PWR Conditions Research Articles

Further improvements of the water-cooled Pb–17Li blanket

The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic–martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept.

Shadow corrosion or crevice corrosion?

Local corrosion phenomena in proximity to other components may have different reasons. In narrow gaps, crevice corrosion can be formed under BWR and to a lesser degree under PWR conditions independent of the counter material. At positions, where dissimilar metals are close to Zry under oxidizing conditions, shadow corrosion, driven by large potential differences, can occur.

Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions.

Numerical Simulation of Two-Phase Boiling Flows and Prediction of DNB under PWR Conditions with a Multidimensional Two-Fluid Model.

The present paper reports numerical modeling of multi-dimensional two-phase boiling flows with a multi-dimensional two-fluid model and prediction of the Departure from Nucleate Boiling (DNB) under PWR accidental conditions. The conventional two-fluid model is adopted and various constitutive equations are incorporated to develop a computer code capable of predicting multi-dimensional void fraction distribution in a flow-boiling channel. The numerical results are compared with the data in literature of the measured profile of void fraction, with the agreement being satisfactory. To predict the occurrence of DNB, the Weisman-Pei model is adopted, which predicts the occurrence of DNB when the local void fraction exceeds a predetermined value in the near-wall buble layer. The predicted DNB power is compared with the results from an experiment of DNB power in an annular channel under PWR flow and pressure conditions, and the agreement is encouraging.

Fuel performance under normal PWR conditions: A review of relevant experimental results and models

Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

Characterization of instabilities during two-phase natural circulation in typical PWR conditions

Strong oscillations in fluid velocities and densities were measured in the LOBI test facility during natural circulation experiments. A sort of “siphon condensation” occurs in the U-tubes of steam generators when primary-side mass inventory reaches roughly 64% of the initial value. The present paper deals with the characterization of the phenomenon considering flooding and condensation dynamics of U-tubes; RELAP5/MOD2 calculations made it possible to select system parameters affecting the oscillation characteristics. An attempt was made to evaluate the possibility of instability occurring in real plant situations.

High spatial resolution SIMS investigation of oxides formed on stainless steel under PWR conditions

High spatial resolution SIMS has been used to measure the variation in cobalt concentrations in oxide layers of stainless steel. Mass spectra from micron-sized crystals formed on the oxidised surface were recorded and depth profiles were taken through the inner oxide layer extending into the base metal. These measurements, used in conjuction with results from other more quantifiable techniques, allowed estimates to be made of the variation in cobalt concentration within the surface layers of the sample.

Characterization by XPS and AES of oxide films formed on stainless steel 304 under various PWR conditions

AbstractIncorporation of 60Co during growth of oxide films on alloys in PWR primary coolant circuits give rise to a significant radiological hazard. These oxide films have a duplex structure consisting of a thin inner layer and an outer layer of crystallites. XPS and AES techniques have been used to examine the oxide layers produced under a variety of controlled PWR coolant conditions. The elemental ratios obtained by the various techniques can be correlated with the extent to which the surfaces are covered with crystallites. The surfaces of coupons with thick outer layers of crystals were depleted in chromium with respect to the base metal composition, while examination of a specimen from which the outer layer had been removed chemically showed a significant enrichment of chromium.The relative differences between the oxides were analysed easily by these surface techniques. However, the absolute accuracy of the elemental composition was in some doubt owing to the lack of suitable, rigorously computed sensitivity factors for use in the quantitative analysis of spectra. An attempt has been made to refine the method for producing sensitivity factors appropriate to these materials.

Effect of single aging on stress corrosion cracking susceptibility of INCONEL X-750 under PWR conditions

Unfavorable morphology of precipitates and inclusions has been thought to be the cause of severe intergranular stress corrosion cracking (IGSCC) in double aged INCONEL* X-750 alloy used in reactor water environments. A single step aging treatment of 200 hours at 811 °C followed by furnace cooling after solution treating for 2 hours at 1075 °C has been found to provide an improved combination of strength, ductility, and resistance to SCC under simulated PWR test conditions. In this single aged condition a reprecipitated secondary carbide, together with γ′ was produced at the grain boundary which resulted in a mixed fracture mode comprising dimple rupture and microvoid coalescence compared with a predominantly intergranular mode for the fully age hardened specimens. This improvement has been explained in terms of the morphology of the second phase precipitates which are produced in these heat treatment regimes.

