Melting attack of solid plates by a high temperature liquid jet — effect of crust formation

When a molten UO₂ jet impinges on a steel structure in a reactor vessel during a severe accident, the erosion rate of the steel by the molten UO₂ jet is expected to be limited considerably by a UO₂ crust layer forming on a molten steel substrate at the jet/steel plate interface. A series of simulation experiments was performed to study the melting behavior of solid plates by high temperature liquid jets and the effects of crust forming at jet/structure interface. In the first series of experiments, salt (NaCl) was selected as the jet material and tin (Sn) as the solid structure. The experiments were conducted with varying the jet diameter (10 30 mm) and jet temperature (900 1100°C). The jets were accelerated to a range of 3 5 m/s at the nozzle outlet by gravitational force and impinged perpendicularly to the solid plate underneath. Furthermore, to check the effects of the thermo-physical properties on the erosion behaviors, preliminary experiments were performed by using a molten Al₂O₃ jet ( 2200°C) impinging on stainless steel plate at room temperature. The erosion rates obtained in the present experiments were far less than the values predicted by an analytical solution that neglects the existence of a crust layer and its thermal effects. With the inclusion of the crust behavior in the model, the experimental results were predicted fairly well. From the present experiments, a Nusselt number of the turbulent heat transfer, which takes into account simultaneous melting and freezing in the impingement region of a molten jet, is correlated by a Reynolds number and a Prandtl number as follows: Nu_m = 0.0033 Re---Pr.In conclusion, the existence of a crust layer plays an important role in the erosion process of a solid plate by the molten fuel jet with high melting point as in a reactor situation.     

Simulations of turbulence, heat transfer and mixing across narrow gaps between rod-bundle subchannels

Turbulent flow and temperature fields were determined numerically in a rectangular duct containing a heated rod. As the spacing δ between the rod and the duct wall decreased from 0.10D (D is the rod diameter) to 0.03D, coherent turbulent kinetic energy and temperature fluctuations dramatically increased in the gap region, but, for δ = 0.01D, coherent fluctuations essentially disappeared. As δ/D → 0, the frequency of coherent fluctuations decreased and cross-gap mixing weakened, contrary to predictions based on extrapolated available empirical correlations.     

Numerical study on mixing of oscillating quasi-planar jets with low Reynolds number turbulent stress and heat flux equation models

A comparative investigation was conducted on the mechanistic numerical simulation of mixing of non-isothermal, quasi-planar jets incorporating low Reynolds number turbulent stress and heat flux equation models (LRSFM) and an experiment. A water test facility with three vertical jets, the unheated in between two heated jets, simulated convective mixing and temperature fluctuations expected at the outlet of a liquid metal fast reactor core. The LRSFM and a two-equation k- turbulence model were applied to the simulation. The LRSFM showed good agreement with the experiment, with respect to mean profiles and reproduced the oscillatory motion of the jetting flow, while the k- model under-predicted the mixing effect, such that a transverse mean temperature difference remained well downstream. Specifically the LRSFM results indicated that the influence of turbulence on mixing was of second order importance in the present flow, while the contribution by coherent phenomena, mainly associated with the periodic oscillation of the present jets, promoted the mixing. These results were evident in both the numerical simulation with LRSFM and in the experiment.     

Numerical analysis of fragmentation mechanisms in vapor explosions

Fragmentation of molten metal is the key process in vapor explosions. However, this process is so rapid that the mechanisms have not yet been clarified in experimental studies. In addition, numerical simulation is difficult because we have to analyze water, steam and molten metal simultaneously with boiling and fragmentation. The authors have been developing a new numerical method, the moving particle semi-implicit (MPS) method, based on moving particles and their interactions. Grids are not necessary. Incompressible flows with fragmentation on free surfaces have been calculated successfully using the MPS method. In the present study, numerical simulation of the fragmentation processes using the MPS method is carried out to investigate the mechanisms. A numerical model to calculate boiling from water to steam is developed. In this model, new particles are generated on water–steam interfaces. A two-step pressure calculation algorithm is also developed. Pressure fields are separately calculated in both heavy and light fluids to maintain numerical stability with the water and steam system. The new model and algorithm are added to the MPS code. Water jet impingement on a molten tin pool is calculated using the MPS code as a simulation of collapse of a vapor film around a melt drop. Penetration of the water jet, which is assumed in Kim–Corradini’s model, is not observed. If the jet fluid density is hypothetically larger, the penetration appears. Next, impingement of two water jets is calculated. A filament of the molten metal is observed between the two water jets as assumed in Ciccarelli–Frost’s model. If the water density is hypothetically larger, the filament does not appear. The critical value of the density ratio of the jet fluid over the pool fluid is ρ_jet/ρ_pool=0.7 in this study. The density ratios of tin–water and UO₂–water are in the region of filament generation, Ciccarelli–Frost’s model. The effect of boiling is also investigated. Growth of the filament is not accelerated when the normal boiling is considered. This is because normal boiling requires more time than that of the jet impingement, although the filament growth is governed by an instant of the jet impingement. Next, rapid boiling based on spontaneous nucleation is considered. The filament growth is markedly accelerated. This result is consistent with the experimental fact that the spontaneous nucleation temperature is a necessary condition of vapor explosions.     

