Experimental investigation of natural circulation instability in a BWR-type small modular reactor

The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV). Specifically, the NMR is one third the height and area of a conventional BWR RPV with an electrical output of 50 MWe. Experiments are performed in a well-scaled test facility to investigate the thermal hydraulic flow instabilities during the startup transients for the NMR. The scaling analysis for the design of natural circulation test facility uses a three-level scaling methodology. Scaling criteria are derived from non-dimensional field and constitutive equations. Important thermal hydraulic parameters, e.g. system pressure, inlet coolant flow velocity and local void fraction, are analyzed for slow and fast normal startup transients. Flashing instability and density wave oscillation are the main flow instabilities observed when system pressure is below 0.5 MPa. And the flashing instability and density wave oscillation show different type of oscillations in void fraction profile. Finally, the pressurized startup procedure is recommended and tested in current research to effectively eliminate the flow instabilities during the NMR startup transients.     

Critical crack depth diagram of reactor vessel for pressurized thermal shock

During the operation of a pressurized water reactor, a certain type of transients could induce rapid cooldown of the reactor pressure vessel (RPV) with relatively high or increasing system pressure. This induces a high tensile stress at the inner surface of the RPV, which is called the pressurized thermal shock (PTS). The structural integrity of the RPV during PTS should be evaluated assuming the existence of a flaw at the vessel. For the quantitative evaluation of the vessel failure risk associated with PTS, the probabilistic fracture mechanics (PFM) analysis technique has been widely used. But along with PFM analysis, deterministic analysis is also required to determine the critical time interval in the transient during which mitigating action can be effective. In this study, therefore, the procedure for the deterministic fracture mechanics analysis of RPV during PTS is investigated using the critical crack depth diagram and the computer program to generate it is developed. Four transients of typical PTS, steam generator tube rupture, small break loss of coolant accident and steam line break are analyzed, and their response characteristics such as critical crack depth and critical time interval from the initiation of the transient are investigated.     

Dependence of irradiation creep on temperature and atom displacements in 20% cold worked type 316 stainless steel

Irradiation creep studies with pressurized tubes of 20% cold worked Type 316 stainless steel were conducted in the Second Experimental Breeder Reactor. These studies have shown that as atom displacements are extended above 5 dpa and temperatures are increased above 375°C, the irradiation induced creep rate increases with both increasing atom displacements and increasing temperature. The stress exponent for irradiation induced creep remained near unity. Irradiation induced effective creep strains up to 1.8% were observed without specimen failure.     

NHR—5负荷自跟随特性与机理的分析

用带有高过冷沸腾模型的ＲＥＴＲＡＮ－０２程序对５ＭＷ核共热核负荷增加５０％和失去全部负荷的自跟随实验进行了分析，其计算结果与实验结果符合得很好。并对额定工况下负荷增加３０％和失去全部负荷两种工况的安全性进行了分析，结果表明反应无所作为理安全的。     

中国试验快堆（CEFR）典型事故工况下的瞬态分析

本文采用计算机仿真的方法，对我国首座试验快堆ＣＥＦＲ在几种设计基准事故下的动态响应过程进行了分析计算。计算结果表明，当保护停堆系统正常工作时，ＣＥＦＲ在所分析的事故下具有良好安全性。     

Features and performance of a coupled three-dimensional thermal-hydraulic/kinetics TRAC-PF1/NEM pressurized water reactor (PWR) analysis code

