Typical Nuclear Power Plant Research Articles

Minimum dose path planning during reactor coolant system maintenance with corrosion product activity

To facilitate the periodic maintenance of the reactor coolant system, the nuclear power plant must be shut down. However, maintenance work on the primary loop components could result in post-shutdown radiation exposure. Activated corrosion products are the main source of radiation inside the containment building during post-shutdown. Traditionally, radiation detectors are used to measure in-containment radiation to protect workers from radiation risk during maintenance. However, these devices cannot predict corrosion product activity and the subsequent radiation dose. This study presents a minimum dose path planning methodology for protecting workers during the reactor coolant system maintenance in a typical pressurized water reactor nuclear power plant. Specifically, the study uses the corrosion product release and activation (CoPRA) code to estimate the corrosion product activity from the primary loop’s three major post-shutdown source terms. Subsequently, the FLUKA Monte Carlo code is used to simulate dose rates from the corrosion product activity, and the Three-Degree of Vertex shortest path algorithm is implemented to find the minimum dose path. Comparison results show that the predicted dose is consistent with the measured dose from the nuclear power plant. The results also show the flexibility and adaptability of the proposed method in finding the shortest path with the minimum dose in different reactor operating conditions.

Fractional order modelling and robust multi-model intelligent controllers’ synthesis for ACP1000 nuclear power plant

In this research work, a third generation Pressurized Water Reactor (PWR) type Nuclear Power Plant (NPP) of 1100 MWe rating called Advanced Chinese PWR (ACP1000) NPP is adopted. The research and development work is mainly encompassed of addition of third steam generator loop, hybrid programming platforms, fractional order modular modelling and intelligent controllers. The ACP1000 NPP is comprised of primary loop, secondary loop and balance of plant (BOP). The ACP1000 nonlinear dynamics is precisely modelled with greater efficiency using fractional order differential equations. The entire plant is modelled into eighteen modules in a fractional order framework in Visual Basic (VB) environment. The eighteen modules are governed by nine controllers. Controllers are configured in a fashion of an adaptive neuro-fuzzy inference system (ANFIS). An innovative multi input single output (MISO) ANFIS virtual instrument (VI) is designed for training and configuring controllers in LabVIEW. The developed MISO ANFIS VI in combination is implemented for multi-input multi output (MIMO) ANFIS controllers in parallel computing framework. The variable transfer between fractional order multi- modular framework in VB (FO- MMF-VB) and robust intelligent ANFIS controllers in LabVIEW (RI-ANFIS-LV) is established in client server configuration and system exec VI for calling external code in VB. This hybrid programming framework of Visual Basic and LabVIEW is actually a development of LabVIEW Visual Basic Integration (LabVb) Toolkit used to evaluate the closed loop transient performance of ACP1000 nuclear power plant. Various parameters are simulated for dynamic transient simulations under load manoeuvring from 100% power to 50% power. The results are analysed and validated against benchmark design data and Preliminary Safety Analysis Report (PSAR). All the results are within the protection systems design limits under normal operating conditions.

Mapping thermal energy storage technologies with advanced nuclear reactors

Advanced nuclear power plants (NPPs) will potentially need to operate in environments where power generation flexibility is more highly valued than the stability or baseload generation capability for conventional demand curves. Thermal energy storage (TES) systems would enable NPPs to respond nimbly to market variability and could also position advanced NPPs to participate differently in restructured markets, thus further enhancing their economic competitiveness. TES systems could also benefit the electric grid by eliminating the need for peaking plants, as well as by improving the economic performance of baseload NPPs. While TES technologies afford a unique opportunity to address many of these challenges, the applicability of these systems is also complicated by the fact that various advanced NPPs are designed differently, each with its own temperature range, size, operating fluids, and operating conditions. Hence, TES systems face significant barriers to investment, as more information on their compatibility and performance metrics is needed to quantify the advantages provided by each, as well as the challenges these technologies might face if coupled with a particular type of advanced NPP. This study explores the possibility of integrating a wide variety of TES technologies with various categories of advanced NPPs, based on their operating characteristics. To help decision makers, users and developers decide which TES technology is best suited to a particular category of advanced NPPs, this research present a Phenomena Identification and Ranking Table (PIRT) analysis of 10 TES systems that could potentially be coupled with advanced NPPs, which themselves are divided into nine categories based on their operating conditions. Each advanced NPP category is evaluated for compatibility with the 10 TES systems by assembling and discussing a database of information concerning 10 engineering questions, defined herein in as figures of merit (FOMs), such as: technology readiness level (TRL), temperature compatibility, energy density, size, cycle frequency, ramp time, realignment frequency, geographic needs, environmental impact, and interventions. By assembling a database of information concerning the TES technologies’ compatibility with various advanced NPP systems, this study can help developers acquaint themselves with a particular TES technology before choosing to build a new integrated installation.