The thermal neutron induced fission cross-section of Am

The σ(n_th,f) of ²⁴³Am has been measured using the cold neutrons with a 25 K Maxwellian distribution available at the Grenoble high flux reactor. Surface barrier detectors recorded the fission fragments. A value of σ(n_th,f) = (198.3 ± 4.2)mb was obtained for ²⁴³Am relative to σ(n_th,f) = (582.2 ± 1.3)b for ²³⁵U.     

Turbulent heat mixing of a heavy liquid metal flow in the MEGAPIE target geometry—The heated jet experiment

The MEGAPIE target installed at the Paul–Scherrer Institute is an example of a spallation target using eutectic liquid lead–bismuth (Pb⁴⁵Bi⁵⁵) both as coolant and neutron source. An adequate cooling of the target requires a conditioning of the flow, which is realized by a main flow transported in an annular gap downwards, u-turned at a hemispherical shell into a cylindrical riser tube. In order to avoid a stagnation point close to the lowest part of the shell a jet flow is superimposed to the main flow, which is directed towards to the stagnation point and flows tangentially along the shell.The heated jet experiment conducted in the THEADES loop of the KALLA laboratory is nearly 1:1 representation of the lower part of the MEGAPIE target. It is aimed to study the cooling capability of this specific geometry in dependence on the flow rate ratio (Q_main/Q_jet) of the main flow (Q_main) to the jet flow (Q_jet). Here, a heated jet is injected into a cold main flow at MEGAPIE relevant flow rate ratios. The liquid metal experiment is accompanied by a water experiment in almost the same geometry to study the momentum field as well as a three-dimensional turbulent numerical fluid dynamic simulation (CFD). Besides a detailed study of the envisaged nominal operation of the MEGAPIE target with Q_main/Q_jet = 15 deviations from this mode are investigated in the range from 7.5 ≤ Q_main/Q_jet ≤ 20 in order to give an estimate on the safe operational threshold of the target.The experiment shows that, the flow pattern establishing in this specific design and the turbulence intensity distribution essentially depends on the flow rate ratio (Q_main/Q_jet). All Q_main/Q_jet-ratios investigated exhibit an unstable time dependent behavior. The MEGAPIE design is highly sensitive against changes of this ratio.Mainly three completely different flow patterns were identified. A sufficient cooling of the lower target shell, however, is only ensured if Q_main/Q_jet ≤ 12.5. In this case the jet flow covers the whole lower shell. Although for Q_main/Q_jet ≤ 12.5 the flow is more unstable compared to the other patterns most of the fluctuations close to the centerline are in the high frequency range (>1 Hz), so that they will not lead to severe temperature fluctuations in the lower shell material. In this case the thermal mixing occurs on large scales and is excellent.For flow rate ratios Q_main/Q_jet > 12.5 complex flow patterns consisting of several fluid streaks and vortices were identified. Since in these cases the jet flow does not fully cover the lower shell an adequate cooling of the MEGAPIE target cannot be guaranteed and thus temperatures may appear exceeding material acceptable limits.All conducted experiments show a high sensitivity to asymmetries even far upstream. A comparison of the numerical simulation, which assumed a symmetric flow, with the experimental data was due to the experimentally found asymmetry only partially possible.     

Transformation and fragmentation behavior of molten metal drop in sodium pool

In order to clarify the fragmentation mechanism of a metallic alloy (U–Pu–Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 °C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point = 660 °C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T_i) between molten aluminum drop and sodium is lower than the boiling point of sodium (T_c,bp), the molten aluminum drop can be fragmented and the mass median diameter (D_m) of aluminum fragments becomes small with increasing T_i. When T_i is roughly equivalent to or higher than T_c,bp, the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt.     

