Basic investigation of particle swarm optimization performance in a reduced scale PWR passive safety system design

This work presents a methodology to investigate the viability of using particle swarm optimization technique to obtain the best combination of physical and operational parameters that lead to the best adjusted dimensionless groups, calculated by similarity laws, that are able to simulate the most relevant physical phenomena in single-phase flow under natural circulation and to offer an appropriate alternative reduced scale design for reactor primary loops with this flow characteristics.A PWR reactor core, under natural circulation, based on LOFT test facility, was used as the case study. The particle swarm optimization technique was applied to a problem with these thermo-hydraulics conditions and results demonstrated the viability and adequacy of the method to design similar systems with these characteristics.     

Development of passive safety systems for Next Generation PWR in Japan

Our country’s energy demand is expected to increase steadily into the future. When the situation of our country, which is not rich in energy resources, is taken into account, it seems that the importance of nuclear power generation will be heightened. Based on such a background, the basic policy for nuclear power generation is ‘from light water reactors to fast breeder reactors’. However, considering that light water reactors have become common, the recent outlook for the supply and demand for uranium resources, development trends of fast breeder reactor technology, etc., the light water reactor is expected to remain dominant in our country until at least the second half of the 21st century. Therefore, five PWR utilities in Japan (Hokkaido, Kansai, Shikoku, Kyushu, and Japan Atomic Power), Mitsubishi Heavy Industries Ltd and Westinghouse Electric Corporation have jointly started researching the Next Generation PWR (N.G.P) which is expected to be the leading nuclear power plant, taking place of APWR [T. Magari, Development of Next Generation PWR in Japan, Proceedings of the 10th Pacific Basin Nuclear Conference, 1996; K. Fujimura, et al., Proceedings of the Second International Symposium on Global Environment and Nuclear Energy Systems, 1996]. In this program, construction is targeted to start from 2010 based on expected future environmental conditions. Now, the capacity of more than 1500 MWe class PWR concept is investigated and a plant concept which has innovative features of a hybrid safety systems, i.e. an optimum combination of active and passive safety systems, and horizontal steam generators for core cooling at the accidents is developed as a promising candidate. The plant concept and the results of the investigation are presented in this paper.     

Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

Thermal-hydraulic characteristic investigation on passive residual heat removal system (PRHRS) of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features. A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified. On the basis of theory analysis, a correlation of two-phase natural circulation was obtained, and relative errors of 95% test data were less than ±16%. There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink, and its correlation of two-phase natural circulation system has been obtained. The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.     

Addressing the challenges posed by advanced reactor passive safety system performance assessment

Passive safety features play an essential role in the development of nuclear technology and within advanced water cooled reactor designs. The assessment of the reliability of such systems in the frame of plant safety and risk studies is still an open issue. This complexity stems from a variety of open points coming out from the efforts conducted so far to address the topic and concern, for instance, the amount of uncertainties affecting the system performance evaluation, including the uncertainties related to the thermal-hydraulic (T-H) codes, as well as the integration within an accident sequence in combination with active systems and human actions. These concerns should be addressed and conveniently worked out, since it is the major goal of the international community (e.g. IAEA) to strive to harmonize the different proposed approaches and to reach a common consensus, in order to add credit to the underlying models and the eventual out coming reliability figures. The main key points that may influence the reliability analysis are presented and discussed and a viable path towards the implementation of the research efforts is delineated, with focus on T-H passive systems.     

Survey of experiments and code development for the passive residual heat removal system of PWR in China

Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.     

The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

1 Introduction The technology of passive safety is the trend of safety systems in nuclear power plant, and various novel reactor concepts, including AP600, EPP1000, SPWR, WWER1000, and MS600, have adopted pas- sive safety systems [1]. Passive safety system is one of the main features of Chinese advanced PWR, which is different from other conventional PWR [2]. Passive residual heat removal system (PRHRS), which ac- counts for the majority of passive safety systems of Chinese advanced…     

Efficient reliability estimate of passive thermal hydraulic safety system with automatic differentiation

An approach for efficient estimation of passive safety system functional reliability has been developed and applied to a simplified model of the passive residual heat transport system typical of sodium cooled fast reactors to demonstrate the reduction in computational time. The method is based on generating linear approximations to the best estimate computer code, using the technique of automatic reverse differentiation. This technique enables determination of linear approximation to the code in a few runs independent of the number of input variables for each response variable. The likely error due to linear approximation is reduced by augmented sampling through best estimate code in the neighborhood of the linear failure surface but in a sub domain where linear approximation error is relatively more. The efficiency of this new approach is compared with importance sampling MCS which uses the linear approximation near the failure region and with Direct Monte-Carlo Simulation. In the importance sampling MCS, variants employing random sampling with Box-Muller algorithm and Markov Chain algorithm are inter-compared. The significance of the results with respect to system reliability is also discussed.     