This paper summarizes the current status of the Pennsylvania State University (PSU) version of the coupled three-dimensional (3-D) thermal-hydraulic/kinetics TRAC-PF1/NEM code for pressurized water reactor (PWR) transient and accident analysis and describes applications to reactivity insertion accident (RIA) simulations as well as recent developments. The TRAC-PF1/NEM methodology utilizes closely coupled 3-D thermal-hydraulics and 3-D core neutronics transient models to simulate the vessel and a 1-D simulation of the primary system. An efficient and flexible cross-section generation procedure was developed and implemented into TRAC-PF1/NEM. These features make the coupled code capable of modeling PWR reactivity transients, including boron dilution transients, in a reasonable amount of computer time. Three-dimensional studies on hot zero power (HZP) rod ejection and main steam line break (MSLB) transients in a PWR, as well as a large break loss-of-coolant-accident (LBLOCA) and boron dilution transients, were accomplished using TRAC-PF1/NEM. The results obtained demonstrate that this code is appropriate for analysis of the space-dependent neutronics and thermal-hydraulic coupled phenomena related to most current safety issues.     

Temperature dependence of the 20% cold worked 316 stainless steel steady state irradiation creep rate

The irradiation creep data from four completed tests have been analysed to show that the steady state irradiation creep rate exhibits a moderate and complex temperature dependence. The irradiation creep tests were performed in the Experimental Breeder Reactor Number II (EBR-II) using beams and pressurized tubes, and in the Oak Ridge Reactor (ORR) and the High Flux Isotope Reactor (HFIR) using pressurized tubes. The data cover the temperature range from 200°C to 585°C, and show that from 200°C to 330°C, the steady state rate increases moderately with increasing temperature. At about 330°C, the steady state rate peaks and rapidly decreases with increasing temperature from 330°C to 370°C. From 370°C to 585°C the steady state rate moderately increases with increasing temperature.     

A fuzzy modeling approach to the identification of transients in nuclear components

In this paper, a fuzzy-logic modeling approach is adopted for the identification of the causes of transients in a component of a nuclear power plant. The if–then rules, representing the underlying processes are inferred from the available input–output signal data. The method is applied to the early classification of the causes of transients in a steam generator of a pressurized water reactor (PWR). Based on the measured signals, the forcing function responsible for the transient is readily classified and its amplitude is estimated. The case of two concurrent causes of transients is also considered.     

A 10% cold leg break test at ROSA-IV large scale test facility

The Large Scale Test Facility (LSTF) in the ROSA-IV Program is an integral test facility for investigation of pressurized water reactor (PWR) thermal-hydraulic behavior during small break loss-of-coolant accidents (SBLOCAs) and operational transients.A 10% cold leg break test was conducted in the facility shakedown phase to assess and confirm the facility capability and to collect code assessment data. The test conditions, test procedures and test results are described. The test results are compared with a pretest analysis obtained using RELAP5/MOD1 Cy18.     

Simulation of reactivity transients in current MTRs

This work aims at simulation of reactivity induced transients in High Enriched Uranium (HEU) and Low Enriched Uranium (LEU) cores of a typical Material Test research Reactor (MTR) using PARET code. The transient problem was forced through specification of externally inserted reactivity as a function of time. Reactivity insertions are idealized by ramps and steps. Superdelayed-critical transients, superprompt-critical transients and quasistatic transients are selected for the analysis. Ramp and step reactivity functions were employed to simulate these perturbations. The effect of initial power on transient behavior has also been investigated. The low enriched uranium core is analyzed for transients without scram. The magnitudes of maximum reactivity insertions are chosen to be in the range of $0.05 to 2.0 for different reactivity insertion times. Transient simulation with scram reveals that response of both HEU and LEU-cores is similar for selected ‘ramps’ and ‘steps’. The difference is observed in the peak values of power and coolant, clad and fuel temperatures. Trip level is achieved earlier in case of LEU-core. The peak clad temperatures in both LEU and HEU-cores remain below the melting point of aluminum-clad for the selected reactivity insertions. Simulation show that the LEU-core is more sensitive to perturbations at low power as compared to the transients at full power. For reactivity transients at low power level, power rises sharply to a higher peak value. In transients at full power, the peak power barely exceeds the trip level. The power oscillations after the first peak are observed for transients without scram.     