Diffusion of uranium in molybdenum,niobium, zirconium,and titanium

The diffusion of uranium was studied in molybdenum, niobium, zirconium, and titanium. The diffusion coefficients were determined by measuring the over-all activity of the residue of the sample, using the a-radiation of uranium enriched with the U²³⁵ isotope to 90% at temperatures 1500 to 2000° C (molybdenum, niobium) and 915 to 1200° C (zirconium, titanium). The temperature dependence of the diffusion coefficients was given by the equationsD _Mo ^U=7.60.10³ exp (–76 400/RT) cm²/sec;D _Nb ^U=8.90.10^–2 exp (–76 800/RT) cm²/sec;D _Zr ^U=7.77,10^–5 exp (–25 800/RT) cm²/sec;D _Ti ^U=4.90. t0^–4 exp (–29 300/RT) cm²/sec.The considerable differences between the diffusion mobilities and activation energies of molybdenum and niobium on the one hand and zirconium and titanium on the other were probably due to the effects of lattice defects, for example, excess vacancies arising in zirconium and titanium during polymorphic transformations.Translated from Atomnaya Énergiya, Vol. 19, No. 6, pp. 521–523, December, 1965Report read by G. I. Budker at the International Conference on High-Energy Accelerators (Frascati, Italy).     

Fracture mechanics evaluation of small diameter piping considering the latest experimental results

The results obtained from investigations carried out on austenitic piping of small nominal diameter (DN80 and DN50) are introduced and discussed together with their assessment using fracture mechanics methods. Essential results are summarised as following. The pipes with flaws (fatigue crack) down to a depth to a_max/t=0.51 (DN80) as well as a_max/t=0.62 (DN50) and a circumferential extension of results 2α=120° reached bending angles up to 26°. The ASME collapse load (test collapse load) was exceeded considerably and the experimental maximum load could not be reached. Failure due to a leakage or rupture did not occur in any test. The maximum crack extension was 0.69 mm (DN80, a_max/t=0.51) resp. 0.3 mm (DN50, a_max/t=0.62). The experimental maximum load can approximately be assessed by the limit analysis. The fracture mechanics approximation methods GE/EPRI and LBB/NRC calculated a/t=0.4 and 2α=120° initiation loads above the experimental maximum load for pipes containing flaws. These results confirmed the procedures for the proof of integrity of small diameter piping by updating information on load, deformation and failure behaviour of austenitic piping damaged with circumferential flaws. Using these results may formulate a final safety concept for the proof of integrity of small diameter piping by completing the current concepts.     

Effects of an intermediate heat treatment during a cold rolling on the tensile strength of a 9Cr–2W steel

The effects of an intermediate heat treatment during a cold rolling on the tensile strength of a 9Cr–2W steel were evaluated. Before a cold rolling, the steel was normalized at 1050 °C and tempered at 550 °C in order to avoid the formation of M₂₃C₆ and V-rich MX precipitates in the martensitic structure. A 75% cold rolling and a heat treatment at 750 °C for 30 min induced the formation of large M₂₃C₆ carbides in a fully recrystallized structure. However, three cold rollings with an intermediate heat treatment at 750 °C for 10 min after each cold rolling led to the formation of fine and uniform M₂₃C₆ carbides in a partially recrystallized structure, providing an enhanced tensile strength at 650 °C. It is thus concluded that an intermediate heat treatment during a cold rolling could be an effective procedure for fabricating a high strength 9Cr–2W steel at high temperatures.     

European experiences of the proposed ASTM test method for crack arrest toughness of ferritic materials

The proposed ASTM test method for measuring the crack arrest toughness of ferritic materials using wedge-loaded, side-grooved, compact specimens was applied to three steels: A514 bridge steel tested at −30°C (CV30–50°C), A588 bridge steel tested at −30°C (CV30–65°C), and A533B pressure vessel steel tested at +10°C (CV30-12°C) and +24°C (CV30+2°C). Five sets of results from different laboratories are discussed here; in four cases FOX DUR 500 electrodes were used for notch preparation, in the remaining case HARDEX-N electrodes were used. In all cases, notches were prepared by spark erosion, although root radii varied from 0.1–1.5 mm. Although fast fractures were successfully initiated, arrest did not occur in a significant number of cases.The results showed no obvious dependence of crack arrest toughness, K_a, (determined by a static analysis) on crack initiation toughness, K₀. It was found that K_a decreases markedly with increasing crack jump distance, Δα/W. A limited amount of further work on smaller specimens of the A533B steel showed that lower K_a values tended to be recorded.It is concluded that a number of points relating to the proposed test method and notch preparation are worthy of further consideration. It is pointed out that the proposed validity criteria may screen out lower bound data. Nevertheless, for present practical purposes, K_a values may be regarded as useful in providing an estimate of arrest toughness — although not necessarily a conservative estimate.     

