Design and Analysis of Shipping Container for Big Rock Decommissioned Reactor Vessel

A major milestone in 2003 on the Big Rock Point (BRP) decommissioning project involved shipping the Reactor Vessel (RV) in a steel cask for burial. The Reactor Vessel Transport System (RVTS) cask was a sealed integral container, which provided necessary radiological shielding and containment of radioactive waste for shipping and disposal. The RVTS, using the provisions of the ASME BPVC Section III, Subsection NB, was designed as a Type B package in accordance with the requirements of 10 CFR Part 71. This included meeting Normal Condition of Transport (NCT) and the Hypothetical Accident Conditions (HAC) loading per 10 CFR 71, Regulatory Guide 7.6, “Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels,” Regulatory Guide 7.8, “Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material” and Regulatory Guide 7.11, “Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Maximum Wall Thickness of 4 Inches.” The RVTS was designed to withstand accelerations and shocks postulated during highway and rail transit using guidelines from the Association of American Railroads (AAR) and ANSI N14.2. The design analysis methodology, fabrication process and transportation planning for the Big Rock RVTS Cask are presented in this paper.

The Analysis of Acoustic Emission Code and the Research of Acoustic Emission Signal of P355NL1 Under Low Temperature of −40°C

The latest content of acoustic emission code GB/T18182-2012 are shown in this paper. The differences of GB/T18182-2012 and GB18182-2000 are summarized. Acoustic emission method was used to inspect the whole process of tensile specimen test under the temperature of 0°C, −20°C and −50°C. Based on the characteristics of the signal of different temperature condition, some difference of these signals was analyzed and discussed. By using the above result, acoustic emission was applied to inspect a drikold storage vessel.

Review of Zion Station Active Decommissioning Project

Review of Zion Station Active Decommissioning Project The paper will provide a review of project activities at the Zion Station during the forty-five months of active decommissioning. A discussion of project planning strategy, project successes and unanticipated project challenges experienced to date will be provided. A discussion of construction, demolition and waste handling issues, including compliance with railroad car loading rules will also be provided. Innovations and management techniques enlisted to address project challenges will be discussed.

License Renewal and Subsequent License Renewal in the U.S.

As of February 2014, the NRC has renewed the operating licenses for 73 nuclear units, allowing for up to 60 years of safe operation. In addition, the NRC has license renewal applications under review for 18 units and 9 additional units have announced plans to submit applications over the next few years [1]. This brings the total of renewed licenses and plans for renewal to 100% of the operating nuclear units in the U.S. By the end of 2014, there will be 38 nuclear plants that will have operated for more than 40 years and will be eligible to seek a subsequent license renewal (or almost 40% of the nuclear units expected to be operating at the end of 2014). In 2013, nuclear plant owners of 5 units shutdown operation or announced plans to shutdown by the end of 2014. However, most of the remaining operating plant owners are keeping the option open for long term operation beyond 60 years. NRC and the U.S. nuclear industry have made significant progress in preparing the way for subsequent license renewal applications. This paper presents the status of the U.S. license renewal process and issues being addressed for possible applications for subsequent renewals for up to 80 years of operation.

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

Development of Enertech’s Advanced Normally Open NozzleCheck Valves for Generation III+ Nuclear Power Plants

Enertech introduced the first Normally Open NozzleCheck valves to the nuclear power industry nearly 20 years ago. This passive valve design was developed to address reoccurring maintenance and reliability issues often experienced by various check valve types due to low or turbulent flow conditions. Specifically, premature wear on the hinge pins, bushings and severe seat impact damage had been discovered in several applications while the systems were in steady state operating conditions. Over the last two decades, Enertech has continued to improve upon the design of the valve, with the culmination coming most recently in support of Generation III+ passive reactor requirements. This entirely new valve is designed with minimal stroke, ensuring quick closure under low reverse flow conditions which no other check valve design could support. Additionally, features such as first in kind test ports, visual inspection points, and the ability to manually stroke the valve in line have resolved many of the short comings of previous inline welded flow check valves. Most importantly, advanced test based methodologies and models developed by Enertech, allow for accurate prediction of NozzleCheck valve performance. This paper presents the development of Enertech’s advanced Normally Open NozzleCheck Valve for Generation III and III+ nuclear reactor designs. The Valve performance was initially determined by using verified and validated computational fluid dynamic (CFD) methods. The results obtained from the CFD model were then compared to the data gathered from a prototype valve that was built and tested to confirm the performance predictions. Enertech has fully tested and qualified the Normally Open NozzleCheck valve which is specifically designed for applications that require a high capacity in the forward flow direction and a quick closure during low reverse flow condition with short stroke to minimize the hydraulic impact on the system.

