Behavior of zirconium fuel cladding under fast pressurization rates

Rapid pressurization test was carried out to evaluate the mechanical behavior of the zirconium cladding under a fast strain rate as well as a biaxial stress state for simulating an out-of-pile reactivity initiated accident (RIA) behavior. Influence of temperature, hydrogen content and alloying elements have been addressed in the conducted mechanical tests. The results showed that pressurization rates of 5.4 GPa/s at room temperature and 3.1 GPa/s at 350 °C were achieved. The corresponding time to failure was similar to expected power transient duration during a RIA. Maximum hoop stress of Zircaloy-4 at room temperature and 350 °C increased, respectively by 24.3 and 16.8% when compared to the conventional burst test results. Failure mode switched from a ductile ballooning to a brittle failure which leads to an axial split of the cladding when the hydrogen was added at a nominal value of 600 ppm. When the test temperature increased, its effect was diminished. Addition of an alloying element influenced the mechanical property differently. Niobium acted beneficially against hydrogen embrittlement in that it increased the ductility of the metal matrix.     

Investigation of the ductile fracture properties of type 304 stainless steel plate, welds, and 4-inch pipe

J-integral fracture toughness tests were performed on welded 304 stainless steel 2-inch plate and 4-inch diameter pipe. The 2-inch plate was welded using a hot-wire automatic gas tungsten arc process. This weldment was machined into 1T and 2T compact specimens for single specimen unloading compliance J-integral tests. The specimens were cut to measure the fracure toughness of the base metal, weld metal and the heat affected zone (HAZ). The tests were performed at 550°F, 300°F and room temperature. The results of the J-integral tests indicate that the J_Ic of the base plate ranged from 4400 to 6100 in lbs/in² at 550°F. The J_Ic values for the tests performed at 300°F and room temperature were beyond the measurement capacity of the specimens and appear to indicate that J_Ic was greater than 8000 in lb/in². The J-integral tests performed on the weld metal specimens indicate that the J_Ic values ranged from 930 to 2150 in lbs/in² at 550°F. The J_Ic values of the weld metal specimens tested at 300°F and room temperature were 2300 and 3000 in lbs/in² respectively. One HAZ specimen was tested at 550°F and found to have a J_Ic value of 2980 in lbs/in² which indicates that the HAZ is an average of the base metal and weld metal thoughness. These test results indicate that there is a significant reduction in the initiation fracture toughness as a result of welding.The second phase of this task dealt with the fracture toughness testing of 4-inch diameter 304 stainless steel pipes containing a gas tungsten arc weld. The pipes were tested at 550°F in four point bending. Three tests were performed, two with a through wall flaw growing circumferentially and the third pipe had a part through radial flaw in combination with the circumferential flaw. These tests were performed using unloading compliance and d.c. potential drop crack length estimate methods. The results of these test indicate that the presence of a complex crack (radial and circumferential) reduces in the initiation toughness and the tearing modulus of the pipe material compared to a pipe with only a circumferentially growing crack.     

Mechanical response of LMFBR and P1 cladding tubes under transient heating

Type 316 stainless steel tubing specimens comparable to LMFBR cladding were burst tested with relatively constant internal pressure in the 219–836 psi range and with increasing temperature. Continuous measurements of diameter change, temperature, and pressure were recorded as the samples were heated to temperatures near the melting point at rates from 10–1800°F/sec. The effects of varying initial wall thickness, cold work level, length, and thermal experience were explored. Ductile failures were observed at 10°F/sec, and stable strains at time of failure were greater than those reported by HEDL. At 200°F/sec the initially 20% cold-worked, 15-mil wall tubing produced brittle failures; while initially 40% cold worked, 10-mil wall samples displayed a mixture of ductile and brittle features. At 1000°F/sec the behavior of the latter material was prodominately brittle, although stable strains as large as 6% were observed. Failure temperatures were generally above 2000°F. When substantial ductility was displayed, an exponentially increasing stable strain was recorded as temperature and time progressed: from such curves temperatures corresponding to 1% strain were derived. Factors controlling the mechanical response appear to be separable by analysis based on the recorded data and the variety of materials and conditions of the tests.     