Simulation tests for temperature mixing in a core bottom model of the HTR-Module

Interatom and Siemens are developing a helium-cooled Modular High Temperature Reactor. Under nominal operating conditions temperature differences of up to 120°C will occur in the 700°C hot helium flow leaving the core. In addition, cold gas leakages into the hot gas header can produce even higher temperature differences in the coolant flow. At the outlet of the reactor only a very low temperature difference of maximum ±15°C is allowed in order to avoid damages at the heat exchanging components due to alternating thermal loads. Since it is not possible to calculate the complex flow behaviour, experimental investigations of the temperature mixing in the core bottom had to be carried out in order to guarantee the necessary reduction of temperature differences in the helium. The presented air simulation tests in a 1:2.9 scaled plexiglass model of the core bottom showed an extremely high mixing rate of the hot gas header and the hot gas duct of the reactor. The temperature mixing of the simulated coolant flow as well as the leakage flows was larger than 95%. Transfered to reactor conditions this means a temperature difference of only ±3°C for the main flow at a quite reasonable pressure drop. For the cold gas leakages temperature differences in the hot gas up to 400°C proved to be permissible. The results of the simulation experiments in the Aerodynamic Test Facility of Interatom permitted to design a shorter bottom reflector of the core.     

Socket weld integrity in nuclear piping under fatigue loading condition

The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading. The integrity of socket weld is regarded as a safety concern in nuclear power plants because many failures have been world-widely reported in the socket weld. Recently, socket weld failures in the chemical and volume control system (CVCS) and the primary sampling system (PSS) were reported in Korean nuclear power plants. The root causes of the socket weld failures were known as the fatigue due to the pressure and/or temperature loading transients and the vibration during the plant operation. The ASME boiler and pressure vessel (B & PV) Code Sec. III requires 1/16 in. gap between the pipe and fitting in the socket weld with the weld leg size of 1.09 × t₁, where t₁ is the pipe wall thickness. Many failure cases, however, showed that the gap requirement was not satisfied. In addition, industry has demanded the reduction of weld leg size from 1.09 × t₁ to 0.75 × t₁. In this paper, the socket weld integrity under the fatigue loading was evaluated using three-dimensional finite element analysis considering the requirements in the ASME Code. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P = 0 to 15.51 MPa, and the thermal transient ranging from T = 25 to 288 °C were considered. The results are as follows; (1) the socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME operation and maintenance (OM) code. (2) The effect of pressure or temperature transient load on socket weld in CVCS and PSS is not significant owing to the low frequency of transient during plant operation. (3) ‘No gap’ is very risky to the socket weld integrity for the systems having the vibration condition to exceed the requirement specified in the ASME OM Code and/or the transient loading condition from P = 0 and T = 25 °C to P = 15.51 MPa and T = 288 °C. (4) The reduction of the weld leg size from 1.09 × t₁ to 0.75 × t₁ may induce detrimental effect on the socket weld integrity.     

Conservatism of ASME KIR-reference curve with respect to crack arrest

The conservatism of the RT_NDT temperature indexing parameter and the ASME K_IR-reference curve with respect to crack arrest toughness, has been evaluated. Based on an analysis of the original ASME K_Ia data, it was established that inherently, the ASME K_IR-reference curve corresponds to an overall 5% lower bound curve with respect to crack arrest. It was shown that the scatter of crack arrest toughness is essentially material independent and has a standard deviation (S.D.) of 18% and the temperature dependence of K_Ia has the same form as predicted by the master curve for crack initiation toughness. The ‘built in’ offset between the mean 100 MPa√m crack arrest temperature, TK_Ia, and RT_NDT is 38°C (TK_Ia=RT_NDT+38°C) and the experimental relation between TK_Ia and NDT is, TK_Ia=NDT+28°C. The K_IR-reference curve using NDT as reference temperature will be conservative with respect to the general 5% lower bound K_Ia(5%)-curve, with a 75% confidence. The use of RT_NDT, instead of NDT, will generally increase the degree of conservatism, both for non-irradiated as well as irradiated materials, close to a 95% confidence level. This trend is pronounced for materials with Charpy-V upper shelf energies below 100 J. It is shown that the K_IR-curve effectively constitutes a deterministic lower bound curve for crack arrest The findings are valid both for nuclear pressure vessel plates, forgings and welds.     

