Improved Monte Carlo-perturbation method for estimation of control rod worths in a research reactor

A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. Perturbation method is used to obtain the equation for the relative efficiency of control rod insertion. A series of coefficients, describing the axial absorption profile are used to correct the equation for a composite rod, having a complicated burn-up irradiation history. These coefficients have to be determined – by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross-sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn-up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct MCNPX evaluations of control rod worths is also presented.     

稳压器压力水位控制系统建模与仿真

通过对压水堆核电站稳压器实际运行特性的分析研究,在合理简化与假设的基础上分别对稳压器蒸汽区以及液体区建立质量和能量守恒方程,建立一个两区不平衡的稳压器模型。然后通过模块封装组建成稳压器水位和压力控制系统,最后通过仿真对稳压器主要参数进行动态特性分析,仿真结果符合理论分析,所建模型合理。     

Measuring and analysis methods applied to on-line vibration and noise monitoring in PWR power plants

Measurements and analyses performed in a PWR plant are discussed in the light of their relevance for application to an on-line monitoring system. This is achieved by extensive investigation during the preoperational tests. The results gained are transferred to the situation existing during full power operation by means of spectral analysis and correlation techniques. Theoretical models provide further proof of the results. The final conclusions show the consequences for establishing an on-line monitoring system.     

A new method to predict Grid-To-Rod Fretting in a PWR fuel assembly inlet region

Grid-To-Rod Fretting (GTRF) is one of the main causes of leaking fuel in a Pressurized Water Reactor (PWR). GTRF is caused by grid-to-rod gap, secondary flow, and axial/lateral turbulence caused pressure fluctuations within the fuel assembly, which produces rod vibration and wear. The cross flow and vortex shedding phenomenon produce low frequency vibration forces on fuel rods. In some plants, leaking fuel has been detected at the fuel inlet region of fuel assembly designs that do not have Protective Grid (P-grid) which, in addition to providing debris protection, also provides lateral stability against vibration. In order to understand the root cause of the fuel leaks, a thorough investigation of the flow field at the fuel inlet region is required. Leaking fuel has also been detected in the fuel inlet region in transition cores. In the transitional core arrangement, there are different fuel assembly designs next to each other. Due to the structure difference, there will be cross flow between fuel assemblies, which may be the initiating factor for fuel leaks.A method based on Computational Fluid Dynamics (CFD) has been developed in Westinghouse to predict the GTRF in the fuel inlet region. The fuel inlet region consists of the lower core plate, the bottom nozzle, the fuel rods, the thimble rods, the P-grid, and the bottom grid. This study employed CFD to investigate the unsteady forces on the fuel rods under typical reactor in-core conditions. Two fuel assembly (FA) inlet regions with and without the P-grid were simulated. The time history of the unsteady force components on fuel rods was recorded. Fast Fourier Transform (FFT) analyses were carried out for the force history. Compared to the data from operating plants, the new method predicted synchronized excitation forces on the rods that leaked in real operation. The CFD results also demonstrated the advantage of using the P-grid. GTRF at the fuel inlet region can be significantly reduced when the P-grid is used in Westinghouse fuel assembly designs.     

Thermal-hydraulic calculations using MARS code applied to low power and shutdown probabilistic safety assessment in a PWR

The methods developed for full-power probabilistic safety assessment, including thermal-hydraulic methods, have been widely applied to low power and shutdown conditions. Experience from current low power and shutdown probabilistic safety assessments, however, indicates that the thermal-hydraulic methods developed for full-power probabilistic safety assessments are not always reliable when applied to low power and shutdown conditions and consequently may yield misleading and inaccurate risk insights. To increase the usefulness of the low power and shutdown risk insights, the current methods and tools used for thermal-hydraulic calculations should be examined to ascertain whether they function effectively for low power and shutdown conditions. In this study, a platform for relatively detailed thermal-hydraulic calculations applied to low power and shutdown conditions in a pressurized water reactor was developed based on the best estimate thermal-hydraulic analysis code, MARS2.1. To confirm the applicability of the MARS platform to low power and shutdown conditions, many thermal-hydraulic analyses were performed for the selected topic, i.e. the loss of shutdown cooling events for various plant operating states at the Korean standard nuclear power plant. The platform developed in this study can deal effectively with low power and shutdown conditions, as well as assist the accident sequence analysis in low power and shutdown probabilistic safety assessments by providing fundamental data. Consequently, the resulting analyses may yield more realistic and accurate low power and shutdown risk insights.     

Optimization method to determine mass transfer variables in a PWR crud deposition risk assessment tool

Optimization numerical method was implemented to determine several mass transfer coefficients in a crud-induced power shift risk assessment code. The approach was to utilize a multilevel strategy that targets different model parameters that first changes the major order variables, mass transfer inputs, then calibrates the minor order variables, crud source terms, according to available plant data. In this manner, the mass transfer inputs are effectively simplified as “dependent” on the crud source terms. Two optimization studies were performed using DAKOTA, a design and analysis toolkit, with the difference between the runs, being the number of model runs using BOA, allowed for adjusting the crud source terms, therefore, reducing the uncertainty with calibration. The result of the first case showed that the current best estimated values for the mass transfer coefficients, which were derived from first principle analysis, can be considered an optimized set. When the run limit of BOA was increased for the second case, an improvement in the prediction was obtained with the results deviating slightly from the best estimated values.     

Unitary theory of xenon instability in nuclear thermal reactors—I. Reactor at ‘zero power’

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback.It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.     

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.     

