Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

Probabilistic safety analyses developments resurgence in Bulgaria—Kozloduy nuclear power plant example

This paper presents some of the main technical features and insights of the Kozloduy nuclear power plant (NPP) units 5 and 6 probabilistic safety analysis (PSA) level 1. Probabilistic analyses and their applications in Bulgaria were given further impetus in recent years. More than 17 years after the first PSA study in Bulgaria in 1992 today probabilistic analyses receive increasing attention and application than ever before. The Bulgarian regulatory body (BNRA) is also interested in expanding their capability of reviewing and using PSA in plant safety assessments. In November 2008 within the framework of the program financed by European Union (PHARE), a project for assisting the BNRA in establishing the regulatory requirements on the base of PSA was completed. One of the objectives of this project was performance of the independent review of Kozloduy NPP units 5 and 6 PSA. This review was a new impulse for the authors to present in more details of Kozloduy NPP probabilistic assessment studies in the present paper.     

Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies—the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA.To perform this investigation it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding.It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP.     

Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during cooldown to cold shutdown, and in the validity of a two-tier calculational method. The results have been directly used in updating the plant shutdown PSA, by changing the success criteria for core cooling during cooldown of the plant and showing a reduction in overall risk.     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).     

ARTIST: introduction and first results

Aerosol Trapping In a Steam Generator (ARTIST) is a seven-phase international project (2003–2007) which investigates aerosol and droplet retention in a model steam generator under dry, wet and accident management conditions, respectively. The test section is comprised of a scaled steam generator tube bundle consisting of 270 tubes and three stages, one 1:1 separator unit, and one 1:1 dryer unit.As a prelude to the ARTIST project, four tests are conducted in the ARTIST bundle within the 5th EU FWP SGTR. These first tests address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities are sonic, and far away from the break, where the flow velocities are three orders of magnitude lower. With a dry bundle and the full flow representing the break stage conditions, there is strong evidence that the TiO₂ aerosols used (AMMD 2–4 μm, 32 nm primary particles) disintegrate into much smaller particles because of the sonic conditions at the break, hence promoting particle escape from the secondary and lowering the overall DF, which is found to be between 2.5 and 3. With a dry bundle and a small flow reproducing the far-field velocities, the overall bundle DF is of the order of 5, implying a DF of about 1.9 per stage.Extrapolating the results of the dry tests, it turns out that for steam generators with nine or more stages, it is expected that substantial DF’s could be achieved when the break is located near the tube sheet region. In addition, better decontamination is expected using more representative proxies of severe accident aerosols (sticky, multi-component particles), a topic which is yet to be investigated.When the bundle is flooded, the DF is between 45 and 5740, depending on the mass flow rate, the steam content, and the water submergence. The presence of steam in the carrier gas and subsequent condensation inside the broken tube causes aerosol deposition and blockages near the break, leading to an increase in the primary pressure. This has implications for real plant conditions, as aerosol deposits inside the broken tube will cause more flow to be diverted to the intact tubes, with a corresponding reduction in the source term to the secondary.     

An analytical comparative exercise on the OECD-SETH PKL E2.2 experiment

The “First Workshop on Analytical Activities related to the SETH-OECD project” was held in Barcelona at the UPC's Institute of Energy Technologies (INTE), from 2nd to 3rd September 2003. The workshop gave the participants an opportunity to present the main results of the calculations performed as pre- and post-test simulations of SETH experiments. Among all the post-tests that were both presented and discussed, PKL experiment E2.2 holds special interest as it has been widely studied. Test E2.2 examined the most conservative case in terms of the maximum size that condensate slugs can reach and how far boron concentration can drop on resumption of natural circulation following a cold-side SB-LOCA. The analyses were performed by different working groups belonging to different countries and different codes were used.This paper goes deeper into the comparison of results of the different authors. Its aim is to both show and compare the results obtained by different working groups in their simulation of the experiment and to analyse the main parameters involved in order to draw conclusions on improvements that can be made in the analytical approach to such tests. All the participants managed to successfully predict the overall thermal-hydraulic system behaviour. Vessel fill-up together with slug build-up by reflux-condensation are phenomena that were correctly predicted, while simulation of natural circulation restart and transport of low-borated water slugs still need some improvement.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

Development of supercritical water heat-transfer correlation for vertical bare tubes