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters in (D–T) driven hybrid reactor system

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters are investigated for different coolants under various first wall loads (P_w=2, 5, 7, 8, 9, and 10 MW m⁻²) in (D, T) (deuterium and tritium) driven and fueled with UO₂ hybrid reactors. Plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. The fissile fuel zone is considered to be cooled with four different coolants with various volume fractions, the volumetric ratio of coolant-to-fuel [(Vm/V_f) = 1:2, 1:1, and 2:1], gas (He, CO₂), flibe (Li₂BeF₄), natural lithium (Li), and eutectic lithium (Li₁₇Pb₈₃). Calculation in the fuel rods and the behavior of the fissile fuel have been observed during 4 years for discrete time intervals of Δt=15 days and by a plant factor (PF) of 75%. As a result of the calculation, cumulative fissile fuel enrichment (CFFE) value indicating rejuvenation performance has increased by increasing P_w for all coolants and . Although CFFE and neutronic performance parameter values increase to the higher values by increasing P_w, the maximum temperature in the centerline of the fuel roads has exceeded the melting point (T_m>2830°C) of the fuel material during the operation periods. However, the best CFFE (11.154%) is obtained in gas coolant blanket for =1:2 (29.462% coolant, 58.924% fuel, 11.614% clad), under 10 MW m⁻² first wall load, followed by flibe with CFFE=11.081% for =2:1 (62.557% coolant, 31.278% fuel, 6.165% clad), under 7 MW m⁻², and flibe with CFFE=9.995% for =1:1 (45.515% coolant, 45.515% fuel, 8.971% clad), under 7 MW m⁻² during operation period without reaching the melting point of the fuel material. While maximum CFFE value has been obtained in fuel rod row#10 in gas, natural lithium, and eutectic lithium coolant blankets, it has been obtained in fuel rod row#1 in flibe coolant blanket for all and P_w. At the same condition, the best neutronic performance parameter values, tritium breeding ratio (TBR)= 1.4454, energy multiplication factor (M)= 9.2018, and neutron leakage (L)= 0.0872, have been obtained in eutectic lithium coolant blankets for the =1:2, followed by gas, natural lithium, and flibe coolant blankets. The isotopic percentage of ²⁴⁰Pu is higher than 5% in all blankets for P_w 7 MW m⁻², so that plutonium component in all blankets can be never reach a nuclear weapon grade quality during the operation period.     

Hydrogen influence on mechanical and fracture mechanics characteristics of zirconium Zr–2.5Nb alloy at ambient and elevated temperatures

This article deals with the investigation of the hydrogen concentration and temperature influence onto mechanical and fracture mechanics characteristics of RBMK-1500 Ignalina NPP unit 2 reactor fuel channel material—Zr–2.5Nb zirconium alloy (TMT-2) at temperatures from ambient up to 300 °C. The investigation of mechanical characteristics was performed on tensile specimens, fracture mechanics characteristics K_Q, , J_IC—on compact specimens (B = 4 mm) of hydrogen-free and saturated by hydrogen (52, 100 and 140 ppm) at 20, 170, 200 and 300 °C. The investigation showed that temperature increasing calls mechanical strength decreasing, whereas the reductions of area increase. Stronger influence of hydrogen concentration onto mechanical characteristics is noticed only at 20–170 °C temperature, however this influence diminishes as the temperature increases and weakest hydrogen influence is given at 300 °C. Fracture toughness characteristics K_Q, more depends on temperature than on hydrogen concentration. Critical J_IC integral values for the specimens containing hydrogen were given lowest at 20 °C, increases when temperature were raised up to 140 °C and were given highest when it reaches 300 °C.The analysis of and J_IC dependence due to the mechanical characteristics of zirconium alloy has showed that the modified plasticity Z_mod = (R_p0.2/R_m)Z satisfactorily approximates the influence of temperature and hydrogen concentration on variation of these characteristics.     

