Pressurized Water Reactor Coolant Research Articles

Towards modelling defective fuel rods in TRANSURANUS: Benchmark and assessment of gaseous and volatile radioactive fission product release

This work presents the results of a collaborative benchmark activity between different organizations towards the use of the TRANSURANUS code to estimate the release of gaseous and volatile radioactive fission products from defective fuel rods, into the primary coolant of pressurized water reactors. First, the radioactive release from the fuel to the gap is evaluated according to three approaches: the coupling between TRANSURANUS and SCIANTIX, the development of TRANSURANUS devoted subroutines, and the use of the ANS 5.4–2010 methodology. Fuel-to-gap release calculations are benchmarked and assessed against measured data from the CONTACT1 irradiation experiment. Then, TRANSURANUS has been used to estimate the radioactive release into the primary coolant by applying a first-order phenomenological rate theory and tested against measured data of fission product coolant concentrations from irradiation experiments of the CRUSIFON program.

Numerical analysis on subcooled boiling in PWR coolant channel based on a modified multi-scale interface model

A modified multi-scale interface model was employed with the Eulerian two-fluid framework to investigate the three-dimensional characteristics of subcooled boiling characteristics in a typical Pressurized Water Reactor (PWR) coolant channel with mixing grid. Based on the morphology identification by adopting an advanced interface topology map, the phase-average method and the interface-reconstruction method can be applied for dispersed bubble and sharp interface, respectively, to improve the calculated precision of interfacial transfers. Models were validated by using the dam break data for sharp interface topology and employing the flow boiling experimental data for the phase distribution in both radial and axial directions. The calculated results agreed well with the experimental data. Thermohydraulic characteristics in the coolant channel were obtained, including void fraction, phase velocity and flow streamline, based on which, the effects of mixing vane on the localized field were analyzed. The vapor volume fraction was decreased in downstream of spacer grid due to the mixing effect, while the lateral flow was enhanced. The vapor accumulation phenomenon was captured and the sharp interface between the liquid and the accumulated vapor was identified by the modified model. Based on this, the interfacial transfer characteristics of the sharp interface was considered in simulations.

Nanofluid applications in pressurized water reactors: A review

The main purpose of this paper was to review nanofluid applications as the main coolant of pressurized water reactors (PWRs) with both solid and dual-cooled annular fuel. In this regard, first, theoretical and experimental models for calculating thermo-physical properties of nanofluids including density, specific heat, thermal conductivity, viscosity, and heat transfer coefficient were reviewed and the effects of different variables such as particle volume concentration, particle size, shape, and temperature on these parameters were determined. In the following, research related to neutronic, thermal-hydraulic, and safety investigations of nanofluids as the main coolant of PWRs with solid and dual-cooled annular fuel were reviewed. Corresponding methods, codes, and results were discussed and compared. Based on studies, it was concluded that nanofluids' effects on key neutronic parameters (keff, neutron flux and radial and axial power peaking factor) depend on the used nanoparticle, volume concentration, and nanoparticle deposition on the cladding, while nanofluids can increase heat transfer, critical heat flux, and minimum departure from nuclear boiling ratio (MDNBR). Also, it was found that nanofluids with low volume concentration in reactor core with dual-cooled annular fuel can replicate the same advantages. Furthermore, using dual-cooled annular fuel instead of solid fuel increases MDNBR margin and reduces coolant flow which leads to a more compact design of the reactor core and a reduction of capital cost. Finally, incoming challenges and future works were presented.

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.

Magnetite (Fe3O4) and nickel ferrite (NiFe2O4) zeta potential measurements at high temperature: Part I—Design, materials and preliminary characterization of an apparatus implementing the streaming potential method

