Steady-state thermal-hydraulic analysis of SCWR assembly

Among the six gen-IV reactor concepts recommended by the gen-IV international forum (GIF), supercritical water-cooled reactor (SCWR), the only reactor with water as coolant, achieves a high thermal efficiency and, subsequently, has economic advantages over the existing reactors due to its high outlet temperature. A thermal-hydraulic analysis of the SCWR assembly is performed in this paper using the modified COBRA-IV code. Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow and increasing the thermal resistance between moderator channel and its adjacent sub-channels. It is shown that heat transfer deterioration cannot be avoided in SCWR fuel assembly. It is, therefore, highly required to calculate the cladding temperature accurately and to preserve the fuel rod cladding integrity under heat transfer deterioration conditions. __________ Translated from Nuclear Power Engineering, 2007, 28(5): 18–21, 58 [译自: 核动力工程]     

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

In this paper, three‐dimensional (3D) power distribution of newly designed small nuclear reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is pressurized water reactor‐based small nuclear reactor in which plate type fuel element has been used and the core of the reactor has hexagonal type geometry. This paper depicts the design of the reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross‐section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the reactor core by NK/TH coupling approach. Copyright © 2012 John Wiley & Sons, Ltd.     

Experimental study of critical flow of water at supercritical pressure

Experimental studies of the critical flow of water were conducted under steady-state conditions with a nozzle 1.41 mm in diameter and 4.35 mm in length, covering the inlet pressure range of 22.1–26.8 MPa and inlet temperature range of 38–474°C. The parametric trend of the flow rate was investigated, and the experimental data were compared with the predictions of the homogeneous equilibrium model, the Bernoulli correlation, and the models used in the reactor safety analysis code RELAP5/MOD3.3. It is concluded that in the near or beyond pseudo-critical region, thermal-dynamic equilibrium is dominant, and at a lower temperature, choking does not occur. The onset of the choking condition is not predicted reasonably by the RELAP5 code.     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR ZPR ‐ LFR analysis

Studies related to severe core accidents constitute a crucial element in the safety design of Gen‐IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero‐power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen‐IV cores by using critical facilities and surrogate models. The current study introduces the European lead‐cooled system (ELSY) as an additional Gen‐IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium‐cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.     

Combining the nuclear power plant steam cycle with gas turbines

Nuclear steam power plants (NPP) are characterized by low efficiency, compared to steam power plants using fossil fuels. This is due to the relatively low temperature and pressure-throttling conditions of the NPP compared to those using fossil fuel. The light water pressurized water reactor (LW PWR) commercially known as AP600 was suggested for Kuwait cogeneration power desalting plant (CPDP). It has 600 MW nominal power capacity and 33% overall efficiency. Meanwhile, the Kuwaiti Ministry of Electricity and Water (MEW) installed plenty of gas turbines (GTs) to cover the drastic increase in the peak electrical load during the summer season. Combining some of these GTs with the AP600 can increase the capacity and efficiency of the combined plant, compared to either the GT open cycle or the NPP separate plants. This paper investigates the feasibility of utilizing the hot gases leaving the GT to superheat the steam leaving the steam generator of the AP600 NPP, as well as heating the feed water returning to the steam generator of the NPP condenser. This drastically increases the power output and the efficiency of the NPP. Detailed modifications to the NPP power cycle and the resulting enhancement of its performance are presented.     

A review on hydrogen generation,explosion, and mitigation during severe accidents in light water nuclear reactors

Progress of severe accident (SA) can be divided into core degradation and post core meltdown. An important phenomena during severe accidents is the hydrogen generation from exothermal reaction between oxidation of core components, and molten core concrete interaction (MCCI). During the severe accidents, a large amounts of hydrogen is produced, deflagrated and consequently the containment integrity is violated. Therefore, the main objectives of this study is to highlight the source of hydrogen production during SA. First, a thorough literature review and main sources of hydrogen production, hydrogen reduction systems are introduced and discussed. Based on the available results, the amount of produced hydrogen in a typical pressurized water reactor (PWR) and a boiling water reactor (BWR) are estimated to be 1000 and 4000 kg, respectively during in-vessel phase. The average rate of hydrogen production is about 1 kg/s during reflooding of a degraded core. Also, about 2000 kg hydrogen is produced during MCCI for a PWR. The lower and upper range of hydrogen required to initiate combustion is 4.1 and 74 vol percent, respectively. In this paper a review is provided of what has been done in the literature with regard to hydrogen generation in severe accidents of nuclear power plants. In addition, the review identifies the literature gaps and underlines the need of developing a systematic hydrogen management strategy. A hydrogen management strategy is proposed in order to maintain the containment integrity against the probable combustion or hydrogen explosion loads.     

