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.     

Mössbauer study and structural characterization of UO2–Gd2O3 sintered compounds

Samples of UO₂and up to 10 wt% of Gd₂O₃ were prepared by solid-state reaction under a reducing atmosphere, in a thermal path comprising ramps and dwell times in the temperature range of 900–1750 °C. The sintered material was analyzed by X-ray diffraction and ¹⁵⁵Gd Mössbauer spectroscopy. The results showed that for samples annealed up to 900 °C, the gadolinium sesquioxide remained unreacted. However, when the temperature was increased to 1300 °C, a solid-state reaction took place forming mixed oxides. For the more severe sintering condition, at 1750 °C, gadolinia left urania partially unreacted producing a material consisting of two compositions, UO₂ (with no dissolved gadolinium) and (U, Gd)O₂. The proposed heating cycle provided pellets free from Gd₂O₃ phase and may be used by the nuclear fuel industry as a suitable sintering process.     

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.     

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state (T_ko) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the study.Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C.The estimated value of T_ko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel.The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T_f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T_ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T_ko = 50 °C.The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.     

Prediction of minimum UO2 particle size based on thermal stress initiated fracture model

An analytic study was employed to determine the minimum UO₂ particle size that could survive fragmentation induced by thermal stresses in a UO₂---Na fuel-coolant interaction (FCI), based on a brittle fracture mechanics approach. Solid and liquid UO₂ droplets were considered, with perfect wetting by the sodium or finite heat transfer coefficient. The analysis indicated that particles below the range of 50 μm in radius could survive an FCI under the most severe temperature conditions without thermal stress fragmentation, and seemed to verify the experimental observations as to the range of the minimum particle size due to thermal stress fragmentation by FCI. The basic complexities in fracture mechanics make further investigation in this area interesting but not necessarily fruitful for the immediate future.     

Stress waves in elastic rods induced by radiation heating

Consider a stress-free, thin, elastic rod, freely suspended. At time t = 0, the rod is rapidly exposed to a high temperature tube field inside, uniform over the cross-section T(x, t), t 0. Such a case arises in an external target when bombarded with a fast extracted proton beam of high energy and high intensity. In contrast to the quasi-static solution σ_x ≡ 0, stress waves are created at both free end-faces, at time t = 0, which propagate into the rod. Closed-form solutions are presented for a fast rise of a homogeneous temperature, during rise time t₀ to the final value T₀. Maximum compressive and tensile stresses will occur which may easily lead to failure of the rod. In particular, for an extremely fast rise t₀ < t_m, t_m being the characteristic mechanical time, |σ_x|max = EαT₀, which is the absolute value of a compressive stress in a rod fixed at the ends. For t₀ > t_m, |σ_x|max = EαT₀t_m/t₀, i.e. the maximum stresses are proportional to the length of the rod and inversely proportional to the rise time.     

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.     

An experimental investigation on thermal striping: Mixing phenomena of a vertical non-buoyant jet with two adjacent buoyant jets as measured by ultrasound Doppler velocimetry

An experimental investigation on the thermal mixing phenomena of three quasi-planar vertical jets, with the central jet at a lower relative temperature than the two adjacent jets, was conducted. The central jet was unheated (‘cold’), while the two adjacent jets were heated (‘hot’). The temperature difference and velocity ratio between the heated (h) and unheated (c) jets were, ΔT_hc=5°C, 10°C and r=V_cold,exit/V_hot,exit=1.0 (isovelocity), 0.7, 0.5 (non-isovelocity) respectively. The typical Reynolds number was Re_D=1.8×10⁴, where D is the hydraulic diameter of the exit nozzle. Velocity measurement of a reference single-jet and triple-jet arrangement were taken by ultrasound Doppler velocimetry (UDV) while temperature data were taken by a vertically traversed thermocouple array. Our UDV data revealed that, beyond the exit region, our single-jet data behaved in the classic manner. In contrast, the triple-jet exhibited, for example, up to 20 times the root-mean-square velocity values of the single-jet, especially in the regions in-between the cold and hot jets. In particular, for the isovelocity case (V_exit=0.5 m/s) with ΔT_hc=5°C, we found that the convective mixing predominantly takes place at axial distances, z/D=2.0–4.5, over a spanwise width, x/D|2.25|, centered about the cold jet. An estimate of the turbulent heat flux distribution semi-quantitatively substantiated our results. As for the non-isovelocity case, temperature data showed a localized asymmetry that subsequently delayed the onset of mixing. Convective mixing however, did occur and yielded higher post-mixing temperatures in comparison to the isovelocity case.     