An apparatus for measuring the streaming potential of particles at high temperature has been designed and tested in a recirculating water loop operating at temperatures up to 300 °C. The ultimate objective is to derive values of the zeta potential of corrosion-product particles under the conditions of the primary coolant of pressurized water reactors (PWRs), quantities that until now have been unavailable. They are required for modeling the transport and radioactivation of particulate oxides around the coolant system and the consequential growth of radiation fields around components. The experimental technique employed, measuring the potential difference across a packed bed of particles through which the coolant flows, involved fabricating a tube of temperature-resistant polymer that insulated the particles from the containing stainless steel pressure boundary of the loop. The particle bed was contained between porous ceramic membranes and the whole sealed with shrunk-on PTFE to avoid by-passing flows. Platinum wires, inserted into the loop through Conax fittings and insulated from contact with it by coils of oxidized zirconium wire, served as electrodes upstream and downstream of the bed to measure the potential difference The detailed design and testing of the apparatus containing magnetite powder to verify its correct functioning under operating conditions are described. Since measuring streaming potential involved obtaining the voltage difference across the particle bed alone at the same time as the pressure drop through it from the flowing coolant was measured, the preparatory tests first established baseline values, from measurements on an empty test section, which were subtracted from the measurements on the full test section. Also, immediately before each streaming potential determination, a rapid measurement provided the overall electrical resistance of the test section. The initial measurements of the streaming potential of magnetite, between 20 and 240 °C in a solution containing 1800 ppm of boron and 3.5 ppm of lithium, led to corresponding values of zeta potential. The values show that the apparatus provided reproducible results. The results of the complete series of tests under conditions of the primary system of PWRs, involving nickel ferrite as well as magnetite particles, are the subject of a separate paper.

Effect of zinc injection on the electrochemical behavior and crack growth rate of a 316 L stainless steel in simulated primary coolant of pressurized water reactors

Zinc was added to high-temperature borated and lithiated water and its effect on the oxide film formation, electrochemical behavior and crack growth rate of 316 L stainless steel was investigated. The actual concentrations of zinc in the coolant solution at 300 °C and 350 °C were determined by ICP-Mass spectrometry. Zinc improved the electrochemical corrosion parameters of steel and reduced the crack growth rate of the sample by more than 13 times. Zinc also reduced the thickness of the formed oxide layers on steel and modified the chemical composition of the oxides to a mainly zinc and iron rich spinel.

Evidence of hydrogen trapping at second phase particles in zirconium alloys

Zirconium alloys are used in safety–critical roles in the nuclear industry and their degradation due to ingress of hydrogen in service is a concern. In this work experimental evidence, supported by density functional theory modelling, shows that the α-Zr matrix surrounding second phase particles acts as a trapping site for hydrogen, which has not been previously reported in zirconium. This is unaccounted for in current models of hydrogen behaviour in Zr alloys and as such could impact development of these models. Zircaloy-2 and Zircaloy-4 samples were corroded at 350 °C in simulated pressurised water reactor coolant before being isotopically spiked with 2H2O in a second autoclave step. The distribution of 2H, Fe and Cr was characterised using nanoscale secondary ion mass spectrometry (NanoSIMS) and high-resolution energy dispersive X-ray spectroscopy. 2H− was found to be concentrated around second phase particles in the α-Zr lattice with peak hydrogen isotope ratios of 2H/1H = 0.018–0.082. DFT modelling confirms that the hydrogen thermodynamically favours sitting in the surrounding zirconium matrix rather than within the second phase particles. Knowledge of this trapping mechanism will inform the development of current understanding of zirconium alloy degradation through-life.

Research On Optimization Method of Fatigue Analysis for Nuclear Pipeline

The initial design life of nuclear power plant is 40 years. In 60 year life extending license application, the fatigue of component should be evaluated under the influence of the fatigue factors of the pressurized water reactor coolant environment. Because the original design used a more conservative analysis method, the result could not meet the requirement of Cumulative usage fatigue factor of RCC-M. An optimizing analysis method is studied, and as an example of application, optimizing fatigue analysis of Safety Injection Nozzle of Main Coolant Line is performed. The evaluation results show that the optimized fatigue analysis results meet the requirements of RCC-M.

Behavior of Zinc Incorporation into Fuel Deposits in Borated and Lithiated Water with Dissolved Zinc.

The objective of this study was to investigate the behavior of zinc incorporation into newly forming fuel deposits and pre-formed deposits in a simulated pressurized water reactor coolant including 1000 ppm of boron and 2 ppm of lithium at 328 °C. Zinc was incorporated into fuel deposits that were being newly nucleated and grown on nuclear fuel cladding tubes in a zinc-containing coolant. The zinc incorporation resulted in a decrease in the lattice constant of the deposits, which was attributed to the decrease in larger iron content and the corresponding incorporation of smaller zinc in the deposits. However, zinc incorporation was not found, even after the fuel deposits pre-formed before zinc addition were subsequently exposed to the 60 ppb of zinc coolant for 500 h.