煤矿瓦斯事故中的人因可靠性分析

基于人因可靠性对我国煤矿瓦斯事故的分析,认为在我国煤矿生产过程中存在安全监管不到位、规章制度执行能力差、安全意识差、人员素质低下等问题,本文通对我国近几年发生的煤矿事故的统计,利用故障树方法研究人因可靠性对煤矿瓦斯事故的影响,得到加强监管、规范企业内部管理、做好宣传、进行人员培训和校企合作等解决措施。     

Integrated framework for the external cost assessment of nuclear power plant accident considering risk aversion: The Korean case

Recently, the estimation of accident costs within the social costs of nuclear power plants (NPPs) has garnered substantial interest. In particular, the risk aversion behavior of the public toward an NPP accident is considered an important factor when integrating risk aversion into NPP accident cost. In this study, an integrated framework for the external cost assessment of NPP accident that measures the value of statistical life (VSL) and the relative risk aversion (RRA) coefficient for NPP accident based on an individual-level survey is proposed. To derive the willingness to pay and the RRA coefficient for NPP accident risks, a survey was conducted on a sample of 1550 individuals in Korea. The estimation obtained a mean VSL of USD 2.78 million and an RRA coefficient of 1.315. Based on the estimation results in which various cost factors were considered, a multiplication factor of 5.16 and an external cost of NPP accidents of 4.39E−03 USD-cents/kW h were estimated. This study is expected to provide insight to energy policy decision-makers on analyzing the economic validity of NPP compared to other energy sources by reflecting the estimated external cost of NPP accident into the unit electricity generation cost of NPP.     

Core design of long‐cycle small modular lead‐cooled fast reactor

A core design of small modular liquid‐metal fast reactor (SMLFR) cooled by lead‐bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off‐grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low‐enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S‐CO₂) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the core with low‐enriched uranium (LEU) performs the function of core ignition. The center region plays the role of a breeding blanket to increase the core lifetime for long cycle operation. The core working fluid inlet and outlet temperatures are 300°C and 422°C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi‐static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal‐hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the conceptual core is favorable in terms of self‐controllability, which is the first step towards inherent safety.     

Integral measurements in a simulated fast breeder reactor blanket

Much of the breeding of fissile material and a significant fraction of the energy production in proposed fast breeder reactors will occur in a blanket of fertile material surrounding the reactor core. Current uncertainties in neutron cross sections and calculational methods have much greater significance for blanket design than for core design. The Purdue University Fast Breeder Blanket Facility (FBBF), designed to simulate the neutron behavior in fast breeder reactor blankets, has recently been completed. The facility is now being used to perform integral neutron reaction rate measurements. Such measurements provide tests of nuclear cross sections and calculational methods used in the design of fast breeder reactor blankets.     

Exergy analysis of a pressurized-water reactor nuclear-power plant

An exergy analysis based on the second law of thermodynamics is performed to evaluate the plant and subsystem irreversibility of a nuclear power plant (NPP) with a pressurized-water reactor (PWR). The construction of such a system having a maximum reactor core thermal power of 4250 MW is proposed in Turkey and China. This study concentrates on the questions of where and how much of the available work is lost in such a plant. The evaluated exergy destruction of this plant indicates that the reactor pressure vessel including PWR is the most inefficient equipment in the whole NPP, while the turbines take the second place.     