An evaluation of a general venting strategy in a CANDU 6 reactor building

If the reactor building sprays or local air coolers are not available, depressurization by reactor building venting is considered as a useful mitigation strategy for a severe accident management of the Wolsong plants. As the containment filtered vent system is not established in the Wolsong Units, the reactor building isolation system can be a substitute for reactor building venting. The D₂O vapour recovery system which has a 0.76 m (30 in.) diameter penetration is expected to meet the NRC requirements. To investigate the effectiveness of the Reactor Building Venting Strategy, three kinds of accidents are analyzed: a SBO, a Small LOCA and a Large LOCA. The reactor building pressure behavior was analyzed with the ISAAC computer code for four different cases: without venting, 379 kPa(g)/345 kPa(g) (55 psig/50 psig), 345 kPa(g)/276 kPa(g) (50 psig/40 psig) and 345 kPa(g)/207 kPa(g) (50 psig/30 psig) valve open/close pressures. When the reactor building spray or local air coolers can not be operated, a depressurization strategy by using the D₂O Vapour Recovery System could prevent a reactor building failure and reduce the amount of CsI released to the environment. The present study shows that the operation of valves at a pressure of 379 kPa(g)/345 kPa(g) (55 psig/50 psig) is safe and effective. Based on the current study, the strategy of reactor building venting is involved in severe accident management guidance-5.     

Temperature effect on IG-11 graphite wear performance

IG-11 graphite, used in the 10 MW high temperature gas-cooled test reactor (HTR-10), was tested under different temperatures on an SRV standard wear performance tester. The experiment temperatures were room temperature, 100, 200, 300 and 400 °C. According to the reactor structure, the experiments were designed to test graphite–graphite and graphite–stainless steel wear. The wear debris was collected, and the worn surfaces and debris were observed under scanning electronic microscope (SEM). It was found that there were different wear mechanisms at different temperatures. The main wear mechanism at room temperature was abrasive wear; at 200 °C, it was fatigue wear; at 400 °C, adhesive wear was observed. This difference was mainly due to the change of stress distribution at the contact area. The distribution of wear debris was also analyzed by EDX particle analysis software.     

Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m⁻¹ °C⁻¹ under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.     

Liquid metal embrittlement of T91 martensitic steel evidenced by small punch test

In this work, the sensitivity of liquid metal embrittlement of the T91 martensitic steel is investigated with the small punch test (SPT). The material was studied in three tempering conditions (as quenched, tempered at 500 and 750 °C), at 300 °C in air and in the liquid lead bismuth eutectic (LBE). The load–displacement curves (four stages, low maximum force and large displacement to fracture) obtained for one test condition of the 750 °C tempered material is in general very different from those of the two other materials. An effect of LBE has been observed for the as quenched and 500 °C tempered steels. For these materials, the curves tend to be linear with a reduced displacement to fracture suggesting a brittle behavior. This ductile to brittle transition induced by liquid metal has been confirmed from the fracture surface analysis where cleavage was observed. In comparison with conventional tensile tests, small punch tests appear to be more sensitive to evidence liquid metal embrittlement.     

III. PWR steam generator response to a scram at 50% load and an open grid transient (bugey-4 nuclear power plant)

Two transients, an open grid and a scram at 50% load, were conducted on unit 4 of the PWR power plant Bugey. The thermal hydraulic response of the steam generator was recorded. For the open grid test, the following observations are noted:No alarming phenomena are observed in the steam generator during the transient. Primary pressure oscillations were very mild, and did not exceed about 4.8 bar/min with a maximum amplitude of ±8 bar. This condition should not result in significant stress levels. Steam generator outer shell metal temperature gradients remained within very acceptable limits; a maximum amplitude of about +13°C and a rate not exceeding 0.8°C/min are obtained. This slow rate is explained by a fall in primary water temperature that allows for a temperature decrease inside the U-tube bundle. Similarly, the temperature rise on the tube sheet does not exceed an amplitude of 20°C with a rate of about 2°C/min. Again these conditions do not lead to any significant thermal shock on the tube sheet. The steam generator feed controls maintain the level within the normal operation range and the small addition of colder feedwater does not lead to great temperature changes because of the large mass of the recirculation water in the steam generator.For the scram at 50% load, the following observations are noted: no severe thermal or pressure transients are observed in this test. Fluid temperature fluctuations occur with rates not exceeding 1°C/s and a maximum amplitude of about 20°C in the downcomer and 10°C on the tube sheet. Steam generator outer shell temperature varies at a rate of about ±0.8°C/min with a maximum amplitude of about 16°C. These thermal transients should lead to thermally induced stresses of acceptable levels.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