III. PWR steam generator response to a scram at 50% load and an open grid transient (bugey-4 nuclear power plant)

Two transients, an open grid and a scram at 50% load, were conducted on unit 4 of the PWR power plant Bugey. The thermal hydraulic response of the steam generator was recorded. For the open grid test, the following observations are noted:No alarming phenomena are observed in the steam generator during the transient. Primary pressure oscillations were very mild, and did not exceed about 4.8 bar/min with a maximum amplitude of ±8 bar. This condition should not result in significant stress levels. Steam generator outer shell metal temperature gradients remained within very acceptable limits; a maximum amplitude of about +13°C and a rate not exceeding 0.8°C/min are obtained. This slow rate is explained by a fall in primary water temperature that allows for a temperature decrease inside the U-tube bundle. Similarly, the temperature rise on the tube sheet does not exceed an amplitude of 20°C with a rate of about 2°C/min. Again these conditions do not lead to any significant thermal shock on the tube sheet. The steam generator feed controls maintain the level within the normal operation range and the small addition of colder feedwater does not lead to great temperature changes because of the large mass of the recirculation water in the steam generator.For the scram at 50% load, the following observations are noted: no severe thermal or pressure transients are observed in this test. Fluid temperature fluctuations occur with rates not exceeding 1°C/s and a maximum amplitude of about 20°C in the downcomer and 10°C on the tube sheet. Steam generator outer shell temperature varies at a rate of about ±0.8°C/min with a maximum amplitude of about 16°C. These thermal transients should lead to thermally induced stresses of acceptable levels.     

Monitoring temperature reactivity coefficient by noise method in a NPP at full power

A new method based on noise measurement was used to estimate the temperature reactivity coefficient of a PWR Unit during the entire fuel cycle. This made it possible to measure its dependence on boron concentration. The method gives good results and can thus be applied for permanent monitoring of this safety parameter.     

Development of a bundle correction method and its application to predicting CHF in rod bundles

A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first derivatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of CHF data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions.     

Development and assessment of fire-related risk unavailability matrices to support the application of the maintenance rule in a PWR nuclear power plant

Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The development of these methods is restricted to the compulsory use of fire probabilistic safety assessment (PSA) models. The first method is a fire protection systems and key safety functions unavailability matrix which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. Specific selection and quantification methodologies have been developed to obtain the matrices. The Monte Carlo method has been used to assess the uncertainty of the unavailability matrix. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building.     

Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the ¹³¹I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent ¹³⁵I, ¹³⁵Xe and ⁸⁹Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.     

Probe method to measure the electromagnetic field in a plasma by Poynting vector conversion

In this paper,a high-precision electromagnetic measurement system suitable for a high-temperature and high-speed plasma is built to provide a platform for scientific research on the interaction mechanism of the electromagnetic fields and a plasma.This paper presents a method to measure the electromagnetic field inside a plasma by using a probe and Poynting vector conversion,which is a new and completely different method from the traditional method of measuring the electromagnetic field inside plasma.In addition,for this system and method,this work designs a microstrip antenna probe that can suppress multipath effects.This method is confirmed to be valid and usable after closed-loop verification by the CST software.     

New proposal of reactivity coefficient estimation method using a gray-box model in nuclear power plants

A new method for estimating reactivity parameters, such as moderator temperature coefficient (MTC) and void reactivity coefficient (VRC), is proposed using steady-state noise data. In order to solve the ill-posed problem of reactivity parameter estimation, a concept of a gray box model is newly introduced. The gray box model includes a first principle based model and a black-box fitting model. The former model acts as a priori knowledge based constraints in a parameter estimation problem. After establishing the gray box and noise source models, the maximum likelihood estimation method based on Kalman filter is applied. Furthermore, it is shown that the frequency domain approach of the gray box model is useful in the case of VRC estimation. The effectiveness of the proposed algorithms is shown through numerical simulation and actual plant data analysis.     

Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

Neutron spectrum should be measured before test samples are irradiated.Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW.Sixteen kinds of non-fission foils(19 reaction channels) were selected,of which 10 were sensitive to thermal and intermediate energy regions,while the others were of different threshold energy and sensitive to fast energy regions.By measuring the foil radioactivity,the neutron spectrum was unfolded with the iterative methods SAND-Ⅱ and MSIT.Finally,shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-Ⅱ.     

Characteristics of test and maintenance human errors leading to unplanned reactor trips in nuclear power plants

The test and maintenance (T&M) human errors involved in unplanned reactor trip events in Korean nuclear power plants were analyzed according to James Reason's basic error types, and the characteristics of the T&M human errors by error type were delineated by the distinctive nature of major contributing factors, error modes, and the predictivity of possible errors. Human errors due to a planning failure where a work procedure is provided are dominated by the activities during low-power states or startup operations, and human errors due to a planning failure where a work procedure does not exist are dominated by corrective maintenance activities during full-power states. Human errors during execution of a planned work sequence show conspicuous error patterns; four error modes such as ‘wrong object’, ‘omission’, ‘too little’, and ‘wrong action’ appeared to be dominant. In view of a human error predictivity, human errors due to a planning failure is deemed to be very difficult to identify in advance, while human errors during execution are sufficiently predictable by using human error prediction or human reliability analysis methods with adequate resources.     

Variability in dynamic characteristics and seismic response due to the mathematical modeling of nuclear power plant structures

Calculated variations in dynamic characteristics and seismic response of nuclear power plant structures caused by different modeling assumptions are quantified. Four different mathematical models were created to describe the same structure, the Zion nuclear power station auxiliary/fuel-handling/diesel-generator complex. The modeling idealizations are a detailed finite element model, a detailed finite element model with masses lumped at selected nodes, a detailed finite element model with the constraint of rigid floors, and an equivalent beam model. Dynamic characteristics and response quantities are determined for the models and compared. Results indicate that large variations in dynamic characteristics and response can be introduced by modeling assumptions when a need exists to reduce the number of dynamic degrees of freedom.