Development of a Lumped Parameter Dynamic Degassing Model for Spray-Heating Degasser and Its Application in the Pressurizer of a Pressurized Water Reactor

The pressurizer of a pressurized water reactor (PWR), as a spray-heating degasser, has been widely used to remove dissolved gas in the primary coolant of PWRs. In the real degassing process, the boundary conditions of the pressurizer may change, causing fluctuations in the degassing state and affecting the efficiency of degassing. However, open-published studies have focused mainly on the steady-state degassing characteristics of the pressurizer. This paper studies the dynamic characteristics of a spray-heating degasser as applied to the pressurizer of a PWR. First, a lumped parameter dynamic degassing model for the spray-heating degasser is proposed based on basic gas dissolution and transport theory. Second, this model is extended, and a dynamic degassing model for the pressurizer is obtained. Third, two sets of numerical hydrogen degassing tests are carried out using the pressurizer dynamic degassing model. These two sets of numerical tests take the Shippingport pressurizer as the research object and integrate the structure and operating parameters of the Shippingport pressurizer with the system parameters of a Bettis Atomic Power Laboratory hydrogen degassing test as the numerical test condition. The spray-heating degasser degassing model is universal and applicable to this pressurizer as well as other devices with similar structures. The first set of numerical tests carried out reveals the physical mechanism of degassing with the spray-heating degasser. The pressurizer degassing model can be used for transient degassing analysis, and it also provides a basis for the subsequent design of the control system of pressurizer degassing.

The Trimming-Based Design Method for PWR Coolant Flow Distribution Device

In order to resolve the situations of nonuniform coolant flow distribution and insufficient vortex suppression, the existing Pressurized Water Reactor (PWR) usually adopts complex coolant mixing structures. However, those structures will greatly increase the complexity and maintenance cost of the system. To solve this problem, a trimming-based design method is proposed in this paper for the complex system and applies it to the design process of the PWR coolant flow distribution device. The function model of the coolant flow distribution system is built based on its functional analysis, and, according to the result of the component feature analysis, the columns and part of the basket are trimmed in order to simplify the overall structure of the system. To further solve the technical contradictions occurred in the simplified system, the contradiction solving tools of TRIZ theory are adopted. By setting the stereo flow equalizing plate, which can strengthen the function of flow distribution and vortex suppression, a coolant flow distribution device for PWR based on dome structure is obtained finally. This device owns a simple structure with good effect on coolant flow distribution and vortex suppression, which can achieve the goal of uniform coolant flow distribution of the system effectively.

Water chemistry, solution resistivity, and oxygen reduction reaction in simulated primary coolant of pressurized water reactors

AbstractThe effect of temperature (25 °C–300 °C) on the ohmic potential drop (ΔEΩ) and resistivity (Rsol) of a simulated primary coolant of the pressurized water reactors was investigated using electrochemical impedance spectroscopy (EIS) and galvanostatic potentiometry. The variations in the conductivity and pH of the solution showed that by increasing the temperature the magnitude of the Rsol and therefore ΔEΩ of the system first decreases rapidly (until ~150 °C) and then reaches a relatively stable plateau value. Increasing the pressure of the system decreased the magnitude of the Rsol in the oxygenated coolant. Investigation of the oxygen reduction reaction (ORR) on platinum (Pt), 316L stainless steel (SS), Alloy 690, and carbon steel revealed the great effect of water chemistry, in particular pH on the rate of the ORR on these materials. OH− adsorption on Pt reduced the rate of the ORR on this metal at temperatures higher than ~100 °C. Finally, it was found that higher H3O+ concentrations reduce the rate of the ORR on iron‐ and nickel‐based alloys, while increasing the OH− concentration at temperatures higher than ~150 °C accelerates the rate of the ORR on these alloys.