承压热冲击下压力容器断裂力学分析

依据美国核管会（NRC）最新法规要求和研究进展,阐述了压水堆核电厂反应堆压力容器（RPV）承压热冲击（PTS）最新评估方法。基于热工水力系统程序RELAP5和有限元分析软件ANSYS,针对某传统二代压水堆核电厂模拟在PTS典型瞬态过程下热工响应行为及压力容器模型断裂力学分析,并评估不同瞬态的危险性及其随压力容器材料脆性的变化。分析表明：表面裂纹和靠近内壁面的埋藏裂纹比深埋裂纹更易发生开裂;同等条件下轴向裂纹较环向裂纹更易开裂,且大中破口事故下轴向裂纹远较环向裂纹更易贯穿壁厚。     

Gas-cooled reactor thermal hydraulic analyses with MELCOR

Pursuant to the Energy Policy Act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the reference design for the Next Generation Nuclear Plant (NGNP). Stemming from a U.S. Nuclear Regulatory Commission (NRC) HTGR research initiative, a need was identified for validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform licensing analyses. Because the NRC has used MELCOR for light water reactors (LWR) in the past and because MELCOR was recently updated to include gas-cooled reactor (GCR) physics models, MELCOR is among the system codes of interest to the NRC. This paper describes MELCOR modeling of the General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHGTR is a suitable design for demonstration of MELCOR GCR modeling competency for two reasons: 1) the MHTGR is a predecessor to the more advanced General Atomics’ Gas-Turbine Modular High Temperature Reactor (GTMHR), and 2) experimental data useful for benchmark calculations may soon become available. Using the most complete literature references available for the MHTGR design, researchers at Texas A&M University (TAMU) constructed a MELCOR input deck for the MHTGR to partially validate MELCOR GCR modeling capabilities. Normal and off-normal system operating conditions were modeled with appropriate boundary and initial conditions. MELCOR predictions of system response were obtained for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) scenarios. Code results were checked against nominal MHTGR design parameters, physical intuition, and anticipated GCR thermal hydraulic response. No inherent deficiencies in MELCOR modeling capability were observed, suggesting that the newly-implemented GCR models are adequate for systems-level analysis. If and when experimental benchmark data becomes available, further validation activities may proceed given the modeling efforts discussed herein.     

Containment concepts for high temperature reactors

Different containment concepts have been proposed for High Temperature Reactors. In the paper the confinement, the gastight pressurized containment and the vented confinement are discussed. For a small HTR such as the Modul it seems to be possible to provide a vented confinement instead of a gastight containment. The German Reactor Safety Commission has given a positive statement. Due to the specific safety characteristics of the HTR the safety concepts can differ in part quite considerably from current LWR standard solutions.     

Experimental study of natural circulation instability with void reactivity feedback during startup transients for a BWR-type SMR

The natural circulation boiling type SMR can experience flow instability during the startup transients due to the void reactivity feedback. A BWR-type natural circulation test loop has been built to perform the nuclear coupled startup transient tests for Purdue Novel Modular Reactor (NMR). This test loop is installed with different instruments to measure various thermal hydraulic parameters. The testing process can be monitored and controlled through PC with the assistance of LabVIEW procedure. The effects of power ramp rate on the flow instability during the nuclear coupled tests were investigated by controlling the power supply based on the point kinetics model with coolant void reactivity feedback. Two power ramp rates were investigated and the results were compared with those of the thermal hydraulic startup transients without void reactivity feedback. The time trace of power supply, system pressure, natural circulation rate, and void fraction profile are used to determine the flow stability during the transients. The results show that nuclear coupled startup transients also experience flashing instability and density wave oscillations. The power curves calculated from point kinetics model for startup transients show some fluctuations due to void reactivity feedback. However, the void reactivity feedback does not have significant effects on the flow instability during the startup procedure for the NMR.     

