Primary Coolant Piping Research Articles

Link between b.c.c.-f.c.c. orientation relationship and austenite morphology in CF8M stainless steel.

Slow-cooled CF8M duplex stainless steel is used for critical parts of the primary coolant pipes of nuclear reactors. This steel can endure severe service conditions, but it tends to become more brittle upon very long-term aging (tens of years). Therefore, it is essential to understand its specific microstructure and temporal evolution. As revealed by electron backscatter diffraction (EBSD) analyses, the microstructure consists of millimetre-scale ferritic grains within which austenite lath packets have grown with preferred crystallographic orientations concerning the parent ferritic phase far from the ferrite grain boundaries. In these lath packets where the austenite phase is nucleated, the lath morphology and crystal orientation accommodate the two ferrite orientations. Globally, the Pitsch orientation relationship appears to display the best agreement with the experimental data compared with other classical relationships. The austenite lath packets are parallel plate-shaped laths, characterized by their normal n. A novel methodology is introduced to elucidate the expected relationship between n and the crystallographic orientation given the coarse interfaces, even though n is only partly known from the observation surface, in contrast to the 3D crystal orientations measured by EBSD. The distribution of retrieved normals n is shown to be concentrated over a set of discrete orientations. Assuming that the ferrite and austenite obey the Pitsch orientation relationship, the determined lath normals are close to an invariant direction of the parent phase given by the same orientation relationship.

Ultrasonic wave propagation analyses in centrifugally cast stainless steel using solidification grain structure models

Several components of nuclear power plants are made from cast austenitic stainless steel (CASS) because of its high corrosion resistance and strength. CASS that is centrifugally fabricated in a rotating mold is used in primary coolant piping. In-service inspection based on ultrasonic testing (UT) has to be conducted for primary coolant piping in accordance with fitness-for-service codes. However, a high-accuracy evaluation of flaws in CASS components through UT is difficult because the ultrasonic waves are scattered and attenuated by the coarse grains, and the ultrasonic beam is distorted by the anisotropy resulting from the grain orientations. Some works have attempted to improve UT capability for CASS components. For more improvement, it is essential to understand how ultrasonic waves propagate in the solidification structure of CASS. Numerical simulations are a useful and reasonable way for this purpose. The simulation model should account for a three-dimensional grain structure. In this study, we prepared three different models of solidification structures for centrifugally CASS, and we fed these structures into an explicit finite element model to simulate the propagation of ultrasonic waves. Afterward, we compared the simulated wave propagations with those measured by a laser Doppler vibrometer to determine an appropriate model that showed good agreement between the simulated and experimental results.

Comparison of grain structure models for wave propagation analysis in centrifugally cast stainless steel

Centrifugally cast stainless steel (CASS) is widely used in primary coolant piping of pressurized water reactor plants because of its high corrosion resistance and high strength. An in-service inspection based on ultrasonic testing (UT) has to be conducted for weld joints of primary coolant piping on the basis of JSME Rules on Fitness-forService for Nuclear Power Plants. However, it is difficult to detect and size flaws in CASS components with high accuracy because of the following reasons: Ultrasonic waves are scattered and attenuated due to coarse grains, and anisotropic and heterogeneous properties in CASS lead to ultrasonic beam skewing. Numerical simulations are useful and reasonable ways for better understanding the ultrasonic wave propagation behavior in CASS. To effectively achieve this, the simulation model should include a three-dimensional (3D) grain structure. In this study, we modeled three kinds of the solidification grain structures of centrifugally CASS. One is obtained by using a cellular automaton method, another consists of many hexagonal columns with the same dimensions, and the other is transversely isotropic material. Then these structures were fed into an explicit finite element model for simulating wave propagation and the simulated results were compared with those measured by a laser Doppler vibrometer. Through the comparison, we investigated the applicability of these three kinds of solidification grain structure models to simulation for wave propagation.