Zinc addition and its effect on the corrosion behavior of a 30% cold forged Alloy 690 in simulated primary coolant of pressurized water reactors

The effectiveness of zinc on the electrochemical behavior and oxide film formation of a 30% cold forged Alloy 690 exposed to high-temperature borated and lithiated water was investigated. The morphology and chemical composition of the formed oxide films with and without zinc injection were analyzed and compared by TEM–EDS and XPS techniques. Zinc clearly improved the electrochemical corrosion parameters of the alloy. Cathodic polarization improved the passivation of the alloy especially in presence of zinc. It was also found that actual zinc concentrations higher than 10 ppb is needed to observe a significant improvement in the corrosion resistance of the alloy in the pressurized water reactor operating conditions. Finally, further investigations of the effect of cold work on the high–temperature electrochemical performance of Alloy 690 and other iron and nickel based alloys in zinc injected solutions is recommended for the future studies.

Effect of electropolishing on general corrosion of Alloy 690TT tubes in simulated primary coolant of pressurized water reactors

The purpose of this study is to investigate the effect of electropolishing (EP) time on the general corrosion of Alloy 690TT steam generator tubes in a simulated PWR primary water of at 330 °C. The surface roughness and nanohardness of the tubes gradually decreased, and the electron work function and water contact angle gradually increased with increasing the EP time. However, the effect of EP time on the surface properties mainly changed after 8 min. The corrosion rate was reduced through EP treatment by approximately 60% compared to that of the commercial as-received tube.

Methodology for Stress-Controlled Fatigue Test Under In-Air and Pressurized Water Reactor Coolant Water Condition and to Evaluate the Effect of Pressurized Water Reactor Water and Loading Rate on Ratcheting

This work investigates the behavior of 316 stainless steel (SS) under stress-controlled low cycle fatigue loading. Several fatigue experiments are conducted under different environment such as in air at 300 °C and primary loop water conditions for a pressurized water reactor (PWR). Two different loading conditions are also employed to examine the effect of stress rate on material hardening and ratcheting. During PWR water test, actuator position measurements are used to determine the strain of the specimen. Under PWR environment, 316 SS is found to ratchet to a significantly greater degree compared with in air. At slow stress rate, higher amount of cyclic hardening is observed in 316 SS, and slow stress rate increases the rate of ratcheting. Results also indicate that 316 SS exhibits asymptotic strain response at higher stress loading which can cause material to behave very differently under same stress cyclic loading.

Reassessment of oxidation-induced amorphization and dissolution of Nb precipitates in Zr−Nb nuclear fuel cladding tubes

The surface oxide film of a Zr−2.5Nb alloy subjected to long term corrosion at 633 K in simulated primary coolant of pressurized water reactors has been analyzed. The primary concerns were whether Nb precipitates exhibit amorphization upon oxidation, and whether they dissolve into the matrix, as suggested by previous studies. Their behavior is of particular interest, from the viewpoint of engineering, as the mechanism of improving corrosion resistance of Zr fuel cladding by Nb addition, and from the viewpoint of basic materials science, as the critical condition of solid-state amorphization. If amorphization and dissolution proceed simultaneously, it would follow that amorphization occurs at conditions where both O and Nb atoms are mobile; under such conditions diffusion-induced amorphization has never been observed. It was found that the Nb precipitates exhibited amorphization without dissolution. Some of the inconsistencies among the previous studies were found to be artifacts of materials characterization methods. The final configuration of precipitates was amorphous Nb2O5, which is distinct from the other Nb oxides in terms of its dielectric nature with a wide band gap. The matrix initially contained a large amount of Nb greater than the solubility. Although the excess Nb atoms did not precipitate by thermal aging alone, oxidation was found to enhance their precipitation at this temperature. It appears that amorphization can occur even when the motion of atoms is not frozen-in.