Quantitative and qualitative analysis of safety parameters in nuclear power plants

After the Fukushima accident, it is necessary to develop some technique that can monitor the progression of severe accidents in nuclear power plants (NPPs). It is therefore very important for an operator to monitor safety related parameters for the diagnosis of severe accidents and to manage it properly. So to monitor and to check the availability of plant instrumentation during severe accidents, this paper presents quantitative and qualitative analyses of safety parameters by using online risk monitor system (ORMS). An ORMS considers the increasing potential for failure for a working component due to aging, which appears in the form of component's performance degradation. ORMS therefore requires a continuous feedback regarding performance and failure probabilities of components, which directly or indirectly contributes to the failure of a system. ORMS has been designed to automatically update the online risk models and reliability parameters of equipment. A case study of emergency diesel generator (EDG) of Daya Bay NPP has been performed, and operational failure rate and demand failure probability of EDG have been calculated with the help of ORMS. The results of ORMS are well matched with data obtained from Daya Bay NPP. ORMS can support in decision‐making process for operators and managers at NPPs. Copyright © 2013 John Wiley & Sons, Ltd.     

On the design of the tubesheet and the tubesheet-to-shell junction of a fixed tubesheet heat exchanger

For the (mechanical) design of an existing fixed tubesheet heat exchanger, a C2-Hydrogenation reactor in a petrochemical plant, various code solutions are compared with each other and with a Finite Element solution based on the Direct Route in Design by Analysis (EN 13445-3, Annex B). The codes and standards used in the investigation are ASME Section VIII, Division 1 and EN 13445-3, Clause 13 and Annex J.     

The development of structural integrity criteria for austenitic components

The high ductility and resistance to tearing of austenitic stainless steels, allied to the geometry and loading conditions of the European fast reactor (EFR), result in the prediction of large critical through-thickness crack sizes for EFR components. Based on this understanding of large critical crack sizes, simplified methods have been developed for the assessment of the leak-before-break behaviour of these components. The methods described avoid the need for detailed analysis of crack shape development during cyclic loading, and thus greatly reduce the calculational requirements compared with more common methods used for the assessment of known or postulated defects in high energy pressurised components. Simplifications used in the assessment of leak-tightness in EFR components are also identified.     

Development and application of CFD and subchannel coupling analysis code for lead‐cooled fast reactor

It is of necessity and importance for the simulation of the three‐dimensional thermal hydraulics problem of the pool type fast reactor. However, because of current computing power limitations and the complexity of the reactor core structure, for conventional reactor applications, it is still not possible to directly simulate the entire reactor flow with sufficient fine meshes for detailed pin geometry. Until now, there is a multiscale coupling method which is suitable to deal with this type of simulation challenge. Through the user‐defined function (UDF) of FLUENT, the coupling code FLUENT/KMC‐sub for thermal hydraulic (TH) analysis by coupling the dynamic link library (DLL) complied by the transient subchannel code KMC‐sub is developed by University of Science and Technology of China (USTC). As a code validation case, the steady‐state simulation of a 19‐rod assembly has been carried out by using coupling codes of FLUENT/KMC‐sub, FLUENT and KMC‐sub, and consequently good consistency has been achieved by comparison with experiment results. And coupled code is further tested by comparison with the transient‐state 19‐pin assembly test results of KMC‐sub and FLUENT simulation. This coupling code is then used for TH of M2LFR‐1000 (medium‐size modular lead‐cooled fast reactor) in unprotected loss of flow (ULOF) accident. The transient temperatures of coolant and fuel and multidimensional TH phenomena and safety analysis are presented and discussed in this article.     

A novel cross-linked anion exchange membrane with conjugated and non-conjugated pyridine groups

A novel cross-linked pyridine anion exchange membrane was synthesized via thermal cross-linking methods. 2,5-bis(2,3,4,5,6-pentafluorophenyl)-1,3,4-oxadiazole (FPOx) and diallyl bisphenol A (DABPA) were reacted to obtain the main polymer chain (MP). MP was substituted with 4-bromopyridine and 4-hydroxypyridine to generate two pyridine polymers conjugated pyridine polymer (CPP) and non-conjugated pyridine polymer (NPP). The different conjugation statues of CPP and NPP led to distinct mechanical and electrochemical properties. The diallyl groups in MP acted as the cross-linking structures which strengthened the mechanical properties of CPP and NPP and provided more flexible side chains. The ultimate membrane CPP-membrane and NPP-membrane were prepared by solution casting methods. Compared with NPP-membrane, the ionic conductivity of CPP can achieve to 20.1 mS/cm at 20 °C with excellent mechanical and thermal characteristics. Quantum theory computation and AFM morphology were carried out to figure out the reason of the difference in ionic conductivities and physical properties.     

