ERIN Engineering and Research

Walnut Creek, CA, United States

ERIN Engineering and Research

Walnut Creek, CA, United States
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Patent
Wilton Industries Inc., Deszcz, ERIN Engineering, Research, Chapple, Starr and Somers | Date: 2017-08-23

A cake stacking system for supporting a multi-tiered cake includes a center rod that extends from a bottom of a base tier of the multi-tiered cake to an interior portion of an uppermost tier of the multi-tiered cake, and a plurality of peripheral rods arranged around the center rod, the plurality of peripheral rods disposed in each tier of the multi-tiered cake except for the uppermost tier of the multi-tiered cake. The center rod includes at least one first rod having a first height and one second rod having a second height. The first height is greater than the second height. Each peripheral rod includes a tubular member and a flange coupled to one end of the tubular member. The flange maintains a vertical position of the tubular member and supports a bottom surface of an above tier of the multi-tiered cake.


Sherry R.R.,ERIN Engineering and Research | Gabor J.R.,ERIN Engineering and Research | Hess S.M.,EPRI
Reliability Engineering and System Safety | Year: 2013

In this paper we present the results of application of a risk-informed safety margin characterization (RISMC) approach to the analysis of a loss of feedwater (LOFW) event at a pressurized water reactor (PWR). This application considered a LOFW event with the failure of auxiliary feedwater (AFW) for which feed and bleed cooling would be required to prevent core damage. For this analysis the main parameters which impact core damage for the scenario were identified and probability distributions were constructed to represent the uncertainties associated with the parameter values. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the MAAP4 code. Simulation results were evaluated to determine the safety margins relative to those obtained using typical probabilistic risk assessment (PRA) modeling (success criteria) assumptions. © 2013 Elsevier Ltd.


Wolfgang R.J.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, PSA 2013 | Year: 2013

It is desired that a consistent and reasonable set of assumptions and methods be established in order to adequately evaluate various internal flood hazards without imposing excessive conservative treatments. Using a few basic kinematic principles and fluid flow analysis, the basis behind the distance from a pipe break to a particular target for a spray hazard scenario can be evaluated given the water pressure and size of break. The evaluation of the resistance afforded insulated pipe enclosed by aluminum sheet metal lagging can be analyzed in order to determine whether a pipe break resulting in a flow rate on the order of 100 gpm poses a spray hazard for nearby components. Given a typical piping arrangement for fire protection piping, the forces resulting from a full diameter pipe rupture can be analyzed to determine whether existing pipe restraints are sufficient to prevent the pipe from exceeding its tensile strength due to imposed bending stresses, which could then cause damage to other nearby equipment and structures due to "pipe whip" effects. For typical hollow metal doors found throughout a nuclear facility, the use of a simplistic finite element model can be employed to estimate the height of water required for door failure due to hydrostatic pressure.


Pupek C.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, PSA 2013 | Year: 2013

Now, in the third decade of Probabilistic Risk Assessment (PRA) usage in the industry, senior leaders and regulators are challenging the PRA community to drive an understanding of the insights and methodology that will allow further enhancement to plant safety. This next step is to develop an understanding of key characteristics unique to PRA techniques that can aid communication of insights throughout the organization. Such focused communication can help drive more thoughtful consideration of issues affecting decision-making on a variety of plant issues, from financial management to maintenance tool selection. To accomplish this lofty goal, a new paradigm in PRA education is required. In this new model, education starts with the results and creates an understanding of their derivation, which facilitates insights. This process is best served with the multi-discipline interaction required for successful PRA application and plant performance improvement. This paper discusses several risk-informed applications as tools for PRA socialization and education. It identifies the various disciplines involved with the applications and what they bring to the process with the goal of elaborating on how these experiences provide an opportunity to create understanding of PRA capability and insight that nuclear power plant personnel of all levels can relate to without turning into a training course on PRA techniques. Examples are discussed on how these processes create a reinforcing feedback loop that improves overall plant safety and key indicators. The importance of communications skills and tactics is also discussed.


Addis H.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, PSA 2013 | Year: 2013

Recent industry events have shown that a periodic PRA model update can cause a plant to cross the Mitigating System Performance Index (MSPI) green to white performance indicator threshold resulting in significant additional resources to both investigate the change in the indicator and support an additional NRC inspection. MSPI is an application that requires the PRA modeler to explicitly consider the model results prior to approval to ensure that MSPI fulfills the intended function as an ROP indicator. If a plant has low MSPI margin, the PRA modeler must use extreme caution to ensure the assumptions and modeling are as exact as necessary to preserve MSPI margin, not cause a MSPI color change and comply with all requirements of NEI 99-02. This must be done with integrity and precision to ensure that the indicator is not biased by the plant's desire to stay within the MSPI green threshold. The unique PRA modeling considerations, in light of MSPI insights and lessons learned, are explored including evaluating the impact of the PRA model changes on MSPI, timing of PRA model issuance, and MSPI-specific model reviews. The final insight is that the PRA Modeler should fully understand the impact of model changes on the MSPI value as exceeding the exact value of 1E-06 can result in a white performance indicator.