A new correlation for convective heat transfer coefficient of water–alumina nanofluid in a square array subchannel under PWR condition

The computational fluid dynamic (CFD) simulation is performed to determine on the thermo- and hydrodynamic performance of the water–alumina (Al2O3) nanofluid in a square array subchannel featuring pitch-to-diameter ratios of 1.25 and 1.35. Two fundamental aspects of thermal hydraulics, viz. heat transfer and pressure drop, are assessed under typical pressurized water reactor (PWR) conditions at various flow rates (3×105⩽Re⩽6×105) using pure water and differing concentrations of water–alumina nanofluid (0.5–3.0vol.%) as coolant. Numerical results are compared against predictions made by conventional single-phase convective heat transfer and pressure loss correlations for fully developed turbulent flow. It is observed that addition of tiny nanoparticles in PWR coolant can give rise to the convective heat transfer coefficient at the expense of larger pressure drop. Nevertheless, a modified correlation as a function of nanoparticle volume fraction is proposed to estimate nanofluid Nusselt number more precisely in square array subchannel.

Plurality of radiation-chemical steady states during the radiolysis of water coolant in nuclear reactor

A computational study has been performed on the radiolysis of an aqueous solution that has the chemical composition and irradiation parameters typical of the primary coolant of pressurized water reactors. In this work, in contrast to previous studies on this subject, it has been taken into account that temporary fluctuations of the reactor radiation parameters, such as the composition of reactor n, γ]-radiation and the absorbed dose rate in the coolant, may occur in an actual real reactor during irradiation. The feasibility of spontaneous switching of the radiation-chemical system from one steady state with low concentrations of oxidizing products of radiolysis to another state with high concentrations resulting from a spike of the reactor radiation dose rate against the background of constant dose rate before and after the spike has been shown. The event triggering the switch is an increase in the concentration of oxygen and HO2 (O2–) radicals during the spike, after which the positive feedback does not allow the system to return back to the initial steady state. In practice, this effect can be caused by phenomena associated with local irregularities in the reactor core: the composition and dose rate of reactor radiation, the presence of steam–gas bubbles due to local subcooled nucleate boiling of the coolant in the core, and the perturbation of the water radiolysis mechanism at the surface of fuel rods because of an increased contribution of heterogeneous reactions etc.

Nanostructure and local properties of oxide layers grown on stainless steel in simulated pressurized water reactor environment

The morphology and local electrical resistance of duplex oxide films formed on 316L stainless steel at 325°C in simulated primary coolant of pressurized water reactor have been investigated at the nanometre scale with Conductive Atomic Force Microscopy. The electrical resistance varies over ∼1 order of magnitude for most oxide grains and over 2–3 orders of magnitude locally at grains and grain boundaries. This is rationalized in terms of local variation of the composition and thus resistivity of the mixed Fe(II)–Cr(III) barrier inner layer of the oxide films with grain boundaries of the outer layer possibly promoting Cr(III) enrichment.

Characteristics of Composite Silicon Carbide Fuel Cladding after Irradiation under Simulated PWR Conditions

Silicon carbide possesses a high melting point, low chemical activity, no appreciable creep at high temperatures, and a low neutron absorption cross section, making it an attractive material to investigate for use as fuel cladding in light water reactors. The cladding design investigated herein consists of three layers: an inner monolith of SiC, a central composite layer of SiC fibers infiltrated with SiC, and an outer SiC coating to protect against corrosion. The inner monolith provides strength and hermeticity for the tube, and the composite layer adds strength to the monolith while providing a pseudo-ductile failure mode in the hoop direction. The tube may be sealed by bonding SiC end caps to the SiC tube. A number of samples were irradiated in a test loop simulating pressurized water reactor coolant and neutronic conditions at the Massachusetts Institute of Technology research reactor. Postirradiation hoop stress testing via internal pressurization revealed 10% to 60% strength reduction due to physical properties mismatches between the three layers and corrosion. Weight loss measurements indicated that some irradiation-assisted corrosion occurred. Scanning electron microscope analysis allowed determination of the fracture mechanisms for specimens ruptured during hoop testing. The thermal diffusivities of the as-fabricated three-layer tube samples were measured to be roughly three times lower than those of the as-fabricated monolith layer. With irradiation, the thermal diffusivities decreased by factors of 14 and 8 for the monolith and three-layered samples, respectively. This change may be attributed to radiation damage and the formation of a silica layer on the sample surface. Anisotropic swelling of the bonded α-SiC blocks was sufficient to fail five of the six bond test specimens after a 1.5-month irradiation. Two of each of the calcium aluminate and Ti foil bonded samples failed. One of two TiC/SiC bond samples survived.