Determination of PAR configuration for PWR containment design: A hydrogen mitigation strategy

This paper deals with determination of the minimum number and identification of the best configurations of passive autocatalytic recombiners (PAR) for the effective design of the containment in a pressurized water reactor (PWR). It considers the current design of PAR in the containment of a PWR and for that tries to identify, through a large number of sensitivity analyses, the minimum required number of PARs in different compartments. In this regard, a qualified nodalization has been developed for best estimate modeling by MELCOR integrated code. The developed model includes primary and secondary systems, containment, and related safety systems. A large number of simulations including the plant specific probabilistic safety assessment and success criteria analysis are used to identify the accident scenario with the highest amount of hydrogen production and risk. We first screened postulated accidents based on the PSA results and then based on the deterministic severe accident computations. It is found that the large break loss of coolant accident (LB-LOCA) without emergency core cooling system (ECCS) actuation is the bounding case from the hydrogen hazard point of view. To find the optimal configuration with minimum number of PARs in the containment, 40 different configurations are analyzed for the selected accident for a Westinghouse type PWR. The main finding of this work is identification of the minimum required number of PARs and their best distribution among the associated compartments. The obtained configuration is equally effective for the hydrogen risk mitigation with 36% reduction in the number of PARs in comparison to the base case design. The methodology of the analysis can be used for other plants.     

A review of fault ride through of PV and wind renewable energies in grid codes

This paper presents a comprehensive review of fault ride through (FRT) in the grid code of 38 selected countries with an emphasis on renewable energy (REN) sources–related rules. Grid codes are the rules legislated usually by the transmission system operators (TSOs) to determine the grid integration requirements of electrical power generators. Each country establishes its grid code for satisfying the minimum required technical criteria and revises it frequently to cope with new modifications of the utility. Growing the penetration of REN sources have influenced many operational aspects of the power system such as protection, power quality, reliability, and stability. Thereupon, regulations must ensure the power system's secure and controllable operation of REN sources. FRT is one of the main parts of the grid code, and its characteristics affect the performance and rating of power system apparatus. FRT defines the performance of electric power generators during and in postfault conditions. FRT of solar photovoltaic (PV) and wind turbines (WTs) as the main REN sources of energy has great importance in the grid codes. In this paper, a comparison of FRTs in the grid code of 38 countries, including low‐voltage ride through (LVRT), zero‐voltage ride through (ZVRT), and high‐voltage ride through (HVRT) are provided and surveyed.     

Evaluation of reaction kinetics for the catalyst used in PCRD and study of channel affect on the same

Hydrogen mitigation strategies have gained importance for Nuclear reactors owing to damage caused to integrity of reactor containment by hydrogen fire in three major nuclear accidents of Three Miles Island, Chernobyl and Fukushima. One promising technology for hydrogen mitigation is deploying Passive Catalytic Recombiner Devices (PCRDs). Principle involved here is recombining hydrogen released during accident with oxygen from ambient air inside reactor containment on catalyst surface to form steam. Present work focuses on experimental evaluation of reaction kinetics associated with hydrogen-oxygen recombination on surface of indigenous PCRD catalyst developed for Indian Nuclear Power Plants. Behavior of catalyst plates stacked in parallel inside PCRD has also been evaluated. This effect is due to difference in migration mechanisms of reactants and products to and from the catalyst surface. Overall affect has been empirically approximated as single step Arrhenius equation. This is significant in modelling of PCRDs for faster containment analysis using CFD.     

Phenomena identification and ranking table exercise for thorium based molten salt reactor-solid fuel design

Thorium based molten salt reactor-solid fuel (TMSR-SF) design is an innovative reactor concept that uses high-temperature tristructural-isotropic (TRISO) fuel with a low-pressure liquid salt coolant. In anticipation of getting licensed applications for TMSR-SF in the future, it is necessary to fully understand the significant features and phenomena of TMSR-SF design, as well as its transient behavior during accidents. In this paper, the safety-relevant phenomena, importance, and knowledge base were assessed for the selected events and the transient of TMSR-SF during station blackout scenario is simulated based on RELAP/SCDAPSIM Mod 4.0.The phenomena having significant impact but with limited knowledge of their history are core coolant bypass flows, outlet plenum flow distribution, and intermediate heat exchanger (IHX) over/under cooling transients. Some thermal hydraulic parameters during the station blackout scenario are also discussed.