Crevice Repassivation Potential of Alloy 22 in High-Nitrate Dust Deliquescence Type Environments

The nitrate ion (NO3−) is an inhibitor for crevice corrosion of Alloy 22 (N06022) in chloride (Cl−) aqueous solutions. Naturally formed electrolytes may contain both chloride and nitrate ions. The higher the ratio R = [NO3−]/[Cl−] in the solution the stronger the inhibition of crevice corrosion. Atmospheric desert dust contains both chloride and nitrate salts, generally based on sodium (Na+) and potassium (K+). Some of these salts may deliquescence at relatively low humidity at temperatures on the order of 150°C and higher. The resulting deliquescent brines are highly concentrated and especially rich in nitrate. Electrochemical tests have been performed to explore the anodic behavior of Alloy 22 in high chloride high nitrate electrolytes at temperatures as high as 150°C at ambient atmospheres. Naturally formed brines at temperatures higher than 120°C do not induce crevice corrosion in Alloy 22 because they contain high levels of nitrate. The inhibitive effect of nitrate on crevice corrosion is still active for temperatures higher than 100°C.

Computer Simulations to Address Pu-Fe Eutectic Issue in 3013 Storage Vessel

On November 22, 2005, the Manager of the Plutonium Finishing Plant (PFP) in Richland, WA issued an Occurrence Report involving a potential Pu-Fe eutectic failure mechanism for the stainless steel (SS) 3013 cans containing plutonium (Pu) metal. Four additional reports addressed nuclear safety concerns about the integrity of stainless steel containers holding plutonium during fire scenarios. The reports expressed a belief that the probability and consequences of container failure due to the formation of a plutonium-iron eutectic alloy had been overlooked. Simplified thermal models similar to the HAC thermal models used in the 9975 SARP were created and analyzed to address the Pu-Fe eutectic concerns. The model uses Rocky Flats configuration with 2 stacked Pu buttons inside a 3013 assembly. The assembly has an outer can, an inner can, and a convenience can, all stainless steel. The boundary conditions are similar to the regulatory 30 minutes HAC fire analyses. Computer simulations of the HAC fire transients lasting 4 hours of burn time show that the interface between the primary containment vessel and the Pu metal in the 9975 package will not reach Pu-Fe eutectic temperature of 400°C.

Properties of Fiberboard Overpack Material in the 9975 Shipping Package Following Thermal Aging

Many radioactive material shipping packages incorporate cane fiberboard overpacks for thermal insulation and impact resistance. Mechanical, thermal and physical properties have been measured on cane fiberboard following thermal aging in several temperature/humidity environments. Several of the measured properties change significantly over time in the more severe environments, while other properties are relatively constant. These properties continue to be tracked, with the goal of developing a model for predicting a service life under long-term storage conditions.

Erosive Wear of Sharp-Edged Orifice Plates

Sharp-edged orifice plates are extensively used in flow measurement and flow trimming. In many plant applications where they are used, orifices are installed in systems which are generally clean and debris free. However, there are applications where the system may be either debris laden or exposed, in accident conditions, to debris that could challenge the effectiveness of orifices. This paper presents the results of an accelerated erosive wear test of aluminum and steel orifice plates to determine firstly, how sharp-edged orifices wear in such medium, and secondly, how the wear affects flow measurements.

Data Analysis for RPV Life Management Studies of Temperature Measurements From Monitoring of Safety Injection Fluid in Refueling Water Storage Tank

In Electricite´ de France (EDF), it’s used to taking as a conservative constant value for the temperature of Refueling Water Storage Tank (RWST) in the deterministic and probabilistic analyses for Reactor Pressure Vessel (RPV) life management. The water contained in this storage tank supplies Security Injection during accidental conditions, such as loss-of-coolant accident (LOCA), so the temperature of this water is a very important input parameter for fracture mechanics analyses. In the continuation of [1], the aim of our study is to evaluate the variability of this temperature. Since 1999, EDF has been collecting the RWST temperature for several sites and several nuclear plant units. Using statistical analyses, this study aims at first to identify the most important RPV exploitation life events that influence this temperature and, at secondly, to obtain statistical density distributions in order to describe its variability in the most realistic way possible.

“Adaptative” User Interface for Expert Knowledge Validation and Evaluations Interpretation in an Asset Management Process

The long term management of a production asset raises several major issues among which rank the technical management of the plant, its economics and the fleet level perspective one has to adopt. Decision makers are therefore faced with the need to define long term policies (up to the end of asset operation) which take into account multiple criteria including safety (which is paramount) and performance. In this paper we first remind the reader of the EDF three-level methodology for asset management. We then focus on the knowledge model and on the software tools that implement this methodology in order to gather, preserve, share, maintain and exploit the expert knowledge needed for asset management and to allow decision makers to define, evaluate and analyze long term plant operation and maintenance policies. Lastly, as the quality of the processed plant level evaluations (operation & maintenance strategies are evaluated, at a plantlevel, through a set of technical and economic indicators) and their interpretation relies on the quality of the knowledge captured in the tools, we focus on the definition of a “adaptative” user interface — based on Electronic Structured Documents — that allows technical/strategic experts and decision makers to consult the useful pieces of knowledge in a context dependent way. Such an interface, which, in a near future, should be fully implemented in the tools will facilitate the validation of the knowledge-base content and the analysis of the processed results.

