Pressurized Water Reactor Units Research Articles

Discovering the causes for the change of the vibration characteristics of the core support barrel in PWR nuclear power plants: A combined investigation based on ex-core neutron noise analysis and numerical modal analysis

As a key component of pressurized water reactor (PWR) nuclear power plants, the motion of the core support barrel (CSB) has a great impact on the safety of the nuclear reactor. However, the vibration characteristics of the CSB have not yet been fully investigated and may change due to the high doses of radiation and high temperature in the reactor. In this paper, the ex-core neutron noise signals of two PWR units throughout the multiple fuel cycles were analyzed and the vibration frequency of the CSB beam mode was found, for the first time, to drop with time during the operation of PWRs. Then, the statics analysis of the CSB was conducted to preliminarily identify the potential causes. For the further clarification of such vibration frequency drop phenomenon, a numerical assembly model of PWR reactor internals, including the CSB, hold-down ring (HDR), fuel assembly hold-down spring (FAHDS), etc., was proposed with consideration of the frictional contact and fluid-structure interaction for the modal analysis of the CSB. The results showed that the decrease of the FAHDS stiffness is the main cause for the drop of the vibration frequency of the CSB beam mode in real conditions.

Online estimation approach of the steam specific enthalpy for wet steam turbines in nuclear power plants

Wet steam turbines are widely used in pressurized water reactor nuclear power plants. The specific enthalpy of the wet saturated steam is an indispensable variable for the thermodynamic balance and performance monitoring of the wet steam turbines, but it is rarely online measured in practice. In this study, a model-based online estimation approach of the steam specific enthalpy for the wet steam turbine in nuclear power plants is proposed. The wet steam turbine is divided into several stage groups. In each stage group, the steam expansion process, the moisture separation process, and the steam extraction process are captured in mechanism modeling at first. Then, by solving the model, the specific enthalpy and the dryness fraction of the wet saturated steam are obtained. Using the historical operation data of a 1000 MW pressurized water reactor unit, the model-based pseudo-online simulations and the measured data are compared which suggests a high accuracy of the proposed approach.

Estimation of public exposure during normal operation of unit-1 Barakah Nuclear Power Plant using GALE and HOTSPOT

In August 2020, the first Barakah Nuclear Power Plant (BNPP) unit was connected to the grid and started supplying electricity. The BNPP site consists of 4 pressurized water reactor units, and each unit generates electricity up to 1400 MWe power. Concerning safety, the radiological assessment of the radioactive release and the corresponding public exposure during normal operation is essential. In the present work, the radioactive release during normal operation, in gaseous form, is calculated considering the operating condition of the BNPP by using the GALE code for gaseous forms. Then, the HOTSPOT code is used to simulate the radiological dispersion, including the public exposure and Total Effective Dose Equivalent (TEDE). HOTSPOT code uses the Gaussian dispersion model to provide near-surface releases, short-range dispersion, and short-term releases. Finally, the results in the present work are compared with the values given in the Final Safety Analysis Report of the BNPP. For typical operating circumstances, this study evaluated gaseous source terms such as iodine, noble gases, radioactive particles, tritium, C-14, and Ar-41 and found that TEDE values are within the authorized limits for distances up to 80 kms from the power plant.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. Electric Power Research Institute (EPRI) has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicates that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from other thermal transients, especially those resulted from outflow activities, may also contribute to the failure of reactor coolant system (RCS) branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME section III class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients, particularly the transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

The preparation of oxidative resin loaded with Fe3+ and their removal performance for 110 mAg in colloidal form

110 mAg, comes from pressurized water reactor units, has become one of the major radionuclides affecting the surrounding environment. The chemical species of 110 mAg are controlled by additives or impurities during operation process of nuclear power plant. Colloidal state (named Ag0) and ionic state are main chemical forms. It is difficult to remove Ag0 by commercial ion exchange resin because the colloidal particles have different sizes and low charge. The oxidative resin loaded with Fe3+ were prepared and used for removal of colloidal Ag from aqueous solution. The influence factors such as resin dosage, adsorption time and initial concentration of Ag0 were investigated through batch experiments. The adsorption efficiency of the oxidative resin towards Ag0 in aqueous solution could reach 92.5% under pH 7. The adsorption process corresponds to the Langmuir adsorption model and pseudo-second-order kinetic equation. XPS analysis clearly shows the presence of binding energy peaks of Fe2+ and Ag+, which indicates the adsorption mechanism towards Ag0 is based on the combined actions of redox and ion exchange.

