Pressurized Water Reactor Conditions Research Articles

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.

Qualification of a full plant nodalization for the prediction of the core exit temperature through a scaling methodology

System codes and their necessary power plant nodalizations are an essential step in thermal hydraulic safety analysis. In order to assess the safety of a particular power plant, in addition to the validation and verification of the code, the nodalization of the system needs to be qualified. Since most existing experimental data come from scaled-down facilities, any qualification process must therefore address scale considerations. The Group of Thermal Hydraulic Studies at Technical University of Catalonia has developed a scaling-up methodology (SCUP) for the qualification of full-scale nodalizations through a systematic procedure based on the extrapolation of post-test simulations of Integral Test Facility experiments. In the present work, the SCUP methodology will be employed to qualify the nodalization of the AscóNPP, a Pressurized Water Reactor (PWR), for the reproduction of an important safety phenomenon which is the effectiveness of the Core Exit Temperature (CET) as an Accident Management (AM) indicator. Given the difficulties in placing measurements in the core region, CET measurements are used as a criterion for the initiation of safety operational procedures during accidental conditions in PWR. However, the CET response has some limitation in detecting inadequate core cooling simply because the measurement is not taken in the position where the cladding exposure occurs. In order to apply the SCUP methodology, the OECD/NEA ROSA-2 Test 3, an SBLOCA in the hot leg, has been selected as a starting point. This experiment was conducted at the Large Scale Test Facility (LSTF), a facility operated by the Japanese Atomic Energy Agency (JAEA) and was focused on the assessment of the effectiveness of AM actions triggered by CET measurements. The steps of the SCUP methodology are presented: post-test calculation, scaling effect analysis and design effect analysis. The AscóNPP nodalization has shown to be qualified for the simulation of the involved phenomenology. The final step of the work presented here was to adapt the boundary conditions to a more realistic situation taking place in the AscóNPP. CET and Peak Cladding Temperature (PCT) readings were seen to present large differences similarly as it occurred in the ROSA-2 Test 3. The PCT when all the CET readings were credited was 872K, however, when only the minimum CET reading was credited the maximum temperature in the core rose to 1053K.

Helium Effects on 316L Austenitic Stainless Steel Fracture Mechanism

The most common fracture mechanism of nuclear reactor internals is irradiation-assisted stress corrosion cracking (IASCC). Its susceptibility at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation and radiation hardening. However, the aging of the nuclear fleet combined with the increase of their life-span reveals other mechanisms that could play an important role on IASCC susceptibility. A large amount of helium (He) can be accumulated in reactor internal components of pressurized water reactors (PWR) after long term operation. This occurrence could significantly increase (or even dominate) the IASCC susceptibility at high doses. He has been homogeneously implanted in an especially designed miniaturized specimen at 300°C up to 1000 appm. Slow strain rate tests (SSRT) results in high temperature air and in simulated PWR conditions indicate that homogenized, as-implanted He does not have a significant effect on IASCC up to 1000 appm under these test conditions.

CFD simulation of a four-loop PWR at asymmetric operation conditions

The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

Tribocorrosion in Pressurized High-Temperature Water: A Mass Flow Model Based on the Third-Body Approach

Pressurized water reactors (PWR) used for power generation are operated at elevated temperatures (280–300 °C) and under high pressure (120–150 bar). In addition to these harsh environmental conditions some components of the PWR assemblies are subject to mechanical loading (sliding, vibration and impacts) leading to undesirable and hardly controllable material degradation phenomena. In such situations wear is determined by the complex interplay (tribocorrosion) between mechanical, material and physical–chemical phenomena. Tribocorrosion in PWR conditions is at present little understood and models need to be developed in order to predict component lifetime over several decades. This paper present an attempt to model PWR tribocorrosion through the combination of a tribological third-body approach with a mechanistic description of the involved flows and the mass balance compartments corresponding to well-defined loci of the contact. The obtained model permits to gain better insight in the phenomenology and in the mechanisms of tribocorrosion of metals in PWR conditions. It also allows assessing the relative role of a variety of materials, mechanical and electrochemical parameters affecting the entire system. Quantitative predictions of the model were found to fit reasonably well experimental observations

Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol–gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.Between 2004 and 2006, four fuel segments were irradiated, with on-line recording of centerline temperature and rod pressure of the two instrumented rods and intermittent non-destructive hot-cell investigations of the other two non-instrumented rods. At the end of 2006, the instrumented rods were unloaded for hot-cell investigations. The hot-cell investigations reduced uncertainties in the power history to build a reliable and consistent irradiation history which can be used to assess and validate fuel performance codes. The on-line recorded temperatures of the instrumented rods are presented in this paper and are compared to corresponding calculations on the basis of the power history.One of the non-instrumented rods was re-inserted in the reactor in 2012 and attained a peak burnup level of 37 GWd/tHM by the end of 2014. The combined data set of on-line measurements and post irradiation examinations enables further code validation. In this context, the results of the in-house MACROS code of SCK·CEN have been compared with the experimental results. The code contains dedicated (Th,Pu)O2 models for the calculation of the thermal conductivity as a function of the burnup and models that determine the radial power profile within the pellet.

