System for the Automated Mitigation of Hypergolic Propellant Leaks

Hypergolic liquid rocket propellants consist of fuel and oxidizer liquids stored in separate tanks that ignite when mixed. They are favored for propulsion systems where precise pulsing or throttling is required. High reliability is also insured because the liquid chemicals are stable for long periods of time, and no ignition system is required — the fuel and oxidizer simply need to be brought together. These propellants have relatively high vapor pressures, are toxic, and may be carcinogenic. A pool of the liquid will pose a health, fire, or explosion hazard. Therefore, missiles containing these propellants are stored and shipped in canisters that contain a sensor to detect the leakage of these propellants that will sound an alarm at a set concentration. This paper discusses the development of an automated system that is designed to mitigate a leak of the fuel or oxidizer within the canister in order to render the system safe enough to allow access to the missile. The mitigation system neutralizes the leaked propellant. The design and testing of the system with representative propellants is discussed.

Hydro-Abrasive Wear Damage at Reactor Recirculation Pump Bearing Journals

During the refueling and maintenance outage in August 2011 at Leibstadt Nuclear Power Plant in Switzerland, the inspection of the hydrostatic bearings of the two identical recirculation pumps revealed a deep circumferential erosion groove on the inside surface of each of the bearing journals. The bearing journals are made of austenitic stainless steel. The cylindrical journal is welded to the back shroud of the impeller and surrounds the internal stationary heat exchanger of the pump by forming a narrow fluid filled annulus. The location of material removal was the same as in the year 2004 when similar wear damage was fixed by build-up welding. The plant decided to repair the damage during the subsequent outage in 2012. However, the Swiss Federal Nuclear Safety Inspectorate in return required the plant to identify the precise erosion mechanism, to ensure the structural integrity of the journals by taking into account the rate of material removal from 2004 up to the 2012 outage, and to include provisions for the early detection of a journal failure. This paper summarizes the previous as well as the latest results of different inspections, investigations, evaluations, and analyses done to meet the requirements of the Swiss regulatory authority. The results show that, from a safety-related and an operational availability perspective, it is acceptable not to repair the damaged bearing journals prior to the 2012 outage.

Optical Strain Monitoring Techniques

The Auto-Reference Creep Management and Control (ARCMAC) system is being developed as a technique to evaluate the remaining life of power plant components. The system consists of a pair of Inconel plates with a configuration of silicon nitride (SiN) spheres on them, and a camera system used to take images of the gauge during the component’s deformation. The purpose of the system is to measure the creep strain accumulated by a component at regular intervals, tracking the relative motion of the spheres in order to measure a point-to-point value of strain. The system is currently used to capture images of gauges already installed on power plants in the UK as part of scheduled maintenance during plant outages. It is also possible to use the ARCMAC system to capture speckle paint pattern data used in digital image correlation (DIC) in order to visualise the strain field across the heat affected zones (HAZ) in welds and around other strain concentration features. A newer version of the system: the Digital Single Lens Reflex (DSLR) ARCMAC is being developed specifically to capture this kind of data in order to complement the point-to-point strain measurements obtained. This article presents results of experiments performed at room temperature with the purpose of establishing the basic accuracy of the conventional ARCMAC and the DSLR ARCMAC in order to compare their performance. It also intends to evaluate the performance of the latter when used for digital image correlation. The results showcase the accuracy of the technique at high strains using the DSLR camera, showing its usefulness as a tool to measure creep strain.

Pu/SS Eutectic Assessment in 9975 Packagings in a Storage Facility During Extended Fire

In a radioactive material (RAM) packaging, the formation of eutectic at the Pu/SS (plutonium/stainless steel) interface is a serious concern and must be avoided to prevent of leakage of fissile material to the environment. The eutectic temperature for the Pu/SS is rather low (410°C) and could seriously impact the structural integrity of the containment vessel under accident conditions involving fire. The 9975 packaging is used for long term storage of Pu bearing materials in the DOE complex where the Pu comes in contact with the stainless steel containment vessel. Due to the serious consequences of the containment breach at the eutectic site, the Pu/SS interface temperature is kept well below the eutectic formation temperature of 410°C. This paper discusses the thermal models and the results for the extended fire conditions (1500°F for 86 minutes) that exist in a long term storage facility and concludes that the 9975 packaging Pu/SS interface temperature is well below the eutectic temperature.