Assessment of radiological safety and emergency response of VVER-1200 type reactor

We studied the radiological safety and emergency response of the VVER-1200 type nuclear power plant using the RASCAL 4.3 and HOTSPOT 3.1.2 codes for INES levels 5, 6, and 7 reactor accidents due to SBO concomitant with LOCA for both dry and rainy seasons. The simulation was performed using Gaussian Plume Model and Lagrangian Gaussian Plume Model, respectively. The meteorological data for dry and rainy seasons were considered for the RNPP site of Bangladesh. The amount of total radioactivities of radiological equivalence to I-131 that released to the atmosphere for INES levels 5, 6, and 7 nuclear reactor accidents were 4011.6 TBq, 25,110 TBq, and 129,650 TBq, respectively. Radiological doses such as TEDE, TED, Thyroid CDE, Inhalation CEDE were evaluated for 40 km radial distance from the reactor site. The total effective dose for both RASCAL and HOTSPOT code predicts the higher dose value during the rainy season compared to the dry season. In the rainy seasons, TEDE and TED were very high about several ten times of permissible limit within 16 km radial distance. Therefore, immediate radiological protective measures are required to be implemented based on the emergency response plan of the local regulatory authority.

Fractional order multi-scheduling parameters based LPV modelling and robust switching H∞ controllers design for steam dump system of nuclear power plant

In this research work, the highly challenging problem of novel modelling and nonlinear control of steam dump system of Pressurized Water Reactor (PWR) type Nuclear Power Plant (NPP) is attempted. The Fractional Order Multi- Scheduling Parameters based Multi-Input Single- Output Linear Parameter Varying (FO-MSP-MISO-LPV) model of Steam Dump System (SDS) is estimated with uncertain dynamics under sudden load variation transients. MSP for uncertain dynamics of SDS in FO framework is the most challenging problem and attempted in a novel fashion for the first time in nuclear industry. Scheduling parameters are dynamic in nature that makes the control problem more challenging. The Model is estimated experimentally by least square method using innovative plant operational data of opening positions of different valves as input variables and steam pressure as an output variable and cold leg coolant temperature coefficient of reactivity, hot leg coolant temperature coefficient, steam flow rate and turbine power as dynamic scheduling parameters. A switching controller is designed to address variable conditions of steam pressure for the actuation of dump valves, relief valves and safety valves in SDS. A robust fractional order LPV switching H∞ (RFO-LPV-SWH∞) controllers are formulated and designed for FO-MSP-MISO-LPV model. The design of RFO-LPV-SWH∞ controllers is another significant contribution in switching mode with non-integer and LPV hybrid framework. RFO-LPV-SWH∞ controllers are tested, simulated and validated against benchmark transients as laid down in Final Safety Analysis Report (FSAR) of PWR-type NPP. The input and output variables at first and second vertex of polytope are fast reference tracking under highly nonlinear uncertain dynamics of SDS. Closed loop simulation experiments are conducted and proved that the proposed closed framework is robust in performance under parametric uncertainty.

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Abstract Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the nuclear power plant (NPP) Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 libraries are considered. In the first part of this work the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few 100 pcm in multiplication factor for a fresh fuel due to differences in 235U and 238U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout, and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1%. Some differences in isothermal temperature coefficient were observed, however, they only marginally affect core power defect going from zero to full power.

Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

V&V Methods Used in the Development of FPGA-based I&C Systems for HTR-10

Currently, in few countries, there have been many projects related to the development of Field Programmable Gate Array (FPGA)-based Instrumentation and control (I&C) system designs for their nuclear power plant (NPP), whether for modernization or the new development of NPPs. Indonesia already develops a study related to High Temperature Gas Cooled Reactor as the generation IV for NPPs type for these few years with a specific type which is High Temperature Reactor–10 MW (HTR-10). Since the FPGA-based technology is new in the field, and both the international standards and guidance, from national and international has not defined provisions this issue, therefore verification and validation (V&V) activities for this HTR-10 are mandatory to be included in the design life cycle of I&C system development to get license approval from the national regulatory body. The FPGA-based system needs concentration both in hardware and software development, but commonly only software-related standards that usually included in the design and V&V in the previous computer-based I&C system. Therefore, in this study the V&V plan needs to be developed to state the methods to be required regarding the correctness of the applications of FPGA based I&C system. By arranging this V&V method, we will see whether they are easily applicable to actual systems, had the result to reach the requirement to have reliability, availability, maintainability and safety aspects, and has shortened and easier V&V step than computer-based I&C system. From the results of this study, it can be seen that the results of the method used for V&V between Programmable Logic Controller (PLC)-based and FPGA are almost the same because both have almost the same requirements, especially in software design. Some testing is not mandatory for FPGA-based because of the reduced system complexity. The use of FPGA can shorten the V&V process which can also reduce costs.

The helium bubble: Prospects for 3He-fuelled nuclear fusion

The helium bubble: Prospects for 3He-fuelled nuclear fusion

Segmented mandrel tests of as-received and hydrogenated WWER fuel cladding tubes

The mechanical interaction between the fuel pellet and the cladding tube of a nuclear fuel rod is a very important for safety studies as this phenomenon could lead to fuel failure and release of radioactivity. To investigate the ductility of cladding tubes used in WWER type nuclear power plants, several mandrel tests were performed in the Centre for Energy Research (EK). This modified mandrel test was used to model the mechanical interaction between the fuel pellet and the cladding using a segmented tool. The tests were conducted at room temperature and at 300 °C with inactive as-received and hydrogenated cladding ring samples. The results show a gradual decrease in ductility as the hydrogen content increases, the ductile-brittle transition was seen above 1500 ppm hydrogen absorbed.

LQGI/LTR based robust control technique for a pressurized water nuclear power plant

This work proposes a new hybrid control strategy for a pressurized water type nuclear power plant by integrating linear quadratic integrator (LQI), linear quadratic Gaussian (LQG), and loop transfer recovery (LTR) approaches. The multi-input multi-output nuclear plant model adopted in this work is characterized by 38 state variables. The nonlinear plant model is linearized around steady-state operating conditions to obtain a linear model for the controller design. The proposed LQGI/LTR technique designs state-feedback assisted output control using the estimated states. The control architecture offers robust performance and tracks the reference set-point with zero steady-state error in the presence of uncertainties and disturbances. The effectiveness of the proposed technique is demonstrated by simulations on different subsections of a pressurized water nonlinear nuclear power plant model. The control performance of the proposed technique is further compared with other classical control design schemes. Statistical measures are employed to quantitatively analyse control performance.

Robust-optimal integrated control design technique for a pressurized water-type nuclear power plant

A control design scheme is formulated for a pressurized water type nuclear power plant by integrating the optimal linear quadratic Gaussian (LQG) control with the robust integral sliding mode (ISM) technique. A novel robust-optimal hybrid control scheme is further proposed by integrating the LQG-ISM technique with the loop transfer recovery approach to enhance the effectiveness and robustness capability. The control architecture offers robust performance with minimum control efforts and tracks the reference set-point effectively in the presence of disturbances and parametric uncertainties. The multi-input-multi-output nuclear power plant model adopted in this work is characterized by 38 state variables. The nonlinear plant model is linearized around steady-state operating conditions to obtain a linear model for the controller design. The efficacy of the proposed controllers is demonstrated by nuclear power plant subsystem simulations. The control performance of the proposed technique is also compared with other classical control design schemes. Numerical measures are employed to quantitatively analyse and compare the performance of the different controllers that are studied in the work.

Evaluation of core-wide fuel pin burst during LOCA in APR1400

Core-wide fuel pin burst fractions in a typical APR1400 nuclear power plant during a loss-of-coolant accident (LOCA) are assessed. For the assessment, fuel pin based power distribution at an initial and an equilibrium core of APR1400 is analyzed. Fuel cladding burst probability curves represented as a local peak power are established up to 70 MWd/kgU fuel burnup. Main uncertainty parameters for pin burst related to the fuel and thermal–hydraulic (TH) performance are identified. Monte Carlo method is used for combining uncertainties. FRAPTRAN and MARS-KS integrated code is employed in this study. In the APR1400 core, power ratio of the fuel pin between a target fuel pin and surrounding ones is ranging 0.899–1.131 within ±2 standard deviation. This power ratio has induced about 1.0–2.1 kW/ft burst power change. In the fuel uncertainties, fuel thermal conductivity, fission gas release (FGR), cladding yield stress shows a relatively strong influence on the burst power change. Groeneveld critical heat flux (CHF) and Chen transition boiling correlation shows a strong influence among TH uncertainty parameters. Established cladding burst probability curves reveal that fresh fuel has a relatively low burst power. But, it increases rather steeply until burnup reaches 5–10 MWd/kgU. Then, the burst power is reduced slowly until the lowest power is attained at the fuel burnup of 70 MWd/kgU. Evaluation of fuel pin burst fraction shows that the most susceptible fuel burnup for the core-wide pin burst is fresh one. Beginning-of-cycle (BOC) at the initial core seems to be the most vulnerable during a reactor life time.

