more passive and inherently safety features as well as other advantages such as rapid and more flexible installations than the conventional large nuclear reactors. In order to attain an optimum thermal-hydraulic design condition and analyze the transient phenomena in a small size reactor, a modified single-phase model is applied to simulate primary loop and estimate the main loop parameters of an advanced SMR especially for stability analyses. In this paper, the modified model is considering flash-out phenomena through the core and riser parts in addition of the single-phase model, besides using a four-region model for the integral pressurizer zone. Except some simplifying assumptions in governing correlations and fluid properties that do not make a major source of error, the results demonstrate acceptable accuracy in testing intervals. Particularly, stable states are studied during natural convection phenomenon. Firstly, the self-developing thermal-hydraulic code takes into account the parameters and analyses using the MATLAB software. Then, the RELAP5/MOD3.2 is also used to simulate and analyze the system. It uses a complete two-phase system of continuity equations. It is worth mentioning that, the RELAP5 code was especially proposed to be considered and initialized based on some main active safety system systems particularly pumps. A comparison of the results from both models for stability tests (maximum affordable core power) shows that the modified single-phase model makes suitable prediction during the steady state and also during small and quick deviations from the steady state conditions. On the other hand, the RELAP5 code can initialize and follow the transients of the natural cooling system without any pump. A parametric study has been performed to search the preliminary domain of the design. Then, a typical periodic oscillation is used to take system behavior and responses against possible perturbations or instabilities. Results are fairly promising to reach a stable core cooling system based on a complete natural convection. Finally, results show that including and using a 100 KW heater can stabilize the induced non-linear instabilities. It could also be noted that, the ability of a complete passive heat removal system could be mentioned as the most important passive safety feature of the future nuclear reactors. Academics still remember the Fukushima-Daiichi disaster. Accordingly, in this study, only a small heater has been added to the initial pattern. Then the developed model enhances the stability and safety performance of such an integral Pressurized Water Reactor (iPWR).

Thermal Energy Storage Configurations for Small Modular Reactor Load Shedding

The increased penetration of intermittent renewable energy technologies such as wind and solar power can strain electric grids, forcing carbon-based and nuclear sources of energy to operate in a load-follow mode. For nuclear reactors, load-follow operation can be undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various methods of thermal energy storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid variability caused by daily load demand changes and renewable intermittency. Two TES techniques are investigated as candidate thermal reservoirs to be used in conjunction with a small modular reactor (SMR): a two-tank sensible heat storage system and a stratified chilled-water storage system. The goal when coupling the two systems to the SMR is to match turbine output and demand and bypass steam to the TES systems to maintain reactor power at approximately 100%. Simulations of integral pressurized water reactor dynamics are run in a high-fidelity FORTRAN model developed at North Carolina State University. Both TES systems are developed as callable FORTRAN subroutines to model the time-varying behavior associated with different configurations of these systems when connected to the SMR simulator. Simulation results reveal the sensible heat storage system is capable of meeting turbine demand and maintaining reactor power constant while providing enough steam to power four absorption chillers for chilled-water production and storage. The stored chilled water is used to supplement cooling loads of an adjacent facility.

A steam or gas pressurizer effect on the system pressure characteristics for an integral pressurized water reactor

A pressurizer is a major component used to control a system pressure in a primary coolant system of a pressurized water reactor. To identify the system pressure characteristics on nitrogen gas-water and steam-water pressurizers for an integral pressurizer water reactor, an experimental study is carried out on the integral test facility, VISTA-ITL, which is a scaled down facility of the SMART plant. According to the experimental results, the pressurization rate and peak pressure are affected by the pressurizer type. The pressurization and depressurization rate with the steam pressurizer is smaller than the gas pressurizer since water has its inherent characteristic having latent heat while a nitrogen has ideal gas properties. The safety valve in both the gas and steam pressurizers is not opened during the transient. It is one of the characteristics of the integral reactor SMART design with a large pressurizer volume. The cooldown rate of the primary coolant system during the later part transient depends on the decay heat level regardless of the pressurizer type.

Neutronic/thermal-hydraulic coupling analysis of natural circulation IPWR under ocean conditions

In this paper, a thermal-hydraulic analysis code under non-inertial system was developed based on a RELAP5/MOD3 code by establishing dynamic simulation models of typical ocean conditions. The simulation results under periodic force field were in good agreement with the reference data, proving the accuracy of the program. In order to analyze the reactor core flow distribution and power distribution under different ocean conditions, the thermal-hydraulic code was coupled with a two-group three-dimensional neutron kinetics code. Based on the modified RELAP5 code, the effects of heeling, heaving and rolling conditions on neutronic/thermal-hydraulic coupling characteristics of natural circulation integrated pressurized water reactor (IPWR) were studied. The results indicate that the uneven distribution of coolant flow increases with the increasing inclination angle, but the reactor power distribution is subject to little changes. This trend leads to an uneven distribution of coolant temperature at the core outlet. The flow or power fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. Furthermore, the impact of the heaving amplitude is more significant than the heaving period, whether on the flow or power.

Model-free adaptive control law for nuclear superheated-steam supply systems

Nuclear steam supply system (NSSS), which provides saturated or superheated steam to the secondary loop system for electricity or cogeneration, is central in a nuclear plant. The control of NSSS is important for the safe, stable and efficient plant operation. Nuclear superheated-steam supply system (Su-NSSS), which is equipped with an under moderated fission reactor and a once-through steam generator (OTSG) for producing superheated steam flow, is an important type of NSSS that widely applied in the nuclear plants based on the common large pressurized water reactors (PWR) and small modular reactors (SMRs) such as integral PWR (iPWR) and modular high temperature gas-cooled reactor (MHTGR). There is still very limited results for the control of Su-NSSS. Motivated by the gap between the importance and the lack of results in the field of Su-NSSS control, a model-free adaptive power-level control law is developed for Su-NSSSs with forced primary circulation in this paper. This new control law is free from physical and thermal-hydraulic parameters, is in the simple proportional-integral (PI) or proportional-differential (PD) form. To verify the feasibility of the newly-built Su-NSSS controller, it is applied to the NSSS control of a two modular high temperature gas-cooled nuclear plant with comparison to a specific MHTGR-based NSSS control law. Numerical simulation results show that this new control law can improve the transient performance of neutron flux and steam temperature in normal operation case, and can realize system resilience in some abnormal cases.

Study on natural circulation characteristics of an IPWR under inclined and rolling condition

An ocean-based thermal-hydraulic system analysis code was developed based on RELAP5/MOD3 code by adding additional force model of ocean condition and control volume coordinate solver model. The natural circulation operation characteristics of integrated pressurized water reactor (IPWR) under ocean conditions were studied and the effects of inclination and rolling motions were analyzed. The results conclude that, the inclination condition can reduce the mass flow rate of reactor core and lead to inconsistent coolant flow rates of the left and right loops, furthermore, it affects the heat transfer of once-through steam generators (OTSGs). In the case of rolling motion, the additional pressure drop of the loop is dominated by tangential force, and flow oscillation of different loops cancel each other due to the symmetrical arrangement of the reactor. The off-center roll axis location, the combination of the inclination and rolling motion, both can break the thermal-hydraulic symmetry among different loops and enlarge fluctuation amplitude of the core flow rate.

