Case Study of Certification of Small Quantities of RAM: Comparison With SGQ Concept

The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials, are significantly less hazardous than large amounts of the same materials. The essential functional requirements for RAM packaging are containment of the material, ensuring sub-criticality, and ensuring that the radiation hazard of the package, as represented by the radiation dose for the package, is within the regulatory limits. Knowledge of the composition of the material being shipped is also required. By placing the contents in a containment vessel which is helium leaktight, and limiting the mass so that subcriticality is ensured, the first two requirements are readily met. Some materials emit sufficiently strong photon radiation that a small amount of material can yield a large dose rate. Foreknowledge of the dose rate which will be present for a proposed content is a challenging issue for the SGQ approach. Issues associated with certification for several cases of contents which fall within the SGQ envelop are discussed.

Very High Cycle Fatigue Behavior and Life Prediction of a Low Strength Weld Metal at Moderate Temperature

The fatigue properties of a low strength weld metal in a dissimilar welding joint in high cycle and very high cycle regimes were investigated by fully reversed axial tests in air at room temperature and 370°C. A clear duplex S-N curve existed as a result of the transition of fatigue failure mode from surface-induced failure to internal-induced failure at 370°C, while the S-N curve was continuously decreased at room temperature. A new model was successfully proposed to predict fatigue life, and interpret the crack initiation modes transition from surface inclusion to interior inclusion. It was concluded that cracks were initiated by competition among non-metallic inclusions, welding pores and discontinuous microstructures in high cycle regime. While in the very high cycle regime, non-metallic inclusions were the dominant crack initiation mechanism which depended on stress level, inclusion size as well as inclusion depth.

Operating Nuclear Plant License Renewal Program Status and Trends

As of February 2011, the NRC has renewed the operating licenses for 62 nuclear units, which will allow for up to 60 years of safe nuclear plant operation. In addition, the NRC has license renewal applications under review for 20 units and nuclear plant owners of more than 17 units have announced plans to submit license renewal applications over the next few years. This brings the total of renewed licenses and announced plans for license renewal to over 95% of the 104 currently operating nuclear units in the U.S. This paper presents the status of the U.S. license renewal process, the positive trend in regulatory stability through 2007, and the negative trend in regulatory stability after 2007. From 2000 through 2007, the NRC was able to complete the license renewal review and issue renewed licenses in 30 months or less for 100% of the license renewal applicants. In fact, approximately 77% of the reviews were completed in 22 months or less. Since 2007, NRC reviews have become much less predictable, with 21% of the reviews exceeding 30 months and only 7% being completed in 22 months or less. In fact, some reviews currently underway have exceeded 60 months and the reviews remain incomplete. One of the main factors leading to the loss of timely regulatory reviews has been the NRC adjudicatory process for license renewal, although the safety and environmental review processes have also become less timely since 2007. The factors that contributed to the positive and the negative trends are presented.

Guidelines for Repair Welding of Pressure Equipment in Refinery and Chemical Plants: Part 3—Repair Welding for Specific Materials - Carbon Steel, High Tensile Steel and Cr-Mo Steel

It has been recognized that repair welding plays an important role in the long term, safe operation of pressure equipment. In 2009, The Japan Welding Society (JWES) published guidelines for repair welding of pressure equipment [1], to meet the great need for the safe operation and proper maintenance of aging plants. Pressure equipments made from carbon steel, high tensile steel and Cr-Mo steels are utilized for high pressure services. The subject equipments are pressure vessels, heat exchangers, piping, and storage tanks for petroleum, petrochemical and liquefied natural gas industry. This paper summarizes category and property of these steels and repair welding method including special attention. Especially, weld cold cracking for these steels could be prevented by controlling the repair welding and post-weld heat treatment process to reduce the hydrogen content, hardness and weld residual stress.

Guidelines for Repair Welding of Pressure Equipment in Refineries and Chemical Plants: Part 2—Repair Welding Method

It has been recognized that repair welding plays an important role of in the long term, safe operation of pressure equipment. In 2009, the Japan Welding Engineering Society (JWES) published guidelines for repair welding of pressure equipment, to meet the great need for the safe operation and proper maintenance of aging plants. This paper gives brief descriptions of three types of repair welding method, “Flaw excavation and repair welding”, “Butt-welded insert plates”, and “External fillet welded patches”, which welding methods are based on Chemical Plant Welding Research Committee of JWES.

Inadvertent Blowdown Event in a Nuclear Boiling Water Reactor

On March 6th 2007, the Leibstadt Nuclear Power Plant in Switzerland experienced an automatic blowdown of eight safety/relief valves installed on the main steam lines caused by a faulty electrical manipulation while performing planned maintenance during full power operation. Due to the temperature measurements inside the reactor recirculation system and the reactor pressure vessel this event, at a first glance, appeared to be Event No. 23 (Automatic Blowdown event) as an Emergency (Service Level C) Condition in accordance with the relevant reactor pressure vessel Thermal Cycle Diagram. According to the ASME Code Section III, Service Level C limits permit large deformations in areas of structural discontinuity which may necessitate the removal of a component from service for inspection or repair. This paper presents a summary of thermal-hydraulic, stress, fatigue, and fracture mechanical evaluations as well as plant inspections performed to demonstrate the impact of the event on the reactor pressure vessel and associated components and to fulfill the requirements of the Swiss Federal Nuclear Safety Inspectorate. It is shown that the primary circuit of the plant was not inadmissibly stressed by the event and that it was acceptable from a safety-related point of view to return the plant to service. Corresponding to the 7-level International Nuclear and Radiological Event Scale this event was rated afterwards as level 1 (anomaly) by the Swiss Federal Nuclear Safety Inspectorate.

Analysis of Cured-in-Place Piping for a Nuclear Plant Application

Cured-in-place piping (CIPP) is used to repair existing pressure pipe that has compromised structural integrity and is no longer capable of holding operating pressure without leaking. It is often used to repair buried piping where digging the piping up to replace it would be inconvenient and/or cost prohibitive. CIPP is routinely used to repair water and sewer lines, and an ASTM specification exists to guide the design of the pipe repair for these applications. CIPP can also be used as a repair technique for piping at nuclear power plants; however, such use must be approved on a case-by-case basis. This paper discusses some of the design challenges associated with designing the CIPP for a nuclear plant application. It presents an overview of the analytical approach and the results.

Guidelines for Repair Welding of Pressure Equipment in Refineries and Chemical Plants: Part 4—Weld Repair for Specific Materials - Stainless Steel, Clad Steel and Dissimilar Joints

It has been recognized that repair welding plays an important role in the long term, safe operation of pressure equipment. In 2009, the Japan Welding Engineering Society (JWES) published guidelines for repair welding of pressure equipment to meet the great need for the safe operation and proper maintenance of aging plants. This paper describes Part 4 of the guideline, which covers repair welding of stainless steel, clad steel, weld overlay and dissimilar joints.

Continuous Creep Damage Monitoring Using a Novel Potential Drop Technique

A new directional low-frequency Alternating Current Potential Drop (ACPD) technique has been developed for continuous in-situ monitoring of creep strain and damage in alloys. The sensor relies on a modified ACPD technique that measures simultaneously both values of resistance in the axial and lateral directions using a square electrode configuration. The technique monitors the variation in the ratio of the measured axial and lateral resistances, therefore can efficiently separate the mostly isotropic common part of the resistivity variation caused by reversible temperature variations from the mostly anisotropic differential part caused by direct geometrical and indirect material effects of creep. Initially, this ratio can be considered proportional to the axial creep strain, while at later stages, the resistance ratio accelerates due to the formation of directional discontinuities such as preferentially oriented grain boundary cavities and micro-cracking in the material. This ACPD technique has been applied to a series of accelerated creep tests on 2.25CrMoV Steel at 650 °C. The results are presented and the application of the method for online component monitoring is discussed.

Pre Test Calculations on TOPFLOW PTS Experiment With NEPTUNE_CFD Code

Hypothetical Small Break Loss Of Coolant Accident is identified as one of the most severe transients leading to a potential huge Pressurized Thermal Shock on the Reactor Pressure Vessel (RPV). This may result in two-phase flow configurations in the cold legs, according to the operating conditions, and to reliably assess the RPV wall integrity, advanced two-phase flow simulations are required. Related needs in development and/or validation of these advanced models are important, and the ongoing TOPFLOW-PTS experimental program was designed to provide a well documented data base to meet these needs. This paper focuses on pre-test NEPTUNE_CFD simulations of TOPFLOW-PTS experiments; these simulations were performed to (i) help in the definition of the test matrix and test procedure, and (ii) check the presence of the different key physical phenomena at the mock-up scale.