Design and application of supervisory control based on neural network PID controllers for pressurizer system

The pressurizer system (PRZ) plays a vital role in the operation of pressurized water reactors (PWRs). The PRZ is inherently a non-linear, time-variant system, and hence the PRZ modeling with transfer function or linear time-invariant state-space forms is not accurate enough. In this paper, the thermal-hydraulic behavior of the PRZ is modeled by RELAP5 thermal-hydraulic code that is one of the best modeling codes offered for the PRZ system. Although the RELAP5 best-estimate code is appropriate for PRZ modeling, the RELAP5 is not capable to implement the advanced controllers. As first-time, to eliminate this issue, the RELAP5 code and MATLAB software are coupled to use the capabilities of MATLAB (the ability to implement advanced controllers) and RELAP5 (the best model for the PRZ system) simultaneously. Accordingly, this coupling provides a new platform for designing and implementing various intelligent and advanced controllers for PRZ pressure and level in RELAP5 code. Likewise, after providing the platform for implementing intelligent controllers, a novel supervisory control based on neural network PID (NN-PID) controllers is designed for the PRZ pressure and level control. Also, the PRZ system is modeled as a MIMO system with consideration of the PRZ pressure and level interactions. Furthermore, as a case study, the advanced designed control system is applied to adjust the pressure and level of PRZ in a VVER-1000 type nuclear power plant. The study results show that this newly designed control system is able to control the PRZ pressure and level efficiently and effectively in several conditions. Therefore, the proposed methodology as a novel design of PRZ control can be applied and developed in new generations of nuclear power plants and pressurized test loops.

Three-Dimensional Profilometry Utilization on VVER Type Reactors

Abstract The paper is focused on the application of three-dimensional (3D) profilometry on water–water energetic reactor (VVER) type nuclear power plant (NPP) equipment. This method is becoming increasingly used in the power industry and replaces conventional methods such as micrometer measurements. One of the greatest benefits is the accurate recording of the 3D profile of the measured surface and the possibility of its comparison with the production documentation or with the results from previous measurements. Centrum Výkumu Řež, s.r.o. (Research Center Řež) uses 3D laser scanner with a measuring arm. This method was, for example, successfully used for reactor pressure vessel (RPV), steam generator (SG), and bolts. The results are used by the NPP operator for the lifetime management of the primary circuit components.

Assessing the factors affecting the water chemistry parameters in the auxiliary water system of a nuclear power plant

Cooling towers are used as a final heat sink which cools the reactor auxiliary systems and transfers the heat energy to the atmosphere. The reliability endowed upon the cooling tower operation in a nuclear power plant is considerably high when compared to other power plants since the proper functioning of auxiliary systems is crucial for the decay heat removal from the core. Hence, appropriate chemistry control is mandatory in the operation of cooling towers for corrosion control, scale prevention, and control of biological growth. The operating data and water chemistry parameters like conductivity, turbidity, free residual chlorine, pH, cycles of concentration, chlorides, hardness, and alkalinity are analyzed for a typical nuclear power plant. The effect of the individual water quality parameter is correlated by the Pearson matrix to find its influence on the system water chemistry parameter. It is observed that the build-up of chloride ions in the cooling water is the major contributing factor for the cooling tower feed and bleed operations. The aspect of reducing the feed and bleed operation is analyzed, and suitable modifications are suggested to reduce water consumption.

Influence of frequency content of ground motions on seismic fragility of equipment in nuclear power plant

The objective of this study is to investigate the influence of frequency content of ground motions on the seismic fragility of nuclear power plant (NPP) equipment. Transfer functions of a steam generator, a reactor vessel and an auxiliary building of a Korean APR1400 type NPP are extracted from finite element models. The transfer functions are used to investigate the influence of frequency contents of input ground motion on the seismic fragilities of the equipment. A methodology is proposed to transform fragility functions that are defined in terms of peak ground acceleration (PGA) of the design response spectrum to those that are defined in terms of realistic response parameters. The method is applied to a three-dimensional finite element model of the auxiliary building as a case study. The results show that all equipment has a higher probability of failure when the auxiliary building is subjected to ground motions with high-frequency content.