Code validation on a passive safety system test with the SMART-ITL facility

A thermal-hydraulic integral effect test facility, SMART-ITL, was constructed to examine the system performance of SMART, a 330 MWt integral type reactor, and to provide data for validation of related thermal-hydraulic models in the system analysis codes. SMART is equipped with various passive systems such as a passive residual heat removal system (PRHRS), a passive safety injection system (PSIS), and an automatic depressurization system (ADS). The PSIS of SMART is made up of four core makeup tanks (CMTs), four safety injection tanks (SITs), and related piping. Over 10 tests have been performed to investigate the behavior of a single train of a PSIS (a CMT and a SIT) in connection with PRHRSs and an ADS. Using a system analysis code, MARS-KS, we validated the experimental results for a representative test. All geometrical and thermal-hydraulic conditions of SMART-ITL were reflected in the code input construction. Through the validation process, several models, including a break flow model, heat transfer models, and pressure drop models, were examined. Overall, the major system parameters were well reproduced.     

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.     

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.     

发展无严重事故风险核电站的曙光具有完全非能动安全冷却系统的压水堆核电站

本文提出了用自然力直接触发启动压水堆核电站一整套完全非能动的停堆安全冷却系统.这里的自然力主要是指一回路运行工况转换时由于其压力分布变化所形成的压差力.在这一系统中,当进行停堆或发生某种一回路事故工况时,相应的安全冷却系统便自然地投入运行,立即缓解事故后果,将事故时一回路释放的能量及堆芯余热非能动地排入最终热阱.在全过程中不依靠自动控制系统、能动设备及任何人为因素的介入,即可确保对堆芯余热无限期的安全冷却能力,完全避免压水堆核电站发生向环境泄漏放射性物质的严重事故,排除对核电站周围居民进行事故疏散的必要性,彻底解除公众对核电安全的疑虑.实施本文中提出的压水堆核站完全非能动停堆安全冷却系统,完全立足于采用现有成熟技术,因而在近期内即可应用于无严重事故风险新型第三代压水堆核电站的设计与对现有第二代压水堆核电站的技术改造项目.立足于现有压水堆核电站的运行经验,只要进一步采用完全非能动原则,实现固有安全,排除严重事故风险,那么压水堆核电站在近期内即能够成为电网的主要支柱,为破解全人类所面临的严酷环境问题做出重大贡献.     

SOPRAG: a system for boiling water reactors reload pattern optimization using genetic algorithms

Genetic Algorithms (GA) are used in combination with the steady state nodal core simulator PRESTO-B to create a system for the optimization of reload patterns for Boiling Water Reactors (BWR). The system uses the basic GA operators, crossover, mutation and selection over the loading pattern (LP) represented by a combination of fresh and burned fuel assemblies, as well as an objective function taking into account cycle length and radial peaking factor, to obtain improved loading patterns compared with real BWR loadings. The system takes advantage of the efficient quarter core two dimensional (2D) calculations, using the Haling technique to perform thousands of LPs evaluations and obtain the better candidates in a reasonably computer processor (CPU) time.     

AXIAL: a system for boiling water reactor fuel assembly axial optimization using genetic algorithms

A system named AXIAL is developed based on the genetic algorithms (GA) optimization method, using the 3D steady state simulator code Core-Master-PRESTO (CM-PRESTO) to evaluate the objective function. The feasibility of this methodology is investigated for a typical boiling water reactor (BWR) fuel assembly (FA). The axial location of different fuel compositions is found in order to minimize the FA mean enrichment needed to obtain the cycle length under the safety constraints. Thermal limits are evaluated at the end of cycle using the Haling calculation; the hot excess reactivity and the shutdown margin at the beginning of cycle are also evaluated. The implemented objective function is very flexible and complete, incorporating all the thermal and reactivity limits imposed during fuel design analysis; furthermore, additional constraints can be easily introduced in order to obtain an improved solution. The results show a small improvement in the FA average enrichment obtained with the system related to the reference case that has been studied. The results show that the system converge to an optimal solution, it is observed that the mean fuel enrichment decreases while all the constraints are satisfied. A comparison was also performed using one-point and two-points crossover operator and the results of a sensitivity study for different mutation percentage are also showed.     

Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1

A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions.     

Stability of the water-chemistry regime in the loop of a passive safety system of nuclear power reactors

The organization of the water-chemistry regime in the loop of a passive safety system, whose purpose is emergency removal of heat from the core of a nuclear power reactor, is examined. It is shown that a selfregulated water-chemistry regime in which gaseous products of radiolysis can be dissolved in water coolant and recirculated into the irradiation zone, which will intensify liquid-phase radiation-chemical reactions of hydrogen with oxygen and organic release of gases from the liquid phase into the vapor-gas phase of the coolant, can arise in the loop of a passive safety system. This will result in the establishment in the loop of dynamic equilibrium between the release and dissolution of gases and will enable prolonged functioning of the safety system without intervention from the outside. The physicochemical and technical criteria for the appearance of a self-regulated water-chemistry regime for closed loops with natural circulation of the two-phase coolant are formulated and substantiated.     

Power density axial oscillations induced by Xenon dynamics: Parameter identification via genetic algorithms

In this paper we describe the axial flux oscillations in PWRs by means of the Onega and Kisner model (1978), a two-point xenon oscillation model based on the one-group, one-dimensional neutron diffusion equation with nonlinear power reactivity feedback and on the nonlinear xenon and iodine balance equations. We investigate the feasibility of using genetic algorithms for estimating the effective nuclear parameters involved. This approach has the advantage of allowing the periodic re-estimation of the effective parameter values pertaining to each reactor on the basis of its recent history. By so doing, other effects, such as the burn up, are automatically taken into account.     

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.

For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.     

New genetic algorithms (GA) to optimize PWR reactors: Part I: Loading pattern and burnable poison placement optimization techniques for PWRs

The objective of this study was to develop a unique scientific methodology as well as a practical tool for designing the loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and corresponding BP placement design that will achieve maximum cycle length while satisfying the safety constraints. To solve this optimization problem, a core reload optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed. This code is applicable for all types of PWR cores having different geometries and designs with an unlimited number of FA types in the inventory. GARCO has three modes: the user can optimize the core configuration (LP pattern) with or without BPs in the first mode; the second mode is the optimization of BP placement in the core and the last mode is the user can optimize LP and BP placements simultaneously in mode 3. In this study, the first mode finds the optimal LPs using the Haling Power Depletion Method (HPD) for placing BPs in the core. The second mode, which depletes the core accurately, places BPs in the selected optimum LP pattern. This methodology is applied only to the TMI-1 PWR. However, the improved Mode 1 GA option was applied to both the VVER-1000 and the TMI-1 to demonstrate and verify the advantages of the new enhancements in optimizing the LP pattern only. The “Moby-Dick” code is used as reactor physics code for VVER-1000 analysis in this research. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1. The libraries of the BP designs used in SIMULATE-3 in this study were produced by Yilmaz (2005) [Yilmaz, S., 2005. Multilevel optimization of burnable poison utilization for advanced PWR fuel management. Ph.D. Thesis in Nuclear Engineering. the Pennsylvania State University].     

New genetic algorithms (GA) to optimize PWR reactors: Part III: The Haling power depletion method for in-core fuel management analysis

The Haling Power Distribution (HPD) has been applied in a unique process to greatly accelerate the in-core fuel management optimization calculations. These calculations involve; the arrangement of fuel assemblies (FAs) and the placement of Burnable Poisons (BPs) in the fresh FAs. The HPD deals only with the arrangement of FAs. The purpose of this paper is to describe past uses of the HPD, provide an example selected from many similar calculations to explain why and how it can be used, and also to show its effectiveness as a filter in the GARCO GA code. The GARCO (Genetic Algorithm Reactor Core Optimization) is an innovative GA code that was developed by modifying the classical representation of the genotype and GA operators. A reactor physics code evaluates the LPs in the population using the HPD Method, which rapidly depletes the core in a single depletion step with a constant power distribution. The HPD is used basically in GARCO as a filter to eliminate invalid LPs created by the genetic operators, to choose a reference LP for BP optimization, and to create an initial population for simultaneous optimization of the LP and BP placement into the core. The accurate depletion calculation of the LP with BPs is done with the coupled lattice and reactor physics CASMO-4/SIMULATE3 package. However, the fact that these codes validate safety of the core with the added BP placement design also validates the use of the HPD method. The calculations are applied to the TMI-1 core as an example PWR providing concrete results.