Stress corrosion cracking of nuclear pressure vessel steels in pressurized high temperature waters.

The quantitative evaluation of the subcritical crack growth behavior of nuclear pressure vessels and piping steels in pressurized high temperature water has been receiving extensive attention to ensure the structual integrity of nuclear pressure vessels and piping. Recent research progress in this field suggests the significance of the sulfur content of steels in assessing susceptibility to environmentally assisted cracking of these steels in high temperature water from the view point of solution chemistry at a crack tip, where dissolution of MnS, non-metallic inclusion, results in an enrichment of S anion, such as SO42-. Hence, it is neccesary to simulate the solution which has been commonly used as a useful tool to see stress corrosion cracking characteristics in a short time. In this material/environments systems are examined quantitatively as a function of temperature, sulfur content in steels and also SO42- contents in solution, simulating crack tip solution chemistry. Cracking potential showed a lower value of about 250°C, rather than 288°C. Sulfur content of steel and SO42- concentration in solution can drastically influence the cracking potential and some amount of sulfur in steels or SO42- anion at a crack tip make it possible to crack at a BWR or PWR condition.

Structural behaviour of a piping system at blowdown: Significant parameters for a comparison of measurement and calculation

As part of the German Reactor Safety Program, blowdown tests were conducted on a feedwater line in the HDR superheated steam reactor, which had been shut down, under PWR conditions. This work included the pre- and postcalculations of the dynamic structural behaviour of piping systems on the basis of measured blowdown pressure/time histories. In the present paper it is shown that the degree of agreement of the calculated results with the measured structural response is determined primarily by the establishment of equivalent force/time functions from the measured pressures at blowdown and by the application of these forces to the structural model of the piping system. In the precalculation the fluid-dynamic forces were determined by means of time-related extrapolation of the pressure/time histories measured in the straight pipe sections to the elbow centre points. When the measured and computed structural responses were compared, deviations were revealed. The reason for this was that the pressure/time functions measured during the blowdown test included significant portions from the reflection of pressure waves. These portions are not correctly covered by the time-related extrapolation method. In a postcalculation we selected a geometrical interpolation method to determine the fluid-dynamic elbow forces. This method does not give an exact reproduction of the frequency content of the pressure/time functions but does cover the magnitude of impulse for both the incoming and the reflected pressure waves. The clear improvement in compound results shows that this simple method is acceptable for determining structural responses. In contrast to the sensitivity of the computed results when the fluid-dynamic forces are determined in different ways, it became evident that the piping codes in common use today (which are based on the beam theory) make it possible to give an adequate description of the dynamic structural response with pressure waves arising from blowdown.

Piping response testing associated with pipe rupture

EPRI has sponsored an experimental program in the pipe whip impact and pipe rupture and depressurization areas. Sixteen pipe whip tests were performed with 3 in Schedule 80 (or 10) carbon steel pipes impacting on rigid target or concrete slab. The major testing parameters include distance, impact location, pipe rupture location, and concrete slab thickness and strength. The piping crushing at impact correlates with impact force and target response behavior. Conservatism was established by comparing measured and calculated impact forces. The pipe rupture and depressurization tests were carried out using 6 in stainless steel and carbon steel pipes under either PWR or BWR fluid conditions. These tests are of axial crack with initial machined-in surface flaw. It was found that pipe rupture would occur only if a long unstable through-wall crack was embedded in a sufficiently long unstable part-through crack (in the pipe wall). All other flaw configuration tested led to pipe leakage only. Reaction forces were measured which show conservatism of simplified method for fully ruptured condition. No good crack propagation information was obtained.

A theoretically based critical heat flux prediction for rod bundles at PWR conditions

A theoretically based procedure developed for round tubes has been applied to the prediction of DNB heat fluxes in rod bundles at PWR conditions. State-of-the-art subchannel analysis procedures were used to determine local flows and enthalpies. Very good comparison between DNB predictions and experimental observations are found for rod bundles which both uniform and non-uniform axial heat fluxes.