Implementing a test facility in reduced scale for analysis of boron dispersion in a pressurizer of an integral compact and modular reactor

Small Modular Reactors, SMRs, are variants of compact modular reactors of generation III+ that present attractive features such as simplicity, enhanced safety and are economically competitive. In an iPWR (integral pressurized water reactor), the pressurizer is located at the top of the reactor vessel; this configuration requires the investigation of the mechanisms adopted for the homogenization of boron. The present work shows the contributions to the design of an experimental installation proposed to provide relevant data to study the boron homogenization phenomena in a pressurizer of an SMR. For this, an experimental installation was built with a test section made of stainless steel with a volumetric scale equal to 1:200, representing one-fourth of the pressurizer of an iPWR. The experimental installation has a test section, a heat exchanger, two pumps, five flow meters and three stainless steel tanks with capacities of 0.143, 0.069 and 0.216m3. The test facility layout was mounted so that the test section was fed with water or saline solutions with different concentrations, representing scenarios of boration or deboration. To determine the concentration at the inlet and outlet of the test section, samples have been collected every 10min during the experiments representing some scenarios (boration/deboration). The electrical conductivity of these samples was determined by a conductivity meter and converted to the concentration of the tracer element (NaCl) used. Three representative experiments were conducted and it was possible to prove the viability of using the experimental setup in a reliable form, for the range of values occurring in the boron homogenization phenomenon in a pressurizer of a compact modular reactor.

Comparison and analysis on transient characteristics of integral pressurized water reactors

In the present work, the similarities and differences of representative IPWRs (integral pressurized water reactor) are studied, and two typical reactor design schemes are summarized. To get a comprehensive understanding of their transient characteristics, SBO (station blackout) and SBLOCA (small break LOCA) are simulated and analyzed respectively by using Relap5/Mod3.2. The calculation results show that, both designs are effective in keeping reactor safe. However, the transient features of the two designs show significant differences. In the primary side passive safety system (PSS) connection design, PRHRS (passive residual heat removal system) shows a roughly congruent performance in removing residual heat under various accidents. While in secondary side PSS connection design, the capability of PRHRS is closely related to primary coolant circulation condition. In SBLOCA analysis, different design approach shows different primary coolant water inventory change trend. And primary PSS connection design could potentially keep reactor core well covered for a longer time.

Screening the design space for optimized plutonium incineration performance in the thorium-based I2S-LWR

This paper presents an optimization of a thorium-plutonium (Th-Pu) fuel cycle by screening various design options for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The I2S-LWR is an advanced 2850MWt integral pressurized water reactor with enhanced safety beyond that of Gen-III+ reactors. The features of this reactor, such as material choice, make it attractive for alternative fuel cycles including the use of thorium. Recently, the feasibility of the Th-Pu cycle was studied and the benefits associated with it were demonstrated. More specifically, the Pu incineration performance was enhanced by adopting multi-batch (i.e. more than 3-batch) schemes and extended burnup (above 100MWd/kg). The optimized design with the most favorable loading pattern was obtained by applying the Simulated Annealing optimization technique. This paper demonstrates further plausible modifications to the Th-Pu cycle design that may enhance its performance considerably. The paper seeks to identify the contributory factors, such as cladding types, plutonium vectors and initial plutonium loadings, with major impact on the incineration performance. The post-irradiation characteristics are also analyzed and suggest that such a cycle may simplify the design and operation of the waste repository.

Numerical Study on Coolant Flow Distribution at the Core Inlet for an Integral Pressurized Water Reactor

Abstract When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

Development of a LOCA safety analysis evaluation model for the Westinghouse Small Modular Reactor