The feasibility of cooling heavy-water reactors with supercritical fluids

It appears technically feasible to use supercritical carbon dioxide as a coolant for a CANDU-type reactor. A new supercritical loop is proposed in which the reactor is cooled by a single-phase fluid pumped in a high density liquid-like state. The supercritical fluid-cooled reactor has the advantage of gas-cooled reactors of avoiding dryout, and of liquid-cooled reactors of low coolant-circulation power. By eliminating dryout, the maximum operating temperature of the fuel sheath can be increased to 1021°F (550°C) for existing Canadian fuel bundles, with a coolant exit temperature of 855°F (458°C) producing steam comparable to that of conventional fossil-fuel plants. Since the reactor coolant exit temperature from the steam generator may be as high as 280°F (138°C) low-pressure steam may also be produced. A new dual-reheat cycle is proposed with an ideal overall plant efficiency of 33%, comparable to the Pickering generating station.     

The effects of heat treatments on the microstructure and mechanical properties of the Ti–2.19Al–2.35Zr alloy

Ti–2.19Al–2.35Zr alloy is one of the candidate materials for the steam generator tubing of an integrated reactor, System Modular Integrated Advanced ReacTor (SMART) being developed in Korea. In this study, the effects of heat treatments on the mechanical properties of Ti–2.19Al–2.35Zr alloy were evaluated. Mechanical tests were implemented to examine the effects of an annealing, cooling rate and re-annealing temperature/time on the mechanical properties of the alloy. The annealing temperatures ranged from 600 to 1050 °C and the cooling rates were controlled by introducing a water-quenching (WQ), air-cooling (AC), and furnace-cooling (FC). As for the re-annealing heat treatment, after a β water quenching, the re-annealing temperature was selected as 800 °C for the α-phase heat treatment and 940 °C for the α + β-phase heat treatment with various time intervals (1, 10 and 24 h). The results showed that an increase of the annealing temperature to above the β-region temperature induced an increase of the tensile strength and a decrease of the elongation in the 25 and 300 °C tests. A decrease of the cooling rates from water-quenching to a furnace-cooling revealed a decrease of the tensile strength and an increase of the elongation. Also an increase of the re-annealing time with different phase regimes exhibited a decrease of the strength and an increase of the elongation. These tendencies were more dominant in the 300 °C test rather than the room temperature test from the characteristics of the microstructures which were affected by the heat treatments.     

Boiling characteristics of dilute polymer solutions and implications for the suppression of vapor explosions

Quenching experiments of hot solid spheres in dilute aqueous solutions of polyethylene oxide polymer and surfactant have been conducted for the purpose of investigating the physical mechanisms of the suppression of vapor explosions in this polymer solutions. Two spheres of 22.2 and 9.5 mm-diameter were tested in the polymer solutions of various concentrations and pool temperatures from 30°C to its boiling point. The minimum film boiling temperature in 30°C liquid rapidly decreased from over 700°C for pure water to about 150°C as the polymer concentration was increased up to 300 ppm for a 22.2 mm sphere, and it decreased to 350°C for a 9.5 mm sphere. This trend is observed consistently in the heated pool up to its boiling temperature, while the tests with surfactant solutions do not show an appreciable reduction in the minimum film boiling temperature. The ability of suppression of vapor explosions by dilute polyethylene oxide solutions against an external trigger pressure was tested by dropping molten tin into the polymer solutions at 25°C. It was observed that in 50 ppm solutions more mass fragmented than in pure water, but it produced weaker explosion pressures. The explosion was completely suppressed in 300 ppm solutions with the external trigger. The debris size distributions of fine fragments smaller than 0.7 mm were shown to be almost identical regardless of the polymer concentrations.     