Fuel elements of the high temperature pebble bed reactor

This work is devoted to spherical fuel elements for the high temperature pebble bed reactor, their manufacture and the conditions which they must satisfy for use in a process-heat reactor with an average gas outlet temperature T_{G, out} of 950°C. The positive results known from the operation of the AVR with T_G,out = 950°C and from extensive irradiation tests of the THTR-300 element with BISO coated mixed-oxide particles, even beyond the range of design specifications, and possible damage mechanisms are described in detail. They show that a spherical fuel element already exists, for which only a short-term development is needed to produce a coolant temperature of 950°C in a process-heat reactor. Further developments will be characterized by the use of a pebble bed HTR for high conversion rates (c ≈ 0.95) or for average gas outlet temperatures of more than 950°C. At higher temperatures the increased demands, mainly with regard to the release of fission products, can be fulfilled through the application of TRISO-coated fuel particles and the doping of the fuel kernels with . The reprocessing programme for fuel elements in the Federal Republic of Germany is mentioned briefly.     

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.     

LWR fuel rod behavior during severe accidents

Chemical interactions between UO₂ fuel and Zircaloy cladding up to 2350°C are described. UO₂/Zircaloy single effects tests have been performed with short LWR fuel rod segments in inert gas and under oxidizing conditions. The reaction kinetics of molten Zircaloy cladding with solid UO₂ fuel has been investigated with UO₂ crucibles containing molten Zircaloy. The UO₂/Zircaloy reactions obey parabolic rate laws. The oxygen uptake by solid Zircaloy due to chemical interaction with UO₂ occurs nearly as quickly as that from the reaction with steam or oxygen.To study the competing effects of the external and internal cladding oxidation under realistic boundary conditions and the influence of the uncontrolled temperature escalation due to the exothermic steam/Zircaloy reaction on the maximum cladding temperature, single rod and bundle experiments have been performed. Electrically heated fuel rod simulators, including absorber rod material (Ag, In, Cd alloy), guide tubes and grid spacers are used. The maximum measured cladding temperature during the temperature escalation was about 2200°C. The failure temperature of the absorber rods and the extent of bundle damage depends on the guide tube material (Zircaloy or stainless steel) and varies between 1200 and 1350°C. The molten materials and liquid reaction products can relocate and form large coherent lumps on solidification, which may result in complete blockage of the fuel rod bundle cross section. In the future, 7 × 7 bundle experiments of 2 m overall length will be performed in the new CORA facility to study, in addition, the influence of quenching on fuel rod integrity.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

Non-stationary outflow and rarefaction waves in flashing liquid

The outflow of high pressure liquid (in particular, water) to the atmosphere from a closed tube (of length a few metres and diameter more than a few centimetres) because of sudden destruction of one bottom is theoretically investigated. Evaporation takes places on the nucleus bubbles. The number of nuclei depends on the quality of the liquid or its purification. The process involves flashing evaporation of the liquid.There are two rarefaction waves at the initial stage. The velocity of the first wave (elastic forerunner) is sound speed in the one phase liquid and equals about 1000 m s⁻¹. After the elastic forerunner the liquid becomes superheated because the pressure drops and evaporation begins.The velocity of the second rarefaction wave is about 1–10 ms s⁻¹. There is intensive bubbly evaporation on and after the second wave. Intensity of the outflow is determined by the intensity of evaporation on the interface of the bubbles and by intensity of fragmentation of the bubbles because of their relative slip velocity in the liquid (0.1–1 m s⁻¹). The fragmentation of the bubbles significantly intensifies the evaporation because of augmentation of the bubbly interface.The degree of non-equilibrium or superheating behind the forerunner in water grows with the increasing initial temperature T₀. For T₀<530−540 K this superheating is negligible and the process may be described by an equilibrium scheme. For T₀ above 0.95T_cr≈605 K homogeneous nucleation is possible.After forerunner reflection from the closed bottom, intense evaporation is initiated near the bottom. Then the equalization of the pressure along the tube occurs (quasi-static homobaric stage).There is good correlation with experimental data.     