The Westinghouse Small Modular Reactor (W-SMR) is an integral pressurized water reactor (PWR) with all major components located inside the reactor vessel, with a passive safety system that was derived from the AP1000® PWR passive safety design. The integral design of the reactor coolant system (RCS) effectively removes the postulated large break loss-of-coolant accident (LOCA). Therefore, small break LOCA (SBLOCA) is the most limiting design basis accident for the W-SMR. The LOCA safety analysis of the W-SMR requires an evaluation model (EM) that is adequate for a reactor design with passive safety systems. Development of the EM follows the Evaluation Model Development Assessment Process (EMDAP) by the United States Nuclear Regulatory Commission (US NRC). The first step in this process is the development of a SBLOCA Phenomena Identification and Ranking Table (PIRT), which identifies the figures of merit for the LOCA safety analysis and key thermal hydraulic phenomena pertinent to the passive safety design of the W-SMR. Next, an assessment matrix that consists of experiments from the AP600™ and AP1000 PWR development programs, University of California Berkeley (UCB) tube condensation tests, and the integral effects and separate effects tests for the W-SMR testing program is established to assess the evaluation model. The systems code used in the W-SMR LOCA evaluation model is the WCOBRA/TRAC-TF2 systems code, which has been extensively validated for previous LOCA applications. WCOBRA/TRAC-TF2 is being further developed to address the W-SMR EM PIRT. Among key thermal hydraulic phenomena, the assessment of the condensation on the containment wall and natural circulation/draining through the CMT are presented here as examples to demonstrate the W-SMR SBLOCA EM development, while all important phenomena are expected to be assessed using the WCOBRA/TRAC-TF2 systems code. Finally, the reference SBLOCA transient of W-SMR is presented to demonstrate the ability of the systems code to properly capture the key phenomena identified in the PIRT. It is expected that the development of an evaluation model for safety analysis of the Westinghouse SMR will facilitate its licensing and eventual commercialization.

Neutronics and thermal hydraulics analysis of a small modular reactor

The small modular reactor (SMR) offers many feasible pathways for the construction of more nuclear power plants. A physics model of a near term deployable SMR of the integral pressurized water reactor (IPWR) design is developed. Fuel depletion simulations are performed to optimize the active fuel length, fuel enrichment and core loading pattern in order to achieve a uniform core power distribution. The optimized core can produce 500MW of thermal power with a four year core life-time at a capacity factor of 87%. The core consists of 69 uranium dioxide (UO2) fuel assemblies; 5 assemblies at 4.4at% 235U enrichment and 64 assemblies at 4.95at% 235U enrichment. The active fuel length is 200cm and the core diameter is 194.55cm for an active core height-to-diameter ratio of 1.03. As part of the study the active fuel length is increased to 240cm resulting in an increased capacity factor of 95% at 530MW of thermal power output for an active core height-to-diameter ratio of 1.23. Rod cluster control assemblies (RCCAs) are placed strategically to reduce the overall core power peaking factor to 1.3. Estimated reactor kinetics parameters such as the delayed neutron fraction and mean neutron generation time are typical of existing larger pressurized water reactors (PWRs) from which much of the IPWR based SMR design is derived. This study showed that Doppler, moderator temperature, void and power reactivity coefficients are all negative over the core life-time of four years indicating the possibility of safe reactor operation. A semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin from industry standards on peak fuel and clad surface temperature limits. The critical heat flux (CHF) is calculated and is not exceeded even in 10% overpower conditions. In addition the nucleate boiling ratio (DNBR) is calculated and found to be above 4.8 for the entirety of the active core region. These parameters further engender confidence in the safety of the SMR design.

Enhancing plutonium incineration in the thorium-based I2S-LWR design with loading pattern optimization

This paper presents an optimization of a thorium–plutonium fuel cycle, through a multi-batch reloading scheme, developed for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The I2S-LWR is an advanced 2850MWt integral pressurized water reactor with enhanced safety beyond that of Gen-III+ reactors. Its baseline fuel and cladding materials are U3Si2 and advanced FeCrAl steel, respectively. The advanced steel cladding can withstand longer exposure periods with significantly lower degradation rates compared to traditional Zr-based alloys. In principle, increasing the number of batches allows higher discharge burnups and thus deeper Pu and transuranic elements incineration to be achieved. Therefore, various refuelling strategies were considered in this study, namely 3-, 5- and 7.56-batch schemes. The Simulated Annealing optimization technique was applied for the different batch schemes to obtain the most favourable loading pattern with respect to cycle length performance. The results confirm that increasing the number of batches allows the discharge burnup to be increased by 20% (above 100MWd/kg), which improves the Pu incineration performance. In addition, the increased number of batches improves the reactivity coefficients without violating the power peaking factors limits.

