Assessment of the Fatigue Life of a PWR Surge Line Including the Effect of Coolant Environment

In Germany, the integrity concept is applied to important piping systems in most of the nuclear power plants. Regarding the framework of this concept, those damage mechanisms that cannot be controlled by analysis have to be excluded using appropriate measures. In most of the cases, these damage mechanisms are a result of local effects (like loads, medium, material characteristics) that cannot be determined exactly in advance and thus cannot be controlled by analysis, reliably. Examples are strain induced corrosion (LCF area) and corrosion fatigue (HCF area). For cases like these and given medium, suitable materials have to be chosen in combination with optimized design, appropriate manufacturing procedures (incl. welding), construction and operation. The loads and the water chemistry in operation have to be monitored and the effectiveness of the measures has to be verified, regularly, taking into account the actual state of knowledge. Regarding these boundary conditions the fatigue evaluations that have been performed until today seem to be sufficient, as experience shows with piping systems where this procedure has been applied. There are usually no significant failures (indication of failures); failures detected have been attributed to violation of the boundary conditions. With this background, there seems to be no need to change this procedure to safeguard the effect of environment. In this paper, the measures to guarantee integrity in design and operation state are discussed, first. Using the example of a surge line and the comprehensive monitoring results of this system the evaluation of fatigue usage and the assessment of the effect of coolant environment is discussed with reference to the ANL approach. Where the ANL approach is meant to be applied only in the design phase of a new reactor its relevance for the operation phase is cross-checked with real life measurement data. The conclusion summarizes where the effect of coolant environment has to be taken into account and gives advice how to find realistic transients for the design phase of new reactors.

Materials Aging Degradation of Reactor Vessel Internals: Part III—Structural Evaluation and Local Finite Element Models

Pressurized water reactor (PWR) internals components can experience material aging and degradation due to irradiation. The purpose of the functionality analysis is to provide a best-estimate evaluation of the reactor internals core barrel assembly for materials degradation to see if the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis [1] representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox-type (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the core provide essential information on radiation dose and heat rates of the internals. The computational fluid dynamics domain (CFD) allows evaluation of the internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the internals components and bolted connections is the third major physics field involved, which facilitates the development of operating stress fields within the internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections. Therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [2] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, to be presented at PVP 2009 [2] describes global structural finite element models. Part III, presented in this paper, presents a description of local models of bolted connections, results, and conclusions.

Sloshing Characteristics of Single Deck Floating Roofs in Aboveground Storage Tanks: Natural Periods and Vibration Modes

The floating roofs are widely used to prevent evaporation of content in large oil storage tanks. The 2003 Tokachi-Oki earthquake caused severe damage to the floating roofs due to liquid sloshing. The structural integrity of the floating roofs for the sloshing is urgent issue to establish in the petrochemical and oil refining industries. This paper presents the sloshing characteristics of the single deck floating roofs in cylindrical storage tanks. The hydrodynamic coupling of fluid and floating roof is taken into consideration in the axisymmetric finite element analysis. It is assumed that the fluid is incompressible and inviscid, and the floating roof is linear elastic while the sidewall and the bottom are rigid. The basic vibration characteristics, natural periods and vibration modes, of the floating roof due to the sloshing are investigated. These will give engineers important information on the floating roof design.

The Effect of Post-Irradiation Annealing on VVER-440 RPV Materials Mechanolocal Properties and Nano-Structure Under Re-Irradiation

The present work provides the analyses of embrittlement behavior and atom probe tomography study of nano-structure evolution of VVER-440 RPV materials under irradiation and re-irradiation. Specimens from VVER-440 weld with high level of cupper (0.16 wt.%) and phosphorus (0.027–0.038 wt.%) were irradiated in surveillance channels of Rovno Nuclear Power plant unit 1 (Ro-1). The embrittlement behavior has been assessed by transition temperature shift.

Radiation Heat Transfer Environment in Fire and Furnace Tests of Radioactive Materials Packages

The Hypothetical Accident Conditions (HAC) sequential tests of radioactive materials (RAM) packages includes a thermal test to confirm the ability of the package to withstand a transportation fire event. The test specified by the regulations (10 CFR 71) consists of a 30 minute, all engulfing, hydrocarbon fuel fire, with an average flame temperature of at least 800°C. The requirements specify an average emissivity for the fire of at least 0.9, which implies an essentially black radiation environment. Alternate tests which provide equivalent total heat input at the 800°C time averaged environmental temperature may also be employed. When alternate tests methods are employed, such as furnace or gaseous fuel fires, the equivalence of the radiation environment may require justification. The effects of furnace and open confinement fire environments are compared with the regulatory fire environment, including the effects of gases resulting from decomposition of package overpack materials. The results indicate that furnace tests can produce the required radiation heat transfer environment, i.e., equivalent to the postulated pool fire. An open enclosure, with transparent (low emissivity) fire does not produce an equivalent radiation environment.

Dynamic Analysis of Shipping Cask Subjected to Sequential Bolt Preload and Impact Loads

This paper presents an improved methodology for evaluating the dynamic responses of shipping casks subjected to the sequential HAC impact loads. The methodology utilizes the import technique of the finite-element mesh and the analytical results form one dynamic analysis using explicit numerical integration scheme into another dynamic analysis also using explicit numerical integration scheme. The new methodology presented herein has several advantages over conventional methods. An example problem is analyzed to illustrate the application of the present methodology in evaluating the structural responses of a shipping cask to the sequential HAC loading.

A Radiofrequency Identification (RFID) Temperature-Monitoring System for Extended Maintenance of Nuclear Materials Packaging

A temperature-monitoring system based on radiofrequency identification (RFID) has been developed for extending the maintenance period of the nuclear material packaging for storage and transportation. The system consists of tags, readers, and application software. The tag, equipped with a temperature sensor, is attached to the exterior of a package. The application software enables remote reading, via radio waves, of the temperature from the sensor in the tag. The system reports any temperature violations immediately via e-mail or text message, and/or posts the alarm on a secure website. The system can monitor thousands of packages and record individual temperature histories in a database. The first type of packaging that will benefit from the RFID technology is Model 9977, which has been certified by the U.S. Department of Energy (DOE) to ship and store fissile materials such as plutonium and uranium. The recorded data can be correlated to the temperature of the containment O-ring seals, based on the decay heat load of the contents. Accelerated aging studies of the Viton® GLT O-rings have shown that temperature is one of the key parameters governing the life of the O-ring seals, which maintain the integrity of the containment boundary of the package. Use of the RFID temperature-monitoring system to verify that the surface temperature remains below a certain threshold will make it possible to extend the leak-test period of the package from one year to up to five years. The longer leak-rate testing interval will yield a cost savings of up to $10,000 per package over five years. This work was conducted by Argonne National Laboratory in support of the DOE Packaging Certification Program, Office of Environmental Management, Office of Packaging and Transportation (EM-63).

Security in the Transport of Radioactive Materials

The United States Department of Energy National Nuclear Security Administration’s (DOE/NNSA) Global Threat Reduction Initiative (GTRI), the International Atomic Energy Agency (IAEA) and active IAEA Donor States are working together to strengthen the security of nuclear and radioactive materials during transport to mitigate the risks of theft, diversion, or sabotage. International activities have included preparing and publishing the new IAEA guidance document Security in the Transport of Radioactive Material while ensuring that security recommendations do not conflict with requirements for safety during transport, and developing and providing training programs to assist other countries in implementing radioactive material transport security programs. This paper provides a brief update on the status of these transportation security efforts.

Extending the Internal Examination Intervals for Pressure Vessels Using an RBI Approach

For pressure equipment containing clean, dry and non corrosive products under stable and benign conditions, there may not be any degradation mechanisms affecting the containment material over a considerable length of time. Taking account of the low risk of failure, it may then be possible to justify a longer interval between shutdowns for internal examination. Nonetheless, it is important that this judgement has been made carefully and correctly, and that adequate safeguards are in place so appropriate action may be taken if the expected conditions are subject to change. This is the theme of a recent report by TWI [1] commissioned by the UK Health and Safety Executive (HSE), which has outlined a six-step approach for assessing and justifying internal examination requirements for process plant at high hazard sites. The work has been developed partly through consultation with leading UK petrochemical companies and inspection bodies that are proactive in developing risk-based inspection methodologies.

Effects of Cladding on Reactor Pressure Vessel Integrity Based on the Revised PTS Rule (10 CFR 50.61)

This paper provides a historical perspective of the effects of cladding and the analyses techniques used to evaluate the integrity of an RPV subjected to pressurized thermal shock (PTS) transients. A summary of the specific requirements of the draft revised PTS rule (10 CFR 50.61) and the role of cladding in the evaluation of the RPV integrity under the revised PTS Rule are discussed in detail. The technical basis for the revision of the PTS Rule is based on two main criteria: (1) NDE requirements and (2) Calculation of RTMAX-X and ΔT30. NDE requirements of the Rule include performing volumetric inspections using procedures, equipment and personnel qualified under ASME Section XI, Appendix VIII. The flaw density limits specified in the new Rule are more restrictive than those stipulated by Section XI of the ASME Code. The licensee is required to demonstrate by performing analysis based on the flaw size and density inputs that the through wall cracking frequency does not exceed 1E−6 per reactor year. Based on the understanding of the requirements of the revised PTS Rule, there may be an increase in the effort needed by the utility to meet these requirements. The potential benefits of the Rule for extending vessel life may be very large, but there are also some risks in using the Rule if flaws are detected in or near the cladding. This paper summarizes the potential impacts on operating plants that choose to request relief from existing PTS Rules by implementing the new PTS Rule.