Feed-Water Enthalpy Rise Optimum Distribution Mechanism Model of PWR Nuclear Power Unit Based on Genetic Algorithm

This paper presented a novel method for using Genetic Algorithm (GA) to solve the problem of Feed-water Enthalpy Rise Optimum Distribution in nuclear power unit. According to the characteristics of the pressurized-water reactor unit's secondary circuit, the calculation equations for the power and the heat rate of steam turbine are given in matrix form, the rate of extraction steam is calculated by the general matrix thermal balance equation, and then receive the equation of the efficiency of steam turbine which has relative with parameters of every node in the thermodynamic system in PWR unit. Taking the efficiency equation as aim function, an effective fitness-model was established with the appropriate optimum parameters under the restriction condition, the feed-water enthalpy rise distribution is optimized under the restriction condition by means of Genetic Algorithm. Through the calculation of a 600MW PWR nuclear power unit, the results indicate that Genetic Algorithm has global searching performance, gives a better global optimum than original design value rapidly and shows the relations between the optimum parameters (even the temp parameters) and the optimization easily. Genetic Algorithm is a promising method for analyzed and solved the feed-water enthalpy rise optimum distribution in PWR nuclear power unit.

Developing a Tool for Simulating the Behavior of Axial Flux Difference during Power Reduction for Maanshan Pressurized Water Reactor Units

A tool, AFDCONTROL, is developed to help the operators of pressurized water reactors predict axial flux difference (AFD) and find the control rod positions for power maneuvers. AFDCONTROL has used an off-line procedure to simulate AFD.After adopting a long-cycle strategy as the fuel management policy of Maanshan nuclear power plant (MNPP), AFD control during the power reduction near end of cycle (EOC) becomes more difficult than before. In this study, we have used AFDCONTROL to simulate this trend. AFDCONTROL suggests that AFD (even at higher burnup conditions) can be controlled by extending the power reduction rate. After implantation of the AFDCONTROL tool in MNPP, the operators of MNPP units have used AFDCONTROL to find the control rod sequences and positions under certain power reduction rates to control AFD without violating limits, especially at EOC of the high-burnup cycle.

Evaluation of the Boron Dilution Method for Moderator Temperature Coefficient Measurements

A measurement of the at-power moderator temperature coefficient (MTC) at the pressurized water reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed. The measurement was performed when the boron concentration decreased under 300 ppm in the reactor coolant system, by using the boron dilution method. Detailed calculations were made to estimate all reactivity effects taking place during such a measurement. These effects can only be accounted for through static core calculations that allow calculating contributions to the reactivity change induced by the moderator temperature change. All the calculations were performed with the Studsvik Scandpower SIMULATE-3 code. Analysis of the measurement showed that the contribution of the Doppler effect (in the fuel) was almost negligible, whereas the reactivity effects due to other than the Doppler fuel coefficient and the boron change were surprisingly significant. It was concluded that due to the experimental inaccuracies, the uncertainty associated with the boron dilution method could be much larger than previously expected. The MTC might then be close to –72 pcm/°C, whereas the main goal of the measurement is to verify that the MTC is larger (less negative) than this threshold. The usefulness of the boron dilution method for MTC measurements can therefore be questioned.

Methods and Performance of the Three-Dimensional Pressurized Water Reactor Core Dynamics SIMTRAN On-Line Code

New reactor physics and computation methods have been developed in our three-dimensional pressurized water reactor (PWR) core dynamics SIMTRAN code for on-line surveillance and prediction. The accuracy of the coupled neutronic thermal-hydraulic solution is improved, and its scope is extended to provide, mainly, the calculation of the fission reaction rates at the in-core minidetectors, the responses at the ex-core detectors, and the in-vessel coolant flow and temperature distributions.The functional capabilities implemented in the on-line SIMTRAN code include on-line surveillance, in-core-ex-core calibration, evaluation of peak power factors and thermal margins, nominal cycle follow, prediction of maneuvers, and diagnosis offast transients and oscillations. The new code has been operating on-line at the Vandellós-II PWR unit in Spain since the startup of its cycle 7 in mid-June 1994, including the machine-man interfaces for on-line acquisition of measured data and interactive graphical utilization. The agreement of the simulations with the measurements, along the full cycle 7 and the first months of cycle 8 operation, is well within the accuracy requirements. The performance and usefulness for operational support shown during the demo and routine use phases have proved that the on-line SIMTRAN code has the qualities for the accurate, reliable, comprehensive, and user-friendly on-line core surveillance and prediction.

