Experimental determination of the boron concentration distribution in the primary circuit of a PWR after a postulated cold leg small break loss-of-coolant-accident with cold leg safety injection

It is known that under-borated coolant can accumulate in the loops and that it can be transported towards the reactor core during a loss-of-coolant-accident. Therefore, the mixing of weakly borated water inside the reactor pressure vessel was investigated using the ROCOM test facility. Wire-mesh sensors based on electrical conductivity measurement are used to measure in detail the spreading of a tracer solution in the facility. The mixing in the downcomer was observed with a measuring grid of 64 azimuthal and 32 vertical positions. The resulting distribution of the boron concentration at the core inlet was measured with a sensor integrated into the lower core support plate providing one measurement position at the entry into each fuel assembly.

The boundary conditions for this mixing experiment are taken from an experiment at the thermal hydraulic test facility PKL operated by AREVA Germany. The slugs, which have a lower density, accumulate in the upper part of the downcomer after entering the vessel. The ECC water injected into the reactor pressure vessel falls almost straight down through this weakly borated water layer and accelerates as it drops over the height of the downcomer. On the outer sides of the ECC streak, lower borated coolant admixes and flows together with the ECC water downwards. This has been found to be the only mechanism of transporting the lower borated water into the lower plenum. In the core inlet plane, a reduced boron concentration is detected only in the outer reaches of the core inlet. The minimum instantaneous boron concentration that was measured at a single fuel element inlet was found to be 66.3% of the initial 2500 ppm.     

Assessment methodology for confidence in safety margin for large break loss of coolant accident sequences

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis (BEPU) is increasingly being used for deterministic calculation in NPPs. The PSA methodology integrates information about the postulated accident, plant design, operating practices, component reliability and human behavior. The deterministic and probabilistic methodologies are combined by analyzing the accident sequences within design basis in the event trees of a postulated initiating event (PIE) by BEPU. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event.     

Quantifying reactor safety margins Part 6: A physically based method of estimating pwr large break loss of coolant accident pct

The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling. Applicability and Uncertainty (CSAU) Evaluation Methodology. This effort lead to increased understanding of the phenomena, and their relative dominance, during Large Break Loss of Coolant Accidents (LBLOCAs) in Pressurized Water Reactors (PWRs). Consequently, it became possible, as is done in this paper, to develop a method for establishing clad temperature history by using physically based arguments and engineering correlations. The results from this method are compared with similar uncertainty estimates based on large computer codes. These comparisons provide a rationale, based on physical arguments, for evaluating the large computer code based estimates of uncertainty.     

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.     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

Seismic analysis of a reactor coolant pump by the response spectrum method

Current nuclear steam supply systems (NSSS) are designed to remove the heat of fission by circulating coolant in closed loops from the reactor. For water reactors, this prime function is designated to the reactor coolant pump (RCP). The Westinghouse Type 93A RCP is analyzed for seismic response. Briefly described, this RCP is a vertical, single-stage, centrifugal pump designed to move 90 000 gpm (568 m³/sec) of water and driven by a 6000 hp motor for use in the PWR primary system. The RCP assembly is generally axisymmetric and is modeled using three-dimensional finite elements of the types normally found in general-purpose computer programs such as ANSYS or NASTRAN. The structural frame and the rotating shaft are the principal branches of the model. Each consists of a series of pipe elements complemented by mass elements. Orthogonal sets of linear spring elements connect the branches at the bearings and possibly at each labyrinth. Fluid elements are added to include the interaction between the shaft and the pump case through the intervening water mass. Beam elements are used to account for unsymmetry of the motor stand. To complete the model, stiffness matrix elements representing the support structure and the neighboring loop piping are attached. It is impractical to idealize faithfully each geometric irregularity. Several adjacent sections are combined into one suitable element with total stiffness and equivalence. The number of elements in the model is thus minimized. Shear deflection of the pipe elements is considered; mass and mass inertia are lumped at nodal points, as needed to compensate for the actual material distribution. The RCP model contains 82 nodes, 155 elements and 140 master dynamic degrees of freedom. A modal frequency analysis is first run to identify the mode shapes.The seismic analysis is performed by the response spectrum method in ANSYS, with seismic velocity as the input excitation parameter. The model is excited by a set of three orthogonal spectra. For each load excitation, the modal displacements, forces and moments are computed at each node. A post-run subroutine calculates the absolute sum of nodal response quantities at each mode for one horizontal and the vertical seismic excitations. The resultant modal values are then combined using the square root of the sum of the squares (RSS) to record the final values: SSE X-Y and SSE Y-Z. Nodal stresses are computed; absolute displacements are reviewed for selected nodes along the model branches. The relative displacements at bearings and labyrinths are determined. Finally, the accelerations of nodes previously chosen are found.This paper assesses the effects of a given seismic excitation on the overall structural integrity of an RCP. The in-depth analysis has found the RCP adequate to withstand the imposed seismic loading. All component stresses are within the applicable faulted criteria and the relative movements between closely mated parts fall inside their nominal clearance limits.     