The effect of temperature on fatigue crack growth behaviour of a low alloy pressure vessel steel in a simulated BWR environment

Fatigue crack growth tests for a low alloy steel equivalent to ASTM A533 Grade B Class 1 were conducted under a simulated BWR environment in a temperature range from room temperature to 320°C. The frequency and the stress ratio were 1 cpm and 0.1, respectively. Two types of CT specimens with L-S and T-S orientations were employed. In situ monitoring crack length was continuously carried out by compliance method using LVDT. Complicated temperature dependences of fatigue crack growth rates were observed: the minimum growth rates were at 175°C, while those at 100°C showed the maximum enhancement by the environmental effect, which was approximately four times the maximum compared to the current air default line in ASME Code Section XI. Even the maximum enhancement of the crack growth rates at 100°C did not exceed the current water default lines. The discontinuities of the temperature dependences of fatigue crack growth rates under a simulated BWR condition were basically consistent with those under simulated PWR conditions.

On the mechanisms of environment sensitive cyclic crack growth of nuclear reactor pressure vessel steels

An analysis of the cyclic crack growth rate data generated so far for pressure vessel materials in simulated light water reactor environments suggests a strong dependence on frequency, load ratio, waveform, temperature and material composition. To account for these observations a mechanistic crack growth model has been advanced based on hydrogen-induced cracking, where anodic dissolution creates the conditions for hydrogen absorption at the crack tip. Hydrogen-induced cracking starts from the manganese sulfide inclusions, which act as strong hydrogen traps. The hydrogen-induced crack growth around the manganese sulfide inclusions generally spans several prior austenite grains. At high hydrogen input rates brittle crack growth also occurs, this being unrelated to inclusions. When the crack growth exposes manganese sulfide inclusions, these dissolve and the crack tip environment becomes aggressive and conducive to hydrogen absorption. This hydrogen-induced cracking model explains why inclusions form a preferred crack path, and accounts for the effect of sulfur on crack growth rate both in PWR and BWR conditions. Based on the model, the observed crack growth rate dependence on different testing variables can also be explained.

An Experimental Study of the Rapid Depressurization of Hot Water

New measurements of the pressure history in 5.08 and 1.27 cm i.d. tubes during extremely rapid depressurization from BWR and PWR conditions are presented. The pressure to which the present system (as well as the systems of other investigators) returns, is successfully correlated using a suggestion by Stuhmiller. New pressure undershoot data are given here, but they are rationalized elsewhere. The rate of opening, and the rarefaction wave speed, in the present system are also presented and correlated. The present study suggests that scaled replications of the early process of depressurization are reliable.

Enthalpy transfer between PWR fuel assemblies in analysis by the lumped subchannel model

Assessment is made of the error in enthalpy predictions in PWR cores resulting from various assumptions in the size of homogenized calculation regions. A coefficient is developed from the differential conservation equations and numerically determined to minimize these errors. It was found that these errors are only of significance for regions comprising three to seven subchannels, region sizes which may be employed in coupled neutronic-thermal hydraulic calculations. A correlation for coefficient values for a range of PWR conditions is presented.

Early Response of Hot Water to Sudden Release from High Pressure

The very early stages of blowdown are observed with the aid of a novel quick-opening mechanism. Water at pressures and temperatures up to PWR conditions is released from a long, one-half in. (1.27 cm) ID pipe. Depressurization rates up to 1.62 × 106 atm/s are achieved. As a consequence of the rapidity of depressurization, the water reaches pressures far below saturation (even approaching the spinodal line) before nucleation occurs. The performance of the opening device is described quantitatively, and implications of the response of the water are discussed in relation to nuclear safety problems.

A comparison between code calculation and blowdown experiments simulating a loss-of-coolant accident in a pressurized water reactor

Within the reactor safety programme of the EURATOM Joint Research Centre at Ispra the transient heat transfer phenomena during depressurization are experimentally investigated under PWR conditions. The special closed loop DHT-1 essentially represents one subchannel and the upper and lower plenum of a pressurized water reactor. A test series simulating rupture in the hot leg of a primary cooling circuit was carried out. Pressure and test tube temperatures were measured at various rupture cross-sections. Independently from these experiments, a blowdown computer code was developed by the Groupement Atomique Alsacienne Atlantique (GAAA). The core part of this code allows calculation of the thermohydraulic history of the coolant within the core after a rupture in the primary cooling circuit. It has been checked with regard to the hypothesis and correlations applied; the experiments and calculations are compared.

Properties of zircaloy cladding

This lecture is concerned with the mechanical aspects of fuel cladding, especially the zircaloys', and is related to performance requirements under BWR and PWR conditions, which, in this area, are not greatly dissimilar.