Unmitigated severe accident analysis for a PWR using MELCOR

After the two major nuclear accidents in history considered to be beyond-design-basis accidents (BDBA) – the Chernobyl Reactor Four explosion in Ukraine (1986) and the Fukushima Daiichi accident in Japan (2011), the United States Nuclear Regulatory Commission and other countries have included in the Final Safety Analysis Report (FSAR), a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), with significant damage to the reactor core. The Safety Analysis Report (SAR) is the most important document used by a regulatory body to assess the adequacy of plant safety at all stages of the life time of a Nuclear Power Plant (NPP), and presents the licensing basis of a plant.This study aims at comparing the progression of severe accidents during a Total Loss of Feed Water Accident (TLOFW) and a TLOFW combined with An Anticipated Transient Without SCRAM (ATWS) reactivity accident for a Pressurized Water Reactor (PWR) type NPP. This paper takes into account the results of the PSA Level 1 for the reference NPP. The calculations were performed with the deterministic computer code MELCOR version 2.2 used to analyze severe accident progression in NPP with a Light Water Reactor (LWR). The paper presents a hydrogen risk assessment – the assessment is deterministic, s a prerequisite before performing PSA Level 2. The hydrogen source is determined, too. Saphiro's diagram is applied to investigate the containment atmosphere condition (inert or combustible) for two scenarios. The main results of the study reveal that the containment integrity of the reference PWR NPP when both accidents occurred is maintained. However, the core is severely damaged. A future work will investigate mitigation strategies to limit the hydrogen risk during accident.

A wet-oxidation procedure of radioactive waste resin and waste concentrated liquid for 3H and 14C analysis

3H and 14C are two of the principal isotopes that should be considered in the characterization of radioactive waste from the nuclear power plant. In this paper, an analytical method of 3H and 14C in radioactive waste resin and waste concentrated liquid with a wet-oxidation process was presented, which allowed us to obtain both samples of 3H and 14C simultaneously. The chemical yields were optimized based on an orthogonal experiment with fresh resin. The analytical precision and the minimum detectable concentration (MDC) were provided. The chemical yield of 14C during the process for resin was about 96.8%. There was no significant difference between the analytical results of 14C with the wet-oxidation method and those with the oxygen-bomb combustion method. The MDCs of activity of 3H and 14C were 41 Bq/gww and 1.3 Bq/gww for waste resin, and 1.3 Bq/gww and 0.02 Bq/gww for waste concentrated liquid. The method was applied to determine the radioactivity concentrations in the waste from a typical nuclear power plant. It was shown that the averaged concentrations of 3H and 14C in the radioactive waste resin were 6134 Bq/gww and 2724 Bq/gww, and the averages in waste concentrated liquid were 104.7 Bq/gww and 27.3 Bq/gww, respectively.

Multi-objective-based seismic fragility relocation for a Korean nuclear power plant

An optimal risk-informed seismic fragility relocation is proposed for structures, systems, and components (SSCs) of nuclear power plants (NPPs) that mitigates the core damage risk of the NPP and raises the cost performance. The proposed approach employs a seismic probabilistic safety assessment technique to effectively distribute seismic fragilities of SSCs from the perspective of overall plant risk. The main objectives of the optimal seismic fragility relocation are to mitigate the core damage risk and minimize the total cost for the SSCs of the NPP. To deal with such conflicting purposes, a genetic algorithm-based multi-objective optimization scheme is adopted. This approach systematically obtains a set of Pareto optimal solutions that are neither inferior nor superior to each other. The process for choosing the most appropriate seismic fragility distribution plan from the optimal front of Pareto solutions is also discussed with an emphasis on the study of the uncertainty quantification on such optimal Pareto solutions. As an example for such an optimal fragility relocation study, this study considers a typical type of Korean NPP. The numerical results confirm that the proposed approach provides diverse optimal seismic fragility relocation plans which can effectively reduce the core damage risk as well as the cost compared to the current fragility distribution of this example NPP. Finally, the best optimal alternatives for the fragility relocation are selected based on the Pareto optimal solution’s robustness of performances against their perturbations.