Investigation of the impact of coatings on corrosion and hydrogen uptake of Zircaloy-4

In this study, commercially available coatings of CrN, TiAlN and AlCrN were applied by physical vapour deposition to obtain 2–4 μm thick coatings on Zircaloy-4. The corrosion resistance, as well as the impact on hydrogen permeability of these coatings, was investigated. The latter is important with respect to hydrogen embrittlement of reactor components exposed to water with a certain dissolved hydrogen content. The corrosion resistance was investigated under pure water at 350 °C for 24 h and under CANDU conditions (300 °C and pH 10.5) for 30 days. Limited experiments were also performed in superheated steam (atmospheric pressure) at temperatures up to 1100 °C. Surface imaging (scanning electron microscopy) was performed before and after the exposure of the samples, with additional transmission electron microscopy analysis on a limited selection of samples. The composition of the coatings was determined by energy dispersive x-ray spectroscopy and depth characteristics were determined by secondary ion mass spectrometry. The CrN-coated samples have shown the best corrosion resistance to date, under pressurized water reactor, CANDU, and superheated steam conditions. Additionally, oxidation of the CrN and TiAlN coatings caused a significant reduction in the hydrogen uptake.

XPS study of Ni‐base alloys oxide films formed in primary conditions of pressurized water reactor

The oxide scales formed on alloy 690 during its exposure to pressurized water reactor primary simulated media at 325 °C for different periods, ranging between 24 and 858 h have been characterized through XPS analysis. These results have been compared with the data obtained on Cr2O3, NiCr2O4, NiFe2O4, and Ni(OH)2 reference compounds. This comparison leads to the following conclusions about the scale formed on alloy 690: (i) the deconvolution of Cr‐2p3/2 core level spectra seems to show that the internal oxide scale is made of a mix of chromite and chromia; (ii) the evolution of the height ratio between satellite and principal peaks, on the Ni‐2p3/2 core level, has highlighted the presence of nickel hydroxide for all corrosion durations; and (iii) XPS analysis can be used to study the kinetics of nickel ferrite precipitation. In order to obtain a more complete description of such oxide scales based upon XPS measurements, reference data, obtained on nickel ferrites and chromites in which Fe is partially substituted to Ni, are needed. The comparison of published data and reference spectra performed in this work also leads to conclude that core level fits are dependent on the energy resolution, which can be achieved with specific spectrometer and analysis procedures. Therefore, reference compound spectra need to be recorded in the very identical analytical configuration before dealing with results obtained on corroded specimen. Copyright © 2015 John Wiley & Sons, Ltd.

Investigation of two-phase flow instabilities under advanced PWR conditions

The characteristics of two-phase flow instability under advanced pressurized water reactor (PWR) conditions are explored in this study. Density wave oscillations in vertical parallel tubes are experimentally investigated by employing three different test sections to observe the parametric effects of the axial power shape, unheated riser, and channel inlet throttling on the onset of flow instability (OFI). A critical heat flux (CHF) accompanied by flow oscillation is indicated under certain test conditions with high pressure and/or low inlet subcoolings. A frequency-domain linear stability analysis code, called ALFS_F, was developed on the basis of one-dimensional two-phase mixture equations and a reactivity feedback model with point neutron kinetics and lumped parameter heat conduction equations. The mass flux at OFI conditions and the oscillation periods are calculated by the ALFS_F to compare with the experimental data. The thermal margin characteristics are illustrated with a comparative analysis of OFI and CHF margins under the core conditions of SMART, which is an integral-type advanced PWR being developed at KAERI.