Characterizing fluid structure interactions of a helical coil in cross flow

Mechanical vibrations compromise the integrity of key components of thermal power plants. Without careful design, strong resonances during steady state operation can wear these components to the point of failure, leading to an unsafe situation that may force a plant to shut down. The purpose of this research is to further the understanding of the vibrations induced in a helical coil subject to steady fluid flow. A helical coil steam generator, such as that found in most integral pressurized water reactors, appears to eliminate many flow-induced vibration concerns when compared to traditional steam generators; however this has yet to be clearly demonstrated experimentally. The objective of this study is to detail and demonstrate a new method to quantify the motion of a helical coil in an annulus subject to external axial flow of water and further characterizing the influence of pitch-to-diameter ratio on the fluidelastic instability of a helical coil. This is accomplished by observing the motion of a helical coil mounted to an inner opaque cylinder through an outer glass tube using a high speed video camera. A mirrored image-pair is used to observe this structure from two perspectives simultaneously, allowing for three-dimensional characterization of the coil motion. The experimental facility is described in detail. The method developed herein for identifying specific points on the coil from images and mapping them to the coil location using the law of refraction is described. An uncertainty analysis of the coil position measurement is conducted based on geometry and refractive index which can be readily applied to measurements obtained using this method. The outcome of empirical observations shows these helical coils to hold a slightly higher resonance frequency than that of cylinders in cross flow, their mechanical stiffness approximated through analytical means shows to produce relatively accurate natural frequencies when compared to the empirical data – 14.1Hz and 12.5Hz, respectively for first mode vibration. This study”s contributions present a new method for metering fluid structure interactions with high-fidelity, provide new empirical data which has not previously been produced, and make observations to the response of a helical coil which are new to the field of fluid-structure interactions.

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.

Analysis of natural circulation operational characteristics for integrated pressurized water reactor

The present work discusses the natural circulation operational characteristics of integrated pressurized water reactor (IPWR). A thermal hydraulic model of IPWR was established using RELAP5 code. Under the ideal steady-state control scheme, IPWR can realize the reactor power change with steam flow. The conversion process from forced circulation to natural circulation has been studied. The results show that the reactor power has large fluctuations during the conversion process based on the traditional control scheme. Two optimized control schemes were applied to achieve a smooth conversion process. From the perspective of reducing power oscillation, the result obtained by canceling the reactor power control at initial conversion process is better than the results obtained by shutting down each pump one by one. Under natural circulation conditions, the stable operation boundary of IPWR system was determined. The steady state operation characteristics and rapid load change process under natural circulation show that primary coolant can achieve stable natural circulation operation at low load conditions. The obtained analysis results are significant for improving operation characteristics of the IPWR.

The optimum design and arrangement of a steam generator in an integral pressurized water reactor

This paper presents a double-tube once-through steam generator (DOTSG), whose tube unit includes an outer straight tube and an inner helical tube, in an integral pressurized water reactor (IPWR). To obtain the optimum structure of the inner helical tube and the arrangement of DOTSGs in the reactor pressure vessel, a two-level optimization method is used, aiming at the lower pumping power needed, and a smaller reactor pressure vessel volume is used. The pitch of inner helical tubes and the central distance of outer tubes are considered design parameters when minimizing the pumping power with the genetic algorithm in the bottom level, while the number of tube units in a single DOTSG and the number of DOTSGs in the reactor pressure vessel (RPV) are optimized to obtain the minimum volume of the IPWR, which is conducted in the top level. The optimum pitch of the helical tube varies in the sub-cooled region, boiling region and superheated region. The results show that the smaller pitch brings a shorter tube length and a higher pressure drop, and the effects are strong in the sub-cooled region and the superheated region but weak in the boiling region. In this way, the optimal structure of the inner helical tube is a small pitch in the single-phase region and a large pitch in the boiling region. According to the bottom level results, the optimum arrangement of DOTSGs in the pressure vessel is determined.