Effect of Sigma Phases on Moderate-Temperature Tensile Properties of Z3CN20.09M CASS Used for Primary Coolant Pipe of Nuclear Power Plant

The effect of sigma phases on the moderate-temperature tensile properties of Z3CN20.09M casting austenite stainless steel was investigated by means of isothermal treatment, scanning electron microscopy, transmission electron microscopy, instrumented nanoindentation, tensile testing, and finite element simulation. The results show that the yield strength and ultimate tensile strength of aged specimens tensile tested at moderate temperature increase remarkably with an increasing sigma phase, while the elongation at break decreased. The strain-hardening rate of aged specimens with sigma phases is higher than that of unaged specimens without sigma phases at a certain low-strain range. However, the value of the strain-hardening rate of aged specimens is lower than that of unaged specimens when the strain exceeds a certain value. The effect of the sigma phase on the tensile properties at moderate temperature is also more significant. This can be attributed to the many high-energy σ/γ2 and α/σ/γ2 incoherent phase boundaries caused by the precipitation of sigma phases. On the one hand, these boundaries hinder the movement of dislocations and subsequently accumulate dislocations to some extent, so strength is enhanced and the strain-hardening rate is improved. On the other hand, microcracks at these interfaces initiate and propagate more easily when the strain exceeds a certain value. Thus, the elongation value and the strain-hardening rate decrease.

Effect of Sigma Phases Precipitation on the Moderate Temperature Mechanical Properties of Fe20cr9ni Cass Used for Primary Coolant Pipe of Nuclear Power Plant

Effect of Sigma Phases Precipitation on the Moderate Temperature Mechanical Properties of Fe20cr9ni Cass Used for Primary Coolant Pipe of Nuclear Power Plant

Application of fracture evaluation method for aged elbow pipes

This paper evaluated a closed formed solutions of stress intensity factor K and J-integral for a flawed elbow of French Code, RSE-M, in order to introduce a flaw evaluation for aged elbows into the JSME Rules on Fitness-for-Service (JSME Rules), which have flaw evaluation procedures using K and J for only straight pipes and cylinder-shaped vessels in the current version. The stress distribution of an elbow is more complex than straight pipe and depends on not only flaw shape, flaw direction and radius/thickness but also bending radius, which may result in smaller allowable flaw size than that of a straight pipe. The authors performed K and J-integral calculation using FE analysis (FEA) focusing on an aged primary coolant pipe of PWR plant in Japan and compared with the solutions of the RSE-M code. As a result, it was confirmed that K and J solutions by the RSE-M code gave generally good result with conservativeness. Some cases of K showed large unconservative predictions by the RSE-M code and it was found that those were for the elbows with geometries out of the application range of the RSE-M code. The authors will develop a flaw evaluation for elbows with larger application range than that of the RSE-M code.

Calculation of dose rates in loss of coolant accident due to double ended rupture of the experimental tangential irradiation beam tube of MTR reactor.

A MTR reactor is an open water pool reactor type. The water pool serves as a shield from radioactive radiations. The most serious accident in this type of reactor is the Loss of Coolant Accident (LOCA) due to rupture either of a primary coolant pipe or of any experimental beam tube. In the present work it has been assumed that pool water drains out due to double ended rupture of the tangential irradiation beam tube (TIC) which has a diameter 150 mm. For an operating power level of 22 MW, the equilibrium core would enter into melting conditions if the pool drain time is less than one hour. It was also assumed that Emergency Core Cooling System (chimney water injection system and siphon effect breaker were not working. The reactor interiors have been modeled using the Monte Carlo N-Particle Transport code MCNPX 2.7.0. The source term has been determined using the ORIGEN-2 code. The doses and dose rates calculations in different places of operator (as phantom of Tissue-Equivalent Material) inside of the reactor building were determined by using Monte Carlo N-Particle Transport (MCNPX).The results show that the dose rate in the control room would be 5.24557 SV/h, the dose rate in the reactor hall above the pools on the gate would be7.20137 SV/h and the dose rate in the Emergency control room would be 2.68239 SV/h. Those dose rates are extremely high and would lead to fatal doses in short time.

Calculation of dose rates in loss of coolant accident due to double ended rupture of the experimental tangential irradiation beam tube of MTR reactor

The doses and dose rates following a LOCA in MTR reactor have been studied. A MTR reactor is an open pool type. The water pool serves as a shield from radioactive radiations.The most serious accident in this type of reactor is the Loss of Coolant Accident (LOCA) due to rupture either of a primary coolant pipe or of any experimental beam tube. In the present work it has been assumed that pool water drains out due to double ended rupture of the tangential irradiation beam tube (TIC) which has a diameter 150 mm. For an operating power level of 22 MW, the equilibrium core will enter into melting conditions if the pool drain time is less than 1 h.It was also assumed that Emergency Core Cooling System (chimney water injection system and siphon effect breaker) were not working. Therefore, conservatively a severe damage (∼80%) is expected to occur to the core, either by a core uncover situation following extended boiling operation, or by a core covered situation with extended boiling.The reactor interiors have been modeled using the Monte Carlo N-Particle Transport code MCNPX 2.7.0. The source term has been determined using the ORIGEN-2 code. The doses and dose rates calculations in different places of operator (as phantom of Tissue-Equivalent Material) inside of the reactor building were determined by using Monte Carlo N-Particle Transport (MCNPX). The results show that the dose rate in the control room is 5.24 Sv/h, the dose rate in the reactor hall above the pools on the gate is 7.20Sv/h and the dose rate in the Emergency control room is 2.68Sv/h. Those dose rates are extremely high and would lead to fatal doses in short time.

The Microstructure and Mechanical and Corrosion Behaviors of Thermally Aged Z3CN20-09M Cast Stainless Steel for Primary Coolant Pipes of Nuclear Power Plants

The effects of thermal aging time at 400 °C on the microstructure and mechanical and corrosion behaviors of Z3CN20.09M cast stainless steel were investigated; and the corresponding thermal aging mechanism was studied. It was revealed that the changes in mechanical properties after thermal aging were mainly caused by the iron-rich phase (α) and the chromium-rich phase (α’) produced by the amplitude-modulation decomposition of ferrite. A similar trend of thermoelectric potential during thermal aging was determined in relation to the Charpy impact energy. However, the corrosion resistance of Z3CN20.09M cast stainless steel deteriorates as thermal aging time increases. When the thermal aging is longer than 3000 h, the precipitation of G phase has a great influence on the corrosion resistance. The interfacial matching relationship between G phase and the surrounding ferrite was established by selected area electron diffraction of HRTEM. The relationship is of cube-on-cube phase boundary type. The impact fracture mechanisms in relation to thermal aging time were also studied and compared.

Tolerant fuel for VVER reactors

The main factor of destruction of fuel rods in accidents with loss of coolant is associated with the vapor-zirconium reaction occurring between the fuel rod shell and the coolant (water). Improving the reliability of fuel cells can be obtained by modifying or replacing the fuel shell, materials that do not interact with the coolant during normal operation and in emergencies. The loss of coolant accident is a design-basis accident in light water reactors. The postulated accident requires an analysis of the double guillotine break in a main primary coolant pipe, which allows the coolant to freely discharge out of the primary system into the containment of the building. In the present work, the heat stored in the active zone of the reactor, the heat capacity of the fuel and the shell temperatures are determined for ATF materials U-10Mo and U3Si2. The calculated values are compared with the usually used fuel UO2. The results showed beneficial effect of using ATF materials. The above quantities are decreased dramatically for ATF, promising to replace UO2 with ATF fuels. This reduces the probabilities of accidents and oxidation in the nuclear reactors. The calculations are done for nuclear power plant, type VVER-1200.

Long-term thermal aging performance of welding structures in primary coolant pipes for nuclear power plants

Welding structures of primary coolant pipes were subjected on a series of accelerated thermal aging test at 400 °C until 30,000 h to investigate the resulting microstructure, impact and nanoindentation properties change. The results indicated that after long-term thermal aging, a mottled structure and precipitated G-phases were observed in ferrite phases both of straight pipes and welded joints, leading to a decrease in impact energy and an increase in nanohardness. Moreover, until thermal aging saturation, impact energy of welding structures was still higher than initial design value (80 J), thus demonstrating a high toughness reserve, but saturated impact energy in welded joints was higher than that in straight pipes owing to lower ferrite content. Furthermore, there were higher-density spinodal decompositions and more precipitated G-phases in welded joints than those in straight pipes owing to micro-defects generated from welding process, resulting in a higher nanohardness in ferrite phases of welded joints.

Influence of the dissolved hydrogen concentration on the radioactive contamination of the primary loops of DOEL-4 PWR using the OSCAR code

Corrosion products are generated in the primary circuit during normal operation and are activated in the core. Those activated corrosion products, mainly 58Co and 60Co (coming respectively from the activation of 58Ni and 59Co), are then transported by the primary fluid and deposited on the out-of-flux surfaces (steam generators, primary coolant pipes…). To minimize this radioactive contamination, one needs to understand the behavior of corrosion products by carrying out measurements in PWRs and test loops combined with a reactor contamination assessment code named OSCAR. The aim of this article is to evaluate the influence of the change in the Dissolved Hydrogen (DH) concentration on the contamination of the primary loops of DOEL-4 PWR, a Belgian unit. After the description of the principle of the OSCAR V1.3 code, its use is illustrated with the simulation of DOEL-4. Finally, those calculations are compared to autoclave experiments called DUPLEX with thermodynamic and chemical conditions closed to those observed in PWRs. OSCAR V1.3 calculations show that an increase in the DH concentration results in a decrease in 58Co surface activities. These results are consistent with those from the DUPLEX experiments. Finally, an increase of the DH concentration is then recommended in operating PWRs to reduce the 58Co surface contamination.

Investigation of heterogeneous ratcheting of a GTAW welded joint for primary coolant piping

Abstract Uniaxial tensile and ratcheting tests were performed with a 3D DIC system to investigate the tensile strain evolution and heterogeneous ratcheting of each area in a gas tungsten arc welding (GTAW) welded joint. The results indicated that with increasing applied stress, the strain was initially concentrated in the base metal (BM) owing to its low yield stress, and then extended gradually to the heat affected zones (HAZ) and subsequently to the weld metal (WM) until the occurrence of quasi-shakedown. Furthermore, the most remarkable ratcheting strain and ratcheting rate were obtained in the BM but it manifested lower ratcheting deformation in the BM area of the GTAW welded joint as compared with the pure base metal.

Fatigue behavior of 316 L stainless steel weldment up to very-high–cycle fatigue regime

In nuclear power plant, 316 L weldments are widely used in fabricating the primary coolant pipe and in-core components, which usually suffer from very-high-cycle fatigue. Therefore, in this study, the high-cycle fatigue behaviors of 316 L stainless steel with an argon-arc welding joint were studied up to very high-cycle regime by using an ultrasonic fatigue testing system. Microscopic examination indicated that both the base steel and welding seam showed austenitic equiaxed grains. The grain size and micro-hardness values of the welding seam were comparable to those of the base 316 L steel. And heat affect zone was not evidently observed at the weld joint. Fatigue tests showed that fatigue failure still occurred in 107–109 cycles, indicating that the conventional fatigue limit does not exist for 316 L weldment. The fatigue strengths of 316 L weldments were close to those of the base 316 L steel. Scanning electron microscopic examinations revealed that multiple fatigue cracks initiated at specimen surface in the high cycle and very-high-cycle fatigue regime. A modified Murakami’s model was suggested to take into account the effect of surface roughness and Vickers hardness on fatigue life/strength of 316 L weldment. The fatigue strength predicted using the proposed model showed good agreement with the experimental results.

Microstructure and micropore formation in a centrifugally-cast duplex stainless steel via X-ray microtomography

Cast duplex stainless steels (CDSS) is being extensively used in many industries. The morphology and distribution of ferrite phase present, as well as the micropores defect, significantly influences the hot cracking susceptibility and corrosion resistance of CDSS. In this work, the microstructure and the micropores of the inner wall and outer wall position of a primary coolant pipe used in the pressurized water reactor (PWR) were characterized and compared using a lab scale X-ray microtomography approach. During centrifugal casting process, the local ferrite grains tend to grow in the same direction. Although the ferrite phases seem separated from each other in a two-dimensional (2D) image, most of them are connected in a three-dimensional (3D) view. Solution treatment can result in homogeneous distribution in both fraction and morphology of ferrite phase. Micropores always form along the ferrite/austenite interface and the inner wall position is more prone to porosity.

A comparison between fracture toughness at different locations of SMAW and GTAW welded joints of primary coolant piping

Two primary coolant pipes were narrow-gap multipass circumferentially butt welded by gas tungsten arc welding (GTAW) and shielded metal arc welding (SMAW) methods separately and then subjected to micro-hardness tests to distinguish the base metal (BM), heat affected zones (HAZs), fusion zones (FZs) and weld metal (WM). Subsequently, uniaxial tensile tests were performed with a 3D DIC system to investigate the strain evolution of each area in GTAW and SMAW welded joints and to compare respective tensile properties. The fracture toughness has been investigated at the above four different locations of the SMAW and GTAW welded joints. Then the 0.2 mm offset line method and the stretch zone width method have been both employed to determine the critical initial fracture toughness Ji. The results indicate that the fusion zones (FZs) have the worst fracture toughness compared with other locations over both two types of weld joints. In addition, the GTAW welded joints have a better comprehensive performance than the SMAW welded joints.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Effect of long-term aging on the fracture toughness of primary coolant piping material Z3CN20.09M

In this study, a series of accelerated thermal aging tests were performed on primary coolant piping materials (Z3CN20.09M) at 400 °C for 0, 2000, 5000 and 18,000 h to investigate the resulting microstructure change and fracture toughness. To further assess their fracture toughness, 0.2 mm offset line method and stretch zone width method were both employed to determine the fracture toughness. The results indicated that after long-term thermal aging, round shaped particles and characteristic morphology of spinodal decomposition were found in ferrite, thus resulting in the thermal aging embrittlement of Z3CN20.09M. Subsequently, the J-R curves and J-T curves all decreased significantly with the increasing thermal aging time. Furthermore, the fracture toughness parameters J50, JSZW(2D), JSZW(3D) and dJ/dΔa were first decreased rapidly and then slow saturated. In addition, the JSZW(2D) and JSZW(3D) values were significantly lower than JQ while the reduction of the critical crack size based on JSZW(2D) and JSZW(3D) parameters was about 30% compared with JQ, and thus JSZW was relatively close to the onset of crack initiation.

Fatigue crack growth of 316NG austenitic stainless steel welds at 325 °C

316NG austenitic stainless steel is a commonly-used material for primary coolant pipes of pressurized water reactor systems. These pipes are usually joined together by automated narrow gap welding process. In this study, welds were prepared by narrow gap welding on 316NG austenitic stainless steel pipes, and its microstructure of the welds was characterized. Then, fatigue crack growth tests were conducted at 325 °C. Precipitates enriched with Mn and Si were found in the fusion zone. The fatigue crack path was out of plane and secondary cracks initiated from the precipitate/matrix interface. A moderate acceleration of crack growth was also observed at 325°Cair and water (DO = ∼10 ppb) with f = 2 Hz.

Technical Note: Corrosion Fatigue Crack Growth of Forged Type 316NG Austenitic Stainless Steel in 325°C Water

Forged Type 316NG austenitic stainless steel is commonly used to fabricate primary coolant pipe in advanced pressurized water reactor systems, but it is subject to corrosion fatigue cracking in primary coolant environments. In this study, the fatigue crack growth of Type 316NG austenitic stainless steel in 325°C water was investigated. The corrosion fatigue effect in 325°C water was mainly correlated to the crack growth rate and load frequency. A maximum Fen (environment corrected factor) value up to 100 was observed at f = 0.005 Hz, ΔK = 13 MPa√m. The dissolved oxygen tended to have little influence on the fatigue crack growth. The crack path was deflected and branched. Most secondary cracks tended to be initiated from grain boundaries.