Moment modified compact tension specimen for measuring crack arrest toughness

Crack arrest toughness in reactor vessel steels in the transition and Charpy upper shelf energy temperature range are of particular interest to the nuclear industry to aid with the analysis of the phenomenon known as pressurized thermal shock (PTS). A test specimen and analysis technique have been developed to measure crack arrest toughness at temperatures from the transition region up to and beyond the Charpy upper shelf energy level. The moment modified compact tension (MMCT) specimen combines a thermal gradient with mechanical loadings to initiate a crack in brittle material below NDT and then have arrest take place in hot, ductile material. A finite element model was used to help design the specimen and fixturing geometry as well as calculate the arrest toughness. Tests have been conducted on ASME SA533 Grade B Class 1 steel plate with a variety of loadings confirming the veracity of the technique and developing valuable data. Crack arrest toughness has been measured from 0°F to 110°F (−18°C to 43°C). This work has been part of a research program performed by C-E, Windsor and funded by the Electric Power Research Institute.     

Fretting-wear of zirconium alloys

Fretting tests of Zircaloy fuel sheath bearing pads in contact with zirconium alloy (Zr–2.5Nb) pressure tube specimens were conducted at temperatures varying from 25 to 315 °C. The effects of motion type and amplitude, water chemistry, fuel sheath manufacturer and pressure tube surface finish were also investigated. The effect of temperature is the most significant. The pressure tube wear coefficient in the 225–286 °C for all four motions studied is considerably greater than that above 300 °C. The fretting rate for small amplitude motion representative of flow turbulence excitation is about equal at temperatures below 150 °C and above 300 °C, but is five to ten times greater in the 250–286 °C range.     

Cyclic deformation behaviour of AISI 321 austenitic steel and its characterization by means of HTC-SQUID

AISI 321 austenitic steel forms martensite due to quasi-static and cyclic loading. This presupposes the exceeding of the threshold value of cumulated plastic strain. The main aim is to determine the fatigue damage of austenitic steel by characterizing the martensitic structure with the help of the SQUID measuring technique. Several specimen batches were evaluated and thereby the load amplitudes and the test temperatures were varied (room temperature and 300°C). The experiments result in characteristic curves of the SQUID signals according to the fatigue damage which could be confirmed with comparative measurements with different methods, such as, e.g. ultrasonic absorption measurements. The extremely sensitive SQUID measuring technique allows also detection of information in specimens fatigued at a temperature of 300°C in which the phase fractions of strain-induced martensite are extraordinarily low.     

秦山核电厂安全壳系统B、C类密封性试验

叙述了秦山核电厂安全壳系统B、C类密封性能试验概况,主要包括试验范围、泄漏率分配、试验结果和总体评价等。     

Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 °C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 °C, respectively, at full power operation and at the scram from the operation.     

The bending and pressurizing of pipe bends with flanged tangents

Pipe bends with flanged tangents are analyzed, by linear thin shell theory, for the stresses and flexibilities by pure inplane bending and pressurization, assuming that they are not interactive. Results are presented for flexibilities, decay constants and stresses. Zero tangent lengths, that is, flanges directly on the bend, substantially reduce the stresses and bend deflections due to applied moments, but tangents longer than one pipe circumference effectively allow 90° bends to act as if they are unterminated. With pressurization, flanges on 90° bends only marginally reduce stresses and bend deflections compared to those for an unterminated bend.     

Experimental and computational investigation of LBE–water interaction in LIFUS 5 facility

The interaction between heavy liquid metal, such as LBE, and pressurized water has been analyzed at ENEA Brasimone Research Centre, in order to investigate the evolution of this phenomenon over a wide range of conditions. The study includes an experimental campaign on LIFUS 5 facility and a numerical simulation activity performed with SIMMER III code.The first test of the experimental program was carried out injecting water at 7 MPa and 235 °C in a reaction vessel containing LBE at 350 °C. A pressurization up to 8 MPa was observed in the test section during the short term (about 2 s) of the transient.In the post-test analysis performed with SIMMER III code, two different geometrical models were developed in order to reproduce in the best manner the experimental results and, therefore, to confirm the code's capabilities of reproducing the phenomenology of the LBE–water interaction.The data calculated through both models agreed in a good way with the experimental results, despite the necessary simplifications adopted in the models due to the 2D features of the code.