Break up time of fragmentating-solidifying UO2 spheres when quenched in sodium

When molten UO₂ is quenched in sodium, a sand-like debris results containing about 80% of fractured particles and 20% of smooth particles and spheres. The production of the fractured particles is normally explained by the thermal stress fragmentation model. Previously brittle fracture mechanics was applied to the complete solid shell of a freezing UO₂ drop, i.e. where 954°C < T < 2850°C; a calculation of fragmentation time was not possible. In this contribution the solid shell is continuously subdivided in a plastic or ductile layer for 1300°C < T < 2850°C and a brittle one for 945°C < T < 1300°C. Cracking occurs in the brittle layer only. In the present model a layer of a predescribed depth is assumed to ablate instantaneously, when the temperature reaches the transition point of elastic of ductile behavior (T = 1300°C) at its inner boundary. A new layer is formed within a time step, governed by the heat conduction equation. The discontinuous ablation process is thus related to the continuous progression of the solidification front. A calculation of the fragmentation time is possible: in principle it comprehends the summation of a large number of time steps for the formation of brittle layers. The thickness of the cracked brittle layer is parametrized to 20, 10, 5 and 1 μm. The concept of instantaneous ablation was suggested by the experience that the violent boiling forces of sodium are very effective on the UO₂ surface. The introduction of these minor changes makes the thermal stress model more realistic, because it can explain now, why UO₂ does not fragment in argon and water. The fragmentation time assessed for a UO₂ drop of 7.2 mm diameter in sodium, brittle layer 10 μm, is 250 ms.     

Results on research of templates from Kozloduy-1 reactor pressure vessel

The studies on the specimens manufactured from the templates cut out from the weld 4 of Kozloduy NPP Unit 1 reactor vessel have been conducted. The data on chemical composition of the weld metal have been obtained. Neutron fluence, mechanical properties, ductile to brittle transition temperature (DBTT) using mini Charpy samples have been determined. The phosphorus and copper content averaged over all templates is 0.046 and 0.1 wt.%, respectively. The fluence amounted up to 5×10¹⁸ n cm⁻² within 15–18 fuel cycles, and about 5×10¹⁹ n cm⁻² for the whole period of operation. These values agree well with calculated data. DBTT was determined after irradiation (T_k) to evaluate the vessel metal state at the present moment, then after heat treatment at the temperature of 475°C to simulate the vessel metal state after thermal annealing (T_an), and after heat treatment at 560°C to simulate the metal state in the initial state (T_k0). As a result of the tests the following values were obtained: T_k, +91.5°C; T_an, +63°C; and T_k0, 54°C. The values of T_k and T_an obtained by measurements were found to be considerably lower than those predicted in accordance with the conservative method accepted in Russia (177°C for T_k and 100°C for T_an). Thus, the obtained results allowed to make a conclusion that it is not necessary to anneal Kozloduy NPP Unit 1 reactor vessel for the second time. The fractographic and electron-microscopic research allowed to draw some conclusions on the embrittlement mechanism.     

An analytical model for the analysis of BWR/4 long-term cooling with either intact or broken jet pump seals

An analytical model for the long-term emergency core cooling (ECC) of a boiling water nuclear reactor (BWR) has been developed. This one dimensional drift-flux model, is an extension of a previous study by Lahey and Kamath [1]. It considers both subcooled and bulk boiling in the core, allows the drift-flux parameters, C₀ and V_gj, to be functions of void fraction (α), and can accommodate both broken and intact jet pump seals. The results of this analytical model compare well with data from simulated full scale BWR fuel rod bundles, and experiments in the PCE facility at RPI.It has been found that the unlikely failure of jet pump seals can have a detrimental effect on the long term cooling capabilities of a BWR/4.     

Radiation damage of α-Al2O3 in the HVEM: II. Radiation damage at high temperature and high dose

The temperature dependence of the apparent displacement threshold energy has been measured for single crystal α-Al₂O₃ in a high voltage electron microscope, over the temperature range 320–1120 K. Below 570 K the threshold energy for observable damage remains constant at 390 ± 10 keV, then rises to a peak of 440 ± 20 keV at 695 K followed by a rapid fall to ~200 keV at 870 K and above. Measurement of the displacement threshold energy, using the optical absorption and luminescence of oxygen and aluminium electron irradiation induced vacancies at several temperatures, gives a value of 440 ± 25 keV for the oxygen ion and 175 ± 25 keV for the aluminium ion. These results show that oxygen and aluminium ions have widely different displacement energies, E_d, of ~75 and 18 eV respectively.