Energy Saving for a Chinese Crude Oil Pipeline

The Dongying-Linyi Crude Oil Pipeline, transporting the crude oil from Shengli Oilfield to the outside, is one of the most important pipelines in East China. It is of great significance for saving energy, reducing emissions and improving the economy by optimizing pipeline operation. By analyzing the daily operation reports and monthly energy consumption reports, the energy consumption situation of the pipeline is revealed. Then the crude oil from Shengli Oilfield is collected, and its physical property is tested and researched. Based on the topological structure, the technological process and the operation philosophy of the pipeline, the mathematical models for the optimal operation and the energy consumption are built and the corresponding algorithm is introduced. With the models and the algorithms mentioned above, a computer software is developed for the operation simulation and optimization of the pipeline. Combining with practice, three energy-saving measures for the pipeline are analyzed, including reduction of oil heating temperature, operation matching with peak-valley electricity price and operation matching with the combination of flow rates in a planned operation period. In addition, these energy-saving measures are evaluated by the software with several examples. It turns out that: the first measure can considerably reduce the energy consumption and the energy consumption cost of the pipeline; the second could reduce the energy consumption cost of the pipeline a lot, while may make the energy consumption go higher a little; the third could slightly lower the equivalent energy consumption and the energy cost at the same time, but the effect of it is not obvious.

MAI Benchmark Campaign of International Software for Reactor Pressure Vessel Integrity Assessment

Nuclear Reactor Pressure Vessel (RPV) integrity is a major issue concerning plant safety and this component is one of the few within a Pressurized Water Reactor (PWR) whose replacement is not considered as feasible. To ensure that adequate margins against failure are maintained throughout the vessel service life, research engineers have developed and applied computational tools to study and assess the probability of pressure vessel failure during operating and postulated loads. The Materials Ageing Institute (MAI) sponsored a benchmark study to compare the results from software developed in France, Japan and the United States to compute the probability of flaw initiation in reactor pressure vessels. This benchmark study was performed to assess the similarities and differences in the software and to identify the sources of any differences that were found. Participants in this work included researchers from EDF in France, CRIEPI in Japan and EPRI in the United States, with each organization using the probabilistic software tool that had been developed in their country. An incremental approach, beginning with deterministic comparisons and ending by assessing Conditional Probability of crack Initiation (CPI), provided confirmation of the good agreement between the results obtained from the software used in this benchmark study. This conclusion strengthens the confidence in these probabilistic fracture mechanics tools and improves understanding of the fundamental computational procedures and algorithms.

Numerical Modeling of Turbulent Cavitation Flows in Safety Relief Valves

Safety relief valve (SRV) is still the ultimate security component of pressure vessels or piping equipment. It does not take the place of a regulating or control valve but it aims to protect devices and people by preventing damage due to overpressure in the system. This is ensured by discharging an amount of fluid when excessive rising pressure occurs. For the incompressible flows, the discharge coefficient of the relief valve can be modified by cavitation development under specific operating conditions. Then, the sizing of the valve doesn’t correspond to the flow discharged resulting in severe damage. This study aims to demonstrate the capability of numerical modeling to predict the evolution of discharge coefficient under cavitation conditions. URANS simulations have been performed with ANSYS CFX 13.0 using Shear Stress Transport modeling. A Rayleigh-Plessey model is used to predict the development of cavitation in the relief valve. A modification of the saturation vapor pressure is proposed in the cavitation model to take into account the turbulent effects on the cavitation development. Additionally a Plexiglas mock-up has been built for flow visualization and two valve discs are used to measure the discharge coefficient or the flow force acting on the valve. Numerical and experimental approaches are compared first by analyzing qualitative results through flow visualization and also by evaluating hydraulic characteristics of the SRV.

Analysis of the Pressure Transient and Burst Pressure for Exposure of the 9975 Primary Containment Vessel to Fire

Bare shipping package containment vessels can be utilized to stage plutonium oxide at the Savannah River Site. Pressurization and subsequent release could occur due to a hypothetical facility fire. Pressurization due to adsorbed moisture on the plutonium oxide and plastic packaging materials could result in rupture of the containment vessel. The containment vessel was evaluated to determine rupture pressure when subjected to the fire conditions. The rupture pressure is compared with pressures developed due to radiolytic gas generation.