Dynamic Analysis of Hanford Unirradiated Fuel Package Subjected to Sequential Lateral Loads in Hypothetical Accident Conditions

Large fuel casks present challenges when evaluating their performance in the Hypothetical Accident Conditions (HAC) specified in the Code of Federal Regulations Title 10 part 71 (10CFR71). Testing is often limited by cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing by using simplified analytical methods. This paper presents a numerical technique for evaluating the dynamic responses of large fuel casks subjected to sequential HAC loading. A nonlinear dynamic analysis was performed for a Hanford Unirradiated Fuel Package (HUFP) [1] to evaluate the cumulative damage after the hypothetical accident Conditions of a 30-foot lateral drop followed by a 40-inch lateral puncture as specified in 10CFR71. The structural integrity of the containment vessel is justified based on the analytical results in comparison with the stress criteria, specified in the ASME Code, Section III, Appendix F [2], for Level D service loads. The analyzed cumulative damages caused by the sequential loading of a 30-foot lateral drop and a 40-inch lateral puncture are compared with the package test data. The analytical results are in good agreement with the test results.

Complex Monitoring Programme for WWER Reactor Pressure Vessels Lifetime Assessment in the Czech Republic

Lifetime of reactor pressure vessels practically depends on a level of degradation of RPV material properties during operation. The most important degradating mechanism of RPV materials is usually radiation damage, characterized by values on neutron fluence on one side and radiation embrittlement of RPV materials on the second side. WWER reactor pressure vessels in the Czech Republic are a subject of a very thorough and complex monitoring program, that includes: • Standard material surveillance program containing of WWER-440 RPV materials — base metal, weld metal, heat affected zone, but irradiated with high lead factor (13 to 18), • Supplementary surveillance program of WWER-440 RPV materials, including additionally austenitic cladding materials, IAEA reference material JRQ irradiated with low lead factor (2 to 3) with parts subjected to annealing and re-irradiation after annealing, • Modified surveillance program of WWER-1000 RPV materials — base metal, weld metal, heat affected zone, cladding materials, IAEA reference JRQ material irradiated in low lead factor (2 to 3) near RPV inner beltline region, • Integrated surveillance specimen program for WWER-1000 reactor including materials from NPP Temelin (Czech Republic), Belene (Bulgaria), Kalinin (Russia) and Ukranian NPPs, • Continous exvessel monitoring of neutron fluence on outer RPV surface for both WWER-440 and WWER-1000 plants, • Neutron fluence determination on inner RPV surface (austenitic cladding) using special technique for removal of specimens from cladding for Nb activity measurements, • Ex-vessel temperature measurements during RPV operation. All these programs serve for precision of operation conditions and determination of degradation of RPV materials for RPV integrity and lifetime assessment.

Development of Testing Methodologies for Onsite Radioactive Material Storage Containers

Based on a recommendation from the Defense Nuclear Facilities Safety Board, the Department of Energy (DOE) Office of Nuclear Safety Policy and Assistance (HS-21) has recently issued DOE Manual 441.1-1 entitled Nuclear Material Packaging Manual. This manual provides guidance regarding the use of non-engineered storage media for all special nuclear material throughout the DOE complex. As part of this development effort, HS-21 has funded the Oak Ridge National Laboratory (ORNL) Transportation Technologies Group (TTG) to develop and demonstrate testing protocols for such onsite containers. ORNL TTG to date has performed preliminary tests of representative onsite containers from Lawrence Livermore National Laboratory and Los Alamos National Laboratory. This paper will describe the testing processes that have been developed.

Analytical Solutions and Tools for Assessment of Surface Cracks in Cylindrical Components

The paper reviews some advanced stress intensity factor solutions derived for analyses of axial and circumferential surface cracks in cylindrical components subjected to variable stress fields. The solutions are examined considering their validity ranges with respect to the crack and cylinder geometry, ability to account for a complex stress distribution in the pipe wall, as well as their accuracy. A method for estimating errors in numerical stress intensity factor solutions is introduced and applied to a particular set of data. Examples of a leak-before-break assessment and crack growth calculations under thermal fatigue loading are included to demonstrate the solutions performance. The considered analytical stress intensity factor solutions yield close results provided that the stress field in the prospective crack plane is described by a smooth function of the radial coordinate. For two-dimensional stress profiles as well as for variable ratios of the cylinder wall thickness to the inner radius, a selective use of the solutions is recommended considering their specific features and validity ranges.