Pressure vessel steel surveillance program: Developments, experience feed-back and perspectives

Aims of such surveillance programs and the reactor operating scenarios which were used to establish them are recalled. The main characteristics of surveillance programs drawn up at the design of each series of French pressurized water reactor units are then presented. Developments in the cumulative dose rate at the design end of life have contributed to a reduction in the number of irradiation capsules. The withdrawal schedule not only takes into consideration the anticipation due to the irradiation location, but also takes into account real operating conditions. A large number of specimens and the presence of a dosimetric and thermal monitoring in each capsule play a great part in the quality of the results. Extensive work aimed at a more accurate definition of reference neutronic data were undertaken to improve the reliability of the results. Experience feed-back from the first capsules used confirms the interest of this work and the weight of parameters such as geometries, spectrum, fissile dosimeter, neutron filter, etc. Developments in fuel design and operation call into question some manufacturers' hypotheses of primary importance. This necessitates the reassessment of the damage to the vessel steel and the operator's becoming aware of the relevant risks and modification of the design of the next programs of reactor surveillance.

Seismic Isolation for Advanced Nuclear Power Stations

Seismic isolation offers an attractive approach for reducing seismic loads in nuclear structures, and more significantly, in reactor components. Isolation will lead to a simplification of designs, facilitate standardization, enhance safety margins, and may potentially reduce cost. To date, six large Pressurized Water Reactor units have been isolated in France and South Africa and several advanced nuclear concepts in the U.S., Japan, and Europe have incorporated this approach. It is recognized that to qualify and license an isolation system in the U.S. and in Japan, a comprehensive testing program of isolation components and systems would be required. A major seven year program was initiated in Japan in 1987 with the objective of establishing a qualified seismic isolation design for a large fast breeder reactor to be constructed at the end of this decade. In the U.S., two concepts which use steel laminated elastomeric bearings for seismic isolation have been developed. One of these concepts is a novel system which provides three-dimensional isolation. An extensive test program of scaled prototype bearings to demonstrate their feasibility and effectiveness has been carried out.

Aerosol Behavior and Light Water Reactor Source Terms

The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered.

Life duration of PWR nuclear power plants

The importance of the Pressurized Water Reactor (PWR) nuclear plants in the whole of the electrical production facilities of EDF is a good reason why a special effort is devoted to assessing their life duration. It seems that the longevity of the PWRs is going to show more favourable technical and economical conditions than did the previous steam generating plants. The EDF ‘life duration project’ includes a four-year task program. First, a full inventory of the available data will be drawn up in order to forecast the ageing of the sensitive materials of a power plant and to follow up its development so as to anticipate or to remedy problems. These sensitive materials are those for which the difficulties and the cost of restoration would not allow the inclusion of those operations within the normal framework of maintenance programs. The first analysis of how much these materials may be damaged leads one to believe that the main results of such ageing, which occur mainly in reactor vessels as embrittlement due to exposure and also in the fatigue of the main components of the boilers, will not limit the life of these sensitive materials, except in the very long term. Overcoming the effects of certain ageing or materials corrosion damage may well give difficult problems to maintenance so that it is vital to forecast and assess them. If this necessary effort is undertaken, the lifetime of the PWR units should be very long.

Chemistry Experience in the Primary Heat Transport Circuits of Kraftwerk Union Pressurized Water Reactors

Chosen for this description of the selected Kraftwerk Union (KWU) pressurized water reactor units were Obrigheim (KWO, 345 MW(e)), Stade (KKS, 662 (MW(e)), Borselle (KCB, 477 MW(e)), and Biblis (KWB-A, 1204 MW(e)). The experience at these plants shows that with a special startup procedure and a proper chemical control of the primary heat transport system that influences general corrosion, selective types of corrosion, corrosion product activity transport and resulting contamination, and radiation-induced decomposition, KWU units have no basic problems.