Sloane B.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015 | Year: 2015

The ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM) is responsible for development and maintenance of standards and related guidance on nuclear facility probabilistic risk assessment (PRA) and risk management. The JCNRM Subcommittee on Standards Development (SC-SD) has responsibility of developing new PRA standards in accordance with JCNRM-adopted guidelines, and there are currently five new PRA standards in development and in various stages of the consensus ballot process. The JCNRM has recently implemented a process whereby a new standard that has been approved by technical committee consensus ballot and received administrative approval from both the ASME and ANS standards boards may be initially issued for a designated trial use period. The purpose of the trial use period, during which the standard, although approved within the ASME and ANS process, is not an ANSI (American National Standards Institute) standard, is to: (a) allow standard users and other stakeholders the opportunity to "test out" the standard and provide feedback to the SC-SD and writing groups regarding technical dfJIculties, application dfJIculties, editorial issues, and so forth, and (b) obtain feedback from stakeholders on selected technical issues that have been identfled by the standards writers as possibly involving limitations related to state of technology, lack of consensus, and so forth hut were deemed not so monumental as to preclude issuance of the standard. As of this submittal, one trial use standard has been issued (Non-Light Water Reactor PRA Standard) andfour others are planned to be issued for trial use within 18-24 months. This paper summarizes the current status of trial use standards and issues the SC-SD is evaluating regarding implementation of the trial use process, and provides initial feedback on how the process has been working to date. © 2015 by the American Nuclear Society.


Baranowsky P.W.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015 | Year: 2015

This paper presents a probabilistic concept for treating early phase fire growth from fire ignition to detection, suppression, and resultant fire severity in a PRA framework that estimates the likelihood of damaging fires in a manner more consistent with the operating experience fire data. The concept and data analysis were developed for fires in electrical cabinets. The data shows that the current approach used in fire PRAs does not realistically represent the early stage fire growth in timing, magnitude and damage likelihood. An event tree approach has been adopted using insights from the data for electrical cabinet fires from EPRI's Fire Events Data Base to estimate the likelihood that early stage fires (pre T- squared fire growth) will develop into more risk significant fires. The conceptual approach comports well with the operating experience data by design, showing a substantial reduction in the likelihood of early stage electrical cabinet fires progressing to more risk significant conditions. © 2015 by the American Nuclear Society.


Wolfgang R.J.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015 | Year: 2015

An internal flood PRA involves scenarios in which a pipe rupture can result in the release of a large volume of water over a short period of time. Given the size of the flow rate and the available space available for water accumulation, the rise in water height as a function of time is necessary to be computed in order to determine the time available for operator intervention before extensive damage is experienced. To determine the available time before a critical water height is achieved, it is desired that a simplified, yet consistent, approach be used to help determine the height of water rise as a function of time due to water inflow and outflow. Accounting for area drainage and propagation of water from one adjoining compartment to another must also be taken into account in order to realistically model a given internal flood scenario. A mathematical model using mass balance equations can be used to model flow of water into and out of multiple adjoining areas subject to an internal flood event using a set of linked differential equations that define the change in water height as a function of time for each affected space. Various restrictions and openings that exist between compartments and allow propagation of water from one area to another can be modeled using a consistent set of fluid flow equations to estimate flow rates and simulate the postulated scenario as a function of time. An example showing the use of such equations for a given pipe rupture is presented to best show how a simplistic model can be created to estimate water height versus time for various rooms connected via propagation flowpaths. The use of such information can yield the available time for operator mitigation before the equipment in affected rooms is damaged due to water submergence. © 2015 by the American Nuclear Society.


Edom J.,ERIN Engineering and Research
International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015 | Year: 2015

This paper provides a method for enhancing the model rollout process using two EPRI Risk and Reliability software tools, PRA DocAssist and SYSIMP. The enhancement establishes an infrastructure for various risk ranking applications such as MOV ranking, AOV ranking and Maintenance Rule Function significance. Once established, this combination of software tools allows for a rapid completion of routine tasks from the Model Of Record (MOR) rollout process. © 2015 by the American Nuclear Society.


Grant
Agency: NSF | Branch: Fellowship | Program: | Phase: GRADUATE RESEARCH FELLOWSHIPS | Award Amount: 50.33K | Year: 2010

None

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