Ways for Extending the Reactor Pressure Vessel Lifetime

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

A Non-Destructive Testing Method and Analysis for Air Pressure Distribution in the Stacks of Building Drainage Systems

A drainage system is one of the most essential facilities in building service engineering. Unfortunately relevant technology used today to analyze it was developed decades ago. This research investigated the case of existing building drainage systems in Taiwan, including our previous studies. The purpose of this paper is the development of a non-destructive testing method of air pressure fluctuation in a stacked building drainage system using field observation and experimental study of stack fluid mechanisms. A portable testing device is developed to execute field testing in existing drainage systems to determine air pressure fluctuation in the stacks of buildings. Meanwhile, the Fourier Transform Process is adopted in this paper to analyze the power spectrum of air pressure fluctuation in a drainage stack and to verify the previous theoretical study. Validation obtained from case-studies can be used to confirm the practicality of this portable and non-destructive testing method. As a result, the proposed testing method can be applied to the diagnosis of existing building drainage systems and improve the design of a drainage system in an existing housing complex.

The Department of Energy Replacement for the 110-Gallon Specification 6M Shipping Container for Radioactive Contents

The Department of Energy (DOE) has been shipping university reactor fuels and other fissile materials in the 110-gallon Department of Transportation (DOT) Specification 6M container for over 20 years. The DOT 6M container has been the workhorse for many DOE programs. However, packages designed and used according to the Specification 6M (U. S. Code of Federal Regulations, 49 CFR 178.354; 2003) do not conform to the latest package safety requirements in 10 CFR 71, especially performance under hypothetical accident conditions. For that reason, the 6M specification containers are being terminated by the DOT. Packages designed to the 6M specification will no longer be allowed for in-commerce shipments after October 1, 2008. To meet on-going transportation needs, DOE evaluated several different concepts for replacing the 110-gallon 6M. After this evaluation, DOE selected the Y-12 National Security Complex for the project. The new Y-12 container, designated the ES-4100 shipping container, will have a capacity of four times the current 6M and will be certified by the Nuclear Regulatory Commission (NRC). The ES-4100 project began in September 2006 and prototypes of the new container are now being fabricated. Details on the design features and the upcoming regulatory testing of this new container are discussed in this paper.

License Renewal: Status of Long Term Operation in the U.S.

As of February 2008, the NRC has approved renewal of the operating licenses for 48 nuclear units and has applications under review for 15 more units. In addition, nuclear plant owners for at least 25 more 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 80% of the 104 currently operating nuclear units in the U.S. This paper presents some of the factors that have made the U.S. license renewal process so successful. These factors include (1) the successful regulatory process and on-going continuous improvement of that process, (2) long-term safe plant operation trends, (3) stable low-cost generation of electricity, (4) high levels of plant reliability, and (5) improving public opinion trends.

Plant Life Management Models: A Comparison With Analysis of Impact on Both Safety and Non-Safety Issues

Due to current social and economical framework, in last years many nuclear power plant owners started a program for the Long Term Operation (LTO)/PLIM (Plant Life Management) of their older nuclear facilities. A PLIM framework requires both a detailed review of the features of the main safety programs (Maintenance, ISI, Surveillance) and a complete integration of these safety programs into the general management system of the plant. Therefore PLiM should address safety as well economics, knowledge management as well as decision making, and provide an overall framework to keep the whole plant in a safe and economically sustainable condition. Moreover, new external factors, such as: large use of subcontractors, need for efficient management of spare parts, request for heavy plant refurbishment programs demand for updated techniques in the overall management of the plant. Therefore new organisational models have to be developed to appropriately support the PLIM framework. In recent years a network of European Research Organisations (SENUF) carried out many R&D tasks aiming at capturing the aspects of the maintenance programs where research is mostly needed and at developing suitable optimised maintenance models. Using the outcome of these initiatives, this paper aims at identifying the technical attributes of the PLIM program more directly affecting the decision for a long-term safe operation of a nuclear facility, and the issues related to its optimal implementation. A comparison of some of the available models is presented and an analysis of the potential impact on safety and non-safety programs is provided in order to support the development of optimised life management models.

Repair Weldability at Overlay/Base Metal Interface of Petroleum Pressure Vessel

In order to extend the life of petroleum pressure vessels operated in long term, it is needed to establish the reliable repair welding technique. Weld cold cracking sometimes occurred in long-term operated petroleum pressure vessels due to hydrogen embrittlement by thermal stress and diffusible hydrogen after repair welding. The cracking was caused by the hydrogen concentration at the base meal of 2.25Cr-1Mo steel/overlaying metal of austenitic stainless steels interface during the service with high temperature and hydrogen partial pressure. The tendency was accelerated by carbide precipitation at the interface due to the post weld heat treatment (PWHT) and the operation with high temperature. That is, the crack susceptibility at the interface became markedly higher owing to the hydrogen embrittlement with metallurgical degradation by thermal embrittlement. To make clear the effect of weld thermal cycles during repair welding on the hydrogen content and weld cold cracking at the interface in the structural material of petroleum pressure vessels, the crack susceptibility was estimated by y-groove weld cracking test with varying overlay thickness and hydrogen exposure conditions. In addition, the hydrogen distribution in the material was calculated by the theoretical analysis using the diffusion equation based on activity. The crack susceptibility was raised with increase in the hydrogen content at the interface. It was concluded that the cracking could be prevented by controlling the repair welding process to reduce the hydrogen content at the interface.