MWe Pressurized Water Reactor Research Articles

Safety margin quantification by integrating probabilistic and deterministic safety assessments — Application to design extension conditions

Safety margin quantification by integrating probabilistic and deterministic safety assessments — Application to design extension conditions

Advanced Multi-Modeling of PWR Dynamics and Deep Learning based Computational Tool in SIMULINK and LabVIEW

The reactivity monitoring, prediction, and investigation is the most important parameter to ensure the safety and reliable operation of a nuclear power plant. This parameter is gained further importance in Pressurized Water Reactor (PWR) due to more sophisticated reactivity insertion mechanisms and innovative reactor core fuel loading scheme. Based on deterministic internal and external dynamics and neutronics analysis of Advanced PWR, all the reactivity feedback effects such as Doppler effect, moderator effect, control rod effect, liquid boron effect and reactor poisons effect are investigated, modeled and stochastically optimized using deep artificial intelligence. Advance Pressurized Water Reactor (APWR) of 600 MWe rating (AP-600) is used as a reference reactor model and based on the dynamics of AP-600, an Advanced Pressurized Water Reactor Dynamics and Intelligent Stochastic Optimization based Deterministic Neutronics Simulation (APD-ISO-DNS) Code is developed in the hybrid SIMULINK andLabVIEW environments. AP-600 reactor model is fine-tuned and adjusted for 300 MWe PWR (P-300) and 1070 MWe Advanced Chinese PWR (ACP-1000) using neutronics parameters and operational dynamic data of operating PWR nuclear power plants in Pakistan. Various load reduction transient experiments are conducted and dynamic transient simulations of P-300, AP-600 and ACP-1000 are evaluated in SIMULINK and in LabVIEW environments and found as per design basis.

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.

PWR core loading pattern optimization with adaptive genetic algorithm

In this paper, a new adaptive genetic algorithm (AGA) is proposed for the loading pattern optimization (LPO) of the pressurized water reactor (PWR) core. The tournament selection, the two-point crossover and the mutation based on randomly swapping positions between two fuel assemblies (FAs) are applied. New calculation formulae are introduced to adjust effectively the crossover and mutation probabilities according to the fitness value of the individual. The proposed algorithm is implemented in the LPO for the first core of the 1000 MWe PWR. The objective function is to minimize the maximum radial power peaking factor (RPPF) at the equilibrium of Xe under the constraint condition for the cycle length. The maximum RPPF of the obtained LP is decreased than that of the loading pattern (LP) by the designer. The results show that the proposed AGA is effective to improve the convergence rate of genetic algorithms (GAs) in the LPO.

NUCLEAR POWER IN PAKISTAN

Pakistan started its nuclear power program by installing a 137 MWe Canadian Deuterium Reactor (CANDU) at Karachi in 1971 which became operational in 1972. The post-contract technical support for the Karachi Nuclear Power Plant (KANUPP) was withdrawn by Canada in 1976 as a consequence of Indian nuclear device test in 1974. Inspite of various difficulties PAEC resolved to continue to operate KANUPP and started a process for the indigenous fabrication of spare parts and nuclear fuel. The first fuel bundle fabricated in Pakistan was loaded in the core in 1980. Since then KANUPP has been operating on the indigenously fabricated fuel. The plant computer systems and the most critical instrumentation and Control systems were also replaced with up-to date technology. In 2002 KANUPP completed its original design life of 30 year. A program for the life extension of the plant had already been started. The second nuclear power plant of 300 MWe pressurized water reactor purchased from China was installed at Chasma in 1997, which started commercial operations in 2001. Another unit of 300 MWe will be installed at Chashma in near future. These nuclear power plants have been operating under IAEA safeguards agreements. PAEC through the long-term performance of the two power plants has demonstrated its competence to safely and successfully operate and maintain nuclear power plants. Pakistan foresees an increasingly important and significant share of nuclear power in the energy sector. The Government has recently allocated a share of 8000 MW for nuclear energy in the total energy scenario of Pakistan by the year 2025.

SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

ABSTRACT This study focuses on the in-vessel phase of severe accident management (SAM) strategy for a hypothetical 1000 MWe pressurized water reactor (PWR). To examine the effectiveness of SAM strategy, it is necessary to identify and assess epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is reactor coolant system (RCS) depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, sensitivity analyses are carried out to assess the impact of the cutoff porosity related to the flow area of core node (EPSCUT), the critical temperature (TCLMAX) and the minimum fraction of oxidized Zr (FZORUP) for cladding rupture, and the flag to divert gas flows in the core to the bypass channel (FGBYPA) on the core melting and relocation process. In this study, the effect of injection time on the integrity of RPV has been examined based on the quantification of the scenario and phenomenological uncertainties.