Fuzzy identification of transients in nuclear power plants

The identification of transients is of fundamental importance for the timely monitoring of nuclear plants operation. The main target is detecting the occurrence of the onset of a transient, in its early stages, and identifying its kind so as to be able to readily act to fix its causes. Given the safety and economical importance of the problem, various approaches have been investigated and applied for transient identification, and many efforts are still devoted to the improvement of the results so far obtained.In this paper, a fuzzy-logic based method for the identification of transients is proposed. The method is ‘model-free’ in that the if-then rules, which constitute the heart of the approach, are inferred only from the available input-output signal data. The method is tested on an example of identification of reactor transients generated by four forcing functions of different nature. The necessary data for the identification have been simulated by the QUAndry-based Reactor Kinetics code (QUARK, distributed freely by NEA) configured so as to model the operations of the Westinghouse Advanced Pressurized water reactor, AP600.     

Simple mixing model for pressurized thermal shock applications

The phenomenon of fluid/thermal mixing in the cold leg and downcomer of a Pressurized Water Reactor (PWR) has been a critical issue related to the concern of pressurized thermal shock. The question of imperfect mixing arises when the possibility of cold emergency core cooling water contacting the vessel wall during an overcooling transient could produce thermal stresses large enough to initiate a flaw in a radiation embrittled vessel wall. The temperature of the fluid in contact with the vessel wall is crucial to a determination of vessel integrity since temperature affects both the stresses and the material toughness of the vessel material. A simple mixing model is described which was developed as part of the EPRI pressurized thermal shock program for evaluation of reactor vessel integrity.     

A quantitative methodology for reactor vessel pressurized thermal shock decision making

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Consideration of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS.A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation.Once the specific event sequences of concern are identified, detailed thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. This paper addresses key aspects of the thermal-hydraulic and fracture mechanics analyses of the reactor vessel. The effects of incomplete mixing of safety injection flow in the primary cold leg and vessel downcomer and the application of warm prestressing are emphasized. The results of these analyses are being used to define further modifications in vessel and plant system design and to operating procedures.Previous design considerations that have evolved as a result of reactor vessel integrity evaluations are mentioned. These include the development of realistic design analysis tools and selection of plant system modifications. Modifications that are being developed or are under consideration are also mentioned. These include vessel fluence reductions, additional modifications to operating procedures, increased use of probabilistic event sequence and fracture mechanics analysis methods, enhanced material fracture toughness, and reductions in the severity or frequency of occurrence of dominant reactor vessel PTS transients.     

Unbiasedness Evaluation of Statistical Error in Staged Monte Carlo Calculation

Fuel rod failure behavior has been studied under a reactivity initiated accident condition in Nuclear Safety Research Reactor (NSRR), JAERI. In the studies, inetallurgical observations showed that the incipient fuel rod failure mode was oxygen-induced embrittlement of the cladding independent of the test conditions such as fuel designs and cooling environments except for pressurized and waterlogged fuels. Development of the oxidation layers and embrittlement of β-Zry were quantitatively evaluated through the metallurgical examinations. A diffusion equation of oxygen was solved under a finite system with moving boundary conditions to obtain the oxygen concentration and evaluate the cladding embrittlement. The calculation showed that the wall thinning due to the cladding melt is needed for the complete embrittlement because the wall thinning enhances the oxygen concentration in the β-Zry, which well explain the experimental results. Therefore the failure threshold energy is determined by the cladding melting temperature. The failure threshold derived from this study is expected to be applicable to predicting the fuel rod failure behavior in computer analyses and also useful to evaluate the failure threshold energy for the new types of fuel rod.     

Round robin analysis of pressurized thermal shock for reactor pressure vessel

A comparative assessment study is performed here for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). A round robin problem is proposed using the data available in Korea and all organizations interested in the PTS analysis are invited. The problems consisting of two transients and 10 cracks are solved and their results are compared to generate a reference solution that could serve as benchmarks for future qualification of analytical method. Nine participants from seven organizations responded to the problem and their results are compiled in this paper.     

Dynamic model of Fast Breeder Test Reactor

Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.