The Use of Flashing Drums and Microchannel Heat Exchangers to Generate Steam in Large Integral Light Water Reactors

Integral pressurized water reactors are innovative reactors in which all of the components typically associated with the nuclear steam supply system of a nuclear power station are located within the reactor pressure vessel. In order to facilitate this modification in large [~1000-MW(electric)] light water reactors (LWRs), compact heat exchangers such as microchannel heat exchangers must be used. Previous attempts at using microchannel heat exchangers were unsuccessful since they are prone to vapor locking and crud blockage when the primary coolant boils. Therefore, the authors propose the use of a flashing drum to facilitate boiling in conjunction with a primary microchannel heat exchanger for a large integral LWR. The integral inherently safe light water reactor (I2S-LWR) is used as a basis for the implementation of this novel concept. The high-temperature, high-pressure secondary water generated in the secondary loop through heating in the microchannel primary heat exchanger of the I2S-LWR is sent to a flashing drum where 99.9% pure vapor is extracted and sent to the turbines. This prevents boiling in the primary heat exchanger that in turn reduces crud deposition, flow instabilities, and the potential for channel blockage or vapor locking in the small channel sizes of microchannel heat exchangers. The benefits and disadvantages of this approach are presented in this paper. Unfortunately, this innovative approach to nuclear steam generation for integral LWRs is challenged by a potential decrease in thermodynamic efficiency. Therefore, a sensitivity study is presented that explores the impact of several design variables on the thermodynamic efficiency of the plant. As part of this study, a simple and a complex Rankine cycle were modeled in order to determine the impact that system design modifications can play in recovering thermodynamic efficiency lost by the steam drum. Both cycles utilize turbines, condensers, and condensate/recirculation pumps, while the complex Rankine cycle utilizes a four-stage turbine with subsequent separation and open feedwater heaters. The optimized efficiencies for the simple and complex Rankine cycles are 31% and 33%, respectively, indicating that additional system enhancements to the power conversion system could compensate for the inclusion of a flashing drum.

The influence of ocean conditions on thermal-hydraulic characteristics of a passive residual heat removal system

There are many differences between the flow and heat transfer characteristics of nuclear reactors under ocean and land-based conditions for the effects of ocean waves. In this paper, thermal hydraulic characteristics of a passive residual heat removal system (PRHRS) for an integrated pressurized water reactor (IPWR) in ocean environment were investigated theoretically. A series of reasonable theoretical models for a PRHRS in an IPWR were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, flow and heat transfer and ocean motion models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code in FORTRAN 90 format has been developed to analyze the thermal–hydraulic characteristics of the PRHRS under ocean conditions. The code was implemented to analyze the effects of different ocean motions on the transient thermal-hydraulic characteristics of PRHRS. It is found that the oscillating amplitudes and periods of the system parameters are determined by those of the ocean motions. The effect of rolling motion is more obvious than that of pitching motion when the amplitudes and periods of rolling and pitching motions are the same. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the IPWR.

A novel coordinated control for Integrated Pressurized Water Reactor

Integrated Pressurized Water Reactor (IPWR) has the characteristic of strong coupling, nonlinearity and complicated dynamic performance, which requires high standards of the control strategy and controller design. Most of IPWR systems utilize control strategy of ideal steady-state and PID controller, even though this strategy causes flow instability in the once through steam generator (OTSG) in low load conditions. Besides, the simple form of PID limits the performance developing which could not appropriately satisfy the requirements for quality. Motivated by these drawbacks, this paper proposes an IPWR coordinated control strategy and adopts PID controller to control each subsystem. The control strategy considers the system as a two-level hierarchical control system, and considers coordinating controller and bottom controllers. In the period of controller design, this strategy utilizes PID controller to control each subsystem, and modifies the controller parameters in real time by Fuzzy-KWNN algorithm, which adaptively achieves the system adjustment. Finally, simulation results are presented to exhibit the effectiveness of the proposed IPWR control system.