Thermal Upgrading of 9977 Radioactive Material (RAM) Type B Package

The 9977 package is a radioactive material package that was originally certified to ship Heat Sources and RTG contents up to 19 watts and it is now being reviewed to significantly expand its contents in support of additional DOE missions. Thermal upgrading will be accomplished by employing stacked 3013 containers, a 3013 aluminum spacer and an external aluminum sleeve for enhanced heat transfer. The 7th Addendum to the original 9977 package Safety Basis Report describing these modifications is under review for the DOE certification. The analyses described in this paper show that this well-designed and conservatively analyzed package can be upgraded to carry contents with decay heat up to 38 watts with some simple design modifications.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

The MacArthur Maze Fire and Roadway Collapse: A “Worst Case Scenario” for Spent Nuclear Fuel Transportation?

In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the “MacArthur Maze.” The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC’s analysis of the consequences of this type of severe accident.

Residual Stress Distributions and Guidelines for Crack Inspection in Pipe Bends

Root cause investigations of feeder pipe cracks detected in one CANDU™ nuclear power plant indicated that the residual stress had a significant role in these failures. As a result, numerous residual stress measurements of pipe bends were obtained and models for predicting the residual stress distributions associated with various bending processes were developed. This paper provides a comprehensive review of pipe bend residual stress data and provides guidelines for identifying the most susceptible locations for targeted crack inspections. Residual stress data, generated since 1997 as part of the CANDU feeder cracking investigation, was compiled and presented for the quick dissemination of information. This information is summarized in quick lookup tables where likely crack locations are identified for pipe bends manufactured using various fabrication techniques.

Evaluation of Cumulative Damage From Drop to Puncture Conditions

During safety assessments of transport packages, cumulative damages are naturally accumulated for assessments performed using physical tests specimens. However, the cumulative damages are not as easily accounted when assessments are by numerical simulations. While analysts are comfortable with simulating packages for single events, it is not yet common practice to incorporate the effect handed over from a former event to the next, in a series of sequential load events. Thus, many numerical simulations in SAR (Safety Analysis Report) represent just a single event in the series of sequential event comprising the required accident condition. These single event numerical simulations are then based on initial conditions different from the analogous physical test specimen, which could contribute to a growing disparity in results between assessments by physically testing compared to numerical simulation. The reason why analyses do not consider the cumulative damage is difficulties in delivering the final result of the previous analysis to the current analysis. The hypothetical accident conditions described in the IAEA regulations include drop, puncture, fire, and water immersion conditions, which should be sequentially simulated. There can be cumulative damage between two accident conditions, such as drop and puncture, puncture and fire, and so forth. In this study, as the first step to consider cumulative damage, an analysis technology to perform a puncture analysis incorporating the final response field from a prior drop analysis is proposed. The necessity and validity of the proposed analysis technology are evaluated by a comparison with the results obtained by performing each analysis independently.

Dynamic Finite Element Analysis of Drum-Type Shipping Packages for Radioactive Materials

The US Department of Energy (DOE) often uses Type AF and Type B drum-type packages for shipment of radioactive materials (RAM), both of which shall be designed and certified to meet the regulatory requirements specified in 10 CFR 71, to ensure safety, public health and protection of environment. In particular for the hypothetical accident conditions (HAC) prescribed in 10 CFR 71.73, RAM packages are subjected to sequential tests of 30-ft drop, crush, puncture, engulfing fire, and water immersions. Packages shall maintain structural integrity of containment, radiation shielding, and criticality control following these HAC tests. The structural evaluation (i.e., drop, crush, and puncture) of packages should address different combinations of test conditions, such as drop orientations, sequence, temperature and payload during the drop, crush and puncture tests. The combinations to be considered are those which would produce most damage to the package, challenge the most vulnerable packaging components, and cause the most cumulative damages. The evaluation of the most damage should also consider the effects of fire and water immersions following the structural tests. In this paper, the structural evaluation details of two drum-type packages, Model 9979 Type AF and Model ES-3100 Type B(U)F, are discussed. The design and performance of these packages were evaluated by physical testing of full-size prototype units. However, it is not practical to account for the worst test conditions and sequence in physical testing. Therefore, confirmatory finite element analyses have been performed to determine whether the cumulative damage resulting from the worst test sequence and conditions is acceptable. It was found for the 9979 package, the corner drop followed by corner crush causes most damage, and most unfavorably challenges its split-ring closures; for the ES-3100 package, the containment vessel (CV) experiences maximum strain following the sequence of bottom-to-lid slapdown and side crush. Although a lower temperature does not compromise their structural performance, the ES-3100 CV does experience slightly more strain because the impact limiter imparts more impact load because of its higher stiffness at lower temperature. In summary, the confirmatory analysis results show that the structural performance of the packages meets the regulatory requirements.