Hot Functional Tests Research Articles

Effects of Zn injection into boric acid and lithium hydroxide solutions on the corrosion behaviors of 316LN stainless steel in simulated hot functional test high-temperature pressurized water

The effects of Zn injection into boric acid and lithium hydroxide solutions on the corrosion behaviors of 316LN stainless steel (SS) in simulated hot functional test (HFT) high-temperature pressurized water at 573 K were investigated. Low concentration of lithium hydroxide solutions with 20–100 ppb Zn injection are recommended for Zn-injected HFT water chemistry and the beneficial effects of Zn injection into both HFT and primary coolant water chemistry are confirmed in view of the protective properties of the oxide films on 316LN SS. The Zn injection involved corrosion mechanism during HFT is discussed.

Effects of Zn Injection into Boric Acid and Lithium Hydroxide on the Corrosion Behaviors of 316ln Stainless Steel in Simulated Hot Functional Test High-Temperature Pressurized Water

Effects of Zn Injection into Boric Acid and Lithium Hydroxide on the Corrosion Behaviors of 316ln Stainless Steel in Simulated Hot Functional Test High-Temperature Pressurized Water

The Evaluation of the Primary Circuit Passivation Layer Formed After Hot Functional Tests at the EMO3

Abstract The presented article outlines and describes the processes involved in the creation and testing of the passivation layer that was created during hot functional tests (HFTs) at the third unit of the nuclear power plant (NPP) Mochovce. The article is divided into several sections. In Sec. 1 is the description of the hot tests on the third block, in particular the chemical regime, i.e., the formation of hydrogen or the necessary adjustment of the pH value. Sec. 2 of the article describes the evaluation of the passivation layer using instrumental techniques, i.e., Raman microscopy, scanning electron microscopy (SEM), and X-ray photoelectron spectroscopy. The results obtained by using these techniques provided a comprehensive view of the resulting layer and clearly defined its' structure. The measurement revealed that a passivation layer was formed, the inner layer of which is enriched with a chrome element. The upper part of the passivation layer was composed of iron oxides, mainly magnetite.

Measurement of vibration and stress for APR-1400 reactor internals

Abstract The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

Effect of water chemistry on the oxide film on Alloy 690 during simulated hot functional testing of a pressurised water reactor

Electrochemical impedance spectroscopic measurements to follow the oxidation of Alloy 690 in high-temperature water environments simulating hot functional test (HFT) chemistries are presented and discussed. The thickness and in-depth composition of the formed oxides is estimated by X-ray photoelectron spectroscopy. Kinetic and transport parameters of the oxidation process are estimated by quantitative comparison of the results with the mixed-conduction model for oxide films. Based on the influence of LiOH and H3BO3 concentrations on the parameter values, conclusions for the relationship between HFT water chemistry and the electrical and electrochemical properties of the passive layer on reactor coolant circuit surfaces are drawn.

Investigation and application of the de-soot technology for stoke boilers : Sun, D. et al. Ranshao Kexue Yu Jishu, 2002, 8, (1), 59–62. (In Chinese)

Large numbers of industrial boilers with lower thermal efficiency are widely distributed in the whole country and polluting the environment.A new combustion technology Coal Smokeless Combustion Technology for vertical and horizontal hand fired boiler are proposed to change the existing situation.A 0.1 t/h hot functional test device was built and the coal smokeless combustion experiments were conducted combining with thin coal streak gasification and stage air input and combustion.The results showed that good smoke self prevent and dust self control effects in furnace could be achieved.A 1 t/h capacity horizontal coal gasification smokeless combustion hand fired boiler was designed and the flue gas blackness and dust discharge could accord with the Ⅰ class region given in GWPB3-1999 environmental protection standard under directing burning raw coal.

Corrosion product measurements at Sizewell B PWR

Corrosion product concentrations and surface activities have been measured during hot functional tests and cycles I to 4 at Sizewell B, in a programme aimed at understanding corrosion product behaviour and radiation field build-up. The results are discussed in terms of primary coolant chemistry, sample line effects, passivation, radiation fields and, for shutdown and start-up, the origins of the corrosion products released into the primary coolant.