Effects of oxide on fatigue crack growth behaviour of type 347 stainless steel in PWR water conditions

AbstractThe fatigue crack growth behaviour of type 347 stainless steel in pressurized water reactor (PWR) conditions was investigated at two different temperatures, 25 and 316 °C. The fatigue crack growth rate was slightly increased at the elevated temperature in air. In the simulated PWR water environment, the fatigue crack growth rate was changed in a rather complicated manner with a change of the dissolved oxygen (DO) content. The DO content did not significantly change the fatigue crack growth rate at 25 °C. With the lower oxygen content of 5 ppb, which is the practical limit of deoxygenated water, the fatigue crack growth rate was similar at both 25 and 316 °C. The fatigue crack growth rate was significantly decreased at 316 °C with the higher oxygen level of 100 ppb. Under 316 °C water conditions, oxides were observed on the fatigue crack surface where the size of oxide particles was about 0.2 µm at 5 ppb and about 1 µm at 100 ppb. The thickness of the oxide layers also increased with the increase of DO. Moreover, the ΔK threshold (ΔKth) also increased as the DO increased from 5 to 100 ppb. The dissolved hydrogen levels did not affect the measured crack growth rate at the given test conditions. The decrease of the fatigue crack growth rate with higher DO content is attributed to a crack closure resulting from the formation of larger oxides near the crack tip at a rather fast loading frequency of 10 Hz that was used in this study.

ICONE23-1841 THERMAL HYDRAULIC ANALYSIS OF WATER COOLED BREEDER BLANKET DESIGNED FOR CFETR

CFETR (Chinese Fusion Engineering Testing Reactor) is a superconducting Tokamak device. The simulation was performed on the cooling circuit inside water cooled breeder blanket which is one of the breeding blanket candidates for CFETR to compare the thermal hydraulic characteristics based on superheated steam and PWR (pressurized water reactor) water conditions. The work is carried out using MELCOR code. As a result, under the optimal flow distribution, it turns out that the thermal hydraulic characteristics meet the design requirements and the nuclear heat inside blanket module as well as heat flux from plasma could be removed completely by the cooling circuits on both conditions. It's also found that the temperature of first wall (FW) is higher on PWR condition than superheated steam. But due to the generation of superheated steam, the wall temperature would grow significantly in the backward zone of cooling circuit. Based on the steady state, LOFA (loss of flow accident) was simulated. It was predicted that FW will melt anyway due to the high nuclear heat without considering thermal radiation. As for different ways of FPSS, a 100s delay is invalid and a 3s delay could buy us time to take measures.

A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {111} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

First-principles investigation of boron incorporation into CRUD under Pressurized Water Reactor conditions

ABSTRACTThe accumulation of boron within the porous nickel ferrite (NiFe2O4, NFO) deposits on nuclear fuel rods is a major technological problem with important safety and economical implications. In this work, first-principles results are combined with experimental thermochemical data to analyze the energetics of vacancy formation in NFO and the possibility of B incorporation into the structure of NFO. Under solid-solid equilibrium conditions, the calculations suggest that vacancy formation and B incorporation into the NFO structure is energetically unfavorable, the main limiting factors being the narrow stability domain of NFO and the precipitation of B2O3, Fe3BO5, and Ni3B2O6 as secondary phases. Assuming solid-liquid equilibrium between NFO and the surrounding aqueous solution saturated with respect to NFO, the calculations predict that in operating PWR environment, Ni vacancies are likely to form. Under these conditions the possibility of B incorporation at the Ni vacancy sites cannot be excluded.

Two-Phase CFD Study of Coolant through a 5×5 Rod Bundles with Grid Spacers

A numerical study using computational fluid dynamics (CFD) under air/water two-phase flow condition without thermal transmission is carried out utilizing the commercial CFD code CFX. A detailed geometry of 5×5 rod bundles with two grid spacers in the Pressurized Water Reactor (PWR) is set up for analyzing. The Multiple Size Group (MUSIG) model based on the population balance equation is employed to describe the characteristic of flow field in which the dispersed phase has a large variation in size, and the breakup and coalescence interaction among different sizes of the bubbles. The effects of the size fraction at inlet, the size group quantities, the breakup and coalescence coefficients, the momentum transfer model, and so on, are studied. The results demonstrate that various parameters and two-phase model have different impacts on the analysis results. The detailed CFD modeling methodology for two-phase flow conditions in PWR fuel assemblies is developed according to the above-mentioned investigations. The flow field patterns and air/water phase distributions in the computational results shows that the two-phase modeling methodology in the present work is reasonable.