IAEA Coordinated Research Projects on Master Curve Technology: An Overview

The IAEA has sponsored a series of Coordinated Research Projects (CRPs) that have led to a focus on reactor pressure vessel (RPV) structural integrity. In the previous CRP on the application of the Master Curve (MC), three key areas were identified to need further works. These three tasks are: (1) test specimen bias, constraint, and geometry effects on the measurement and application of To values; (2) effects of loading rate up to impact loading conditions on To; and (3) potential changes in the shape of the MC for highly embrittled RPV materials. These three areas are focused on the new CRP (here below CRP-8). A new CRP is follow-on to previous successful CRPs on resolving technical issues associated with application of the MC approach. Overall objectives of CRP-8 include: 1) Better quantification of fracture toughness issues related to testing surveillance specimens for application to RPV integrity assessment, and 2) Development of approaches for addressing MC technical issues in integrity evaluation of operating RPVs. 15 organizations from 11 countries participate in the CRP and the final report will be issued in the end of 2008.

Use of a CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressurised Thermal Shock Conditions: Thermal-Hydraulic Studies of a Safety Injection in a PWR Plant — Methodology and Present Limitation

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behaviour of cracks under PTS loading conditions due to the emergency cooling during PTS transient like SBLOCA. This paper explains the Research and Development program started at Electricite´ De France about the cooling phenomena of a PWR vessel after a Pressurised Thermal Shock. The numerical results are obtained with the E.D.F ThermalHydraulic code (Code_Saturne) coupled with the thermal-solid code SYRTHES to take into account the conjugate heat transfer on the cooling of the vessel. We first explain the global methodology with a progress report on the state of the art of the tools available to simulate the different scenari displayed within the frame of the plant life project in order to reassess the integrity of the RPV, taking into account the evolution of some input data, such as the new value of end of life (EOL) fluence, the feedback results of surveillance program and the evolution of the functional requirements. The main results are presented and are related to the evaluation of the RPV integrity during a Small Break Loss Of Coolant Accident transient for 900 and 1300 MWe nuclear plant. On the whole, the main purpose of the numerical CFD studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins. In a second time, a new analysis is performed to assess an accurate temperature distribution in the RPV. Indeed, from a physical phenomena point of view, the EDF thermalhydraulic tool Code_Saturne is now qualified in order to assess single phase transient but in the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur, such as condensation due to the emergency core cooling injections of sub-cooled water. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. In that purpose, a program has been set up to extend the capabilities of the Neptune_CFD two-phase solver which is the tool able to solve two phase flow configuration. In a same time, A simplified approach has showed that for a type of transient weakly uncovered, a free surface calculation was sufficient to respect the necessary criteria of safety. A Qualification study was carried out on the Hybiscus experimental E.D.F facility, representing a cold leg with ECC injection and a third down comer. Temperature profiles have been compared and are presented and analysed here, showing encouraging results.

A Qualitative Risk-Based Assessment Procedure for High Temperature Hydrogen Attack of C-1/2Mo Steel

A qualitative risk-based assessment procedure was developed to determine the relative probability of failure for high temperature hydrogen attack (HTHA) of C-1/2Mo steel. It is well documented that C-1/2Mo possesses a variable resistance to HTHA. Over the years, the Nelson curve limits, as published in API RP 941, Steels for Hydrogen Service at Elevated Temperatures and Pressures in Petroleum Refineries and Petrochemical Plants, have been significantly reduced since originally published. This has created a need for an assessment procedure to evaluate C-1/2 Mo equipment that was designed many years ago when the Nelson curve was at much higher operating conditions than today. This assessment procedure is based on the current ExxonMobil assignment of C-1/2Mo resistance to HTHA on the Nelson curve and the relative dependence of HTHA on temperature, hydrogen partial pressure and time. We use the Pv relationship that appears in literature to define the relative dependence of HTHA on temperature, hydrogen partial pressure and time. This qualitative assignment of probability levels for HTHA was validated against actual HTHA failures of C-1/2Mo in high temperature hydrogen service that have been reported to the API and other internal Company incidents not reported to API. This assessment procedure is used to qualitatively define the probability of failure consistent with the probability levels defined in the ExxonMobil risk matrix which is used to manage all risk based decisions in the Corporation.

Improving Thermal Performance of Radioactive Material Drum Type Packages by Using Heat Pipes

This paper presents a feasibility study to improve thermal loading of existing radioactive material packages by using heat pipes. The concept could be used to channel heat in certain directions and dissipate to the environment. The concept is applied to a drum type package because the drum type packages are stored and transported in an upright position. This orientation is suitable for heat pipe operation that could facilitate the heat pipe implementation in the existing well proven package designs or in new designs where thermal loading is high. In this position, heat pipes utilize gravity very effectively to enhance heat flow in the upward direction Heat pipes have extremely high effective thermal conductivity that is several magnitudes higher than the most heat conducting metals. In addition, heat pipes are highly unidirectional so that the effective conductivity for heat transfer in the reverse direction is greatly reduced. The concept is applied to the 9977 package that is currently going through the DOE certification review. The paper presents computer simulations using typical of-the-shelf heat pipe available configurations and performance data for the 9977 package. A path forward is outlined for implementing the concepts for further study and prototype testing.