This paper presents an analysis of heat-transfer to supercritical water in bare vertical tubes. A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those in supercritical water-cooled nuclear reactor (SCWR) concepts.The experimental dataset was obtained in supercritical water flowing upward in a 4-m long vertical bare tube with 10-mm ID. The data were collected at pressures of about 24 MPa, inlet temperatures from 320 to 350 °C, values of mass flux ranged from 200 to 1500 kg/m² s and heat fluxes up to 1250 kW/m² for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature.A dimensional analysis was conducted using the Buckingham Π-theorem to derive the general form of an empirical supercritical water heat-transfer correlation for the Nusselt number, which was finalized based on the experimental data obtained at the normal and improved heat-transfer regimes. Also, experimental heat transfer coefficient (HTC) values at the normal and improved heat-transfer regimes were compared with those calculated according to several correlations from the open literature, with CFD code and with those of the proposed correlation.The comparison showed that the Dittus-Boelter correlation significantly overestimates experimental HTC values within the pseudocritical range. The Bishop et al. and Jackson correlations tended also to deviate substantially from the experimental data within the pseudocritical range. The Swenson et al. correlation provided a better fit for the experimental data than the previous three correlations at low mass flux (∼500 kg/m² s), but tends to overpredict the experimental data within the entrance region and does not follow up closely the experimental data at higher mass fluxes. Also, HTC and wall temperature values calculated with the FLUENT CFD code might deviate significantly from the experimental data, for example, the k-? model (wall function). However, the k-? model (low Reynolds numbers) shows better fit within some flow conditions.Nevertheless, the proposed correlation showed the best fit for the experimental data within a wide range of flow conditions. This correlation has an uncertainty of about ±25% for calculated HTC values and about ±15% for calculated wall temperature. A final verification of the proposed correlation was conducted through a comparison with other datasets. It was determined that the proposed correlation closely represents the experimental data and follows trends closely, even within the pseudocritical range. Finally, a recent study determined that in the supercritical region, the proposed correlation showed the best prediction of the data for all three sub-regions investigated.Therefore, the proposed correlation can be used for HTC calculations in SCW heat exchangers, for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for future comparison with other datasets and for the verification of computer codes and scaling parameters between water and modelling fluids.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic–ceramic (CERCER) plate-type fuel-elements with (U–Pu) carbide fuel contained within a SiC inert matrix. For the neutronic calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European fast reactor (EFR)-type pattern for the control assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal–hydraulic constraints could be appropriately accounted for. The thermal–hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins.     

MELCOR/MACCS simulation of the TMI-2 severe accident and initial recovery phases, off-site fission product release and consequences

MELCOR has become the preferred code of the Swiss nuclear industry and of PSI for severe accident analysis, on account of its integrated systems-level approach and validation against experiments and more detailed codes, while MACCS is commonly used by safety authorities for independent assessment of off-site consequences, in particular health effects. The present work arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. The MELCOR 1.8.5RD calculations are based on a model for phases 1 and 2 provided by the code developers but with a simplified thermal hydraulic noding in certain regions and the inclusion of a simple representation of the fission product release and transport pathways. The model has also been extended to simulate phases 3, 4, and the continuing initial period of core recovery and stabilisation. These calculations are a first attempt to demonstrate a MELCOR–MACCS capability to simulate the whole plant accident sequence beyond phase 4, including the containment response and off-site consequences arising from fission product release from the containment. Emphasis is placed on the overall accident evolution and whole plant response, rather than the detailed behaviour. Results are compared with observed and deduced data for the major accident signatures and rough estimates for exposure based on off-site monitoring. The results provide a good basis for the NPP analysis foreseen.     

Validation of RELAP5/MOD3.2 model on trip off one main coolant pump for VVER 440/V230

This paper presents the validation of RELAP5/MOD3.2 model of the VVER 440 for Nuclear Power Plant (NPP) in the analysis of the following transient: “Trip off one MCP”.This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with measurement transient data received from Kozloduy NPP Unit #4. The baseline input deck for VVER440 was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events, and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP.The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparison between the RELAP5 calculations and the test data indicates a good agreement.This validation was possible through the participation of leading specialists from Kozloduy NPP and with the support of the Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones.In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes).The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.     

Design evaluation of emergency core cooling systems using Axiomatic Design

In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies.     

Development and diversity and defense-in-depth application of ABWR feedwater pump and controller model

This work developed an advanced boiling water reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element/three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including step change of dome pressure, feedwater pumps transfer, inadvertent closure of all turbine control valves, and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of stand-alone model and incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an emergency core cooling system (ECCS) automatic diverse back-up. In Lungmen nuclear power plant (NPP), a diverse manual initiation means for the high pressure core flooder (HPCF) loop C is designed as the back-up of digitalized engineered safety features actuation system (ESFAS). If the motor-driven feedwater pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common-cause failure (CCF).     

Axial offset control of PWR nuclear reactor core using intelligent techniques

Improved load following capability is one of the main technical performances of advanced PWR (APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to AO control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability.