Calculation of the Hydride Cracking Velocity in Irradiated Fuel Elements of Light-Water Reactors

The velocity of delayed hydride cracking in irradiated É110 alloy (0.01% hydrogen) and zircalloy-2 (0.078% hydrogen) is predicted. It is shown that the velocity of cracking in VVÉR and RBMK fuel-element cladding is lower than in the stronger BWR fuel-element cladding. The maximum predicted velocity of cracking in É110 alloy is 2·10^–7 m/sec and is reached at 533 K. Below this temperature, the cracking process in both materials studied follows the Arrhenius law with activation energy 53–56 kJ/mole.     

Intermediate size pressure vessel programme (ZB 1)

One of the focal points in the discussion about the safety of nuclear power plants is the integrity of the reactor pressure vessel.In order to prove its integrity tests are in progress in an underground test facility of the main power station in Mannheim with an intermediate size vessel from the research programme “Integrity of Components”. Patches of A 533 B and modified A 508 B material were welded into the vessel ZB 1, the test temperatures are approximately 70 and 290°C. The main goal of the tests is to measure the behaviour of artificial and natural flaws during static hydrotests and simulated operational (cyclic) conditions.In the first half of the research programme the objective is to produce a crack growth of some centimetres by cyclic loading between a variable minimum pressure and a maximum pressure of about 24 MPa. The total number of load cycles will be approximately 30 000.In the second half of the tests the vessel will be loaded by a number of pressure cycles which correspond to the loading a reactor pressure vessel experiences during 40 years of operation.During the static and cyclic loading acoustic emission monitoring is being made by German and American laboratories.This paper presents details of the vessel, the test loop, results of the nondestructive examinations conducted to quantify the crack depths and results of the acoustic emission monitoring.     

Effect of pre-strain on creep of three AISI 316 austenitic stainless steels in relation to reheat cracking of weld-affected zones

Microstructural modifications induced by welding of 316 stainless steels and their effect on creep properties and relaxation crack propagation were examined. Cumulative strain due to multi-pass welding hardens the materials by increasing the dislocation density. Creep tests were conducted on three plates from different grades of 316 steel at 600 °C, with various carbon and nitrogen contents. These plates were tested both in the annealed condition and after warm rolling, which introduced pre-strain. It was found that the creep strain rate and ductility after warm rolling was reduced compared with the annealed condition. Moreover, all steels exhibited intergranular crack propagation during relaxation tests on Compact Tension specimens in the pre-strained state, but not in the annealed state. These results confirmed that the reheat cracking risk increases with both residual stress triaxiality and pre-strain. On the contrary, high solute content and strain-induced carbide precipitation, which are thought to increase reheat cracking risk of stabilised austenitic stainless steels did not appear as key parameters in reheat cracking of 316 stainless steels.     

Properties of heavy components of steel grade 91 and their welds

The use of the modified 9% Cr1% Mo steel grade (grade 91) for the manufacturing of heavy products (forgings, plates) for pressure vessels is evaluated on the following aspects: (i) possibility to produce heavy components (base material) with the required properties, (ii) conditions for welding without cracking, and (iii) availability of welding products and determination of their mechanical properties. Heavy plates with thickness ranging from 20 to 300 mm have been manufactured and tested. All mechanical properties (tensile, Charpy V notch impact, creep…) have been obtained in conformity with the requirements of ASTM specifications and literature data on thinner materials. Some laboratory tests have been carried out to determine the sensitivity of this material to cracking phenomena during welding (cold cracking) or during the postweld heat treatment (reheat cracking). They show that this material does not present major welding difficulties in comparison with more conventional materials used for the manufacture of pressure vessels. To obtain high toughness properties of the weld metal and low residual stresses, the postweld heat treatment temperature was found to be high (750 to 770°C). A significant modification of the weld metal deposit chemistry was found to be necessary to achieve the convenient level of impact properties, but correlatively, a decrease of its creep resistance may be expected for this very low Nb weld material.     

Decontamination of the reactor coolant pump in Maanshan nuclear power plant

To reduce the radiation dose that accumulated on the reactor coolant pump, decontamination work was carried out at the Maanshan Nuclear Power Plant. A four-step alkaline permanganate (AP)-CanDecon process was applied to remove the activity on the turning vane diffuser and pump impeller. The first step consisted of 8 h of AP treatment and 7 h of decontamination. It was followed by 2.5 h of AP treatment and 5 h of decontamination. An average decontamination factor of 2.9 was obtained. To understand the corrosion of the decontaminating reagents on the materials, coupons were installed in the decontamination tank. These were as-received and sensitized 304SS, alloy 600, casting stainless steel (CF-8), stellite-6, and carbon steels (A508 and A533). The exposure rates (mR h⁻¹) of the carbon steels were approximately five times higher in magnitude than those of the other materials. The decontamination levels (dpm per 100 cm²) of the A508 and A533 carbon steels were 5432 and 3701 respectively, while most of the rest of the materials were below the low limit of detection. Apparently, the corrosion product on the materials was a major factor in sustaining the exposure rate and the contamination level. The corrosion rate of the materials was also examined and compared with published data. An examination of the surface morphology of the materials after decontamination showed intergranular attack on the 304 stainless steel.     

Fuel rod deformation code FEMAXI-II and its application

This paper describes an axisymmetric, two-dimensional finite element method programme, FEMAXI-II, for fuel rod deformation analysis, and its application to the instrumented fuel assembly, IFA-508, in an HBWR. FEMAXI, the old version, was developed from Yamada's finite element programme by applying a pellet-cladding contact model and a pellet-cracking model and temperature distribution subroutine. Using the “softened” plastic-creep stress-strain matrix and the pseudo-stresses introduced by Cyr and Teter, we improved FEMAXI to FEMAXI-II in order to follow the changes of creep rate and yielding stress affected by the temperature increase. In the application of FEMAXI-II to IFA-508 experiments, the prediction of this programme agrees with results of the in-pile diameter measurements. The Cyr creep model predicts well that the ridge deformation is smaller in the thick cladding than in the thin one.     

Analysis of pressure vessel steel fatigue tests in air

Fatigue data on A533B and A508 steels tested in air at laboratories all over the world are analyzed by the FATigue Data Analysis Code, FATDAC. A model including an R-ratio factor is fitted to the data, and the quality of the fit is verified by several sets of data that were not used for fitting.     

Peridynamic modeling of concrete structures

The peridynamic model described in Silling (Silling, S.A., 1998. Reformation of Elasticity Theory for Discontinuous and Long-Range Forces, SAND98-2176. Sandia National Laboratories, Albuquerque, NM), being a central-force model, is limited to modeling materials with a Poisson's ratio of 1/4. In this paper, the peridynamic model is generalized by adding pairwise peridynamic moments to simulate linear elastic materials with varying Poisson's ratios. The new model is called the “micropolar peridynamic model”. The micropolar peridynamic model is placed within a finite element context to enable efficacious application of boundary conditions and efficient computational solutions using an implicit, rather than an explicit solution algorithm. The implicit solution algorithm is suitable for quasistatic simulation of damage and cracking in concrete structures. With this new model, very simple tensile damage mechanisms at the micro structural (peridynamic) level are sufficient to explain a great deal of the microcracking (damage) and fracture mechanics observed in concrete structures. The new implementation appears to be computationally efficient.     

A simplified model for concrete at low confining pressures

The paper describes a model for the response of concrete that is subjected to essentially monotonic straining at low confining pressures. We assume that, under these conditions, the response of the concrete is dominated by cracking when the stress state is predominantly tensile, and by gross inelastic deformation under compressive stress. The model uses a “crack detection surface” in stress space to determine when cracking takes place and the orientation of the cracking at a point, together with a damaged elasticity approach to describe the post-failure behavior of the concrete with open cracks. A yield/flow surface (associated flow) model is used to define the concrete's response in compressive states of stress. The model is simple enough that it can be implemented so as to operate effectively in an implicit finite element code: modeling accuracy is sacrificed for this purpose. Preliminary studies with the model indicate that it can give useful predictions in cases of interest.     

Welding consumables for nuclear power plants

In nuclear power plants, submerged arc welding and covered arc welding have long been employed especially for main weld seams, including the core region of RPV.This paper investigates the mechanical properties of several welding consumables we have developed for industrial plants — that is, welding consumables which lower the phosphorus and copper content of the welded metal, those for plates possessing particularly high tensile strength and those for the narrow gap welding method.Recent data derived from irradiation embrittlement tests show that these welded metals using a non-copper coating are highly effective in minimizing shifts in the transition curve.Welding consumables for A533B C1.2, A543 C1.1 or A508 C1.4 steels have a higher tensile strength than those for A533B C1.1 or A508 C1.3.We have developed submerged arc and covered arc welding consumables to be used with these kinds of steels, and it was confirmed that these consumables possess excellent tensile strength and notch toughness.Our tests also confirmed that the narrow gap SAW and MIG welds are more efficient than the conventional ones. Moreover, the mechanical properties of the welded metals are also excellent.     

Characterization of industrial NbTi strands at variable field for JT-60SA toroidal field coils

The mission of the JT-60SA Tokamak, to be built in Japan, is to contribute to the early realization of fusion energy by its exploitation in support of the ITER program. JT-60SA project is an important part of the “broader approach” activity as a satellite program for ITER. The toroidal field (TF) coils are a European “in kind” contribution and they will partly be built by France. JT-60SA TF coil uses the Cable In Conduit Conductor (CICC) with NbTi superconductor strands. TF conductors will have to operate at 5.7 T, 5 K and at current density of 450 A/mm² with sufficient margins. In the framework of JT-60SA TF coil manufacture, the variable temperature characterization is an important step to select NbTi strand. At an early stage of design, we had to choose the strand with acceptable performances. During the design qualification and validation stage, it is important to qualify strands in conditions close to the operation conditions. The industry has proposed various strands manufactured with different processes. This work and publication examines a strand with an internal CuNi barrier, which is expected to lead to better current distribution between strands, by more precise calibration and control of the inter-strand resistance. The strands were tested at the Grenoble High Magnetic Field Laboratory facility. The domain (B, T, J) explored was in the range of 4.5–11 T for the magnetic field intensity, 4.2–6.5 K for the temperature and between 40 A/mm² and 1200 A/mm² for the current density. For each strand, “critical current density” and “current sharing temperature” measurements have been carried out, with a temperature precision of few tens of mK. Once the measurements performed, the fitting parameters (of type J_C = f(B, T)) of each strand have been found, by performing regression analysis. This work will lead to select the strand with the best characteristics. In this paper, we present the results of this measurement task, the data and regression analysis (fits, T_cs, etc.) and the conclusion about the strand choice.     

Current status of scientific activity in Russia on high level radioactive waste transmutation and use of weapon grade and power plutonium in subcritical systems driven by proton accelerator

Features of neutron fuel cycles with the accelerator-driven system (ADS) as well as fission product and actinide transmutation in the ADS are analyzed in this paper: fuel type, fertile materials, neutron consumption, secondary radioactivity, change in radiotoxicity of actinides. The use of weapon-grade and power plutonium in the ADS is also considered. Information on various design versions of the ADS blanket including study of the sectioned blanket with neutron valves, its performance and R&D programme, including a neutron source driven by the 56.MeV “Istra” proton linac, is given.     

Strain-induced corrosion cracking of low-alloy steels in LWR systems — case histories and identification of conditions leading to susceptibility

The term “strain-induced corrosion cracking” (SICC) is introduced to describe crack formation involving dynamic straining, but in the absence of obvious, cyclic loading. Its origins in slow-strain-rate testing and in corrosion failures in boiler systems are described and the links with “classical” stress corrosion cracking and low-cycle corrosion fatigue are identified. Four areas, in which SICC of low-alloy steels in LWR systems has occurred, are described in detail and the typical features are used, together with literature data from laboratory testing, to identify conditions leading to susceptibility. Indications are given of remedial measures and of areas in which further work is necessary.     

Irradiation effects on fracture toughness of four nuclear reactor pressure vessel submerged-arc welds

This study applies statistical analyses to fracture toughness results for four irradiated “current practice” submerged-arc welds and an A533 grade B class 1 plate. Charpy V-notch, tensile, and 25 mm thick compact specimens were irradiated at 288°C to neutron fluences of 0.7 to 2.0 × 10²³ neutrons/m² (>1 MeV). The plate material contained 0.14% Cu and 0.67% Ni. The four submerged-arc welds contained 0.04 to 0.12% Cu and 0.10 to 0.63% Ni. The plate material showed a Charpy V-notch impact transition temperature increase of 68°C, and a Charpy V-notch upper-shelf energy drop of 16%. The four submerged-arc welds showed smaller changes than the plate material did. The fracture toughness results from the 25 mm thick compact specimens showed approximately the same temperature shift as the Charpy V-notch results. The results imply that submerged-arc welds with both low-copper and low-nickel contents can exhibit essentially zero radiation embrittlement and that nickel can contribute to radiation embrittlement even when the copper content is low.     

The development of a diagnostic and monitoring system for vibrating pipe branches

This paper deals with a diagnostic and monitoring system for assessing the integrity of pipe branches, during the operation of the nuclear power plant. This system have been developed under the concept of “easy to use without any sophisticated analysis” and “portable”. The accuracy of the diagnosis is based on the model optimization subsystem, which automatically modifies the numerical vibration model so as to fit its natural frequency to the actual natural frequency. The information obtained by this system may be reflected to a maintenance program of the plant to assure more reliable operation of the plant.     

Combined effects of neutron irradiation and hydrogen absorption on tensile properties and fracture mode of steels for nuclear pressure vessel

The combined effect of neutron irradiation and hydrogen absorption on the mechanical properties and fracture modes are investigated for the reactor pressure vessel steels in use and pure iron. The embrittlement of the steel A533B neutron-irradiated up to 4.4 × 10¹⁹n cm⁻² (E > 1 MeV) is not accelerated by the interaction between irradiation-induced defects and hydrogen atoms. For the steel A542 and pure iron, the embrittlement due to the combined effects of the above two factors is a little larger than that of A533B, and the ductility behaviour of these materials is quite similar to that of the unirradiated hydrogen-charged specimens. The change of the tensile properties and fracture modes can reasonably be explained by the interaction between irradiation defects and hydrogen atoms. Sensitivity to the brittle fracture of the steel in use, i.e. A533B, may not be increased by the combination of the above two embrittlement factors.     

Corrosion fatigue behavior of low alloy steels under simulated BWR coolant conditions

The corrosion fatigue crack growth behavior of A533 and A508 low alloy steels under simulated boiling water reactor (BWR) coolant conditions was studied. Corrosion fatigue crack growth rates of A533B3 and A508 cl. 3 steels were significantly affected by the steel sulfur content, loading frequency and dissolved oxygen content of water environments. The data points outside the bound of Eason’s model could be attributed to the low frequency, higher steel sulfur content and high dissolved oxygen in water environments. The sulfur dissolved in the water environment from the higher-sulfur steels was sufficiently concentrated to acidify the crack tip chemistry even in the hydrogen water chemistry (HWC). Therefore, nitrogenated or HWC water showed little or no beneficiary effect on the high-sulfur steels. For the steel specimens of the same sulfur level, their corrosion fatigue crack growth rates were comparable in different orientations, which could be related to the exposure of fresh sulfides to the water environment. The percentages of sulfides per unit area, by quantitative metallography, were comparable for the steel specimens of both orientations. When the steel sulfur content was decreased to a critical sulfur content 0.005 wt.%, the crack growth rates decreased remarkably.     

ANL survey of LWR containment penetrations: A progress report

Argonne National Laboratory is currently working on specific tasks in a containment penetration integrity program funded by NRC and managed by the Sandia National Laboratories. The first of these tasks is called “Characterization of Existing Penetration Designs”. The objective of this task is to identify those penetrations in nuclear reactor containments which, because of historical data or expected behavior under accident loads, are believed to have a relatively high probability of developing leakage when subjected to temperatures and pressures well beyond the containment design basis values. The program focuses on large and operating penetrations — such as personnel airlocks, equipment hatches, and bellows seals — and excludes electrical penetration assemblies and valve penetrations. (Sandia is working on electrical penetrations and EG&G is studying valve penetration assemblies.) This task will determine which penetrations require detailed study to determine leakage characteristics, and will identify which types of penetrations may require specific model and/or large-scale testing to obtain such characteristics. The survey is concentrating on containments built primarily between 1970 and 1982, and includes a comprehensive sample involving not only all types of containment types and materials, but also includes work performed by a large number of A-E design firms. The survey includes a good sample of containment penetration fabrication vendors. About 40 containments have been completed mid-August 1984.     

The use of energy absorbers to protect structures against impact loading

When a flying missible impacts a fixed structure, the interface loading is dependent on the deformation characteristics of both impacting and impacted bodies. If both are too rigid to accommodate the amount of gross deformation required to neutralize the incoming kinetic energy, or if such energy absorption has a chance to proceed in uncontrolled and unreliable ways, then there is a need to interpose a specifically designed “energy absorber” between missile and structure, from which a well-defined load time history can be derived during the course of impact.

The required characteristics of such an energy absorption material are:

• the capability to accommodate large permanent deformation without structural failure; and

• the reliable and controlled “load-deformation” (or “stress-strain”) behaviour under dynamic conditions, with an aim at an optimal square shape curve.

Consideration must also be given to environmental or other disturbing effects, like temperature, humidity, and “out of plane” loading. A short survey is presented of the wide range of energy absorbers already described in technical papers or used in a number of practical safety applications within varied engineering fields (from vehicle crash barriers to high energy pipe whipping restraints). However, with such open a literature, information is usually lacking in the specific data required for design analysis.

The following “energy absorption” materials and processes have thus been further experimentally investigated, with an a aim at pipe whipping restraint application for nuclear power plants:

1. (1) plastic extension of austenitic stainless steel rods;

2. (2) plastic compression of copper bumpers; and

3. (3) punching of lightweight concrete structures.

Dynamic “stress-strain” characteristics have been established for stainless steel bars at several temperatures under representative loading conditions. For this purpose, a test rig has been specifically designed to incorporate a number of adjustable parameters and to behave as a representative “slice” of an actual pipe whipping restraint; typical strain rates are in the 10 sec⁻¹ range. The behaviour of copper bumpers has been compared under static and dynamic conditions (using a conventional drop weight test (DWT) machine); as no significant strain rate effects were emphasized, only static tests have been further developed. The DWT rig was used again to investigate crushing or punching of cellular concrete under varying geometries and loading conditions. To remedy certain deficiencies of the regular commercial grades of cellular concrete, special lightweight mixtures have been studied to optimize material toughness and provide a wider range of specific resistance.Results of this experimental program are presented and discussed. The use of energy absorbers is then illustrated for a few typical pipe whipping restraints. The design of restraints is based on real dynamic characteristics of “energy absorption” material as produced by the test program. To derive design loads of restraints, a number of methods can be used ranging from a simplified “energy balance” graph to sophisticated plastodynamic computer analysis. Typical results are presented and discussed to compare the efficiency of these alternative methods.     

2D simulation of the initiation and propagation of crack array under thermal fatigue

Various components of nuclear reactors are submitted to various thermo-mechanical loadings. Thermal fatigue cracking has been clearly detected in reactor heat removal system (RHRS) of pressurized water reactors (PWRs). The present study focuses on AISI 304 L stainless steel used in PWRs. The thermal fatigue behavior of this steel has been investigated using a specific thermal fatigue facility called “SPLASH”. This test equipment allows the reproduction of multiple crack networks similar to those detected during component inspections. The present study deals with the modeling of crack networks initiation and propagation. It is structured in two parts: (i) experimental details and main characteristics of the cracks networks, and (ii) numerical simulation of multiple cracks initiation and growth problem, using an elastic–plastic thermal–mechanical computation and a generalized Paris’ law. The model presented in this study gives predictions in a good agreement with observations, as far as the evolution of the mean and deepest cracks during cycling is concerned.