Formation of organic iodides from containment paint ingredients caused by gamma irradiation

The formation of volatile alkyl iodides other than methyl iodide during a serious nuclear reactor accident may have radiological significance. The hypothesis that radioactive alkyl iodides, other than methyl iodide, could form from paint solvents under the conditions of a serious nuclear accident in light water reactors (under boiling water reactor (BWR) and pressurised water reactor (PWR) conditions) was tested using stable elemental iodine, a gamma irradiator and gas chromatography equipment. It was found that methyl and isopropyl iodides were formed from the texanol ester, which is used in many modern water-based paints. Methyl, ethyl, propyl and butyl iodides were formed from a hydrocarbon solvent (white spirit) commonly used in paint products used in the past. These results suggest that further work on the formation and behaviour of the higher alkyl iodides (containing more than one carbon atom) under the conditions of a serious nuclear accident is justified.

Corrosion behavior of stainless steel 316 in sub- and supercritical aqueous environments: Effect of LiOH additions

Corrosion control in nuclear power systems is examined via addition of LiOH to the coolant at pressurized water reactor (PWR) and supercritical water-cooled reactor (SCWR) operating conditions in stainless steel tubing. The loss of metal to the coolant is analyzed using the voltammetry method. The SEM/EDX analysis of metallographic cross-sections is performed. The results indicate that an adequate pH control is possible for water temperatures up to 500°C. Above this temperature pH control becomes progressively more difficult as dielectric constant and density of water decrease. Significant hydrogen production is detected at 650°C.

Modeling the critical hydrogen concentration in the AECL test reactor

Hydrogen is added to a pressurized water reactor (PWR) to suppress radiolysis and maintain reducing conditions. The minimum hydrogen concentration needed to prevent radiolysis is referred to as the critical hydrogen concentration (CHC). The CHC was measured experimentally in the mid-1990s by Elliot and Stuart in a reactor loop at Atomic Energy of Canada (AECL), and was found to be approximately 0.5scc/kg for typical PWR conditions. This value is well below industry-normal PWR operating levels near 40scc/kg. Radiation chemistry models have also predicted a low CHC, even below the AECL experimental result. In the last few years some of the radiation chemical kinetic rate constants have been re-measured and G-values have been reassessed by Elliot and Bartels. These new data have been used in this work to revise the models and compare them with AECL experimental data. It is quite clear that the scavenging yields tabulated for high-LET radiolysis by Elliot and Bartels are not appropriate to use in the present context, where track-escape yields are needed to describe the homogeneous recombination kinetics in the mixed radiation field. In the absence of such data for high temperature PWR conditions, we have used the neutron G-values as fitting parameters. Even with this expedient, the model predicts at least a factor of two smaller CHC than was observed. We demonstrate that to recover the reported CHC result, the chemistry of ammonia impurity must be included.

TBR comparison of water-cooled blanket based on PWR and SCWR water conditions

Nuclear analysis results were compared for water-cooled blanket based on PWR (pressurized water reactor) and SCWR (sub-critical water reactor) water conditions. The local TBR (tritium breeding ratio) in outboard zone was discussed in the range of Pn (neutron wall load) from 1MW/m2 to 5MW/m2. It was found that water fraction has little impact on TBR, which is an important factor related to blanket tritium efficiency. It indicated that TBR value of each Pn would be similar under the two kinds of water conditions, but PWR case is a little higher than that of SCWR's. In addition, it was found that beryllium is the dominant factor leading a higher TBR inside blanket. As a result, TBR is an insensitive value with the water condition variation. The results would be important to water condition choice for solid blanket in the future.

Flammability Limits of Hydrogen-Air Mixtures

The aim of the present work is to identify and characterize the type of combustion of hydrogen-air mixtures near the flammability limits for different initial temperatures (from 298 to 423 K) and pressures (100 and 250 kPa) relevant to pressurized water reactor conditions. This experimental study has been carried out using a spherical vessel equipped with a pressure transducer to monitor the pressure increase subsequent to the combustion and with two optical windows to record the flame propagation. From the schlieren images, different regimes of flame propagation have been identified depending on the temperature and pressure. The maximum pressure obtained experimentally has been compared to the theoretical maximum pressure for adiabatic combustion at constant volume. The flammability limits have been determined for different temperatures and pressures and are compared to the literature.

Corrosion fatigue crack growth behaviour of austenitic stainless steels under light water reactor conditions

The corrosion fatigue crack growth behaviour of different wrought low-carbon and stabilised austenitic stainless steels was characterised under simulated boiling water and primary pressurised water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens in the temperature range from 70 to 320°C.The major parameter effects and critical conjoint threshold conditions, which result in relevant environmental acceleration of fatigue crack growth are discussed and summarised. Furthermore, the observed corrosion fatigue behaviour is compared with the corresponding (corrosion) fatigue curves in the ASME and JSME boiler and pressure vessel code or open literature and conclusions with regard to their adequacy and conservatism are given.