An Interval Approach to Nonlinear Controller Design for Load-Following Operation of a Small Modular Pressurized Water Reactor

A pressurized water reactor (PWR) is most suitable for load-following operation because of its self-regulating nature. However, rapid power maneuvering at an enhanced rate is heavily constrained by the overall power coefficient of reactivity which depends on both the reactor neutronics and thermal hydraulics. In this paper, a 100 MWe PWR module with an integrated steam generator, is modeled using the nonlinear equations representing both the neutronics and thermal hydraulics. Such a PWR is typically an integrated structure with a steam generator included in the same casing and can either be connected to the grid or connected with other such modules using a multimodular approach to constitute a Nuclear Steam Supply System (NSSS) driving a common turbine and an alternator. In this paper, Nonlinear Dynamic Inversion (NDI) technique is used to design a controller capable of controlling such a reactor for load-following with frequent changes in demand, in both the modes of operation, namely, reactor-follow-turbine and turbine-follow-reactor. With the assumption that the parameters of the reactor vary within intervals as the power varies, an interval approach coupled with NDI ensures that the controller satisfies the operating constraints. The methodology is established for both the modes with credible real-time Hardware-in-Loop (HiL) simulations.

Transient Diagnosis and Prognosis for Secondary System in Nuclear Power Plants

Abstract This paper introduces the development of a transient monitoring system to detect the early stage of a transient, to identify the type of the transient scenario, and to inform an operator with the remaining time to turbine trip when there is no operator's relevant control. This study focused on the transients originating from a secondary system in nuclear power plants (NPPs), because the secondary system was recognized to be a more dominant factor to make unplanned turbine-generator trips which can ultimately result in reactor trips. In order to make the proposed methodology practical forward, all the transient scenarios registered in a simulator of a 1,000 MWe pressurized water reactor were archived in the transient pattern database. The transient patterns show plant behavior until turbine-generator trip when there is no operator's intervention. Meanwhile, the operating data periodically captured from a plant computer is compared with an individual transient pattern in the database and a highly matched section among the transient patterns enables isolation of the type of transient and prediction of the expected remaining time to trip. The transient pattern database consists of hundreds of variables, so it is difficult to speedily compare patterns and to draw a conclusion in a timely manner. The transient pattern database and the operating data are, therefore, converted into a smaller dimension using the principal component analysis (PCA). This paper describes the process of constructing the transient pattern database, dealing with principal components, and optimizing similarity measures.

Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.

D114 高温配管肉厚連続モニタリング装置を用いたPWR二次系水質改善によるFAC抑制効果の検証(OS6 保全・設備診断技術(2))

Flow Accelerated Corrosion (FAC) of carbon steel (CS) piping is one of main issues in secondary system of Pressurized Water Reactor (PWR) nuclear power plant Wall thinning of CS piping due to FAC increases potential risk of pipe rupture, cost for inspection and replacement of damaged pipes In particular, corrosion products generated by FAC of CS piping brought steam generator (SG) tube corrosion and degradation of thermal performance, when it intruded and accumulated in secondary side of PWR. ^(1) Therefore, we estimated an applicability and effectiveness of Oxygenated Water Chemistry (OWC) as a new approach to FAC suppression for PWR secondary system by experiment in laboratory and demonstration test at Tsuruga-2 (1160 MWe PWR, commercial operation started in 1987). ^<(2)(3)> Based on this result, OWC has been applied to condensate system in Tsuruga-2 from Jan 2011 And, to evaluate the FAC mitigation effect of OWC, wall thickness of actual condensate piping in Tsuruga-2 after OWC application was measured by continuous monitoring system, using high-temperature and high-resolution ultrasonic probe In this paper, behavior of actual wall thinning rate of condensate piping in Tsuruga-2 under various pH and DO concentration is discussed

Multivariable H∞ Loop-Shaping Control for a PWR Nuclear Power Plant

Abstract For a Pressurized Water Reactor (PWR) nuclear power plant, a control-oriented dynamic model, based on lumped-parameter mass, energy and momentum conservation equations, is provided first, with focus on the reactor core, the primary circuit, the steam generators and the steam turbine system. A 900 MWe PWR plant for which operating data are available in the literature is taken as a reference for model parameter tuning. Then, the reactor-turbine control problem is considered, in normal operating conditions, in the reactor-following operating mode: steam pressure at the turbine inlet and the generated electric power are controlled by the reactor control rods and the turbine inlet valve, so as to be able to comply with load variation requests. Primary grid frequency control is considered as well. These problems are tackled by a multivariable H ∞ loop-shaping technique, thus obtaining robustly stabilizing controllers ensuring effective reference tracking and good disturbance rejection. The performance of the proposed control scheme is tested by MATLAB-Simulink® simulations.