Fracture mechanical properties and neutron irradiation effects of fuel compacts for the HTTR

To simulate the nuclear fuel for High Temperature Engineering Testing Reactor (HTTR), fuel compact models using SiC-kernel coated particles instead of UO₂-kernel coated particles were prepared under the same conditions as those for the real fuel compact. The mechanical and fracture mechanics properties were studied at room temperature. The thermal shock resistance and fracture toughness for thermal stresses of the fuel compact were experimentally assessed by means of arc discharge heating applied at a central area of the disk specimens. These model specimens were then neutron irradiated in the Japan Material Testing Reactor (JMTR) for fluences up to 1.7 × 10²¹n/cm² (E ·> 29 fJ) at 900°C ± 50°C. The effects of irradiation on a series of fracture mechanical properties were evaluated and compared with the cases of graphite IG-110 used as the core materials in the HTTR.     

Corium kinematic viscosity measurement

Viscosity is one of the important properties which along with thermal conductivity and coefficient of volumetric expansion, determines convection in molten corium in the case of unfavourable progress of a severe accident in nuclear reactors. The viscosimetric attachment has been designed, developed and built for the purpose of measurement of corium kinematic viscosity with the technique of torsional oscillations. The created experimental devices and procedure for measurement of liquid corium viscosity have been tested and refined over temperature range up to 2800°C. The experimental data of 62UO₂+38ZrO₂ (mol%) melt viscosity have been obtained at temperature ranging from 2600 to 2800°C. The viscosity of 62UO₂+8ZrO₂+30Zr (mol%) melt has been measured at temperatures from 2400 to 2700°C. This work has been performed within OECD RASPLAV Project. The Management Board approves publication of this paper.     

Developing uranium dicarbide-graphite porous materials for the SPES project

Uranium carbide dispersed in graphite was produced under vacuum by means of carbothermic reduction of different uranium oxides (UO₂, U₃O₈ and UO₃), using graphite as the source of carbon. The thermal process was monitored by mass spectrometry and the gas evolution confirmed the reduction of the U₃O₈ and UO₃ oxides to UO₂ before the carbothermic reaction, that started to occur at T > 1000 °C. XRD analysis confirmed the formation of α-UC₂ and of a minor amount of UC. The morphology of the produced uranium carbide was not affected by the oxides employed as the source of uranium.     

Performance of the model fuel pin of the very high-temperature gas-cooled reactor at temperatures above 2000°C

Failure of unirradiated Triso-coated UO₂ particles in the model fuel pin of the very high-temperature gas-cooled reactor has been studied at temperatures exceeding 2000°C. It has been found that the thermal stability of the particles is greatly reinforced by the presence of the graphite sleeve outside the fuel compact in which they are dispersed. The results have been analyzed by a simple mechanistic model which takes into account both the pressure within the particle and the thermal degradation of SiC layer. The model has been further applied to predict the behavior of the irradiated particles. Comparison of the model prediction with the literature data on the irradiated Triso-coated UO₂ particles has shown a reasonable agreement.     

Processing of uranium oxide and silicon carbide based fuel using polymer infiltration and pyrolysis

Ceramic composite pellets consisting of uranium oxide, UO₂, contained within a silicon carbide matrix, were fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, particles of depleted uranium oxide, in the form of U₃O₈, were dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900 °C under a continuous flow of ultra high purity argon. The pyrolysis of AHPCS, at these temperatures, produced near-stoichiometric amorphous silicon carbide (a-SiC). Multiple polymer infiltration and pyrolysis (PIP) cycles were performed to minimize open porosity and densify the silicon carbide matrix. Analytical characterization was conducted to investigate chemical interaction between U₃O₈ and SiC. It was observed that U₃O₈ reacted with AHPCS during the very first pyrolysis cycle, and was converted to UO₂. As a result, final composition of the material consisted of UO₂ particles contained in an a-SiC matrix. The physical and mechanical properties were also quantified. It is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix.     

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.     

The fracture toughness measured on sensitized 304 stainless steel in simulated reactor water

The environmental conditions chemically equivalent to BWR primary water, e.g. 288°C, 0.2 ppm O₂ and/or 98°C, air-saturated, were found to influence considerably the in-water fracture toughness values of furnace-sensitized Type 304 stainless steel.Notched compact tension and three point bend specimens sampled from two heats of standard materials (0.06% C) showed significant reduction in dJ/da values reflecting consistently the effects of loading rate, temperature, dissolved oxygen concentration and degree of sensitization. In particular the crack enhancement with lowering the loading rate was significant. The effect became apparent with dJ dt at and below 1× 10−1 kg·mm/mm²/min (1.6 × 10 J/m²/s) in the typical BWR environment.Based on the results, it is suggested that a critical consideration is needed on the significance of such an environmental effect in the LWR structural safety evaluation, in particular that the probability of instable fracture at the “rings” of sensitized material near welded joints is subject to reviewing.