冷中子源中氢系统的纵深防御设计

反应堆冷中子源装置的安全性包括核安全、辐射安全和氢安全,其中氢安全是最复杂的一方面.冷中子源氢系统一般采取纵深防御的安全设计方法,即通过多重屏障的技术措施尽可能杜绝氢和空气(氧气)直接接触的可能性.本文详细描述了一座建设中的反应堆冷中子源氢系统的纵深防御安全设计特点,这种安全设计可以确保冷中子源的氢安全性.     

Estimation of break location and size for loss of coolant accidents using neural networks

In this work, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to identify the break locations of loss of coolant accidents (LOCA) such as hot-leg, cold-leg and steam generator tubes. Also, a fuzzy neural network (FNN) is designed to estimate the break size. The inputs to PNN and FNN are time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. It is verified that the proposed algorithm identifies very well the break locations of LOCAs and also, estimate their break size accurately.     

Non-dimensional analysis of the transient flow rate and the coolant temperature in a reactor coolant system under different pump failures

The mathematical models are developed to solve the non-dimensional transient flow rates in two loops and a reactor core under different power failures of reactor coolant pumps. Comparison of the experimental results of the Qinshan Nuclear Power Plant and the test results of the nuclear ship reactor in Japan under one pump power failure shows an excellent agreement. The non-dimensional flow rates are determined by the established non-dimensional parameters λ, ?, and α. Under the sequential power failure of two pumps, the non-dimensional flow rates are determined by the established λ, ?, α, and ΔT parameters. λ, ?, α, and ΔT are four important non-dimensional parameters in the prediction of flow transients. λ indicates the resistance coefficient ratio of the single loop to the reactor core, ? indicates the fluid inertia ratio of the reactor core to the single loop, α indicates the ratio of the initial kinetic energy of the single loop coolant fluid to the effective initial kinetic energy of the reactor coolant pump, and ΔT means the non-dimensional time interval of the sequential power failure of two pumps. The effects of λ, ?, α, and ΔT on the non-dimensional flow rates and the temperature change are investigated.     

Passive modular gas safety system for a reactor

Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center, RNTs. Translated from Atomnaya Énergiya, Vol. 75, No. 1, pp. 8–13, July, 1993.     

反应堆冷却剂泵转速测量系统改进设计及实践

反应堆冷却剂泵（主泵）转速是核电站关键设备反应堆冷却剂泵运行状态监测的重要参数，直接反映设备的运行状况，并担负向反应堆保护系统输送反应堆的保护信号。但是该信号一直存在运行过程中测量不稳定的情况。从该测量通道的测量原理、历史状态，结合现场的实际检修过程，对转速测量的缺陷、可能的原因进行分析，同时对以上原因采取改进方式。经过2008年的运行验证，改进的测量方式信号稳定，满足了现场的要求，有利于改进的持续进行。     

AP1000核电厂反应堆冷却剂系统泄漏分析

第三代AP1000非能动核电厂的主要特征是采用非能动安全原理,使核电厂的系统、设备、构筑物大幅度简化,安全性、可靠性、经济性大幅度提高,以满足美国先进轻水堆业主要求文件的基本要求。本文针对美国业主要求文件(Utility Requirements Document,URD)第三卷第五章《专设安全系统》中对非能动先进轻水堆核电厂反应堆冷却剂系统压力控制功能的要求:在很小的反应堆冷却剂系统(Reactor Coolant System,RCS)净泄漏率(不大于2.27 m3·h-1)条件下,具有足够的系统冷却剂装量及补水能力,以保证在8 h(28 800 s)内不会触发自动降压系统而进行计算分析,本分析采用安全分析报告小破口失水事故(Loss of coolant accident,LOCA)分析采用的NOTRUMP程序,分析结果表明AP1000核电厂可满足上述美国URD要求。     

Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences

A coolant injection into the reactor vessel with depressurization of the reactor coolant system (RCS) has been evaluated as part of the evaluation for a strategy of the severe accident management guidance (SAMG). Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feedwater (LOFW) accident in Optimized Power Reactor (OPR)1000 have been analyzed by using the SCDAP/RELAP5 computer code. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 s with indirect RCS depressurization by using one condenser dump valve (CDV) at 6  min after implementation of the SAMG prevents reactor vessel failure for the small break LOCA without SI. In this case, only one train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent reactor vessel failure. Only one train operation of the HPSI at 20,208 s with direct RCS depressurization by using two SDS valves at 40 min after an initial opening of the safety relief valve (SRV) prevents reactor vessel failure for the total LOFW.     

核电站反应堆冷却剂系统主管道安装焊接

以秦山核电二期工程为例,论述了核电站反应堆冷却剂系统主管道安装焊接技术及质量控制要点,并对反应堆冷却剂系统主管道的安装顺序、安装技术要求、焊接质量检验方法以及焊接变形的控制等方面给予了详细的阐述,对核电站反应堆冷却剂系统主管道安装焊接及质量控制具有借鉴作用。     

Modelling the activity of I in the primary coolant of a CANDU reactor

A mathematical treatment has been developed to describe the activity levels of ¹²⁹I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10⁻⁹ has been determined for the coolant activity ratio of ¹²⁹I/¹³¹I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ∼2×10⁻⁸ for ¹²⁹I/¹³⁷Cs. This ratio can be used to estimate the levels of ¹²⁹I in reactor waste based on a measurement of the activity of ¹³⁷Cs.