Structural evaluation of thermal stratification for PWR surge line

Recent observations at operating plants and subsequent US NRC requirements have identified flow stratification in surge lines as a phenomenon that must be considered in the design basis of surge lines. To address these concerns, the stratified loading conditions were included in the design of YGN 3 and 4 surge line as a design basis transient and pipe temperature and displacement measurement were taken during YGN 3 pre-core hot functional testing to determine the degree of surge line flow stratification. The measured displacements and temperatures were extensively reviewed and evaluated in detail: (1) to verify the validity of the thermal hydraulic model used to predict the pipe top-to-bottom temperature differentials; (2) to analytically correlate measured surge line temperatures and displacements; and (3) to confirm the validity of the stratified flow analysis procedure. This paper shows that the stratified flow phenomenon is generic and therefore generic loadings can be developed and evaluated for the surge line analyses.

Acoustic emission system calibration at Watts Bar Unit 1 nuclear reactor

An acoustic emission system has been installed on TVA's Watts Bar Unit 1 reactor to monitor selected areas of the pressure boundary during hydrostatic testing, hot functional testing, and ultimately, during reactor startup and operation. This is a part of a cooperative effort between TVA and NRC Research to test and demonstrate AE technology. This technology has been developed under an NRC Research program to validate the application of AE techniques for continuous, on-line monitoring of reactor pressure boundaries to detect cracking. This report discusses the performance of and results from a special calibration test of the AE system using simulated AE signals to evaluate the accuracy of signal source location by the system.

Vibration monitoring of the primary piping system during the hot functional tests of the Mülheim-Kärlich PWR

During the hot functional tests of the Mülheim-Kärlich plant, which was the first plant of its type, vibration measurements were made on the reactor pressure vessel and its internal parts and on the primary piping system and the main coolant pumps. This paper contains the results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new structural design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. A good correspondence was found. In the course of these comparisons, information about the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained. The vibration of the main coolant pumps was also continuously monitored. The pump surveillance system for each pump includes two non-contacting displacement sensors for measuring the kinetic shaft orbit, as well as velocity sensors for recording the vibrational velocity of the pump motor housing. During the continuous monitoring, it was checked whether the signal amplitudes remained within the allowable limits. In addition the frequency content of the signals was determined periodically. In this way deviations could be detected immediately and be explained by means of subsequent correlation analysis. Thus amplitude changes resulting from resonance effects were identified.

Evaluation of methods for leak detection in reactor primary systems

No currently available, single leak-detection method combines optimal leakage detection sensitivity, leak-locating ability, and leakage measurement accuracy. Technology is available to improve leak detection capability at specific sites by use of acoustic monitoring. However, current acoustic monitoring techniques provide no source discrimination (e.g., to distinguish between leaks from pipe cracks and valves) and no leak-rate information (a small leak may saturate the system). Seven cracks, including three field-induced IGSCC specimens and two thermal-fatique cracks, have been installed in a laboratory acoustic leak detection facility. The IGSCC specimens produce stronger acoustic signals than the thermal-fatigue cracks at equivalent leak rates. Despite significant differences in crack geometry, the acoustic signals from the three IGSCC specimens, tested at the same leak rate, are virtually identical in the frequency range from 300 to 400 kHz. Thus, the quantitative correlations between the acoustic signals and leak rate in the 300–400 kHz band are very similar for the IGSCC specimens. Also, acoustic background data have been acquired during a hot functional test at the Watts Bar PWR. With these data, it is now possible to estimate the sensitivity of acoustic leak detection techniques.

Acoustic emission for on-line reactor monitoring: Results of intermediate vessel test monitoring and reactor hot functional testing

This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using that AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility. A vessel, designated ZB-1, has been tested under fatigue loading with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialprüfungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE. AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions.

Neutron noise measurements on Bugey 2 PWR

Following Bugey 2 PWR hot functional tests, dimension measurements of internals hold down spring led to suspect that vibration levels could change with time. Neutron noise measurement runs during the first fuel cycle enabled describing vibration behaviour of internals. Comparisons with previous analytical and experimental results gained on the Safran model as well as on similar reactors were also made.

The Fort St. Vrain high temperature gas-cooled reactor: II. Helium circulators

In field tests in a fossil-fueled facility, performed concurrently with Fort St. Vrain's construction, data indicated that the helium circulator design was well suited to provide primary coolant circulation for the high temperature gas-cooled reactor. After plant installation, primarily during the hot functional tests, a number of time-consuming delays developed caused by cavitation damage on circulator speed valves, cavitation and fatigue damage on auxiliary water turbine buckets, water turbine nozzle erosion, static shutdown seal cracks and circulator primary closure helium leakage. After extensive analysis and testing, all of these problems were corrected. Circulators have performed satisfactorily at levels up to 70% of rated power.