Particle Population Diagnosis and Euclidean Minimum Spanning Tree in Monte Carlo Calculation of Power Distribution

A mesh-input-free method is not yet established for particle population diagnosis in the power distribution calculation by the Monte Carlo source iteration method. To approach this issue, the Euclidean minimum spanning tree (EMST) in the graph theory was applied to the source particles. A characteristic volume that one particle covers was defined to be the cubic power of the average edge length of EMST. Thirty and one hundred times of the characteristic volume were proposed as weak and strong requirements, respectively, for a minimum tally cell volume since ten particle characteristic volumes can be accommodated within the tally cells producing one-third and 10% of average power density. These requirements were examined against a three-dimensional full-core model of a 1,100MWe pressurized water reactor. The comparison with the population diagnosis with a mesh in Nucl. Sci. Eng., 158, 15 (2008) shows a lot of promise of the EMST-based approach. Further developmental issues are identified concerning computational time and output fitting. In addition, a practically useful result was obtained as follows. If the three-dimensional uniform tally cells have volume larger than the quarter fuel bundle unit, the EMST-based approach yields a less conservative diagnosis than fissile volume per source particle.

Qualification of the CFD code Trio_U for full scale reactor applications

Abstract The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem). 1 The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.

06/00638 De-regulated electric power markets and operating nuclear power plants: the case of British energy: Hewlett, J. G. Energy Policy, 2005, 33, (18), 2293–2297.

A methodology has been developed to evaluate the reliability of passive systems characterised by a moving fluid and whose operation is based on thermal–hydraulic (T-H) principles. The methodology includes:•identification and quantification of the sources of uncertainties and determination of the important variables;•propagation of the uncertainties through T-H models and assessment of T-H passive system unreliability;•introduction of passive system unreliability in the accident sequence analysis.Each step of the methodology is described and commented and a diagram of the methodology is presented. An example of passive system is presented with the aim to illustrate the possibilities of the methodology. This example is the Residual Passive heat Removal system on the Primary circuit (RP2), an innovating system supposed to be implemented on a 900 MWe Pressurized Water Reactor.

Methodology for the reliability evaluation of a passive system and its integration into a Probabilistic Safety Assessment

Abstract A methodology has been developed to evaluate the reliability of passive systems characterised by a moving fluid and whose operation is based on thermal–hydraulic (T-H) principles. The methodology includes: • identification and quantification of the sources of uncertainties and determination of the important variables; • propagation of the uncertainties through T-H models and assessment of T-H passive system unreliability; • introduction of passive system unreliability in the accident sequence analysis. Each step of the methodology is described and commented and a diagram of the methodology is presented. An example of passive system is presented with the aim to illustrate the possibilities of the methodology. This example is the Residual Passive heat Removal system on the Primary circuit (RP2), an innovating system supposed to be implemented on a 900 MWe Pressurized Water Reactor.

Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling

Abstract If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

Experimental research progress on critical heat flux of Chinese PWR

Abstract One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of Φ10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR.

A PSA-based vital area identification methodology development

Abstract This paper presents a vital area identification method based on the current probabilistic safety assessment (PSA) techniques. The vital area identification method in this paper is focused on core melt rather than radioactive material release. Furthermore, it describes a conceptual framework with which the risk from sabotage-induced events could be assessed. Location minimal cut sets (MCSs) are evaluated after developing a core melt location fault tree (LFT). LFT is a fault tree whose basic events are sabotage-induced damages on the locations within which various safety-related components are located. The core melt LFT is constructed by combining all sequence LFTs of various event trees with OR gates. Each sequence LFT is constructed by combining the initiating event LFT and the mitigating event LFTs with an AND gate. The vital area could be identified by using the location importance measures on the core melt location MCSs. An application was made to a typical 1000 MWe pressurized water reactor power plant located at the Korean seashore. The methodology suggested in the present paper is believed to be very consistent and most complete in identifying the vital areas in a nuclear power plant because it is based on the well-proven PSA technology.